[Federal Register Volume 69, Number 16 (Monday, January 26, 2004)]
[Rules and Regulations]
[Pages 3698-3814]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-35]



[[Page 3697]]

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Part III





Nuclear Regulatory Commission





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10 CFR Part 71



Compatibility With IAEA Transportation Safety Standards (TS-R-1) and 
Other Transportation Safety Amendments; Final Rule

Federal Register / Vol. 69, No. 16 / Monday, January 26, 2004 / Rules 
and Regulations

[[Page 3698]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 71

RIN 3150--AG71


Compatibility With IAEA Transportation Safety Standards (TS-R-1) 
and Other Transportation Safety Amendments

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations on packaging and transporting radioactive material. This 
rulemaking will make the regulations compatible with the latest version 
of the International Atomic Energy Agency (IAEA) standards and codify 
other applicable requirements. This final rule also makes changes in 
fissile material exemption requirements to address the unintended 
economic impact of NRC's emergency final rule entitled ``Fissile 
Material Shipments and Exemptions'' (February 10, 1997; 62 FR 5907). 
Lastly, this rule addresses a petition for rulemaking submitted by 
International Energy Consultants, Inc.

EFFECTIVE DATE: This final rule is effective on October 1, 2004. 
Portions of Sec.Sec. 71.19 and 71.20 expire on October 1, 2008.

FOR FURTHER INFORMATION CONTACT: Naiem S. Tanious, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone (301) 415-6103; e-mail 
nst@nrc.gov.

SUPPLEMENTARY INFORMATION:

Contents

I. Background
II. Analysis of Public Comments
III. Discussion
    A. TS-R-1 Compatibility Issues
    Issue 1: Changing Part 71 to the International System of Units 
(SI) Only
    Issue 2: Radionuclide Exemption Values
    Issue 3: Revision of A1 and A2
    Issue 4: Uranium Hexafluoride (UF6) Package Requirements
    Issue 5: Introduction of the Criticality Safety Index 
Requirements
    Issue 6: Type C Packages and Low Dispersible Material
    Issue 7: Deep Immersion Test
    Issue 8: Grandfathering Previously Approved Packages
    Issue 9: Changes to Various Definitions
    Issue 10: Crush Test for Fissile Material Package Design
    Issue 11: Fissile Material Package Design for Transport by 
Aircraft
    B. NRC-Initiated Issues
    Issue 12: Special Package Authorizations
    Issue 13: Expansion of Part 71 Quality Assurance (QA) 
Requirements to Certificate of Compliance (CoC) Holders
    Issue 14: Adoption of the American Society of Mechanical 
Engineers (ASME) Code
    Issue 15: Change Authority for Dual-Purpose Package Certificate 
Holders
    Issue 16: Fissile Material Exemptions and General License 
Provisions
    Issue 17: Decision on Petition for Rulemaking on Double 
Containment of Plutonium (PRM-71-12)
    Issue 18: Contamination Limits as Applied to Spent Fuel and 
High-Level Waste (HLW) Packages
    Issue 19: Modifications of Event Reporting Requirements
IV. Section-By-Section Analysis
V. Criminal Penalties
VI. Issues of Compatibility for Agreement States
VII. Voluntary Consensus Standards
VIII. Environmental Assessment: Finding of No Significant 
Environmental Impact
IX. Paperwork Reduction Act Statement
X. Regulatory Analysis
XI. Regulatory Flexibility Act Certification
XII. Backfit Analysis

I. Background

    Before developing and publishing a proposed rule, the NRC began an 
enhanced public-participation process designed to solicit public input 
on the part 71 rulemaking. The NRC issued a part 71 issues paper for 
public comment (65 FR 44360; July 17, 2000). The issues paper presented 
the NRC's plan to revise part 71 and provided a summary of all changes 
being considered, both International Atomic Energy Agency (IAEA)--
related changes and NRC-initiated changes. The NRC received 48 public 
comments on the issues paper. The NRC enhanced public participation 
process included establishing an interactive Web site and holding three 
facilitated public meetings: a ``roundtable'' workshop at NRC 
Headquarters, Rockville, MD, on August 10, 2000, and two ``townhall'' 
meetings--one in Atlanta, GA, on September 20, 2000, and a second in 
Oakland, CA, on September 26, 2000. Oral and written comments, received 
from the public meetings by mail and through the NRC Web site, in 
response to the issues paper were considered in drafting the proposed 
rule.
    The NRC published the proposed rule in the Federal Register on 
April 30, 2002 (67 FR 21390), for a 90-day public comment period. In 
addition to approving the publication of the proposed rule, the 
Commission also directed the NRC staff to continue the enhanced public 
participation process. The NRC staff held two public meetings to 
discuss the proposed rule. The first meeting was held in Chicago, 
Illinois, on June 4, 2002, and the second was held at the TWFN 
Auditorium, NRC Headquarters, on June 24, 2002. In addition, the 
Department of Transportation (DOT) staff participated in these 
meetings. Transcripts of these meetings were made available for public 
review on the NRC Web site. The public comment period closed on July 
29, 2002. A total of 192 comments were received. Although many comments 
were received after the closing date, all comments were analyzed and 
considered in developing this final rule.

Past NRC-IAEA Compatibility Revisions

    Recognizing that its international regulations for the safe 
transportation of radioactive material should be revised from time to 
time to reflect knowledge gained in scientific and technical advances 
and accumulated experience, IAEA invited Member States (the U.S. is a 
Member State) to submit comments and suggest changes to the regulations 
in 1969. As a result of this initiative, the IAEA issued revised 
regulations in 1973 (Regulations for the Safe Transport of Radioactive 
Material, 1973 edition, Safety Series No. 6). The IAEA also decided to 
periodically review its transportation regulations, at intervals of 
about 10 years, to ensure that the regulations are kept current. In 
1979, a review of IAEA's transportation regulations was initiated that 
resulted in the publication of revised regulations in 1985 (Regulations 
for the Safe Transport of Radioactive Material, 1985 edition, Safety 
Series No. 6).
    The NRC also periodically revises its regulations for the safe 
transportation of radioactive material to make them compatible with 
those of the IAEA. On August 5, 1983 (48 FR 35600), the NRC published a 
revision of 10 CFR part 71. That revision, in combination with a 
parallel revision of the hazardous materials transportation regulations 
of DOT, brought U.S. domestic transport regulations into general accord 
with the 1973 edition of IAEA transport regulations. The last revision 
to part 71 was published on September 28, 1995 (60 FR 50248), to make 
part 71 compatible with the 1985 IAEA Safety Series No. 6. The DOT 
published its corresponding revision to title 49 on the same date (60 
FR 50291).
    The last revision to the IAEA Safety Series 6, Safety Standards 
Series ST-1, was published in December 1996, and revised with minor 
editorial changes in June 2000, and redesignated as TS-R-1.
    Historically, the NRC has coordinated its part 71 revisions with 
DOT, because DOT is the U.S. Competent Authority for transportation of 
hazardous materials. ``Radioactive Materials'' is a subset of 
``Hazardous Materials'' in 49 CFR under DOT authority. Currently,

[[Page 3699]]

DOT and NRC co-regulate transport of nuclear material in the United 
States. The NRC is continuing with its coordinating effort with the DOT 
in this rulemaking process. Refer to the DOT's corresponding rule for 
additional background on the positions presented in this final rule.

Scope of 10 CFR Part 71 Rulemaking

    As directed by the Commission, the NRC staff compared TS-R-1 to the 
previous version of Safety Series No. 6 to identify changes made in TS-
R-1, and then identified affected sections of part 71. Based on this 
comparison, the NRC staff identified 11 areas in part 71 that needed to 
be addressed in this rulemaking as a result of the changes to the IAEA 
regulations. The NRC staff grouped the part 71 IAEA compatibility 
changes into the following issues: (1) Changing part 71 to the 
International System of Units (SI) only; (2) radionuclide exemption 
values; (3) revision of A1 and A2; (4) uranium 
hexafluoride (UF6) package requirements; (5) introduction of 
the criticality safety index requirements; (6) type C packages and low 
dispersible material; (7) deep immersion test; (8) grandfathering 
previously approved packages; (9) changes to various definitions; (10) 
crush test for fissile material package design; and (11) fissile 
material package design for transport by aircraft.
    Eight additional NRC-initiated issues (numbers 12 through 19) were 
identified by Commission direction and NRC staff consideration for 
incorporation in part 71. These NRC-initiated changes are: (12) Special 
package authorizations; (13) expansion of part 71 Quality Assurance 
(QA) requirements to Certificate of Compliance (CoC) holders; (14) 
adoption of the American Society of Mechanical Engineers (ASME) code; 
(15) change authority for Dual-Purpose Package Certificate holders; 
(16) fissile material exemptions and general license provisions; (17) 
decision on petition for rulemaking on PRM-71-12, Double Containment of 
Plutonium; (18) contamination limits as applied to Spent Fuel and High-
Level Waste (HLW) packages; and (19) modifications of event reporting 
requirements. The first 18 issues were published for public comment in 
an issues paper in the Federal Register on July 17, 2000 (65 FR 44360). 
Also, the authority citation for part 71 has been corrected to include 
section 234.
    This final rule has been coordinated with DOT to ensure that 
consistent regulatory standards are maintained between NRC and DOT 
radioactive material transportation regulations, and to ensure 
coordinated publication of the final rules by both agencies. The DOT 
also published its proposed rule regarding adoption of TS-R-1 April 30, 
2002 (67 FR 21328).

II. Analysis of Public Comments

    As previously stated, the NRC held two facilitated public meetings 
in 2002 to discuss and hear public comments on the proposed rule. 
(Three other facilitated public meetings were held in 2000 before 
drafting the proposed rule.) Each of these meetings was transcribed by 
a court reporter. The meeting transcripts and condensed summaries of 
the comments made in the meeting are available to the public on the 
NRC's interactive rulemaking Web site at http://ruleforum.llnl.gov. and 
the Public Document Room (PDR) located at One White Flint North, 11555 
Rockville Pike, Room O-1F23, Rockville, MD. The NRC has made copies of 
publicly released documents available on the Web site at http://
www.nrc.gov/waste/spent-fuel-transp.html.
    This section provides a summary of the general comments not 
associated with the 19 issues but rather with general topics related to 
this rule and the rulemaking process. These are organized under the 
following subheadings: Compatibility with IAEA and DOT standards, 
Regulatory Analysis (RA) and Environmental Assessment (EA), State 
Regulations, Terrorism, Adequacy of NRC Regulations and Rulemaking 
Process, Proposed Yucca Mountain Facility, and Miscellaneous (including 
comments to DOT). A summary of public comments associated with a 
specific issue is included in Section III of this SUPPLEMENTARY 
INFORMATION.

Compatibility With IAEA and DOT Standards

    Comment. Several commenters generally supported NRC's efforts to be 
consistent with IAEA regulations. The particular reasons for this 
support varied among commenters but included such issues as approving 
of harmonization and encouraging NRC's coordination with DOT. For 
example, some commenters stated that harmonization enhances the 
industry's ability to import shipments and conduct business in 
compliance with both national and international regulations. One 
commenter urged the NRC to move swiftly to complete this rulemaking 
effort and to remain consistent with DOT regulations. One commenter 
stated that uniform international regulations were in the public's best 
interest for the safe movement of nuclear materials. Further, this 
commenter urged the NRC to accelerate the ``harmonization'' with 
international regulations to simplify procedures for companies that 
ship nuclear waste both domestically and internationally.
    Response. The NRC acknowledges these comments, and the NRC 
continues to work to finalize this rule as expeditiously as possible. 
As with the issuance of the proposed rule, the NRC will continue to 
coordinate closely with the DOT in this effort to ensure consistency 
between regulations for the transportation of certain radioactive 
materials.
    Comment. A commenter supported harmonization but said that adoption 
of new or modified requirements into the domestic regulations for 
transportation of radioactive materials must be justified in terms of 
cost and the need for improved safety and performance. The commenter 
added that some of the changes, including the additional technical 
complexity of the proposed regulations (e.g., nuclide specific 
thresholds), are not warranted based on the history of performance in 
the transportation of radioactive materials.
    Another commenter noted several areas of incompatibility between 
DOT and NRC proposed rules. The commenter also suggested that NRC work 
with DOT to agree on a consistent approach in organizing the A1 and A2 
values for international shipments in Table A-1. A third commenter 
noted that DOT has already issued a proposed rule, HM 232, which 
focuses on using the registration program to affect the enhancement and 
security of radioactive materials in transport.
    Response. NRC's goal is to harmonize our transportation regulations 
to be consistent with IAEA and DOT, while ensuring that the 
requirements adopted will benefit public health, safety, and the 
environment. The NRC has conducted an evaluation of the radionuclide-
specific thresholds (the exemption values), including a regulatory 
analysis and an environmental assessment, and concluded that adoption 
of these values is warranted, in spite of the technical complexity. NRC 
has been working with the DOT. The NRC has completed a regulatory 
analysis that supports harmonization in terms of cost and regulatory 
efficiency.
    Comment. One commenter stated that NRC should use the latest 
medical knowledge from independent sources (i.e., not IAEA or 
International Commission on Radiological Protection (ICRP) data) 
regarding the medical effects of radiation.
    Response. The NRC considers a variety of sources of information

[[Page 3700]]

concerning the health effects attributed to exposure to ionizing 
radiation. Two primary sources of information are the National Research 
Council/National Academy of Sciences (NAS) and the United Nations 
Scientific Committee on the Effects of Atomic Radiation (UNSCEAR). Both 
groups provide an independent and comprehensive evaluation of the 
health risks associated with radiation exposure. The NRC currently is 
sponsoring an NAS review of information from molecular, cellular, and 
animal studies of radiation, other environmental exposures, and 
epidemiologic studies to evaluate and update previous reviews of the 
health risks related to exposure to low-level ionizing radiation. These 
studies focus on the latest published information available.
    Comment. Several commenters questioned the credibility of the IAEA 
and the ICRP because these organizations are not publicly accountable. 
Three of the commenters further questioned the process of the NRC 
simply accepting what the IAEA does, noting that agencies in Europe 
have challenged ICRP assumptions. One of these commenters stated that 
regulated or potentially regulated bodies should be allowed more 
involvement in the IAEA decisionmaking process. Furthermore, the 
suggested lack of public involvement led one commenter to express a 
general lack of trust for these organizations and question the 
credibility of their conclusions. This lack of public involvement was 
at issue with another commenter who added that the proposal would only 
``make things easier for the transportation and nuclear industries at 
the expense of public health.''
    Response. The United States is represented at the IAEA for 
transportation issues through the DOT acting as Competent Authority 
(the official U.S. representative organization). The NRC consults with 
DOT on issues related to nuclear material transport. NRC disagrees with 
the statement that the NRC simply accepts what the IAEA does. When the 
NRC (and the DOT) seeks to amend its regulations to harmonize with 
IAEA's, it does so through a deliberate and open process via 
rulemaking. The public has been afforded in the past, and will continue 
to be afforded, the opportunity to comment on DOT's and NRC's proposed 
rulemakings. This effort can result in NRC regulations not matching the 
IAEA guidance. Further, the NRC does not ``simply accept'' the IAEA 
standards. In many instances, the NRC has chosen to implement 
regulations that differ from the IAEA's. Issues 7 and 11 of this final 
rule, discussed elsewhere in this SUPPLEMENTARY INFORMATION, are just 
two examples of where NRC has differed from the IAEA requirements by 
implementing more stringent requirements.
    Information on the IAEA and ICRP can be found at their respective 
Web sites: www.iaea.org and www.icrp.org. These Web sites provide 
background on each organization that should address the concerns about 
the credibility of each organization.
    Comment. One commenter stated that the burden of proof for 
departing from IAEA standards is shifted by the regulators to the 
regulated entities. Another commenter suggested that the burden of 
proof for rejecting the proposed regulatory changes is being shifted to 
citizens and stakeholders.
    Response. Both the NRC and DOT are participating members of the 
IAEA and have direct input to the development of new transportation 
standards. Before DOT or NRC proposes U.S. regulations for 
harmonization with IAEA standards, each agency completes a technical 
evaluation and makes a determination if each new standard should be 
adopted by the U.S. The public involvement process for rulemaking 
solicits stakeholders to suggest changes to proposed rule language or 
to suggest the rejection of a proposed regulatory change. With 
sufficient justification, public comments have resulted in modification 
to regulatory text.
    Comment. One commenter asked if either NRC standards or IAEA's 
could protect the public from ``real world'' problems. The commenter 
inquired how NRC accounts for the fact that a cask might burn for 
longer than existing standards require it to withstand fire. The 
commenter believed that such rationales were particularly relevant in 
light of recent incidents, such as the Baltimore Tunnel fire and the 
Arkansas River bridge accident.
    Response. The NRC notes the questions on how realistic the 
transportation standards established by the NRC and the IAEA are. Both 
NRC and IAEA standards require that cask designs be able to withstand 
hypothetical accident conditions. The conditions bound (or are more 
severe than) those conditions that would be expected in the vast 
majority of real world accidents and therefore provide protection for 
the cask designs. Additionally, the NRC has periodically revisited and 
evaluated the effects of actual accidents to look at the forces and the 
challenges that would be presented to casks in ``real world'' 
transportation accidents. For example, in response to the Baltimore 
Tunnel fire, the NRC staff has conducted two sets of independent 
analyses and has determined that the conditions that existed in the 
fire would not have caused a breech of a current spent fuel 
transportation cask design had it been located in the tunnel for the 
duration of the fire.
    Comment. One commenter stated that the timeline by which NRC would 
adopt IAEA requirements should be changed. The commenter also stated 
that the current 2-year cycle for changes is too frequent.
    Response. The timeline for adopting IAEA standards and the cycle 
for making changes at the IAEA are beyond the scope of this rulemaking.
    Comment. One commenter stated that the proposed rule might allow 
weakening of transportation cask safety testing and increase the risk 
of the release of radioactive materials during transportation 
accidents.
    Response. This concern is acknowledged, but the NRC does not 
believe that this rule weakens testing standards.
    Comment. One commenter stated that all radioactive shipments should 
be regulated and labeled so that transportation workers and emergency 
responders are aware of the risk.
    Response. The comments are acknowledged. DOT regulations include 
requirements for labels, markings, and placarding packages and 
conveyances of radioactive materials, and training of Hazmat workers. 
Existing and proposed regulations for the transportation of radioactive 
materials consider the potential risk to workers and emergency 
responders of exposure to these materials. The NRC believes the 
thresholds for regulation of the transportation of radioactive 
materials protect the health and safety of workers and emergency 
responders.
    Comment. One commenter pointed out that due to the increase in the 
number of nuclear shipments, the NRC and DOT must strengthen their 
standards to protect the millions of people, thousands of schools, and 
hundreds of hospitals residing directly along transportation routes.
    Response. The NRC routinely reevaluates the effectiveness of its 
regulations to ensure that it is meeting its mission to protect the 
public health and safety. In regulating safe and secure transport of 
spent nuclear fuel, the NRC has conducted risk studies to consider the 
fact that a large number of shipments might be made to a future 
geological repository using current generation cask designs. These 
studies have confirmed that the current NRC regulations are robust and 
protective of the public during transportation of

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spent fuel. Therefore even with an increase in the number of shipments, 
these shipments can be made safely in large numbers to a centrally 
located storage facility.
    Comment. On behalf of the nuclear industry, one commenter said that 
harmonization is logical in terms of cost and safety. Harmonized rules 
and uniform standards and criteria allow members of the nuclear 
industry to know how safe a package is, regardless of where it comes 
from. Because many other nations have already adopted many of these 
proposed rules, U.S. transporters are already required to meet these 
standards in many cases. The commenter also voiced support for 
exempting certain domestic shipments from these international 
regulations.
    Response. Harmonization with TS-R-1 should maintain the safety of 
shipments of radioactive materials while eliminating the need to 
satisfy two different regulatory requirements (i.e., domestic versus 
international shipments). The NRC believes that by clarifying and 
simplifying shipping requirements, harmonization will help all who are 
involved in the transport of radioactive material to comply 
successfully with regulations.
    Comment. One commenter stated that there has already been much 
deliberation over the proposed regulations. He stated that his 
organization and the industry at large have been looking at these 
proposed changes for well over 10 years.
    Response. The comments are acknowledged.
    Comment. One commenter stated that harmonization is a ``value 
neutral process'' and isn't necessarily good or bad.
    Response. Harmonization can be viewed as a value neutral process, 
although the NRC believes that harmonizing domestic and international 
regulations generally improves efficiency and safety in the transport 
of radioactive material. NRC's proposed changes are based upon the 
careful evaluation of specific issues and provisions in TS-R-1. At this 
level, the NRC believes that the negative (i.e., costs) or positive 
(i.e., benefits) value of a particular change can be assessed 
effectively. These costs and benefits have been carefully evaluated in 
our decisionmaking process.
    Comment. Four commenters opposed harmonizing rules. One commenter 
opposed harmonization because it ``appears to be occurring to satisfy 
demands of the nuclear industry and affected governmental bodies'' to 
facilitate commerce, rather than in the interest of public safety. 
Another commenter noted that the primary objective of these changes 
should be to protect public health, safety, and the environment. 
Another commenter argued that harmonization should not be used as a 
justification for violating a country's sovereignty or a State's right 
to maintain stringent standards. The commenter said that U.S. rules 
were already harmonized before these proposed changes and that the 
authors of international regulations should not dictate U.S. 
regulations. The fact that other countries have adopted the IAEA 
regulations is not sufficient justification for the U.S. to adopt these 
regulations. The commenter agreed that some degree of harmonization 
makes sense but emphasized that the U.S. needs to maintain control over 
its own rules.
    Response. The IAEA periodically updates international regulations 
for the safe transport of radioactive material in response to advances 
in scientific knowledge and technical experience. These changes are 
implemented with the purpose of improving public safety, as well as 
facilitating commerce. The U.S. has substantial input into the IAEA 
development of these periodic revisions through official representation 
by the DOT. While the NRC aims to harmonize its regulations closely 
with those issued by the IAEA, NRC independently evaluates proposed 
changes in the interest of protecting public health, safety, and the 
environment. This rule reflects this extensive process; NRC routinely 
suggests adoption or partial adoption of certain provisions and 
nonadoption of others.
    Comment. Two commenters asked if NRC could quantifiably prove that 
harmonization is necessary. One asked if NRC's failure to comply with 
the IAEA regulations has disrupted commerce or jeopardized public 
safety, and whether members of the international community have accused 
the U.S. of disrupting commerce by not complying with these 
regulations.
    Response. DOT and NRC accomplish harmonization by adopting domestic 
rules that are compatible with international rules. DOT and NRC rules 
may differ from those of IAEA where it is necessary to reflect domestic 
practices. However, these differences are kept to a minimum because 
regulatory differences can lead to confusion and errors and can result 
in unsafe conditions or events. U.S. failure to comply with 
international safety regulations could easily result in disruption of 
U.S. participation in international radioactive material commerce, with 
no commensurate justifiable safety benefit, because other IAEA Member 
States are under no obligation to accept shipments that do not comply 
with international regulations.
    Comment. One commenter wanted to know how the IAEA drafted its 
regulations and statistics. The commenter questioned who the IAEA is 
and why NRC should accept its statistics. The commenter also asked how 
much input the American public has had on these regulations and noted 
that Congress and the public have previously rejected IAEA regulations.
    Response. The comments concerning the IAEA standards development 
process and U.S. citizen input to that process are both beyond the 
scope of this rulemaking. However, as noted in the public meetings held 
to obtain comments on the proposed rule, DOT is mandated by law to help 
formulate international transportation standards, and to ensure that 
domestic regulations are consistent with international standards to the 
degree deemed appropriate. The law permits DOT the flexibility to 
accept or reject certain of the international standards. The NRC/DOT 
evaluation of the IAEA standards has resulted in the two parallel sets 
of final rule changes. Rejection of an IAEA standard could be based on 
technical criteria as well as on public comment on proposed rules. The 
IAEA has Member States that develop standards as a collegial body, and 
the U.S. is one of those Member States.
    Comment. Several commenters urged NRC to improve its scientific 
understanding and basis for the proposed rulemaking. Two commenters 
suggested that NRC complete the comprehensive assessments of TS-R-1 and 
future IAEA standards, the Package Performance Study (PPS), and full-
scale cask tests before proceeding with this rulemaking. A commenter 
stressed that ICRP does not represent the full range of scientific 
opinion on radiation and health and ignores concepts such as the 
bystander effect and synergism of radiation with other environmental 
contaminants. This commenter also stated that the exposure models used 
to justify certain exposure scenarios are inadequate.
    Response. The NRC acknowledges these comments and notes that NRC 
participates or monitors the work of major, national and international, 
scientific organizations in the fields of health physics and radiation 
protection. As such, NRC has access to the latest scientific advances. 
Moreover, the NRC has completed an assessment of TS-R-1 as part of the 
development of this rule. The PPS is a research project independent of 
this rulemaking. Also,

[[Page 3702]]

see the following comment regarding the ICRP.
    Comment. Several commenters stated that the IAEA rulemaking process 
is not democratic, and their documents are not publicly available and 
were developed without public knowledge or input. One commenter 
suggested that the public should have had an opportunity to ``comment 
on or otherwise participate in the earlier formation of the IAEA 
rules.'' Another commenter proposed that the NRC act as an intermediary 
between public opinion and IAEA by improving communications with the 
public and regulated bodies, providing advanced notice of rulemakings, 
and receiving comments on proposed rules.
    Response. The NRC acknowledges the comments about the IAEA 
rulemaking process, the ICRP representation of scientific opinion, and 
the observation on NRC's role as intermediary between the American 
public and the IAEA, but each of these comments brings up issues that 
are beyond the scope of the proposed rulemaking. Therefore, no changes 
were made to this rulemaking. The NRC notes that the IAEA has begun to 
discuss ways to foster public participation in its standards 
development process.
    Comment. Several commenters stated that IAEA and ICRP regulations 
should not dictate domestic U.S.-based regulations. Two commenters 
stated that IAEA does not necessarily consider the risk-informed, 
performance-based standards that are important to rulemaking in the 
U.S. The commenters added that the NRC must recognize that while IAEA 
standards generally have good technical bases, they are consensus 
standards that do not necessarily consider the risk-informed, 
performance-based aspects of regulations that we have developed in the 
U.S.
    Response. The NRC acknowledges the comment about IAEA and ICRP 
regulations dictating U.S. based regulations and notes that this 
comment is not accurate and is considered to be an opinion. The NRC is 
a participating member of both the IAEA and the ICRP, and neither body 
dictates to the NRC what regulations or standards must be adopted. As a 
participant, the NRC suggests transportation standard changes and as 
such, the NRC both proposes and comments on the language of new 
standards. This participation permits the NRC to infuse its ideas on 
risk-informed regulations, when possible.
    Comment. The effort to harmonize regulations was supported by 
several commenters. One commenter spoke for Agreement States and 
expressed support for harmonizing regulations. Two others explained 
that the benefit of harmonization would be consistent national and 
international regulations and improved safety, yet U.S. regulators (and 
regulations) would retain the legal authority to act when and as 
necessary. Another commenter emphasized that given how new information 
is found all the time and the IAEA is on a 2-year standards revision 
schedule, it does not make sense to hold back harmonizing U.S. 
standards with international standards pending the outcome of any 
studies.
    Response. The NRC believes that its effort to promote regulatory 
harmonization will maintain and/or improve safety, increase regulatory 
efficiency and effectiveness, as well as reduce unnecessary regulatory 
burden. The NRC's aim is to harmonize its regulations with IAEA 
regulations by adopting many of the provisions in TS-R-1. However, the 
NRC does not propose wholesale adoption of TS-R-1, but only when 
adoption provides the best opportunity to maintain and/or improve 
public safety, health, and the environment.

Regulatory Analysis (RA) and Environmental Assessment (EA)

    Comment. Several commenters found the RA to be deficient in various 
aspects. One commenter asserted that updated quantitative data should 
be included in the RA that would include the following information: the 
number of exempt and nonexempt packages; the number of exempt and 
nonexempt shipments; the average number of packages per shipment; and 
the detailed information on curie counts by shipment categories. The 
commenter noted that all stakeholders are affected by these 
deficiencies, notably public information groups and Western States.
    Two commenters focused on the RA's cost analysis with one stating 
that no changes should be made without a cost analysis and the other 
stating that the RA had not adequately considered the cost of the 
proposed rule. The second of these commenters stated that specific dose 
information, calculations, and information regarding the impact of the 
new regulations should have been included in the draft RA and EA. They 
found the RA to be deficient because of its failure to recognize likely 
impacts of the changes to the double containment of plutonium 
regulations, particularly regarding the agreement between the Western 
Governors' Association, the individual Western States, and the 
Department of Energy (DOE) for a system of additional transportation 
safeguards.
    Response. Quantitative data was requested throughout the rulemaking 
process. These requests were made during the development of the 
proposed rule, and a request was again made in the proposed rule. Where 
this information was available, it was used in the development of NRC's 
proposed positions. To the extent that information was provided, it has 
been considered in the development of NRC's final position.
    Comment. One commenter asserted that the proposed rule is a major 
Federal action, thus deserving of a full Environmental Impact Statement 
(EIS). The commenter also stated that an EIS dating from 1977 and a 
study dating from 1985 do not suffice as adequate analysis of the 
proposed rule's impact, due to changes ``in population, in land use, in 
the transportation system, in laws, in issues of national security.''
    Response. NRC acknowledges this comment and notes that it has 
prepared an EA. Based on the results of the EA, the NRC staff has 
concluded that this rule is not a major Federal action requiring an 
EIS. As noted in the proposed rule, NRC is interested in receiving 
additional data, and to the extent that the data was received, it was 
included in the analyses leading up to the final rule.
    Comment. One commenter said that the EA and the rulemaking are too 
carefully tied together. The commenter said that this fact precludes 
NRC from actually finding an environmental impact from the rule.
    Response. The draft EA is a study that is required as part of a 
rulemaking to ensure that the potential impacts to public health and 
safety and the environment are adequately evaluated as part of the 
decisionmaking process. As such, the rule and the EA are necessarily 
``tied together.''
    Comment. Two commenters found the EA to be deficient in various 
aspects. One commenter stated that specific dose information, 
calculations, and information regarding the impact of the new 
regulations should have been included in the draft EA and RA.
    A commenter believes that the EA and RA lack the following pieces 
of information: the number of exempt and nonexempt packages; the number 
of exempt and nonexempt shipments; the average number of packages per 
shipment; and the detailed information on curie counts by shipment 
categories. One commenter believes that the EA should include 
transportation scenarios, updated data rather than 1982 data, and a 
quantitative analysis along with a qualitative analysis.

[[Page 3703]]

    The NRC was criticized for a portion of the EA (page 43), which 
first identifies information necessary to make a risk-informed decision 
on the proposed regulation and then discusses the lack of information 
in the EA. The commenters noted a discrepancy in NRC's efforts, 
particularly the number of NRC staff and resources devoted to this 
rulemaking for the past 2 years versus the lack of resources devoted to 
updating the 1982 data. They stated that the costs associated with the 
Type C package changes were not included in the EA and that process 
irradiators are shipping sources equaling about 50 million curies, much 
greater than the curie count listed in the proposed rulemaking.
    Response. The NRC acknowledge the comments regarding the lack of 
information in some portions of the draft RA and EA. The draft EA and 
RA were developed based on the best information available to the NRC at 
the time. Moreover, NRC solicited in the proposed rule FRN, additional 
information on the costs and benefits of the proposed requirements, 
including the Type C package changes. All the information received has 
been considered in NRC's final decision. The NRC staff notes that the 
majority of the proposed changes are such that the specific dose 
information and calculations are not required to determine the 
appropriateness of adopting or not adopting the change being 
considered.
    Comment. One commenter expressed concerns about NRC's findings of 
``no significant impact'' on radionuclide-specific activity values for 
a number of issues. The commenter requested that more detailed 
information be provided ``on how many and which radionuclide levels 
will rise or fall'' as a result of proposed changes. The commenter also 
asked the NRC to define its use of ``significantly'' and to explain how 
it determined the level of ``risk.''
    Response. Detailed information on the identity of radionuclides 
whose specific activity values rise or fall relative to the previous 
definition of 70 Bq/g (0.002 [mu]Ci/g) may be determined by inspection 
of Table A-2. The context for ``significantly'' is provided in the 
background section. NRC has used estimated dose to the public, as 
determined through the use of radionuclide transport scenarios, as an 
indicator of risk.
State Regulations
    Comment. One commenter asked if these new regulations would 
threaten a State's right to regulate radioactive materials that NRC has 
deregulated. Two commenters stated opposition to the proposed rule due 
to their belief that it would lower standards. The first commenter 
stated that the proposed rule would override State and local laws that 
are stricter than Federal regulations while the second commenter stated 
that the proposed rule would reduce environmental protection. Four 
commenters added that ``harmonization'' with international law was a 
poor and ultimately insufficient justification to weaken U.S. 
regulations.
    Response. State and local governments do not have authority to set 
regulations for the transportation of radioactive materials that are 
stricter or more stringent than those of the Federal government. In 
accordance with section 274b of the Atomic Energy Act, as amended, 
Agreement States programs must be compatible with those of the NRC for 
the regulation of certain radioactive materials to assume authority for 
the regulations of these materials from the NRC. Because of this, the 
Commission developed the ``Policy Statement on Adequacy and 
Compatibility of Agreement State Programs'' which became effective on 
September 3, 1997 (62 FR 46517). One of the provisions of this Policy 
Statement is that an Agreement State should adopt program elements that 
apply to activities that have direct and significant effects in 
multiple jurisdictions' elements in an essentially identical manner as 
those of the NRC (see definition of Compatibility Category B in section 
VI of this notice). This is needed to eliminate any conflicts, 
duplications, gaps, or other conditions that would jeopardize an 
orderly pattern in the regulation of radioactive materials on a 
nationwide basis. Those part 71 requirements applicable to materials 
regulated by Agreement States are designated as Category B and must be 
adopted in an essentially identical manner as those of the NRC because 
they apply to activities that have direct and significant effects in 
multiple jurisdictions.

Terrorism Concerns

    Comment. Six commenters expressed concern with the increased threat 
of terrorism and its impact on radioactive material transport. One 
commenter suggested that shipping standards be strengthened due to both 
an increased threat of terrorist attacks and the decline in rail, 
highway, air, and waterway infrastructure. Two commenters stated that 
they were concerned that many of the new regulations would make 
transported radioactive material more vulnerable to terrorist attacks 
and wanted to know how NRC anticipated responding to the threat of 
these attacks. Three commenters mentioned that the threat of terrorism 
should be taken into account when changing container regulations, with 
one commenter highlighting double versus single containment of 
plutonium. The final commenter stated that the NRC should reconsider 
the scope of the proposed rule due to the ``altered circumstances of 
our nation's vulnerability to terrorist attack.'' The commenter also 
suggested that the proposed rule be withdrawn and that the NRC 
``recalculate the full adverse consequences and the full long-term 
financial, health, and environmental costs to the public, the nation, 
and the economy of worst case terrorist actions.'' The commenter also 
stated that in a time of increased national security threats, the 
safety of containerization must be maximized.
    Response. As discussed on the NRC's Web site (see www.nrc.gov/what-
we-do/safeguards/911/faq.html), most shipments of radioactive materials 
involve materials such as pharmaceuticals, ores, low-level radioactive 
waste, and consumer products containing radionuclides (e.g., watches, 
smoke detectors). A variety of Federal and State government agencies 
regulate the shipment of radioactive materials.
    High-level nuclear waste materials, such as spent nuclear fuel, are 
transported in very heavy, robust containers called ``casks.'' Over the 
past 30 years, approximately 1300 shipments of commercially generated 
spent fuel have been made throughout the U.S. without any radiological 
releases to the environment or harm to the public. Federal regulations 
provide for rigorous standards for design and construction of shipment 
casks to ensure safe and secure transport of their hazardous contents. 
Casks must meet extremely demanding standards to ensure their integrity 
in severe accident environments. Therefore, the design of casks would 
make any radioactive release extremely unlikely. After September 11, 
2001, the NRC issued advisories to licensees to increase security 
measures to further protect the transportation of specific types of 
radioactive materials, including spent fuel shipments. Additional 
measures have been imposed on licensees shipping specific quantities of 
radioactive material.
    Comment. Another commenter, who lives near a route proposed for 
shipping nuclear waste across the country, recommended that NRC 
strengthen radioactive transport regulations. One commenter opposed the 
adoption of new transport regulations that reduce

[[Page 3704]]

the protection to the public from transporting nuclear wastes.
    Response. The NRC believes that the regulations contained in part 
71 adequately protect public health and safety. The changes being 
adopted will not result in any undue increase in risk to public health, 
safety, or the environment.
    Comment. Several commenters were concerned that the proposed 
regulations may increase vulnerability to terrorist threats using 
radioactive materials. A commenter believes that labeling radioactive 
materials could aid terrorists by identifying the packages as 
radioactive, while another commenter stated that shipments with or 
without labels provided potential terrorists with the materials for a 
dirty bomb. Another commenter requested that NRC put protective 
measures into place at ports and to guard all nuclear shipments with 
U.S. military forces. One commenter stated that nuclear shipments 
should be transported at off-peak hours while all side roads, tunnels, 
bridges, overpasses, railroad crossings, access to exit ramps, etc., 
should be secured before the transport vehicle arrives, and that NRC 
should create a ``vehicle-free'' buffer zone ahead and behind the 
shipment. This same commenter advocated FBI background checks on all 
transporters, drivers, and crew workers involved with nuclear 
transport. Two commenters asserted that all new rules should be mindful 
to the threat of terrorism, which would be superior to considering 
terrorism in separate rules.
    Response. The NRC acknowledges these comments and notes that NRC 
has taken immediate regulatory actions to address the potential for 
terrorist activities; these include issuing orders and advisories to 
its spent fuel licensees prior to initiating rulemaking which takes a 
longer time, and initiating shipment vulnerability studies. Also, the 
NRC will make the necessary rule changes, based on these studies, as 
appropriate. Moreover, the NRC staff notes that several of the comments 
above were addressed in recent regulations (March and May, 2003), which 
were published jointly by the Department of Homeland Security and the 
DOT requiring shippers and carriers to submit security plans and 
requiring background checks on drivers.

Adequacy of NRC Regulations and Rulemaking Process

    Comment. Three commenters believe that the NRC should better 
account for low-level radiation. One commenter stated that NRC should 
use the latest medical knowledge from independent sources (i.e., not 
IAEA or ICRP data) regarding the medical effects of radiation. Another 
commenter stated that low-level radiation could cause cell death, 
cancer, genetic mutations, leukemia, birth defects, and reproductive, 
immune, and endocrine system disorders. This commenter added that long-
term exposure to low levels of ionizing radiation could be more 
dangerous than short-term exposure to high levels. Another commenter, 
who was similarly concerned with low dose and low dose-rate radiation, 
stated that ``arguments of nuclear industry proponents that new 
information need not be considered is invalid and since the NRC's legal 
mandate is to protect the public's health and safety'' the NRC needs to 
consider ``cautionary information that is now available in the peer 
reviewed literature.'' The commenter suggested that NRC not focus on 
the ``standard man'' but instead focus on the ``most susceptible 
portions of the population--ova, embryo, fetus, rapidly growing young 
child, elderly, and those with impaired health'' when drafting 
regulations. Lastly, the commenter implied that NRC should attempt to 
``assess and incorporate impacts of additive exposures to other forms 
of life and to ecosystems'' as well as the impacts associated with ``an 
individual recipient of the combinations of and synergies among 
radiation and other contaminants to which people are exposed.''
    Response. As discussed on the NRC's Web site (see http://
www.nrc.gov/reading-rm/doc-collections/fact-sheets/bio-effects-
radiation.html, radiation may kill cells, induce genetic effects, and 
induce cancer at high doses and high dose rates. However, for low 
levels of radiation exposure at low dose exposure rates, health effects 
are so small they may not be detected. No birth defects or genetic 
disorders among the children born to atomic bomb survivors from 
Hiroshima and Nagasaki have been observed at low doses of radiation, 
i.e., < 25 rad (Chapter 6, ``Other Somatic and Fetal Effects,'' of Beir 
V, Health Effects of Exposure to Low Levels of Ionizing Radiation; 
National Research Council, 1990). Consequently, few if any similar 
effects are expected from exposure to low doses of ionizing radiation. 
Moreover, there is no epidemiology data, published in peer reviewed 
journals, to support the concern expressed by the commenter that long-
term exposure to low levels of radiation may be more dangerous than 
short-term exposures to high levels. Humans have evolved in a world 
constantly exposed to low levels of ionizing radiation. The average 
radiation exposure in the U.S. from natural sources is 3.0 mSv (300 
mrem) per year. Although radiation can have health effects at high 
doses and dose rates, for low levels of radiation exposure at low dose 
exposure rates, the incidence of biological effects is so small that it 
may not be detected. For example, information developed by the Health 
Physics Society suggests that the incidence of health effects, if they 
exist below 10,000 mrem (100 mSv), is too small to be observed. People 
living in areas having high levels of background radiation--above 10 
mSv (1,000 mrem) per year, such as Denver, Colorado, have shown no 
adverse health effects.
    The NRC actively and continually monitors research programs and 
reports concerning the health effects of ionizing radiation exposure. 
NRC staff monitors the Low Dose and Low Dose Rate Research Program 
sponsored by the Department of Energy (DOE). The research project is 
designed to better understand the biological responses of molecules, 
cells, tissues, organs, and organisms to low doses of radiation. NRC 
also is co-funding a review of the Biological Effects of Ionizing 
Radiation (BEIR) by the National Research Council. The BEIR committee 
will also review and evaluate molecular, cellular, and animal exposure 
data and human epidemiologic studies to evaluate the health risks 
related to exposure to low-level ionizing radiation. Both groups 
provide a comprehensive evaluation of the health risks associated with 
radiation exposure.
    Finally, existing regulatory guidance suggests that protection of 
individuals (humans) is also protective of the environment. IAEA 
Technical Report Series No. 332 (Effects of Ionizing Radiation on 
Plants and Animals at Levels Implied by Current Radiation Protection 
Standards) suggests that, in most cases, the environment is being 
protected by protecting humans.
    Individuals in occupational or public areas may be exposed to 
radiation and chemical exposure which result from materials present in 
these areas. The NRC, however, has no regulatory authority over any of 
the materials present other than source, byproduct, or special nuclear 
material. In many situations, exposures to chemicals and non-NRC 
regulated materials are under the purview of the U.S. Environmental 
Protection Agency (EPA).
    Comment. Seven commenters opposed the proposed rule because of 
increased exposure, danger to public health, and increased public 
health risk.
    Response. The NRC disagrees that the proposed rulemaking will 
result in any significant increase in exposure,

[[Page 3705]]

endangerment to public health, or increase in health risk. See earlier 
comment responses for further details.
    Comment. One commenter stated that U.S. agencies have not 
adequately represented public opinion regarding transportation safety. 
The commenter was concerned that the number of irradiated fuel and 
plutonium shipments in the nation will increase as the proposed 
regulations weaken container safety standards.
    Response. The DOT and NRC represent the United States before the 
IAEA, DOT as the U.S. Competent Authority supported by the NRC. Both 
agencies are aware of public opinion regarding transportation safety in 
the United States. The NRC disagrees with the comment that U.S. 
agencies have not adequately represented public opinion. Additionally, 
NRC and DOT prepare their rules in compliance with Administrative 
Procedure Act (APA) requirements. The APA requires that public comments 
be requested, considered, and addressed before a final rule is adopted 
unless there are exigent reasons to bypass the public comment process.
    Although the number of irradiated fuel and plutonium shipments in 
the future may increase, the number of shipments to be made is 
independent of this final rule. Lastly, the comment that the regulation 
weakens transportation container safety standards is a statement of 
opinion without supporting data or information.
    Comment. One commenter suggested that NRC staff needs to address 
fully any comments submitted by the public, even when the NRC might 
consider these comments beyond the scope of the proposed rule.
    Response. Although NRC is careful to address all comments with the 
scope of the rulemaking, there are instances when a comment is 
sufficiently outside the scope of a proposed action that it need not be 
addressed. NRC resources need to be used to address issues related to 
the rulemaking for efficiency and effectiveness.
    Comment. One commenter stated that the proposed rule did not 
specifically incorporate ``issues to improve the protective adequacy of 
the regulations'' that were raised by the public during meetings held 
in 2000. The commenter stated that ``changes that were adopted in 
response to public comments in 2000 must be specified in a revised 
Proposed Rule.'' The commenter also asked that further public meetings 
be held before DOT and NRC proceed with further revisions of the 
transportation regulations.
    Response. The current rule stems from NRC's scoping efforts in 
2000, and no rule changes were adopted by the Commission at that time. 
For this proposed rulemaking, public meetings were held in Chicago, IL, 
as well as in Rockville, MD (as previously noted). NRC accepted and 
included all comments received, even those received after the July 29, 
2002, deadline. For these reasons, the NRC believes its proposed 
rulemaking meets the intent of conducting an ``enhanced public 
participation process.''
    Comment. Eleven commenters requested an extension to the comment 
period. One commenter said that the proposed rule is written in a 
manner difficult for the public and even watchdog groups to understand. 
Because the proposal would affect large portions of the general public 
by dramatically changing the standards of radioactive transport, the 
commenter urged the NRC to extend the comment period. Two commenters 
suggested that the NRC extend the comment period 180 additional days 
beyond the July 29, 2002, deadline to allow both the public and the NRC 
more time for further consideration. Commenters added that the proposed 
rule was not urgent and required further analysis and research. 
Finally, one commenter stated that the proposed rule's July 29, 2002, 
deadline for receipt of public comments would prevent it from 
accounting for the impact of Yucca Mountain. The commenter suggested 
that a 1- or 2-month rulemaking extension would be beneficial.
    Response. The NRC believes the 90-day public comment period was of 
sufficient length, especially in view of the availability of the 
proposed rule on the Secretary of the Commission's Web site for over a 
year (i.e., the Commission decided to make the proposed rule available 
to the public in March 2001, while it was under consideration). 
Therefore, the public had the opportunity to comment prior to the 
official comment period. Moreover, while not required to do so, the NRC 
chose to accept and consider comments received after the July 29, 2002, 
deadline. Further, as part of the NRC public participation process, NRC 
held two open meetings accessible to the public at which the NRC 
answered questions on the proposed rule and accepted comments. As part 
of the proposed rule, the NRC solicited additional information from the 
public which was considered in the development of the final rule.
    Comment. One commenter suggested that the NRC separate the comment 
period for the EA and RA from the comment period for the proposed rule.
    Response. The commenter's suggestion is noted but is not feasible 
to implement because the proposed rule and its supporting RA and EA 
must be considered concurrently within the rulemaking proceeding.
    Comment. One commenter asked if there is any systematic process by 
which the NRC has performed or will perform a cost-benefit analysis of 
these proposed regulations.
    Response. Whenever the NRC pursues a cost-benefit analysis 
(otherwise known as a regulatory analysis), the NRC works diligently to 
ensure that monetized, quantitative, and qualitative data are included. 
These data are studied to avoid including faulty and/or misleading 
data. The draft regulatory analysis in NUREG/CR-6713 has been revised 
to take into account the quantitative and qualitative data contained in 
the public comments on the proposed rule.
    Comment. Two commenters asked for clarification of the proposed 
rulemaking's scope in light of the May 10, 2002, letter from Commission 
Chairman Richard A. Meserve.
    Response. Former Chairman Meserve's May 10, 2002, letter to Senator 
Richard Durban provides information on questions posed by the Senator 
on transportation of spent fuel and nuclear waste to the proposed 
repository at Yucca Mountain, Nevada. The letter provides information 
on the NRC's certification process of cask designs, the safety record 
of spent fuel casks, and the NRC's authority with respect to 
transportation of radioactive materials and its relationship with DOT 
and DOE. The issues raised by this letter do not affect the amendments 
to part 71.
    Comment. One commenter asked if the NRC was aware that, on February 
23, 2002, Chicago Mayor Richard M. Daley and 17 other mayors signed a 
letter to President Bush that expressed concerns about nuclear waste 
transportation. The commenter also made reference to the fire in the 
Baltimore tunnel and wondered about safety if the fire had involved 
radioactive materials.
    Response. The NRC searched its Agency Wide Document Access and 
Management System (ADAMS), and no record was found for this letter; 
however, the NRC is aware of concerns about spent nuclear fuel 
transportation issues that have been voiced by public officials. There 
has been significant interest in the Baltimore tunnel fire that 
occurred on July 18, 2001, by State and local officials, and the impact 
that such a fire might have had on a shipment of

[[Page 3706]]

spent nuclear fuel, had such a shipment been in the tunnel during the 
time of the fire. In response to the Baltimore Tunnel fire, the staff 
has conducted two sets of independent analyses and has determined that 
the conditions that existed in the fire would not have caused a breech 
of a spent fuel transportation cask of recent design vintage had it 
been located in the tunnel for the duration of the fire.
    Comment. One commenter stated that changes in the scientific 
community's understanding of radiation injury would affect the risk 
assessments and other aspects of the proposed rule. The commenter said 
that both the DOE Biological Effects Division's and NASA's study of the 
impacts of low dose radiation impacts may require that NRC reconsider 
its current standards.
    Response. The DOE is funding a 10-year Low Dose Radiation Research 
Program to understand the biological responses of molecules, cells, 
tissues, organs, and organisms to low doses of radiation. Using 
traditional toxicological and epidemiological approaches, scientists 
have not been able to demonstrate an increase in disease incidence at 
levels of exposure close to background. Using new techniques and 
instrumentation to measure biological and genetic changes following low 
doses of radiation, it is believed that a better understanding will be 
developed concerning how radiation affects cells and molecules and 
provide a more complete scientific input for decisions about the 
adequacy of current radiation standards. These data are reviewed by 
other groups like NAS and UNSCEAR to provide an independent review of 
this health effects information. NRC reviews the programs and data 
being generated by the DOE and NASA-sponsored research as well as the 
reports published by the NAS and UNSCEAR. All of these data sources are 
used by the NRC for estimating radiological risk, establishing 
protection and safety standards, and regulating radioactive materials.
    Comment. Several commenters expressed concern and doubts about the 
data used to develop the proposed rule and the information the NRC 
provided to support its proposal. One commenter urged NRC to ensure 
that the adopted rule represents a risk-informed, performance-based 
approach. Two commenters criticized the proposed rule for not 
accounting for an expected increase in radioactive shipments. Given 
such an increase, one commenter criticized the NRC for using 20-year 
old data to justify rule changes that will reduce public safety. This 
commenter claimed that the data was out-of-date, inaccurate, not 
independently verified, and did not consider the concepts of 
radiation's synergistic effects when combined with other toxins. 
Another commenter argued that DOT and NRC should use more current data 
and future projections including the expected increases in actual 
nuclear shipments to estimate the impacts of the rule change. Realistic 
scenarios and updated data must be used to project doses and thus 
estimate the impacts of the proposed rule's changes, rather than 
relying on old data, ICRP, and reliance on computer model scenarios (or 
simply stating the lack of data). In addition, DOT and NRC should 
include the expected increases in actual nuclear shipments. Another 
commenter expressed doubt that the proposed rule's technical benefits 
are legitimate and stated that these benefits are not supported in the 
draft EA. One commenter stated that the NRC should wait to adopt any 
new regulations until there is more information available about the 
costs and benefits of such regulations.
    Response. The IAEA developed its latest standards through a 
cooperative process where experts from member nations proposed and 
supported changes to the previous version of the safety standards. The 
NRC has provided detail on the justification for the proposed changes 
in the statements of consideration for this rulemaking. The commenter 
did not provide sufficient detail on which data were of concern for NRC 
to further address.
    The comment that the NRC is relying on 20-year old data for 
justification of its regulations is unfounded. The NRC has completed 
risk studies related to the safety of transportation as recently as 
2001 and is currently engaged in a research program that will include 
the full scale testing of casks, to demonstrate the robust nature of 
certified cask designs.
    The comments about the quality of data and benefits are considered 
to be the opinion of the commenter and were not substantiated. Lastly, 
the NRC notes that a cost-benefit analysis has already been conducted 
and is reflected in the NRC's RA.
    Comment. Four commenters expressed concern that there is inadequate 
quantitative data to support the risk-based approach of the proposed 
rule and that some of the provisions are based on incorrect or outdated 
information. Two commenters were specifically concerned that DOE and 
some commercial nuclear facilities are negligent in keeping radiation 
exposure and release records. These commenters questioned how NRC data 
was gathered and noted that a failure to keep accurate records 
constrains NRC's ability to determine whether the proposed 
harmonization is economically justifiable. Furthermore, these 
commenters added that lack of records undermines the NRC claim that 
hundreds of thousands of radioactive material shipments are conducted 
safely every year.
    Response. See response to the previous comment. Also, the NRC notes 
that the commenter's statements regarding DOE and commercial 
facilities' negligence is an opinion and was not supported by factual 
evidence.
    Comment. Three commenters stated that pertinent documents and data 
were not readily available or were too difficult to access for the 
general public. One commenter requested improved public access to 
``sources of codes and IAEA documents that were cited by reference in 
the draft'' rule.
    Response. The NRC staff worked diligently to ensure that rulemaking 
documents, including all supporting documents, were available either 
electronically, over the internet, or in hard-copy upon the public's 
request in a timely fashion. This includes facilitating public access 
to the internet site of the publisher of IAEA documents in the U.S.
    Comment. Four commenters stated that the NRC should finish the PPS 
and consider its results before finalizing the proposed rulemaking as 
well as the rules governing irradiated fuel containers. Another 
commenter requested that the PPS be completed and thoroughly analyzed 
before this rulemaking is carried out because the current design 
requirements for irradiated fuel containers are inadequate and should 
be improved.
    Response. The NRC believes that shipments of spent fuel in the U.S. 
are safe using the current regulations and programs. This belief is 
based on the NRC's confidence in the shipping containers that it 
certifies, ongoing research in transportation safety, and compliance 
with safety regulations and the conditions of certificates that have 
resulted in an outstanding transport safety record. Thus, an 
established system of regulatory controls protects every U.S. shipment 
of spent fuel from commercial reactors. The NRC sponsored PPS is part 
of an ongoing confirmatory research program to reassess risks as 
shipment technologies change and analytical capabilities improve.
    Comment. Three commenters urged the NRC to require more stringent 
testing of transport packages in real-world (not computer-modeled) 
testing.

[[Page 3707]]

    Response. NRC regulations permit certifications through testing, 
analyses, comparison to similar approved designs, or combinations of 
these methods. A full scale testing is not necessary for the NRC to 
achieve confidence that a design satisfies the regulatory tests, as 
long as the analyses are based on sound and proven analytic techniques.
    Comment. One commenter suggested that the NRC ensure that the 
economic value of these regulations is not skewed. That is, the 
commenter does not want the needs of one particular industry to shape 
the regulations, when the regulations could have a greater impact on a 
different industry.
    Response. The overall value or impact of the proposed changes 
results from the interaction of several influencing factors. It is the 
net effect of the influencing factors that governs whether an overall 
value or impact would result for several different attributes (i.e., 
different industries or the public). Similarly, a single regulatory 
option could affect licensee costs in multiple ways. A value-impact 
analysis, such as was undertaken as part of this rulemaking effort, 
quantifies these net effects and calculates the overall values and 
impacts of each regulatory option. A decision on which regulatory 
option is recommended takes into account the overall values and impacts 
of the rulemaking.
    Comment. One commenter stressed that when the NRC has decision 
makers review public comments, the NRC staff should look at primary 
documents instead of summary documents. The commenter cited NUREG/CR-
6711 as an example where the regulator runs the risk of having decision 
makers read summaries of public comments without understanding the 
underlying context and content.
    Response. In our decisionmaking process, the NRC did not rely on a 
summary document to support the development of the proposed rule. NRC 
used primary documents to fully understand the underlying context and 
content of the technical information. The summary documents the 
commenter refers to were developed to provide the public with a 
comprehensive, yet condensed, version of the underlying information. 
Further, these underlying documents were also made available to the 
public on the NRC Web site during the rulemaking process.
    Comment. One commenter asked which countries have already adopted 
the proposed guidelines.
    Response. The IAEA has conducted a survey that provides the status 
(as of July 1, 2003) of each Member State's plans for implementing TS-
R-1. Based on that survey, many States have already implemented the new 
requirements of TS-R-1 (e.g., European Commission, Germany, and 
Australia). Other States have indicated that they are actively 
implementing these requirements and intend to finalize implementation 
by the end of 2003. No State indicated that it would not adopt these 
standards. This survey is available at http://www-rasanet.iaea.org/
downloads/radiation-safety/MSResponsesJuly1 2003.pdf
    Comment. One commenter requested clarification on NRC assumptions 
for future radioactive materials transportation. Specifically, the 
commenter wanted to know whether NRC is assuming the amounts will 
increase or remain consistent with past levels.
    Response. The NRC's draft RA and EA relied on existing information 
to determine the future impacts of the proposed changes. NRC solicited 
information on the costs and benefits for each of the proposed changes 
as part of the proposed rule. The NRC considered available information 
on future radioactive material shipments in its decisionmaking process. 
Information that was received as part of the public comment process was 
considered in developing NRC's final position. The NRC staff conducted 
some sensitivity studies, see for example Comparison of A1 
and A2 new and old values in the EA, Table A-1, Appendix A.
    Comment. Three commenters opposed weakening regulations that would 
reduce the public safety and health through new definitions or accepted 
concentration values. One commenter worried that the proposed rule 
would weaken regulatory control, allowing increased quantities of 
radioactive materials and wastes ``into the lives of individual 
citizens without their knowledge or approval,'' thus violating ``the 
most fundamental premises of radiation protection.''
    Response. The NRC acknowledges the concerns but believes that the 
rule continues to protect the public's health and safety in a risk-
informed manner.
    Comment. One commenter particularly opposed NRC and DOE studies, 
including the EIS to review alternative policies for disposal and 
recycling of radioactive metals. The commenter requested that the NRC 
maintain stringent controls on all materials being recycled, disposed, 
or otherwise reused. Two commenters expressed opposition to the 
proposed rule due to a belief that the proposed rule would deregulate 
radioactive wastes and materials and allow the deliberate dispersal of 
radioactive materials into raw materials and products that are used by 
the public and are available on the market.
    Response. The NRC acknowledges the commenters' references to DOE 
and NRC studies related to the disposal and recycling of radioactive 
metals. This rule is not related to the referenced studies.
    Comment. One commenter expressed concern that NRC's proposed 
regulations could increase the variety of materials that are regulated 
as ``radioactive'' for transportation purposes.
    Response. The rule does not expand the scope of regulated 
radioactive material.
    Comment. One commenter expressed concern that the proposed rule 
enables commercial and military nuclear industries to ``revive and 
expand, thereby generating ever more wastes to be stored, transported 
and ultimately * * * sequestered from the biosystem.''
    Response. The comment is beyond the scope of this rulemaking.

Proposed Yucca Mountain Facility

    Comment. One commenter expressed opposition to sending shipments of 
nuclear materials to the proposed Yucca Mountain facility.
    Response. Potential shipments to the proposed geologic repository 
at Yucca Mountain are beyond the scope of this rulemaking.
    Comment. Two commenters raised issues related to the possible 
approval of the Yucca Mountain site. One commenter expressed concern 
about the safety of dry casks. The commenter asked if the NRC was aware 
of the accident at the Point Beach Nuclear Plant in Wisconsin on May 
28, 1996, and how similar the dry casks that will ship radionuclides to 
Yucca Mountain will be to the casks used at Point Beach. The commenter 
noted that once one buries a dry cask, one cannot change it; therefore, 
the U.S. will have to be sure that it uses safe casks. The second 
commenter urged the NRC to consider the transportation issues 
associated with the possible approval of the Yucca Mountain site as the 
NRC makes rules pertaining to the packaging and transportation of 
radioactive materials.
    Response. The Nuclear Waste Policy Act (NWPA) requires DOE to use 
casks certified by NRC for transport to Yucca Mountain, if licensed. 
Transport casks are generally not the same as storage or disposal 
casks. Issues regarding the licensing of the Yucca Mountain site and 
the safety of spent fuel storage or disposal casks are beyond the scope 
of the proposed rulemaking. The NRC believes compliance with the

[[Page 3708]]

regulations in part 71 provides for safe transport package designs.
    Comment. Three commenters expressed belief that increases in future 
shipments have not been adequately considered in the rulemaking. The 
first commenter stated that these regulations could have important 
implications for the shipment of high-level radioactive waste. The 
commenter asked if NRC had considered the financial impact of the 
opening of the Yucca Mountain facility before proposing the 
regulations.
    Response. This comment is primarily focused on future shipments to 
Yucca Mountain. The Commission has not received any application 
relative to the Yucca Mountain site, and a final decision has not been 
made on opening the site itself. Any conclusion made now by the NRC on 
future shipments would be purely speculative. Moreover, the commenter 
did not specify which aspect of the proposed rule would have a 
significant bearing on the Yucca Mountain facility.
    The NRC did not identify where major impacts would result, none 
were identified that would impact spent fuel shipments. Furthermore, 
the existing regulations pertaining to spent fuel have been in effect 
for a significant time and have resulted in more than 1300 spent fuel 
shipments being conducted without any negative impacts to public health 
and safety.
    Comment. Two commenters asked how NRC factored the possible 
approval of the Yucca Mountain repository into our rulemaking. One 
commenter urged NRC to seriously consider the likely increase of 
radioactive material transportation in Illinois, Michigan, and 
Wisconsin that will occur if the Yucca Mountain repository is approved. 
The commenter also provided data from DOE's Yucca Mountain EIS on 
projected transportation volume through Illinois.
    Response. The comments are acknowledged. However, they are beyond 
the scope of this rulemaking. As part of the rulemaking process, NRC 
solicited information on the costs and benefits, as well as other 
pertinent data, on the proposed changes. NRC appreciates the 
commenter's submission of data related to projected transportation 
volumes of high-level waste. The NRC believes compliance with the 
regulations in part 71 provides for safe transport package designs.

Miscellaneous (including comments to DOT)

    Comment. One commenter opposed any use of radioactive materials 
entirely.
    Response. This comment is beyond the scope of the rulemaking. This 
rule deals solely with regulations that govern the transportation of 
certain types of radioactive materials and does not address issues 
related to the use of radioactive materials in commerce.
    Comment. One commenter included a comment letter that was 
previously submitted in September 2000, discussing all of the issues in 
this rulemaking. The letter was resubmitted because the commenter 
believes that the NRC did not respond to the comments previously and 
might have lost the original comment letter. The commenter also 
included several diagrams and an article entitled ``New Developments in 
Accident Resistant Shipping Containers for Radioactive Materials'' by 
J. A. Sisler. This article discusses the safety tests required for 
shipping containers.
    Response. The current proposal stems from NRC's scoping meetings 
held in August and September 2000, to solicit public comments on the 
part 71 Issues Paper. NRC accepted all verbal and written comments 
received at the meetings or later in a letter form and considered these 
comments in developing the proposed rule.
    Comment. One commenter stated that the public's opinion is that 
nuclear power and weapons should remain sequestered from the 
environment and the public for as long as they remain hazardous.
    Response. The comment is beyond the scope of the rulemaking. This 
rule deals solely with regulations that govern the transportation of 
certain types of radioactive materials and does not address the use of 
nuclear power or weapons.
    Comment. One commenter expressed a general distrust of business and 
urged NRC to consider recent cases of dishonesty in business when 
formulating regulations.
    Response. The comment is beyond the scope of this rulemaking.
    Comment. One commenter expressed concern that inaccurate reporting, 
inspection failures, and faulty equipment all occur in the nuclear 
transport industry and may contribute to mishaps in transit.
    Response. The NRC is aware of the potential for accidents in 
transporting nuclear material and has considered the accident history 
of nuclear transportation in estimating the risks of shipping. The NRC 
believes that this rule provides adequate protection of the public and 
workers in normal transport conditions and in accident conditions.
    Comment. One commenter recommended that all radioactive shipments 
be tracked, labeled, and publicly reported, including shipments being 
made in secret without the consent of the American public.
    Response. The NRC acknowledges the commenter's suggestion about 
tracking, labeling, and reporting shipments. Current regulations 
include requirements for labels and markings for packages that contain 
radioactive materials. There are notification requirements for NRC 
licensees applicable to shipments of spent nuclear fuel. Current NRC/
DOT requirements for tracking and labeling radioactive shipments 
provide adequate protection of public health and safety.
    Comment. Several commenters were concerned about the public 
reporting requirements pertaining to the shipping of radioactive 
materials. Two commenters believe that NRC should publicly report all 
radioactive shipments.
    Response. The NRC has regulations in 10 CFR part 73 (Physical 
Protection of Plants and Materials) that deal with the reporting of 
shipments of spent fuel nuclear fuel. This rule deals only with part 
71; therefore, these comments are beyond the scope of this rulemaking.
    Comment. Several commenters expressed concern with the tracking and 
labeling aspects of the proposed rule. Two commenters urged the NRC to 
track, label, and publicly report all radioactive shipments. One 
commenter believes that the words ``radioactive materials'' should not 
be removed from shipping placards because personnel and volunteers 
understand the plain English warning better than technical language. 
This commenter also suggested that the warnings be written in several 
languages. In addition, one commenter stated that the standard symbol, 
the black and yellow ``windmill'' for radiation, should adorn all 
containers.
    Response. Tracking and labeling shipments are part of the 
responsibility of the shipper of the licensed material in accordance 
with NRC and DOT regulations. Reporting all radioactive shipments would 
be an administrative burden with minimal benefit. The NRC's regulations 
do require a shipper to provide advance notification of a shipment of 
spent nuclear fuel to both the NRC and to the Governor or designee of a 
State through which the shipment would be passing. The information is 
considered safeguards information and cannot be released to the public 
until after a shipment has been completed.
    Comment. One commenter expressed support for NRC's acknowledging 
DOT's responsibility to ensure the safe shipment of spent nuclear fuel.
    Response. The comment is acknowledged. No further response is 
required.

[[Page 3709]]

    Comment. One commenter requested a clarification of the current 
status of DOT's regulations for international shipments regarding 
exempt quantities and concentrations.
    Response. This request has been forwarded to DOT for consideration. 
The commenter should refer to DOT's proposed rule found at 67 FR 21328 
dated April 30, 2002.
    Comment. One commenter expressed concern with how the proposed 
regulations fit into the hierarchy of Federal, State, and local 
regulations. The commenter noted that DOT regulations expressly preempt 
and supersede State and local regulations.
    Response. The State regulations augment the overall national 
program for the protection of public health and safety of citizens from 
any hazards incident to the transportation of radioactive materials. 
States usually adopt the Federal transportation regulations by 
reference. The combined efforts of DOT, NRC, and the Agreement States 
assure that the applicable Federal regulations are observed with 
respect to packaging and transportation of radioactive materials on a 
nationwide basis. This is accomplished through DOT, NRC, and State and 
local government inspection and enforcement efforts.
    Comment. One commenter expressed concern that the DOT definition of 
``radioactive material'' is now defined as ``any material having a 
specific activity greater than 70 Bq per gram (0.002 micro curie per 
gram).'' According to the commenter, the effect of this new definition 
would be to enable much more radioactivity to be exempt, thus allowing 
more radioactive material to move unregulated in commerce.
    Response. This referenced definition change also exists in the NRC 
final rule. As described in the background section of this rule, NRC 
has analyzed the impact on dose to the public from changing the 
definition of ``radioactive material'' from the current definition 70 
Bq/g (0.002 [mu]Ci/g) for all radionuclides to radionuclide-specific 
exemption values. After considering transport scenarios, NRC concluded 
that the new radionuclide-specific definition would result in an 
overall reduction in dose to the public when compared to the current 
definition.
    Comment. One commenter noted that, in Table 1, the listings for Th 
(nat) and U (nat) (68 FR 21482) do not refer to footnote b. Because 
this is inconsistent with the text of the preamble, the commenter 
concluded that it is a typographical error that should be corrected.
    Response. The comment is acknowledged and was considered in 
developing the final rule.
    Comment. One commenter urged the NRC to consider ``the 
relationships between and among the exposures associated with these 
packaging, container, and transportation regulations and all other 
sources of radiation exposures,'' to protect the public from ``adverse 
impacts on their health and genetic integrity.''
    Response. The comment is acknowledged and has been considered in 
developing the final rule.
    Comment. Three commenters expressed concern with the role of State 
and local governments. One commenter believes that certain States are 
already burdened with unusually high concentrations of hazardous and 
radioactive materials transport. Another commenter asked about ``the 
status of non-Agreement States with respect to compatibility'' and also 
wanted further ``explanation of the extent to which a State or 
Agreement State may deviate from NRC program elements, definitions, and 
standards.'' One commenter stated that county sheriffs and the proper 
State officials should be notified in advance of spent nuclear fuel 
shipments scheduled to pass through their jurisdictions.
    Response. It is NRC practice to seek input and comments from State 
and local governments on any NRC proposed rules. For example, in 
December 2000, the NRC staff forwarded the part 71 proposed rule to the 
Agreement States for comment before sending the rule to the Commission. 
Once the rule is published for public comments, NRC considers comments 
from all State and local governments, and as such, they play an 
important role in the NRC regulatory process. State officials 
designated by the Governor are notified in advance of spent nuclear 
fuel shipments made by NRC licensees, which pass through their 
respective States.
    Comment. Several commenters criticized the proposed rule for 
acquiescing to the desires of the nuclear and radiopharmaceutical 
industries to weaken transport regulations at the expense of increased 
public risk.
    Response. The proposed rule was developed to maintain compatibility 
with the IAEA transportation standards as well as to issue other NRC-
initiated changes. Part 71 has been revised twice in the past 20 years 
to stay compatible with IAEA regulations. The risk to the public from 
transportation of radioactive materials were considered in the 
development of the NRC regulations.
    Comment. Two commenters expressed concern over implications for 
worker safety. These commenters asked if workers would be protected 
from and informed of leaks and whether there is sufficient money to pay 
lawsuit damages. They stated that exposure to the transport vehicle 
itself should not exceed 10 millirems/year, and all crew compartments 
should be heavily shielded to reduce exposure. One commenter then 
asserted that workers should be trained to handle radioactive materials 
and informed of the risks involved.
    Response. NRC radioactive material transportation regulations have 
always been issued and enforced to protect the worker and the public 
health and safety. When shippers of radioactive material follow these 
regulations, they are taking the protective measures called for in NRC 
(and DOT) regulations to protect the crew and public. The NRC and DOT 
regulations require worker training.
    Comment. Several commenters believe that the proposed regulations 
increased public risk and weakened protection of public health. One 
commenter stated that additional independent oversight of the transport 
casks should be conducted regarding quality control to determine 
whether they are adequate for cross-country transport. This commenter 
also believes that the testing criteria for containers should be more 
demanding and require real-world conditions. Another commenter stated 
that nuclear shipments should be transported at off-peak hours and also 
supported the creation of a ``vehicle-free'' buffer zone ahead and 
behind the shipment.
    Response. The commenters did not specify how the proposed 
rulemaking would increase public risk and weaken protection of public 
health. When NRC developed the proposed rule, potential impacts were 
carefully considered. NRC does not believe that any part of the 
proposal will result in a significant impact on public health and 
safety. NRC's quality assurance programs and inspections determine when 
additional oversight is warranted. The request for additional and more 
demanding testing is not specific; it does not specify how and why 
particular testing procedures are inadequate. These procedures have 
been carefully verified by NRC to ensure adequate safety.
    NRC does not support the commenter's suggestion to transport at 
``off-peak'' hours and use a buffer zone as an NRC safety requirement. 
There is no safety basis to justify restricting travel only to off-peak 
hours, and creating (and enforcing) buffer zones could result in 
greater traffic impacts

[[Page 3710]]

and safety issues. Moreover, using these restrictions is not warranted 
based on the more than 1300 shipments without incident.
    Comment. One commenter urged the NRC to prohibit transport of long-
lived spent nuclear fuel via air or via barge across large waterways. 
The commenter also urged NRC to disallow the transport of such fuel in 
combination with people, animals, or plants.
    Response. Existing NRC and DOT regulations establish requirements 
that must be met for safe shipment of spent nuclear fuel by 
transportation modes (i.e., truck, barge, or air). The commenter's 
second recommendation is noted, but it is beyond the scope of the 
proposed rule.
    Comment. One commenter stated that dumping radioactive material 
into oceans or landfills and incineration of such materials should 
never be allowed.
    Response. The comment is acknowledged. However, it is beyond the 
scope of this rulemaking, and therefore no further response is 
required.
    Comment. One commenter suggested that NRC, in concert with other 
agencies, identify and recover formerly regulated nuclear materials 
that have been deregulated or have escaped from control in the past.
    Response. This comment is beyond the scope of this rule.
    Comment. One commenter requested an explanation of how NRC's 
official proposal on the changes in packaging and transporting of 
radioactive materials would affect industrial radiology.
    Response. Generally, industrial radiography cameras are designed to 
meet NRC requirements for Type B transportation packages. Of the 11 
IAEA adoption issues and the 8 NRC-initiated issues, none have a 
significant impact upon the transport package design requirements for 
radiography cameras.
    Comment. One commenter expressed support for compatibility among 
the Agreement States. This commenter indicated that it is appropriate 
for States to have the ability to develop materials necessary for 
intrastate shipments. However, for interstate shipments, the commenter 
stated that it is necessary for one State to be compatible with the 
rest of the country for the country to be compatible with the world.
    Response. NRC notes that the commenter's views are consistent with 
the Commission's Policy Statement on the Adequacy and Compatibility of 
Agreement State Programs, which became effective on September 3, 1997 
(62 FR 46517).
    Comment. Several commenters urged NRC to improve its scientific 
understanding and bases for the proposed rulemaking. Two commenters 
suggested that NRC complete the comprehensive assessments of TS-R-1 and 
future IAEA standards, the PPS, and real cask tests before proceeding 
with this rulemaking.
    Response. NRC believes it has an adequate technical basis to make 
determinations on the adoption of regulatory changes to address the 
issues that are the subject of this rulemaking. The ongoing PPS is 
beyond the scope of this rulemaking.

III. Discussion

    This section is structured to present and discuss each issue 
separately (with cross references as appropriate). Each issue has four 
parts: Summary of NRC Final Rule, Affected Sections, Background, and 
Analysis of Public Comments on the Proposed Rule.

A. TS-R-1 Compatibility Issues

Issue 1. Changing Part 71 to the International System of Units (SI) 
Only
    Summary of NRC Final Rule. The NRC has decided to continue using 
the dual-unit system (SI units and customary units) in part 71. This 
will not conflict with TS-R-1, which uses SI units only, because TS-R-1 
does not specifically prohibit the use of a dual-unit system.
    We have decided not to change part 71 to use SI units only nor to 
require NRC licensees and holders and applicants for a Certificate-of-
Compliance (CoC) to use SI units only because doing so will conflict 
with NRC's Metrication Policy (61 FR 31169; June 19, 1996) which allows 
a dual-use system. The NRC did not make metrication mandatory because 
no corresponding improvement in public health and safety would result; 
rather, costs would be incurred without benefit. Moreover, as noted in 
the proposed rule (67 FR 21395-21396), the change to SI units only 
could result in the potential for adverse impact on the health and 
safety of workers and the general public as a result of unintended 
exposure in the event of shipping accidents, or medical dose errors, 
caused by confusion or erroneous conversion between the currently 
prevailing customary units and the new SI units by emergency responders 
or medical personnel.
    Affected Sections. None (not adopted).
    Background. TS-R-1 uses the SI units exclusively. This change is 
stated in TS-R-1, Annex II, page 199: ``This edition of the Regulations 
for the Safe Transport of Radioactive Material uses the International 
System of Units (SI).'' The change to SI units exclusively is evident 
throughout TS-R-1. TS-R-1 also requires that activity values entered on 
shipping papers and displayed on package labels be expressed in SI 
units (paragraphs 543 and 549). Safety Series No. 6 (TS-R-1's 
predecessor) used SI units as the primary controlling units, with 
subsidiary units in parentheses (Safety Series 6, Appendix II, page 
97), and either unit was permissible on labels and shipping papers 
(paragraphs 442 and 447).
    The NRC Metrication Policy allows a dual-unit system to be used (SI 
units with customary units in parentheses). The NRC Metrication Policy 
was designed to allow market forces to determine the extent and timing 
for the use of the metric system of measurements. The NRC is committed 
to work with licensees and applicants and with national, international, 
professional, and industry standards-setting bodies (e.g., American 
National Standards Institute (ANSI), American Society for Testing and 
Materials (ASTM), and American Society of Mechanical Engineers (ASME)) 
to ensure metric-compatible regulations and regulatory guidance. The 
NRC encouraged its licensees and applicants, through its Metrication 
Policy, to employ the metric system wherever and whenever its use is 
not potentially detrimental to public health and safety, or its use is 
economic. The NRC did not make metrication mandatory by rulemaking 
because no corresponding improvement in public health and safety would 
result, but rather, costs would be incurred without benefit. As a 
result, licensees and applicants use both metric and customary units of 
measurement.
    According to the NRC's Metrication Policy, the following documents 
should be published in dual units: new regulations, major amendments to 
existing regulations, regulatory guides, NUREG-series documents, policy 
statements, information notices, generic letters, bulletins, and all 
written communications directed to the public. Documents specific to a 
licensee, such as inspection reports and docketed material dealing with 
a particular licensee, will be issued in the system of units employed 
by the licensee.
    Currently, part 71 uses the dual-unit system in accordance with the 
NRC Metrication Policy.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:

[[Page 3711]]

    Comment. Eight commenters stated they appreciated the NRC's 
decision to maintain both the international and the familiar system of 
becquerels and curies and sieverts and rem.
    Response. No response is necessary.
Issue 2. Radionuclide Exemption Values
    Summary of NRC Final Rule. The final rule adopts, in Sec.Sec. 
71.14, 71.88 and Appendix A, Table A-2, the radionuclide activity 
concentration values and consignment activity limits in TS-R-1 for the 
exemption from regulatory requirements for the shipment or carriage of 
certain radioactive low-level materials. In addition, the final rule 
provides an exemption from regulatory requirements for natural material 
and ores containing naturally occurring radionuclides that are not 
intended to be processed for use of these radionuclides, provided the 
activity concentration of the material does not exceed 10 times the 
applicable values. These amendments conform part 71 with TS-R-1 and 
with DOT's parallel IAEA compatibility rulemaking for CFR 49.
    During the development of TS-R-1, it was recognized that there was 
no technical justification for the use of a single activity-based 
exemption value for all radionuclides for defining a material as 
radioactive for transportation purposes (a uniform activity 
concentration basis) and that a more rigorous technical approach would 
be to base radionuclide exemptions on a uniform dose basis. The values 
and limits in TS-R-1, and adopted in Appendix A, Table A-2, establish a 
consistent dose-based model for minimizing public exposure. Overall, 
NRC's analysis shows that the new system would result in lower actual 
doses to the public than the uniform activity concentration basis 
system. NRC's regulatory analysis indicated that adopting the 
radionuclide-specific exemption values contained in TS-R-1 is 
appropriate from a safety, regulatory, and cost perspective. Moreover, 
the final rule assures continued consistency between domestic and 
international regulations for the basic definition of radioactive 
material in transport.
    Affected Sections. Sections 71.14, 71.88, and Appendix A.
    Background. The DOT previously used an activity concentration 
threshold of 70 Bq/g (0.002 [mgr]Ci/g) for defining a material as 
radioactive for transportation purposes. DOT regulations applied to all 
materials with activity concentrations that exceeded this value. 
Materials were exempt from DOT's transportation regulations if the 
activity concentration was equal to or below this value. The 70-Bq/g 
(0.002-[mgr]Ci/g) activity concentration value was applied collectively 
for all radionuclides present in a material.
    In Sec. 71.10, the NRC used the same activity concentration 
threshold as a means of determining if a radioactive material was 
subject to the requirements of part 71. Materials were exempt from the 
transportation requirements in part 71 if the activity concentration 
was equal to or below this value. Although the materials may be exempt 
from any additional transportation requirements under part 71, it is 
important to note that the requirements for controlling the possession, 
use, and transfer of materials under parts 30, 40, and 70 continue to 
apply, as appropriate, to the type, form, and quantity of material. 
Basically, the radionuclide exemption values mean that licensed low 
radioactivity materials are not required to be handled as hazardous 
materials while they are being transported. These exemption values do 
not mean that these materials are released from other regulatory 
controls, including the controls that apply to the disposal or release 
of radioactive material.
    During the development of TS-R-1, it was recognized that there was 
no technical justification for the use of a single activity-based 
exemption 70-Bq/g (0.002-[mgr]Ci/g) value for all radionuclides. It was 
concluded that a more rigorous technical approach would be to base 
radionuclide exemptions on a uniform dose basis, rather than a uniform 
activity concentration basis.
    By 1994, the IAEA had developed Safety Series No. 115 (also known 
as Basic Safety Standard, or BSS) and a set of principles for 
determining when exemption from regulation was appropriate. One 
exemption criterion was the effective dose expected to be incurred by a 
member of the public from a practice (e.g., medical use of 
radiopharmaceuticals in nuclear medicine applications) or a source 
within a practice should be unlikely to exceed a value of 10 [mgr]Sv (1 
mrem) per year. IAEA researchers developed a set of exposure scenarios 
and pathways which could result in exposure to workers and members of 
the public. These scenarios and pathways were used to calculate 
radionuclide exemption activity concentrations and exemption activities 
which would not exceed the recommended dose.
    To investigate the exemption issue from a transportation 
perspective during the development of TS-R-1, IAEA Member State 
researchers calculated the activity concentration and activity for each 
radionuclide that would result in a dose of 10 [mgr]Sv (1 mrem) per 
year to transport workers under various BSS and transportation-specific 
scenarios. Due to differences in radionuclide radiation emissions, 
exposure pathways, etc., the resulting radionuclide-specific activity 
concentrations varied widely. The appropriate activity concentrations 
for some radionuclides were determined to be less than 70 Bq/g (0.002 
[mgr]Ci/g), while the activity concentrations for others were much 
greater. However, the calculated dose to transport workers that would 
result from repetitive transport of each radionuclide at its exempt 
activity concentration was the same ((10 [mgr]Sv) (1 mrem)) per year. 
For the single activity-based value, the opposite was true (i.e., the 
exempt activity concentration was the same for all radionuclides (70 
Bq/g) (0.002 [mgr]Ci/g)), but the resulting doses under the same 
transportation scenarios varied widely, with annual doses ranging from 
much less than 10 [mgr]Sv (1 mrem) per year for some radionuclides to 
greater than 10 [mgr]Sv (1 mrem) per year for others. A comparison of 
the transportation scenario doses resulting from the single (70 Bq/g 
(0.002 [mgr]Ci/g)) activity concentration value and the radionuclide-
specific activity concentration values shows that the radionuclide 
activity concentration values reduced the variability in doses that 
were likely to result from exempt transport activities.
    The basis for the exemption values indicates that materials with 
very low hazards can be safely exempted from the transportation 
regulations (see draft Advisory Material for the Regulations for the 
Safe Transport of Radioactive Material, TS-G-1.1, paragraphs 107.5 and 
401.3). If the exemptions did not exist, enormous amounts of material 
with only slight radiological risks (materials which are not ordinarily 
considered to be radioactive) would be unnecessarily regulated during 
transport.
    Some of the lower activity concentration values might include 
naturally occurring radioactive material (NORM). As an example, ores 
may contain NORM. Regarding the transport of NORM, one petroleum 
industry representative stated that there are no findings that indicate 
the current standard fails to protect the public, and that there is no 
benefit in making the threshold more stringent. Further, it would have 
a significant impact on their operations. Other similar comments were 
received during the public meetings. The overall impact would be that 
some material formerly not subject to the radioactive material 
transport regulations may need to be transported as radioactive 
material and therefore

[[Page 3712]]

meet the corresponding applicable DOT transport requirements.
    IAEA recognized that application of the activity concentration 
exemption values to natural materials and ores might result in 
unnecessary regulation of these shipments and established a further 
exemption for certain types of these materials. Paragraph 107(e) of TS-
R-1 further exempts: ``Natural material and ores containing naturally 
occurring radionuclides which are not intended to be processed for use 
of these radionuclides provided the activity concentration of the 
material does not exceed 10 times the values specified in paragraphs 
401-406.''
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter opposed the reuse of radioactive materials 
in other products, arguing that this is not based on sound science, but 
on commercial judgment. Several commenters expressed general objections 
to the proposal to exempt certain amounts of radionuclides from 
transportation regulatory control and urged NRC to help prevent more 
radioactive waste from being deregulated. Seven commenters stated that 
adopting these exemptions would remove a significant barrier to the 
purposeful release of radioactive materials from nuclear power and 
weapons production into raw materials that can be used to make daily 
items (e.g., hip replacements, braces, and toothbrushes) that come into 
contact with members of the public.
    Another commenter stated that the exempted levels could potentially 
provide a back door to recycle and release of radioactive material.
    One commenter said that the NRC's stated objectives to facilitate 
nuclear transportation and harmonize international standards should not 
supersede the NRC's mandate to protect public health and safety. The 
commenter also stated that the proposed regulations do not do enough to 
protect public health. The commenter opposed the technically 
significant motive for adopting exemption values, which is to 
facilitate radioactive ``release'' and ``recycling'' or dispersal of 
nuclear waste into daily commerce and household items.
    One commenter stated that NRC regulations should not treat 
radioactive materials like nonradioactive materials. Two other 
commenters criticized the proposed regulations for treating radioactive 
substances as if they were not radioactively contaminated.
    Response. The transportation exemption values do not establish 
thresholds for the release of radioactive material to unlicensed 
parties or to the environment. They do not relieve the recipient from 
regulations that apply to the use or release of that material. Also, 
the transportation regulations do not authorize the possession of 
licensed material (Sec. 71.0(c)). Thus, no unauthorized party may 
receive or possess radioactive material just because the material is 
exempted from transportation requirements. Radioactive material 
transported under the rule remains subject to separate regulatory 
safety requirements regarding possession, use, transfer, and disposal.
    Comment. One commenter stated that the use of ``or'' in proposed 
Sec. 71.14(a)(2) (67 FR 21448) suggests that there is no consignment 
limit if the exempt activity concentration limits are not exceeded. NRC 
was asked to replace ``or'' by ``and'' to prevent deliberate dilution 
of radioactive material to obtain exemption from transport regulations.
    Response. The comment is correct in that the consignment activity 
limit does not apply to materials that do not exceed the exempt 
activity concentration. Under the final rule, the transport regulations 
apply only to radioactive material for which both the activity 
concentration for an exempt material and the activity limit for an 
exempt consignment are exceeded, so the use of ``or'' in the regulatory 
text is correct. When describing materials that are subject to the 
regulations, ``and'' is the correct term; when describing materials 
that are not subject to the regulations, ``or'' is the correct term. 
Because Sec. 71.14 defines materials that are not subject to the 
regulations, ``or'' is the correct term.
    Material consignments that exceed the exempt activity 
concentration, but not the exempt consignment limit, are not regulated 
in transport due to the small quantity of material being transported. 
Material consignments that exceed the exempt consignment limit, but not 
the exempt activity concentration, are not regulated in transport due 
to the low radioactivity concentration of the material being 
transported. The NRC has no information to support the notion that 
radioactive material is diluted to obtain exemption from transport 
regulations. The NRC does not propose any regulatory action in this 
regard.
    Comment. One commenter expressed concern both that the proposed 
rule would exempt radionuclide values at various levels and that an 
international body created these exemption levels.
    Response. The activity concentration exemption values do vary by 
radionuclide. However, the doses to the public estimated to occur from 
using these values under the transport scenarios are low. The U.S. 
participated in assessing the dose impacts from the use of the 
exemption values in transport.
    Comment. Another commenter asked if it is really necessary for NRC 
to adopt the entire IAEA rule to accomplish its goals.
    Response. There are a number of specific goals associated with this 
rulemaking, one of which is harmonization of NRC regulations with 
IAEA's TS-R-1 and DOT regulations. NRC is not adopting TS-R-1 in its 
entirety in this rulemaking. However, with respect to revising 
exemption values, the NRC staff believes adoption of the exemption 
values from TS-R-1 is warranted to maintain consistency between 
domestic and international regulations.
    Comment. One commenter asked if the NRC told DOT that the American 
public has rejected these proposed standards three times in the past 
decade, and if DOT has advised IAEA of these objections. The commenter 
said that if the IAEA has not been informed of the American public's 
resistance to these regulations, NRC needs to inform the agency (DOT 
and IAEA) immediately.
    Response. The NRC acknowledges this comment, including both the 
NRC's and DOT's earlier opposition to the IAEA proposed exemption 
values. This rule is the first time that IAEA exemption values are 
adopted and are being carried out for maintaining compatibility with 
international transportation regulations.
    Comment. One commenter asked about the amount of money being spent 
regulating levels below the exemption values. The commenter asked if 
more money would be spent attempting to verify the proposed exemption 
values than would be saved by deregulating them. The commenter wanted 
to know if there is any guarantee that money saved by deregulating 
levels below the exemption values will be spent on improving public 
safety in other areas.
    Response. The NRC believes the benefits of the exemption values 
will outweigh the costs. NRC analyses lead the NRC staff to believe 
that the increase in regulatory efficiency between regulatory agencies 
and the facilitation of international shipments make the exemption 
values advantageous overall. Further, as part of this rulemaking, NRC 
specifically requested information on the costs and benefits of the 
proposed

[[Page 3713]]

changes. To the extent this information was received, it was considered 
in the development of NRC's position. Lastly, it is beyond the scope of 
this rulemaking to guarantee that any money saved will be spent on 
improving public safety elsewhere.
    Comment. One commenter suggested that the NRC could not determine 
costs or savings from the proposed radionuclide exemption values, in 
part because the NRC does not know what amounts will be exempted. The 
commenter also explained that although NRC could attempt to do 
projections based on the current industry, NRC could not know what 
amounts would be exempted in the future.
    Response. The NRC fully realizes the difficulties associated with 
predicting the impacts of implementing the exemption values. The NRC 
also agrees that it is difficult to predict what amounts would be 
exempted under this final rule, just as it is difficult to assess the 
amount of material exempted under the current regulations. However, a 
large majority of commercial radioactive materials are shipped in 
highly purified forms that far exceed the exemption levels. NRC expects 
this would continue to be the case under the exemption values. For all 
of these reasons, the NRC staff explicitly asked for data on the 
anticipated impacts of the proposed rule. The NRC staff used these data 
to aid decisionmaking. In general, the NRC expects that the increase in 
regulatory efficiency among regulatory agencies and the facilitation of 
international shipments will outweigh any increased costs of shipments 
resulting from the changes in the exemption values.
    Comment. One commenter requested that a cost-benefit analysis be 
done to account for both the proposed rule's complexity and its 
enforcement difficulties. The commenter notes that no cost-benefit 
analysis had been done on this issue and that the NRC chose it 
subjectively.
    Response. The draft regulatory analysis considered the benefits and 
costs associated with adoption of the radionuclide exemption values 
from TS-R-1 using the best available information. In addition, the NRC 
decided to adopt the dose-based exemption values because the NRC 
believes these values would actually reduce exposure in transport by 
establishing a consistent dose-based model for minimizing public 
exposure. This benefit is in addition to the expected harmonization and 
financial benefits. NRC disagrees with the commenter's assertion that 
the exemption values were chosen subjectively. NRC used the best 
available information and gathered as much information as possible from 
the public, the regulated community, and outside experts. The purpose 
of this rulemaking, with its public meetings and public comment period, 
is to ensure that all affected parties have adequate opportunity to 
register their comments and provide supporting materials to justify 
their position (and thus better influence the development of NRC's 
final position).
    Comment. Another commenter stated that the technical benefits of 
the proposed rule do not outweigh the associated costs and efforts.
    Response. Because NRC staff are unclear what the commenter means by 
``technical benefits,'' NRC cannot specifically respond to this 
comment. Overall, NRC believes that the benefits that will accrue with 
adoption of exemption values from TS-R-1 (e.g., harmonization with 
other regulatory agencies and facilitation of international shipments) 
will outweigh the costs (e.g., administrative changes, determining 
whether packages are exempt, and regulating previously exempt 
packages).
    Comment. One commenter opposed the proposed exemption values 
because they were not derived directly and did not directly involve 
public input or a cost-benefit analysis.
    Response. A preliminary RA that evaluated possible costs and 
benefits was conducted as part of the development of this rule. 
Additional information obtained during the rulemaking process was 
considered in determining NRC's final position on adopting the TS-R-1 
exemption values.
    Comment. One commenter stated that, although the revised limits are 
not expected to create any significant burden to the Naval Nuclear 
Propulsion Program, use of the new limits could create a cumbersome 
work practice for some shipments. All low-level shipments that are 
currently exempt will require a detailed evaluation to ensure that 
activity concentrations for each radionuclide are acceptable. For 
example, thoriated tungsten weld rods and soil from site excavations 
would require individual isotope analyses at an additional expense. The 
commenter stated that the current 70-Bq/g activity concentration limit 
for domestic shipments should be retained.
    Response. The comment is consistent with others from the shipping 
community (i.e., the radionuclide activity concentration and activity 
exemption values are likely to be more cumbersome to work with but do 
not pose an excessive burden). The NRC agrees that expenses may be 
involved in achieving compliance with these values but notes that 
expenses are also associated with determining compliance with the 
current 70-Bq/g (0.002-[mu]Ci/g) value. Most shipments of radioactive 
materials involve materials that have been processed to concentrate 
radioactivity. These materials are known by shippers to greatly exceed 
the exemption values, and are packaged and transported in accordance 
with the radioactive material transporation safety regulations. Thus 
the exemption values are irrelevant to the majority of radioactive 
material shipments, such as most shipments in the Naval Nuclear 
Propulsion Program and most shipments in industry as well. The 
exemption values are relevant to shipments of low activity 
concentration. For these shipments, shippers will need to establish 
either by process knowledge or analysis whether a shipment exceeds the 
exemption values and is regulated in transport as a radioactive 
hazardous material, or does not exceed the exemption values and may be 
shipped as non-hazardous material (regular freight). Most shipments 
that minimally exceed the exemption values are likely to be transported 
as limited quantities, which would impose a minimal regulatory burden 
on shippers. Overall, NRC believes that the benefits that will accrue 
with adoption of exemption values from TS-R-1 (e.g., harmonization with 
other regulatory agencies and facilitation of international shipments) 
will outweigh the costs (e.g., administrative changes, determining 
whether packages are exempt, and regulating previously exempt 
packages).
    Comment. Two commenters stated that the proposed rule would 
increase industry's regulatory burden. In particular, the NRC was told 
that the proposed rule is too conservative and would unnecessarily 
burden industry, particularly in the case of bulk shipments of 
contaminated materials. The proposed exemption thresholds would 
increase worker exposure to radioactive materials.
    Response. NRC acknowledges that the exemption values impose some 
new complexity and economic burden on industry. However, NRC believes 
that the increase in costs will be minimal. The NRC believes that the 
exemption values represent a good balance between economic and public 
health interests. From an economic perspective, the increased costs of 
the exemption values are outweighed by the benefits of conforming to 
other regulatory agencies and facilitating international shipments. NRC 
staff recognizes that preshipment requirements under the exemption 
values may increase some low-level exposures, but the NRC still expects 
that

[[Page 3714]]

the shift to a consistent set of dose-based exemption values will 
minimize the potential dose to transport workers.
    Comment. One commenter stated that, although cost reduction was one 
incentive for the rule, the proposed rule as written was so complicated 
that enforcement costs would rise.
    Response. NRC acknowledges the comment and, as previously 
discussed, NRC believes that any additional enforcement or other costs 
will be minimal due to the anticipated benefits of having only one set 
of shipping requirements, as well as the cost savings that would result 
from moving some materials outside the scope of transport regulation.
    Comment. Two commenters stated that the proposed regulations failed 
to properly implement IAEA exemption values regarding naturally 
occurring radioactive material, which would dramatically expand the 
universe of regulated materials and increase the burden on the 
regulated community. One commenter stated that other agencies, such as 
the Occupational Safety and Health Administration (OSHA), afford 
adequate protection from naturally occurring radioactive materials for 
workers and the public, and therefore NRC should not enter this 
regulatory arena. This commenter also stated that the proposed 
exemption values would also lead to a conflict with the Resources 
Conservation and Recovery Act (RCRA), which stipulates that waste 
disposal sites may not accept radioactive materials of more than 70 Bq/
g.
    Another commenter specifically noted that the NRC has not 
implemented the exemption provisions for phosphate ore and fertilizer; 
zirconium ores; titanium minerals; tungsten ores and concentrates; 
vanadium ores; yttrium and rare earths; bauxite and alumina; coal and 
coal fly ash. The commenter urged NRC to consider the activity 
concentration of the parent nuclide in determining exemption values.
    Response. Section 71.14(a)(1) provides the same exemption for low 
level materials (e.g., natural materials and ores) that IAEA provides 
in TS-R-1 paragraph 107(e). The exemption multiple for activity 
concentration (10 times the values listed in 10 CFR part 71, Table A-2) 
applies to natural material and ores containing naturally occurring 
radionuclides which are not intended to be processed for use of these 
radionuclides. If the materials identified in the comment meet the 
definition and are not being processed to use radionuclides, the 
exemption multiple would apply. Thus, the burden indicated by the 
commenter would not occur.
    The activity concentration for exempt material applies to each 
radionuclide listed in Table A-2. For radionuclides in secular 
equilibrium with progeny, the listed activity concentration applies to 
the listed radionuclide (as parent), and was determined considering the 
contribution from progeny. Table A-2, as published on April 30, 2002; 
67 FR 21472, contains several typographical errors, including the 
omission of the reference to footnote (b) for the U (nat) and Th (nat) 
radionuclides. These errors have been corrected in this final rule.
    Comment. One commenter was concerned that the exemption values in 
TS-R-1 could result in the unnecessary regulation of certain materials 
that are currently exempt from NRC regulation under Sec. 40.13. The 
commenter urged NRC to allow unimportant quantities to remain exempt. 
The commenter was concerned that the public and operators of RCRA 
disposal facilities may question the safety of materials that were 
previously exempt but are not exempt under the new regulations. The 
commenter pointed out that the actual risk would not change because 
RCRA will not change.
    Response. Materials that are exempt (i.e., not licensed) under Sec. 
40.13 are not subject to part 71 under the current or final 
transportation regulations. Nothing in this final rule affects the 
exemption status of materials subject to Part 40.
    RCRA sites can continue to use the 70-Bq/g (0.002-[mu]Ci/g) value 
as a material acceptance criterion at their option. The final rule 
establishes new exemption values for radioactive materials in transport 
that differ from 70 Bq/g (0.002 [mu]Ci/g) that might be used (for 
nontransport purposes) at RCRA sites. However, the final rule does not 
preclude the shipment of materials to RCRA sites in a manner that would 
satisfy both transportation and site safety regulations.
    Comment. Ten commenters expressed opposition to the exemption 
values. One commenter argued that the proposed guidelines should allow 
no exemptions. Two commenters stated that the proposed exemptions would 
negatively impact public health. Two commenters argued that the 
redefinition would pose a threat to public health. Two commenters 
opposed weakening regulations that would reduce the public safety and 
health through new definitions or accepted concentration values. Two 
commenters emphasized that there is no justification for increasing 
allowable concentrations because there are ramifications beyond 
transportation, and that using a dose-based system is less measurable, 
enforceable, and justifiable.
    Some commenters added that if NRC needed to adopt risk-based 
standards, NRC should adopt the standards that would reduce the 
allowable exemptions. One commenter criticized the proposed rule for 
increasing the allowable contamination in materials. One commenter 
disagreed with the current 70 bequerels-per-gram exemption level and 
urged NRC to change only the exemption levels to make them more 
protective for isotopes whose exempt concentrations go down.
    One commenter also stated that NRC had not actively participated in 
determining the proposed exemption values.
    Response. NRC disagrees with the comment that no exemptions should 
be allowed. Because almost all materials contain at least trace 
quantities of radioactivity, if there were no exemptions, essentially 
all materials transported in commerce would be treated as radioactive 
materials. This would entail considerable expense and impact on 
commerce without commensurate benefit to public health and safety.
    The NRC disagrees that the proposed exemptions would negatively 
impact public health. The NRC's analysis of the radionuclide-specific 
exemption values indicates the overall dose impact of their adoption 
would be low (much less than background levels), and lower than that of 
the single-value exemption currently in place. Please see the 
Background section under this issue for further details.
    The NRC acknowledges the comment that there is no justification for 
increasing allowable concentrations. However, the NRC believes the 
benefits of the exemption values will outweigh the costs. NRC analyses 
lead the NRC staff to believe that the increase in regulatory 
efficiency between regulatory agencies and the facilitation of 
international shipments make the exemption values advantageous overall. 
The NRC finds the low uniform-dose approach that was used in the 
development of the exemption values to be acceptable.
    Although additional measurements may be necessary under the new 
requirements, the industry has not indicated that these requirements 
pose an excessive burden. The NRC does not believe the radionuclide 
exemption values would be less enforceable than the current single 
exemption value.
    Lastly, as a working participating member of the IAEA, both NRC and 
DOT staff participated in the development of the exemption values.

[[Page 3715]]

    Comment. One commenter requested information on calculations for 
dose impacts to members of the public, particularly regarding recycling 
and the possibility of exempting materials that pose a radiation hazard 
to the public.
    Response. An assessment of public dose that might result from 
adopting the exempt activity concentrations and exempt activities per 
consignment under transportation scenarios may be found at the 
following reference: A. Carey et al. The Application of Exemption 
Values to the Transport of Radioactive Materials. CEC Contract CT/PST6/
1540/1123 (September 1995). The NRC has performed no assessment 
regarding recycling because that is beyond the scope of this 
rulemaking.
    Comment. A commenter requested the risk and biokinetic data 
supporting the proposed exemption values. The commenter also wanted to 
know more about who determines what data NRC uses, including the 
physiological data used to justify the change in dose models.
    Response. The basic radiological protection data used in the 
development of the exempt activity concentrations and exempt activities 
per consignment may be found at the following reference: International 
Basic Safety Standards for Protection Against Ionizing Radiation and 
for the Safety of Radiation Sources, Safety Series No. 115, IAEA 1996.
    Comment. Two commenters stated that it is unclear how or why the 
risk decreases for 222 of the 382 listed radioisotopes, when the 
allowable concentrations for those radioisotopes increase to above 70 
becquerels. The commenters asked how the ``risk or dose goes down'' 
while some exempt quantities could lead to more than the ``worker doses 
to members of the public from unregulated amounts of exempt quantities 
of radioisotopes.''
    Response. Under the previous system, radioactive materials 
exceeding the 70-Bq/g (0.002-[mu]Ci/g) activity concentration were 
regulated in transport. Although the 70-Bq/g (0.002-[mu]Ci/g) value 
applied to all radionuclides, different radionuclides resulted in 
different doses to the public when transported at that activity 
concentration (as calculated using the transport scenarios). The 
transport scenario doses for many radionuclides when transported at 70 
Bq/g (0.002 [mu]Ci/g) are less than the reference dose of 0.01 mSv/y (1 
mrem/y). However, for other radionuclides, the transport scenario doses 
at 70 Bq/g (0.002 [mu]Ci/g) are greater than the reference dose of 0.01 
mSv/y (1 mrem/y). Under the radionuclide-specific approach, the 
calculated doses are more representative, and the average dose 
(considering all radionuclides) is lower than under the 70-Bq/g (0.002-
[mu]Ci/g) approach. Overall, the NRC's analysis shows that the new 
system would result in lower actual doses to the public than the 
current system.
    Comment. Another commenter urged NRC to either make exemption 
values more stringent or not adopt any new values at all.
    Response. The comment provides no justification to make the 
exemption values more stringent. The IAEA and other Member States have 
adopted the new system. Failure to adopt the new system would put the 
U.S. at a competitive disadvantage in international commerce without 
commensurate benefit to public health and safety and would allow the 
continued shipment of exempt materials that are calculated to produce 
higher doses to workers and members of the public.
    Comment. One commenter asked that NRC provide a separate activity 
concentration threshold, and suggested 2,000 picocuries per gram, for 
samples collected for laboratory analysis in situations where relevant 
data is unavailable. The commenter believes that the current proposed 
threshold of 2.7 picocuries per gram is too restrictive for samples 
acquired for laboratory analysis.
    Response. Although data is apparently unavailable for the samples 
the commenter refers to, it appears the samples are minimally 
radioactive and, therefore, could be shipped as a limited quantity, one 
of the least burdensome shipments. As we received no other comment on 
this issue, the commenter's concern does not appear to be widespread. 
The NRC has concluded that the information and justification provided 
do not warrant the introduction of a provision in part 71 that would 
not be compatible with TS-R-1.
    Comment. One commenter asked that NRC provide for expeditious 
transportation of discrete solid sources encountered in public areas. 
The commenter noted that part 71 currently permits a source of up to 
2.7 millicuries to be transported as a limited quantity, even if no 
relevant data about the source is available. The commenter then asked 
NRC to retain this arrangement for sources encountered in public areas 
because it has been a useful provision.
    Response. The quantities involved (2.7 mCi) would not normally 
require NRC-certified packaging, thus the current part 71 rulemaking 
would have little bearing upon them. The NRC understands that DOT has a 
system of exemptions in place, which has been coordinated with State 
regulators, to facilitate the safe and timely transport of sources 
discovered in the public domain.
    Comment. One commenter asked about the proposed mechanism for 
approving nondefault exemption values. Some commenters requested 
further information on how default exemption values could be calculated 
from the A1 and A2 values.
    Response. The scenarios used to develop the exemption values were 
selected to model exposures that could result from relatively close 
distances and long duration exposure times to exempt materials. The 
scenarios used in the Q-system were selected to model exposures that 
could result from shorter-term exposure to the contents of a damaged 
Type A package following an accident. Because of the differences in the 
exposure scenarios and the resulting differences in the equations used 
to calculate the values, the Q-system cannot be used to calculate 
activity limits for exempt consignments or exempt activity 
concentrations.
    Comment. One commenter stated that the landfill disposal of NORM is 
outside NRC jurisdiction when technologically advanced NORM is involved 
with RCRA-regulated hazardous constituents. The commenter explained 
that numerous RCRA landfills around the country have adopted the EPA- 
and State-approved programs for the disposal of NORM. The commenter 
wondered how the proposed changes in radionuclide exemption values 
would affect the regulations governing these landfills.
    Response. Part 71 has no direct effect on the regulations governing 
the licensing or operation of landfills. The comment is beyond the 
scope of this rulemaking.
    Comment. Two commenters opposed the regulation of NORM ores and 
natural materials, including materials derived from those substances, 
because it does not include appropriate exemptions and will result in 
unjustified increased costs and transportation burdens and liabilities.
    Response. This rule does not extend NRC's scope of regulation of 
radioactive material. If a material, such as NORM, was not previously 
subject to NRC regulation, it would not be subject to regulation under 
this final rule. For regulatory consistency, both DOT and NRC publish 
the radionuclide exemption tables, including the 10 times exemptions 
for natural materials and ores containing NORM. Also, part 71 only 
applies to material licensed by

[[Page 3716]]

the NRC, and NRC does not regulate NORM.
    Comment. One commenter suggested that NRC reevaluate the proposed 
factor for the allowance of NORM. This commenter recommended that NRC 
consider using a factor of 100 rather than 10, because many materials 
are not hazardous and do not require more stringent shipping 
regulations.
    Response. The comment does not provide compelling data to support 
the requested change. Furthermore, the requested change would result in 
the U.S. being noncompatible with international transportation 
regulations. Therefore, no change is made.
    Comment. One commenter stated that this rule has taken the focus 
off of more important issues in place of issues that are of less 
concern, such as the regulation of NORM. The commenter stated that 
lowering exemption values could distract attention from materials that 
would otherwise be of concern to law enforcement, particularly 
regarding transportation across U.S. borders.
    Response. The exemption values are considered by shippers when 
preparing radioactive materials for transport. The NRC staff does not 
believe these rule changes will affect law enforcement activities.
    Comment. One commenter was concerned that ``uranium and thorium 
levels in phosphate, gypsum, and coal cannot be considered safe simply 
because they are naturally occurring. The commenter added that from a 
public health point of view, there is no need to determine whether 
alpha emissions above the 70-Bq/g (0.002-[mu]Ci/g) threshold are 
naturally occurring or man-made, their effect on somatic cells and germ 
cells is the same.'' The commenter was concerned that NRC has not 
proposed sufficient regulations regarding the ``shipment of ores and 
fossil fuels with regard to radioactive levels of naturally occurring 
radionuclides.'' The commenter requested that NRC provide an analysis 
of the ``regulatory burden of radionuclide HMR on the fertilizer, 
construction, and fossil-fuel energy industries.''
    Response. NRC's transportation regulations apply to NRC licensees 
that transport licensed material and require that licensees comply with 
U.S. DOT Hazardous Materials Regulations. The DOT regulations 
previously included the 70-Bq/g (0.002-[mu]Ci/g) value in the 
definition of radioactive material, and materials determined to be less 
than that activity concentration did not satisfy DOT's definition of a 
radioactive material and were not regulated as hazardous material in 
transport. The DOT definition applied regardless of whether the 
material was naturally occurring or not.
    With regard to burden, this rule adopts a change in the 
transportation exemption for radioactive materials from a single value 
to radionuclide-specific values. In its proposed rule, NRC requested 
specific information on the impact of that change. The information 
provided to NRC is presented in the regulatory analysis accompanying 
this rule.
    Comment. One commenter suggested that NRC not use the wording in 
Sec. 71.14(a)(1), ``Natural materials * * * that are not intended to be 
processed for the use of these radionuclides * * *,'' because it 
unreasonably requires the shipper to know the intended use of the 
material. The commenter emphasized that NRC should base transport 
regulations solely on the radiological properties of the material 
shipped.
    Response. This provision applies to a subset of the industry that 
processes an ore that contains radioactive material, not for the 
radioactive material, but for some other element, mineral, or material. 
For example, this provision would apply to the processing of an ore 
during which thorium or uranium was produced incidentally in a waste 
stream, but would not apply to the processing of an ore to extract 
thorium or uranium for use or sale. NRC staff believes the industry can 
reasonably be expected to determine the intent for processing the ore 
when that ore is shipped to a consignee.
    Comment. One commenter indicated that, should the exemption values 
be adopted in a way that departs from IAEA, newly regulated entities 
could face high monetary penalties for failure to comply with the 
regulations due to DOT's enforcement penalty policies. The commenter 
noted that DOT regulations preempt and supersede State and local 
regulations, so these regulations make it more difficult for people to 
protect themselves from the dangers of exposure to radiation.
    Response. The NRC staff believes the rule adopts the exemption 
values in a manner that is compatible with the IAEA regulations and 
with a parallel DOT final rule.
    Comment. One commenter asked the NRC if States whose regulations 
are more protective than the proposed rule would have to abandon those 
regulations if NRC adopted the proposed rule.
    Response. States do not have regulations that are more protective 
than those in this rulemaking for the transportation of radioactive 
materials. State regulations in this area are essentially identical to 
those of the Federal government to eliminate any conflicts, 
duplications, gaps, or other conditions that would jeopardize an 
orderly pattern in the regulation of radioactive materials on a 
nationwide basis.
    Comment. One commenter stated that there is no way to know how much 
is being exempted in terms of curies or becquerels because there is no 
limit on the number of negligible doses from exemptions.
    Response. The dose criteria used in determining the activity 
concentrations for exempt materials ensure that the doses (from either 
single or multiple sources) do not reach unacceptable levels, and will 
therefore be far below public dose limits. Quantifying exempted 
materials (i.e., those materials that are not regulated as radioactive 
material in transport) would impose a significant burden without 
commensurate benefit to public health and safety.
    Comment. One commenter expressed concern that, for some members of 
the public, exposure could be over 100 millirem per year. The commenter 
understood from the proposed rule that the dose-based exemption values 
are designed to deal with transport worker exposures in the range of 25 
to 50 millirem per year. The commenter requested information about how 
the expected annual dose to transport workers changes under the 
proposed rule, particularly if it increases or decreases.
    Response. The NRC staff notes that exposures to members of the 
public are more likely to be over 1 mSv (100 mrem) per year under the 
current single exemption value than under the radionuclide-specific 
system. However, these are dose estimates; the transport scenarios used 
to estimate these doses overstate actual doses by overstating exposure 
periods in a year (50-400 hrs/yr) and exposure distances [less than 
1.52 m (5 ft)] to radioactive materials in transport.
    For those radionuclides with a relatively low estimated dose for 
transport at 70 Bq/g (0.002 [mu]Ci/g) under the transport scenarios, 
the estimated dose will increase under the dose-based exemptions; for 
those radionuclides with a relatively high estimated dose for transport 
at 70 Bq/g (0.002 [mu]Ci/g) under the transport scenarios, the 
estimated dose will decrease under the dose-based exemptions. Even in 
those instances where the estimated dose increases under the final 
rule, the dose remains low and the average dose (considering all 
radionuclides) is lower under the radionuclide-specific system.

[[Page 3717]]

    Comment. One commenter questioned the composition of a list of 20 
representative nuclides used to estimate the average annual dose per 
radionuclide. The commenter asserted that, among the 20 representative 
nuclides, a minority of nuclides whose doses decrease in the proposed 
regulations were overrepresented. The commenter stated that most of the 
dose concentrations increase, some of them dramatically.
    Response. The 20 radionuclides referred to were chosen to be 
representative of the radiation types (alpha, betas of various 
energies, and gamma) most commonly encountered in transport and were 
used to provide a representative measure of the proposed rule's likely 
impact.
    Although the radionuclide activity concentration values more often 
exceed 70 Bq/g (0.002 [mu]Ci/g) than fall below it, the distribution of 
all the new exemption values centers just above 70 Bq/g (0.002 [mu]Ci/
g).
    It is recognized that the exempt activity concentration for some 
radionuclides (those radionuclides with very low doses under the 
transport scenarios when transported at 70 Bq/g (0.002 [mu]Ci/g)) will 
increase under a dose-based exemption system. However, the measure of 
impact from the change in exemption values is the estimated dose, and 
that remains low, even for radionuclides where the exempt activity 
concentration increases above 70 Bq/g (0.002 [mu]Ci/g). The radiation 
protection benefit from the radionuclide-specific approach is that the 
highest potential doses are reduced as well as the average dose from 
all radionuclides.
    Comment. One commenter noted that there is no precedent for exempt 
quantities in NRC regulations and that this will create a new category. 
The commenter questioned the logic of creating such a category.
    Response. The DOT transportation safety regulations for radioactive 
materials have always had a de facto ``exemption value'' built into the 
definition of ``radioactive material.'' NRC regulations either 
replicate or include references to DOT regulations. Any material with 
an activity below the 70-Bq/g (0.002-[mu]Ci/g) threshold was not 
defined as radioactive for the purposes of the regulations and 
therefore was not subject to the regulations (i.e., exempt). Without 
the exempt activity for consignments value, any quantity of material 
that exceeded the exempt activity concentration, no matter how small, 
would be regulated in transport as radioactive material. The exempt 
consignment value is included to prevent the regulation of trivial 
quantities of material as hazardous material in transport.
    Comment. One commenter stated that the threat of terrorism should 
be taken into account when exempting radionuclides from transport 
regulations and changing container regulations.
    Response. The nature of exempt materials is that they are either of 
very low activity concentration or very low total activity. In both 
cases, these materials present little hazard and would not be 
attractive as targets for terrorist activities.
    Comment. One commenter expressed concern that the revised exempt 
concentrations in Table A-2 are a significant change in the 
requirements for the transportation of unimportant quantities of source 
materials.
    Response. Although the comment expresses concern that the exempt 
activity concentration values represent a significant change in the 
requirements for unimportant source material, it does not provide data 
or justification for this statement. NRC acknowledges that the 
internationally developed transportation exemption values do not align 
precisely with preexisting, domestic requirements in NRC regulations in 
10 CFR part 30 or part 40 that were developed for other licensing 
purposes. However, the current 70-Bq/g (0.002-[mu]Ci/g) exemption value 
does not align precisely with part 30 or part 40 requirements either. 
In most cases, the differences in the regulatory requirements do not 
appear to be that significant, and the industry has not provided data 
that demonstrate that the impact from the change for actual shipments 
would be significant. NRC has no basis to change its conclusion in the 
final RA that the overall benefits of achieving compatibility by 
adopting the exemption values outweigh the associated costs, or its 
belief that permitting natural materials and ores to be shipped at 10 
times the Table A-2 values minimizes the impacts.
    Comment. Five commenters supported NRC's efforts in the proposed 
rule. One of these commenters supported lower concentrations for the 
radioactive isotopes because the proposed rulemaking increases public 
risk. Another stated that it was important to ensure consistency 
between international and domestic regulations and that while 
individual radionuclide levels may be raised or lowered by the proposed 
rule, overall the estimated dose would be significantly lower. Another 
commenter agreed with NRC's proposal to adopt the radionuclide 
exemption values in TS-R-1, particularly the inclusion of exempt 
consignment quantities in the regulations. Another commenter expressed 
general support for ensuring consistency between domestic and 
international regulations.
    Response. NRC acknowledges the comments on revising radionuclide 
exemption values. NRC staff agrees with the commenters who stated that 
consistency between international and domestic regulations is a high 
priority, and that the exemption values overall will result in lower 
public exposure. However, while promulgating lower exemption levels 
could reduce the already low public health risks, NRC believes that the 
exemption values offer the best balance between economic and public 
health concerns.
    Comment. One commenter stated that the proposed exemption values 
were too complex because it is too complicated to maintain more than 
half of all exemption values at 70 Bq/g (0.002 [mu]Ci/g) and to reduce 
those that are more protective.
    One commenter said that there are no comparable exemptions in 
existing regulations.
    Response. The NRC does not believe that the proposal to maintain 
more than half of the activity concentration exemption values at 70 Bq/
g (0.002 [mu]Ci/g), while reducing the activity concentration exemption 
values for the remaining radionuclides, is warranted because the 
resulting exemption system would be inconsistent, have no defined dose 
basis, and would be incompatible with that of the IAEA and other Member 
States.
    The final rule introduces exemptions from the application of the 
hazardous materials transportation regulations for materials in 
transit. However, the definition of ``radioactive materials'' in the 
transportation regulations has, for decades, contained a minimum 
activity concentration value (i.e., any material with an activity 
concentration less than 70 Bq/g (0.002 [mu]Ci/g)); effectively, the 
definition has contained an exemption value. The final rule changes the 
structure of the exemption from a single activity concentration value 
applicable to all radionuclides to individual activity concentration 
and consignment activity values that are specified for each 
radionuclide.
    Comment. Several commenters expressed concern about the health 
effects of these regulations. One commenter opposed reliance on the 
ICRP arguing that ICRP does not take into consideration important 
information on the health impacts of radiation such as synergism with 
other contaminants in the environment and the bystander effect, in 
which cells that are near cells that are hit, but are not

[[Page 3718]]

themselves hit by ionizing radiation, exhibit effects of the exposure. 
One commenter stated that the NRC did not consider the new evidence 
that low doses of radiation are more harmful per unit dose than was 
previously known. This commenter further noted that there are 
synergistic effects and other types of uncertainties in radiation 
health effects. Three commenters opposed the radionuclide exemption 
value tables citing the use of outdated data, lack of data, and/or the 
lack of calculations for more than 350 radionuclides. One commenter 
stated that NRC radiation standards are outdated and should be subject 
to rigorous review, including independent outside experts. One 
commenter stated that ICRP does not represent the full spectrum of 
scientific opinion on radiation and health and does not take into 
account certain health impacts of radiation. One commenter noted that 
ICRP and IAEA risk models only look at fatal cancers and ignore 
nonfatal cancers, years of lost life, and the bystander effect. The 
commenter also asserted that these agencies' reports do not accurately 
reflect risk and that low levels of radiation are more damaging than 
the models are predicting.
    Response. The Board of Governors of the International Atomic Energy 
Agency stated in 1960, that ``The Agency's basic safety standards * * * 
will be based, to the extent possible, on the recommendations of the 
International Commission on Radiological Protection (ICRP).'' The ICRP 
is a nongovernmental scientific organization founded in 1928 to 
establish basic principles and recommendations for radiation 
protection; the most recent recommendations of the ICRP were issued in 
1991 (International Commission on Radiological Protection, 1990 
Recommendations of the International Commission on Radiological 
Protection, Publication No. 60, Pergamon Press, Oxford and New York 
(1991)). The IAEA Basic Safety Standards (from which the exemption 
values are taken) were developed with full IAEA Member State 
participation (including the U.S.) and have taken the ICRP 
recommendations into account. NRC rejects the comment that the data 
used to develop the exemption values are outdated or inadequate. In 
general, NRC believes ICRP reports provide a widely held consensus view 
by international scientific authorities on radiation dose responses and 
accepts their principal conclusions. Furthermore, the NRC notes that 
fundamental research into radiation dose effects is beyond the scope of 
this rulemaking. For that information, NRC relies on national and 
international scientific authorities.
    Comment. The NRC was criticized by commenters for not having 
developed and pursued actual transport exposure scenarios for every 
radionuclide to justify the exemptions. One commenter also noted that 
although NRC has not carried out calculations for transportation 
scenarios for over 350 of the listed radionuclides, individual exempt 
concentration and quantity values have been assigned to each 
radionuclide. The commenter further concluded that NRC has technical 
data to support the conclusion that these exemption values will pose no 
risk to the public. Another commenter stated that it was unclear why 
NRC performed calculations for only 20 of the 350 isotopes. The 
commenter noted that because NRC only modeled 20 of the radionuclides, 
NRC has not collected complete data for the other radionuclides; 
otherwise, they would have been also modeled. The commenter further 
stated that NRC should either lower the exemption values or withdraw 
the values and perform further studies.
    Response. NRC selected a subset of 20 radionuclides believed to be 
representative of the most commonly transported radionuclides. Exempt 
activity concentration and consignment activity values were calculated 
for all the radionuclides listed in Table A-2, not just the 20 selected 
to be used in NRC's impact analysis. NRC used the 20 radionuclides to 
illustrate that the impact from activity concentration exemption values 
for materials commonly transported in significant quantities is less 
than that from the current single exemption value.
    Comment. One commenter expressed concern that NRC had arbitrarily 
determined the radionuclide values.
    Response. The A1 and A2 values in Table A-1 
and the exempt activity concentration values and exempt activity values 
in Table A-2 are not arbitrary values. The derivation of these values 
is dose based and provided in the references in TS-R-1.
    Comment. One commenter expressed opposition to the exemption values 
because they raised the allowable exempt concentrations and allowed for 
exempt quantities, which are currently not permitted.
    Response. The current definition of radioactive material is 
specified only in terms of a minimum activity concentration. 
Conceivably, this leads to the regulation of any quantity of material 
that exceeds that activity concentration, even minute quantities, as a 
radioactive material in transport. To address this issue, an activity 
limit for exempt consignments has been introduced that specifies a 
minimum activity that must be exceeded for a material to be regulated 
as a radioactive material in transport.
    As with the exempt activity concentration values, the exempt 
activity values in Table A-2 were taken from the BSS exemption values. 
The doses associated with the use of these exempt activity values were 
estimated using the same scenarios used for assessing the impact of the 
exempt activity concentration values. The results are that doses are 
low, and that for 19 of the 20 representative radionuclides examined, 
the dose from the radionuclide exempt activity value is less than that 
from the exempt activity concentration value.
    Comment. One commenter asked if there is any possibility that NRC 
could simply decline to adopt the sections of the proposed rules that 
relate to radionuclide exemption values.
    Response. NRC's and DOT's approach in this compatibility rulemaking 
is to adopt the provisions of IAEA's TS-R-1 as proposed unless adoption 
would pose a significant detriment to radioactive material transport 
commerce, or is unjustified. The NRC has determined that the exemption 
change is justified based on its regulatory analysis and public 
comments.
    Comment. One commenter stated that NRC should ensure that no member 
of the public would receive a dose above 1mrem/year from any practice 
or source, and should clarify what is meant by ``practice'' and 
``source.'' One commenter stated that the current HMR standard of 70 
Bq/g (0.002 Ci/g) should be maintained as the minimum standard for the 
protection of public health and transport worker safety. The commenter 
opposed the replacement of this standard with the radionuclide-specific 
values per the IAEA's TS-R-1 for the following reasons:
    (1) There is no radiation risk level which is sufficiently low as 
to be of no regulatory concern;
    (2) There are no collective radiological impacts which are 
sufficiently low as to be of no regulatory concern; and
    (3) No one will be able to determine if proposed exempt sources are 
safe.
    One commenter noted that the current and proposed regulations have 
50 and 23 millirem being average doses, respectively. To adequately 
protect public health, the average dose should be no more than one 
millirem. One commenter stated the assumptions and

[[Page 3719]]

scenarios that NRC and DOT used to justify the adoption of these 
exemption values fail to prove that these exemptions will have either 
no or an insignificant effect.
    One commenter stated that the proposed exemption values are based 
on unrealistic models. The commenter said that the exempt levels do not 
appear to reflect the material's longevity in the environment and 
hazard to living creatures. One commenter stated that the standards 
should be based on the most vulnerable members of the population, and 
NRC should adopt stricter values. Two commenters argued that, using the 
existing dose models, some of the exempt quantities could lead to high 
public doses from unregulated amounts of exempt quantities of 
radioisotopes. Another commenter opposed reliance on computer model 
scenarios that may not be realistic to project doses, citing that this 
lack of realism to justify certain exposure scenarios is inadequate. 
One commenter stated that it is unclear in the proposed regulations 
what the exact dose impact will be in converting from an empirical 
exemption value to a dose-based exemption value. The commenter's 
understanding is that while there is a reduction in dose for the 
results that were calculated, the standard deviation and median dose 
values both decrease. One commenter was concerned that the proposed 
exemption values are not adequately protective for transportation 
scenarios, because the IAEA transportation exemption values for some 
radionuclides are too high to meet safety goals. The commenter added 
that the average annual dose for a representative list of 20 
radionuclides (see April 30, 2002; 67 FR 21396) is too high to be safe. 
Some commenters stated that NRC should tighten controls on radioactive 
materials instead of loosening them because NRC admitted that the 
proposed increases in exempt concentrations of radioactive materials 
would reduce public safety, One commenter stated that the public is 
told not to worry about the proposed exemption values because it will 
only be exposed to one millirem of radioactive material. However, the 
commenter noted that the 20 most commonly shipped materials with the 
new exemption values are at 23 millirem. Therefore, the commenter was 
confused about what it meant to only be exposed to one millirem of 
radioactive material. One commenter stated that the proposed exemption 
values would not enforce the principle of limiting exposure to less 
than 1 mrem/yr. Four other commenters opposed the proposed definition 
of ``radioactive materials,'' one doing so in the name of national 
security. This commenter argued that there are no low-level nuclear 
wastes and that there is no safe threshold for exposure to radioactive 
materials.
    Response. The terms ``practice'' and ``source'' are used in the 
context of the IAEA's BSS, and have the meanings provided in the 
glossary of that document.
    A criterion for the BSS exemption of practices ``without further 
consideration'' (Schedule I, paragraph I-3) is that the effective dose 
expected to be incurred by any member of the public due to the exempted 
practice is of the order of 0.01 mSv (1 mrem) or less in a year. 
Estimates of doses resulting from the use of the exemption values in 
the transport scenarios have been specifically examined and may result 
in doses that exceed 0.01 mSv/yr (1 mrem/yr) (an average of 0.23 mSv/yr 
(23 mrem/yr) for 20 commonly transported radionuclides). However, the 
dose estimates for the use of the exempt activity concentration values 
are less than those resulting from the use of the current 70-Bq/g 
(0.002-[mu]Ci/g) activity concentration (an average of 0.5 mSv/yr (50 
millirem/yr) for the same 20 radionuclides). The NRC staff notes that 
there have been no adverse public health impacts identified from the 
use of the current exemption value. Because the annual doses estimated 
to result from the use of the radionuclide-specific exemption values 
are low, and on average are lower than the dose estimates for the 
current 70-Bq/g (0.002-[mu]Ci/g) activity concentration, the NRC staff 
believes that changing from the 70-Bq/g (0.002-[mu]Ci/g) value to the 
radionuclide-specific exemption values will result in no adverse impact 
on public health and safety.
    In addition, the transport scenarios are based on exposure periods 
(40-500 hours per year) and exposure distances (less than 1.52 m (5 
ft)) that overstate actual exposures to workers and greatly overstate 
actual exposures to the public. The models used to develop the 
exemption values consider the exposure pathways that are significant 
for assessment of impact on public health and safety, including 
external exposure, inhalation and ingestion, and contamination of the 
skin.
    The length of the exposure periods and the close distance 
assumptions make multiple exposures for the full duration at those 
distances to multiple radionuclides very unlikely. The dose estimates 
are sufficiently low that NRC believes any actual multiple exposures 
would also be acceptably low (well below regulatory limits). Neither 
NRC nor DOT has any information to suggest that multiple exposures to 
materials regulated under the current 70-Bq/g (0.002-[mu]Ci/g) minimum 
activity concentration is of concern.
    The NRC believes that regulatory efficiency requires that exemption 
values be established for determining when material in transport should 
be subject to radioactive material transport safety regulations. The 
NRC believes adoption of the radionuclide-specific exemption values is 
warranted because it achieves international compatibility without 
negative public health impact or undue burden.
    Comment. One commenter stated that the proposed regulations were 
unclear as to the exact definition of ``per radionuclide.''
    Response. The term ``per radionuclide'' means that the doses 
estimated to result from the use of the exemption values were 
determined for each radionuclide.
    Comment. One commenter expressed the lack of understanding of the 
concept of the ``millirem.'' To this end, the commenter said that 
``millirem'' is a fluid, unenforceable, and unverifiable term.
    Response. The term ``millirem'' is a combination of the prefix 
``milli,'' meaning one-thousandth, and ``rem,'' an acronym for Roentgen 
Equivalent Man, a radiation dosimetry unit. Units of radiation doses, 
including rem, are defined in Sec. 20.1004.
    Comment. One commenter requested that NRC track, label, and 
publicly report all radioactive shipments of any kind, and reject the 
exemption tables. The commenter believed that ``harmonization'' was not 
an adequate justification for increasing public risk.
    Response. The NRC believes that the current regulations require 
appropriate measures for hazard communication during transportation. As 
noted previously, the public risk from the transportation of exempt 
materials, as measured by the average dose, will actually decrease.
    Comment. One commenter stated that the new exemption values will 
result in bulk shipments of decommissioning soil and debris being 
classed as LSA (Low Specific Activity) rather than being exempted from 
regulation. The commenter quantified the percentage of his shipments 
that would now be classed as LSA. The commenter stated that the 
increase in LSA-classified shipments will result in minimal additional 
costs.
    Response. No response is required.

[[Page 3720]]

    Comment. One commenter expressed opposition to the changes in 
definitions that could include changing exemption values, particularly 
because this is not subject to an EA.
    Response. This rule adopts the TS-R-1 exempt material activity 
concentrations and exempt consignment activity limits as found in Table 
A-2 of the proposed rule. In essence, use of both of these values will 
replace the current definition for ``radioactive material'' found in 49 
CFR 173.403, and applied in current 10 CFR 71.10. Within the revision 
to part 71, reference to the exemption values will be added to the new 
Sec. 71.14, ``Exemption for low-level materials,'' to provide an 
exemption from NRC requirements during the transportation of these 
materials. Estimated impacts from this revision are included in the EA 
prepared to support this rulemaking.
    Comment. One commenter stated that the redefinition would pose a 
threat to national security.
    Response. NRC does not believe adoption of the exemption values for 
radioactive materials in transport will have any bearing on national 
security.
    Comment. One commenter expressed concern that the NRC proposed 
regulations could increase the variety of materials that are regulated 
as ``radioactive'' for transportation purposes.
    Response. It is possible that materials that were not regulated 
under the previous DOT definition based on 70 Bq/g (0.002-[mu]Ci/g) 
would be newly regulated under the exemption values. However, a 
material consignment must exceed both the activity concentration for 
exempt material and the activity limit for exempt consignment to be 
regulated under the final DOT and NRC regulations. It is NRC's position 
that regulation of such material consignments as radioactive material 
in transport is appropriate.
    Comment. One commenter asked the NRC to explain how NRC's official 
proposal on the changes in packaging and transporting of radioactive 
materials would affect industrial radiography.
    Response. The final rule does not affect the transportation of 
standard industrial radiography devices.
    Comment. One commenter stated that in ``no case should NRC part 71 
definitions be relaxed or downgraded merely to provide ``internal 
consistency and compatibility with TS-R-1.''' The commenter stated that 
those who ``wish to engage in trans-boundary trade in nuclear materials 
can be required to meet stiffer U.S. import requirements'' than those 
elsewhere in the world. The existing NRC staff justification is ``a 
very lame dog that won't hunt,'' and regulatory relaxation is ``both 
arbitrary and capricious and unacceptable.'' The commenter stated that 
NRC should have definitions with full clarity, and no changes should be 
allowed that reduce safety levels or relax requirements. The commenter 
was especially troubled with the proposed change to ``radioactive 
material'' because this change would ``allow shipments of radioactively 
contaminated materials that are declared to be exempted according to 
the concentrations and consignment limits shown in the Exemption 
Tables.''
    Response. NRC believes that the amended definitions and new 
adoptions to support definitions for individual Issues are sufficiently 
justified and not arbitrary and capricious.
Issue 3. Revision of A\1\ and A\2\
    Summary of NRC Final Rule. The final rule adopts, in Appendix A, 
Table A-1 of part 71, the new A1 and A2 values 
from TS-R-1, except for molybdenum-99 and californium-252. The final 
rule does not include A1 and A2 values for the 16 
radionuclides that were previously listed in part 71 but which do not 
appear in TS-R-1.
    The A1 and A2 values were revised by IAEA 
based on refined modeling of possible doses from radionuclides. The NRC 
believes that these changes are based on sound science, incorporating 
the latest in dosimetric modeling and that the changes improve the 
transportation regulations. The regulatory analysis indicates that 
adopting these values is appropriate from a safety, regulatory, and 
cost perspective. Further, adoption of the new A1 and 
A2 values will be an overall benefit to public and worker 
health and international commerce by ensuring that the A1 
and A2 values are consistent within and between 
international and domestic transportation regulations. The NRC is not 
adopting the A1 value for californium-252 because the IAEA 
is considering changing the value that appears in TS-R-1 back to what 
presently appears in part 71. The NRC is not adopting the A2 
value for molybdenum-99 for domestic commerce because this would result 
in a significant increase in the number of packages shipped, and 
therefore in potential occupational doses, due to the lower 
A2 value in TS-R-1.
    Affected Sections. Appendix A.
    Background. The international and domestic transportation 
regulations use established activity values to specify the amount of 
radioactive material that is permitted to be transported in a 
particular packaging and for other purposes. These values, known as the 
A1 and A2 values, indicate the maximum activity 
that is permitted to be transported in a Type A package. The 
A1 values apply to special form radioactive material, and 
the A2 values apply to normal form radioactive material. See 
Sec. 71.4 for definitions.
    In the case of a Type A package, the A1 and 
A2 values as stated in the regulations apply as package 
content limits. Additionally, fractions of these values can be used 
(e.g., 1x10-\3\ A2 for a limited quantity of 
solid radioactive material in normal form), or multiples of these 
values (e.g., 3,000 A2 to establish a highway route 
controlled quantity threshold value).
    Based on the results from an updated Q-system (see draft Advisory 
Material for the Regulations for the Safe Transport of Radioactive 
Material, TS-G-1.1, Appendix I), the IAEA adopted new A1 and 
A2 values for radionuclides listed in TS-R-1 (see paragraph 
201 and Table I). IAEA adopted these new values based on calculations 
which were performed using the latest dosimetric models recommended by 
the ICRP in Publication 60, ``1990 Recommendations of the ICRP.'' A 
thorough review of the Q-system also included incorporation of data 
from updated metabolic uptake studies. In addition, several refinements 
were introduced in the calculation of contributions to the effective 
dose from each of the pathways considered. The pathways themselves are 
the same ones considered in the 1985 version of the Q-system: External 
photon dose, external beta dose, inhalation dose, skin and ingestion 
dose from contamination, and dose from submersion in gaseous 
radionuclides. A thorough, up-to-date radiological assessment was 
performed for each radionuclide of potential exposures to an individual 
should a Type A package of radioactive material be involved in an 
accident during transport. The new A1 and A2 
values reflect that assessment.
    While the dosimetric models and dose pathways within the Q-system 
were thoroughly reviewed and updated, the reference doses were 
unchanged. The reference doses are the dose values which are used to 
define a ``not unacceptable'' dose in the event of an accident. 
Consequently, while some revised A1 and A2 values 
are higher and some are lower, the potential dose following an accident 
is the same as with the previous A1 and A2 
values. The general A value radiological criteria are: effective or 
committed effective dose to a person should not exceed 50 mSv (5 rem); 
the dose or committed dose received by individual organs should not 
exceed 0.5 Sv (50 rem) (see IAEA

[[Page 3721]]

TS-G-1.1 for further details on Q-system dosimetric models and 
assumptions). Changes in the A values do not change the reference dose 
values. The revised dosimetric models are used internationally to 
calculate doses from individual radionuclides, and these refinements in 
the pathway calculations resulted in various changes to the 
A1 and A2 values. In other words, where an 
A1 or A2 value has increased, the potential dose 
is still the same--the use of the revised dosimetric models just shows 
that a higher activity of that radionuclide is actually required to 
produce the same reference dose. Conversely, where an A1 or 
A2 value has decreased, the revised models show that less 
activity of that nuclide is needed to produce the reference dose.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter stated that the NRC should not reduce the 
numbers and types of material subject to shipping regulations. The 
commenter was concerned that the proposed rule would:
    (1) Exempt numerous radionuclide shipments from any regulation;
    (2) Increase worker exposure and the difficulty of enforcement;
    (3) Create an inconsistency with other Federal radionuclide 
standards; and
    (4) Otherwise reduce the protections afforded the public during 
radionuclide transportation.
    Another commenter stated that the revisions' rationale does not 
justify such weakening, that inconsistency with IAEA standards is an 
inadequate justification for the proposed changes because there has 
been no demonstration that inconsistencies have caused any difficulty.
    Finally, one commenter stated that increasing the A1 and 
A2 values should not be allowed and added that conforming 
with IAEA regulations is an insufficient justification to increase 
``levels of exposure to American citizens.'' Further, the commenter 
stated that avoiding ``negative impacts on the nuclear industry are not 
justifiable reasons for NRC to relax any standards for protection of 
the public.''
    Response. The NRC disagrees with the first commenter. The final 
rule does not exempt numerous radionuclide shipments, nor increase 
worker exposure, nor reduce protection to the public, nor create an 
inconsistency with other Federal standards.
    The NRC disagrees with the second commenter that the final rule 
weakens the regulations. Conforming NRC regulations to the IAEA 
regulations is not the sole justification; it is also adopting sound 
science, incorporating the latest in dosimetric modeling and that the 
changes improve the transportation regulations. The regulatory analysis 
indicates that adopting these values is appropriate from a safety, 
regulatory, and cost perspective.
    Comment. One commenter suggested that the NRC organize the 
A1 and A2 tables to be sorted alphabetically by 
name rather than symbol, because the people who will use these tables 
most frequently will be more familiar with the spelling of the name 
rather than the chemical symbol. In addition, using the full name will 
make the tables easier to use and will be more consistent with the June 
1, 1998, Presidential memo, ``Plain Language in Government Writing.''
    Response. The comment is acknowledged; however, the tables will 
remain sorted as proposed to maintain consistency with the current DOT 
and IAEA regulations.
    Comment. One commenter stated that the dose to workers could 
increase due to their need to handle more packages. The commenter also 
stated that the demand for molybdenum-99, the principal isotope used in 
medical imaging, would likely increase with the aging population.
    Response. The proposed A1 and A2 values 
should result in only a minimal change in occupational risk. The 
proposed A1 and A2 values are based on the same 
reference doses as the current values, and only the dosimetric models 
were revised, leading to the updated values. In general, the proposed 
A1 and A2 values are within a factor of about 
three of the current values; very few radionuclides have proposed 
A1 and A2 values that are outside this range.
    Currently in part 71, the A2 value for Mo-99 is 0.5 TBq 
(13.5 Ci) for international transport and 0.74 TBq (20 Ci) for domestic 
transport. The NRC originally proposed an A2 value of 0.6 
TBq (16.2 Ci) for Mo-99, but commenters suggested that adopting the 
lower A2 value for domestic use would only result in an 
increase in the number of packages shipped and, thus, in a potential 
increase in occupational dose. Therefore, NRC will retain the current 
Mo-99 A2 value of 0.74 TBq (20 Ci) for domestic shipments.
    Comment. One commenter indicated that the proposed A1 
and A2 values were ``far reaching.'' The commenter was 
concerned by the lack of data supporting these significant changes but 
generally supported the changes.
    Response. NRC does not believe that the proposed changes to the 
A1 and A2 values are ``far reaching.'' NRC does 
not believe there is a lack of data on the proposed changes to the 
A1 and A2 values. Instead, the information on the 
Q-system, the details of the exposure pathways, and the actual IAEA 
A1 and A2 values are contained in the guidance 
document for TS-R-1, TS-G 1.1, and Safety Series 7.
    The revisions of the A1 and A2 values are 
based on a reexamination/new assessment of the dosimetric models used 
in deriving the content limits for Type A packages. The overall impact 
of the reexamination resulted in improved methods for the evaluation of 
the content limits for special form (denoted by A1) and 
nonspecial form (denoted by A2) radioactive material. 
Internationally, as increased knowledge and scientific methods are 
gained and applied in the areas of health physics, radioactive material 
packaging, and radioactive material transportation, it is appropriate 
to take advantage of that knowledge and information and apply it to the 
IAEA regulations. This has occurred with the revision of the 
A1 and A2 values. The IAEA applied the newly-
revised Q-system to the same uptake scenarios it used for the 1985 
regulations. Thus, the same dose criteria, which were used in the 
assessment of the 1985 A1 and A2 values, were 
also used to determine the new A1 and A2 values 
in TS-R-1.
    While some of the A1 and A2 values have 
increased, some values remain unchanged, and some values decreased, the 
overall safety implications for TS-R-1 remain the same as those used in 
the 1985 IAEA regulations.
    Within the Q-system, a series of exposure routes are considered 
which may result in radiation exposure to persons near a Type A package 
of radioactive material that has been involved in an accident. The 
exposure routes include external photon dose, external beta dose, 
inhalation dose, skin and ingestion dose due to contamination transfer, 
and submersion (exposure to vapor/gas) dose.
    Comment. One commenter requested more explanation of the 
implications of revision of the A1 and A2 values. 
The commenter requested simple summaries for both special form and 
normal materials.
    Response. See response to the preceding comment. Special form 
radioactive material and normal form radioactive material are defined 
in Sec. 71.4. In general, special form radioactive material is 
subjected to various tests found in Sec. 71.75, ``Qualification of 
special form radioactive material.'' These materials

[[Page 3722]]

are known to be nondispersible (will not disperse contamination). Thus, 
in a transportation scenario, special form radioactive material could 
be considered relatively safer in transport by the fact that it poses 
only a direct radiation hazard (and not a contamination hazard). On the 
other hand, radioactive material that has not been tested to the 
requirements of Sec. 71.75 or has not passed these tests has not 
qualified to be considered special form radioactive material. Such 
material is called nonspecial form (commonly known as normal form) 
radioactive material. In general, these materials pose both a radiation 
and contamination hazard in that they are considered to be dispersible. 
As an example, consider the A1 and A2 values for 
actinium-227 (A1 = 9E-1 TBq (2.4E1 Ci); A2 = 9E-5 
TBq (2.4E-3 Ci)). Notice the tremendous difference between 
A1 and A2. This example demonstrates that in 
special form, a much larger amount of activity can be placed in a Type 
A package because the special form material has been sealed or 
encapsulated and has proven its robustness by passing the test 
requirements of Sec. 71.75. The same encapsulation and testing is not 
true for the nonspecial form (A2) value. This is where the 
applicability of health physics and metabolic uptake come into 
consideration for determining the A1 and A2 
values for each individual radionuclide.
    Comment. One commenter asked if the justification for the change is 
the shift in accepted dose models from ICRP 26 and 30 to 60 and 66. The 
commenter requested data supporting the shift in dose models.
    Response. The most recent recommendations of the ICRP were issued 
in 1991 (1990 Recommendation of the International Commission on 
Radiological Protection, Publication No. 60, Pergamon Press, 1991). 
Within TS-R-1, IAEA applied the values from ICRP 60 and 66, thus the 
shift in dose models. This data can be found in the ICRP 60 and 66 
documents.
    Comment. One commenter noted that ICRP and IAEA risk models only 
look at fatal cancers and ignore nonfatal cancers, years of lost life, 
and the bystander effect. The commenter asserted that the ICRP and IAEA 
reports do not accurately reflect risk and that low levels of radiation 
are more damaging than the models are predicting.
    Response. The NRC acknowledges this comment but notes that a 
response to similar concerns expressed is provided in the first comment 
of section II--Analysis of Public Comments, under the heading: Adequacy 
of NRC Regulations and Rulemaking Process.
    Comment. One commenter asked if these revisions would actually 
expand the number of containers that have to meet test standards.
    Response. Within part 71, NRC approves packages and shipping 
procedures for fissile radioactive materials and for licensed materials 
in quantities that exceed A1 or A2. NRC will 
continue to apply the regulations in part 71 to Type B and fissile 
radioactive material packages. NRC is not aware of an expansion of the 
container inventory which will have to meet test standards due to an 
increase in any individual A1 or A2 value.
    Comment. One commenter said that the scientific basis for the 
changes to the A1 and A2 values is understood and 
justified. However, the commenter urged NRC to maintain the exception 
(found in Table A-1 of Appendix A to part 71) to allow the domestic 
A2 limit of 20 Ci for Mo-99, which, the commenter states, is 
necessary to allow domestic manufacturers to continue to provide Mo-99 
generators to the diagnostic nuclear medicine community. The commenter 
said that changing the A2 limit to the TS-R-1 value would 
result in an increase in the number of packages shipped and, thus, an 
increase in the doses received by manufacturers, carriers, and end 
users.
    Response. NRC agrees with this commenter concerning the revision to 
the A1 and A2 values and the scientific 
background used to support the changes. Further, the commenter has 
indicated that the TS-R-1 A2 value for molybdenum-99 would 
increase the number of packages shipped and, thus, an increase the 
radiation exposure to various workers. Accordingly, to reduce these 
concerns NRC will retain the current A2 value for 
molybdenum-99 (7.4E-1 TBq; 2.0E1 Ci) as stated in the proposed rule and 
as found in Table A-1 for domestic transport. NRC is aware that by 
adopting this value (as opposed to the current value for molybdenum-99 
in TS-R-1), the number of shipments of molybdenum-99 and the associated 
radiation exposure may be reduced.
    Comment. One commenter indicated that revising the A1 
and A2 values might have an adverse impact on currently 
certified casks. The commenter stated that the proposed regulation does 
not ensure that transport casks certified under previous revisions will 
still be usable without modification or analysis in the future.
    Response. Although NRC staff could revise cask certificates if 
necessary, no changes are known to be needed to accommodate the revised 
A1 and A2 values.
    Comment. One commenter stated that because DOE is the principal 
shipper of californium-252 under the current exemption value, the 
potential impacts to industry could not be assessed.
    Response. NRC is aware of the limited and safe transportation of 
californium-252 by DOE.
    Comment. One commenter stated that by omitting the A1 
and A2 values for 16 radionuclides, the Commission would 
have to set these values upon future request of a licensee. The 
commenter recommended that the NRC not delete these values from part 
71, Appendix A, to save NRC the cost and resources necessary to 
establish these values in the future.
    Response. NRC agrees that more time and effort may be needed to 
reintroduce these 16 radionuclides into Appendix A at some time in the 
future, as compared to retaining their names and symbols but not 
publishing actual A1 and A2 values for them. 
Instead, the reference to the general values for A1 and 
A2 provided in Table A-3 would be used without NRC approval 
for shipping these radionuclides. Further, to maintain consistency/
harmonization with future IAEA transport standards, NRC may adopt a 
revised list of A1 and A2 values, should there be 
revisions to Table 1 in future editions of the IAEA transport 
standards.
    Comment. Four commenters agreed with NRC's efforts to revise 
A1 and A2 values.
    Response. The NRC acknowledges these comments.
    Comment. Several commenters disagreed with the NRC staff's 
position. One commenter opposed weakening the present standard of 
radiation protection during transportation, particularly because NRC is 
proposing to ship radioactive wastes to a repository. Another commenter 
expressed concern that many, if not most, of the A1 and 
A2 values, both current and proposed in the NRC's part 71 
regulations, appear to have been arbitrarily chosen and are unsafe. 
Another commenter stated that any additional costs ``must be borne by 
licensees and beneficiaries of use of materials.'' Another commenter 
asked the NRC not to adopt the exemption values contained in Table 2 of 
TS-R-1.
    Response. NRC does not consider the adoption of the A1 
and A2 values from TS-R-1 to be a weakening of the present 
standards for packaging and transporting radioactive material. The NRC 
believes the revision of the A1 and A2 values to 
be based on sound science and that it provides adequate protection to 
the public and workers. Furthermore,

[[Page 3723]]

there is not a direct connection between adopting the revised 
A1 and A2 values into part 71 and the package 
standards and safety requirements which will be imposed on the 
transport packages for high-level waste en route to a geologic 
repository.
    The process used to determine the appropriate A1 and 
A2 value assigned to each radionuclide is based on several 
factors. These include the type of radiation emitted by the 
radionuclide e.g., alpha, beta, or gamma), the energy of that radiation 
i.e., strong alpha emitter, strong gamma emitter, weak beta emitter, 
etc.), and the form of the material (nondispersible as applied to 
special form radioactive material, or dispersible as applied to 
nonspecial form radioactive material). All of these factors have been 
modeled in the IAEA's Q-system to determine the appropriate value to be 
assigned to each radionuclide. Thus, the values have not been 
arbitrarily obtained, and they are safe. Further, the revision to the 
A1 and A2 values in TS-R-1 has maintained the 
same level of safety as was applied in determining the A1 
and A2 values for the radionuclides in the 1985 IAEA 
transportation standards. Thus, there is no weakening of the intended 
safety aspects of the new A1 and A2 values.
    Comment. Several commenters noted various typographical errors. The 
first commenter noted that Footnote 2 to Table A-1 is incorrect and 
should instead read, ``See Table A-4.'' The second commenter noted an 
error in the proposed Table A-1 for the A2 (Ci) value for 
Pu-239, suggesting that the correct value should be 2.7 x 
10-2 Ci, as evidenced from the A2 (TBq) value for 
Pu-239 and the similar Table 1 in the IAEA TS-R-1 regulations and Table 
10A in the proposed DOT regulations.
    Response. NRC acknowledges the comment, and corrections have been 
made to the final rule.
    Comment. One commenter addressed changing a number of the 
radionuclide values. The commenter suggested that the radionuclide Al-
26 value for specific activity in 10 CFR part 71, Table A-1, should be 
changed from 190 Ci/g to 0.019 Ci/g. The A1 and 
A2 values in both 10 CFR part 71 Table A-1 and 49 CFR 
173.435 for Ar-39 appear reversed from that listed in IAEA TS-R-1. The 
radionuclide Be-10 value for specific activity in 10 CFR part 71 Table 
A-1 should be changed from 220 Ci/g to 0.022 Ci/g. The radionuclide Cs-
136 value for specific activity in 49 CFR 173.435 should be changed 
from 0.0027 TBq/g to 270 TBq/g. The radionuclide Dy-165 value for 
A2 (Ci) in 10 CFR part 71 Table A-1 should be changed from 
0.16 to 16 Ci. The radionuclide Eu-150 (long-lived) value for 
A1 (TBq) in 10 CFR part 71 Table A-1 and 49 CFR 173.435 is 
not consistent with the IAEA TS-R-1 value of 0.7. The radionuclide Fe-
59 value for A2 (TBq) in 10 CFR part 71 Table A-1 is in 
error. The radionuclide Ho-166m value for A2 (TBq) in 10 CFR 
part 71 Table A-1 should be 0.5. The radionuclide K-43 value for 
A2 (TBq) in 10 CFR part 71 Table A-1 should be 0.6. The 
radionuclide Kr-81 value for A1 (TBq) in 49 CFR 173.435 
should be 40, A1 (Ci) in 49 CFR 173.435 should be 1100. The 
radionuclide Kr-85 value for A2 (TBq) in 49 CFR 173.435 
should be 10; A2 (Ci) in 49 CFR 173.435 should be 270. The 
radionuclide La-140 value for A2 (Ci) in 49 CFR 173.435 
should be 11. The radionuclide Lu-177 value for A2 (TBq) in 
49 CFR 173.435 should be 0.7; A2 (Ci) in 49 CFR 173.435 
should be 19. The radionuclide Mn-52 value for specific activity (Ci) 
in 49 CFR 173.435 should be 4.4E+05. The radionuclide Np-236 (long-
lived) value for A1 (TBq) in IAEA TS-R-1 is 9; A2 
(TBq) in IAEA TS-R-1 is 0.02, different from the values in both 49 CFR 
173.435 and 10 CFR part 71, Table A-1. The radionuclide Pt-197m value 
for A2 (TBq) in 49 CFR 173.435 should be 0.6; A2 
(Ci) in 49 CFR 173.435 should be 16. The radionuclide Pu-239 value for 
A2 (Ci) in 10 CFR part 71, Table A-1, should be 0.027. The 
radionuclide Pu-240 value for specific activity (Ci) should be 0.23 Ci/
g. The radionuclide Ra-225 value for A2 (Ci) in 10 CFR part 
71, Table A-1, should be 0.11. The radionuclide Ra-228 value for 
A2 (TBq) in 10 CFR part 71, Table A-1, should be 0.02. The 
radionuclide Rh-105 value for A2 (Ci) in 10 CFR part 71, 
Table A-1, is in error. The radionuclide Sc-46 value for A1 
(TBq) in 10 CFR part 71, Table A-1, should be 0.5. The radionuclide Sn-
119m value for A2 (TBq) in 10 CFR part 71, Table A-1, should 
be 30. The radionuclide Sn-126 value for specific activity (TBq) in 10 
CFR part 71, Table A-1, should be 0.001. The radionuclide H-3 value for 
A2 (TBq) in 10 CFR part 71, Table A-1, should be 40. The 
radionuclide Ta-179 value for A1 (TBq) in 10 CFR part 71, 
Table A-1, should be 30. The radionuclide Tb-157 value for 
A1 (TBq) in 10 CFR part 71, Table A-1, should be 40; value 
for specific activity (TBq) in 10 CFR part 71, Table A-1, should be 
0.56 TBq/g. The radionuclide Tb-158 value for A2 (Ci) in 10 
CFR part 71, Table A-1, should be 27; value for specific activity (TBq) 
in 10 CFR part 71, Table A-1, should be 0.56 TBq/g.
    The radionuclide Tb-160 value for A1 (Ci) in 10 CFR part 
71, Table A-1, should be 27. The radionuclide Tc-96 value for 
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.4. The 
radionuclide Tb-96m value for A1 (TBq) in 10 CFR part 71, 
Table A-1, should be 0.4; value for A2 (TBq) in 10 CFR part 
71, Table A-1, should be 0.4. The radionuclide Tc-97 value for specific 
activity (TBq) in 10 CFR part 71, Table A-1, should be 5.2E-05; value 
for specific activity in 10 CFR part 71, Table A-1, should be 0.0014. 
The radionuclide Te-125m value for A2 (Ci) in 10 CFR part 
71, Table A-1, should be 24. The radionuclide Te-129 value for 
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.7; value 
for A2 (TBq) in 10 CFR part 71, Table A-1, should be 0.6. 
The radionuclide Te-132 value for A1 (TBq) in 10 CFR part 
71, Table A-1, should be 0.5. The radionuclide Th-227 value for 
A2 (Ci) in 10 CFR part 71, Table A-1, should be 0.14. The 
radionuclide Th-231 value for A2 (TBq) in 10 CFR part 71, 
Table A-1, should be 0.02. The radionuclide Th-234 value for 
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.3. The 
radionuclide Ti-44 value for A1 (TBq) in 10 CFR part 71, 
Table A-1, should be 0.5; value for A2 (TBq) in 10 CFR part 
71, Table A-1, should be 0.4, value for A2 (Ci) in 10 CFR 
part 71, Table A-1, should be 10. The radionuclide Tl-200 value for 
A1 (TBq) in 10 CFR part 71, Table A-1, should be 0.9. The 
radionuclide Tl-204 value for A2 (TBq) in 10 CFR part 71, 
Table A-1, should be 0.7. The radionuclide U-230, U-232, U-233, and U-
234 values for medium and slow lung absorption, and U-236 values for 
slow lung absorption are not consistent with IAEA TS-R-1. The comment 
points out that the Table values published in the Federal Register for 
the proposed rule did not match TS-R-1.
    Response. NRC accepts the comment and has updated the values in the 
final rule, Table A-1, to be consistent with TS-R-1. Appropriate 
changes have been made in the final rule.
    Comment. Three commenters stated that the A2 value for 
molybdenum-99 and the A1 and A2 values for 
californium-252 should be retained for domestic use only packages.
    Response. NRC agrees with the comment. (See 67 FR 21399; April 30, 
2002, for more details.)
Issue 4. Uranium Hexafluoride (UF6) Package Requirements
    Summary of NRC Final Rule. The final rule provides, in new Sec. 
71.55(g), a specific exception for certain uranium hexafluoride 
(UF6) packages from the requirements of Sec. 71.55(b). The 
exception allows UF6 packages to be evaluated for 
criticality safety without considering the in leakage of water into

[[Page 3724]]

the containment system provided certain conditions are met, including 
that the uranium is enriched to not more than 5 weight percent uranium-
235. The rule makes part 71 compatible with TS-R-1, paragraph 677(b). 
Other uranium hexafluoride package requirements in TS-R-1 (paragraphs 
629, 630 and 631) do not necessitate changes for compatibility because 
NRC uses analogous national standards and addresses package design 
requirements in its design review process.
    The specific exception being placed into the regulations for the 
criticality safety evaluation of certain uranium hexaflouride packages 
does not alter present practice which has allowed the same type of 
evaluation under other more general regulatory provisions. NRC has 
decided to provide this specific exception: (1) To be consistent with 
the worldwide practice and limits established in national and 
international standards (ANSI N14.1 and IS 7195) and current U.S. 
regulations (49 CFR 173.417(b)(5)); (2) because of the history of safe 
shipment; and (3) because of the essential need to transport the 
commodity.
    Affected Sections. Section 71.55.
    Background. Requirements for UF6 packaging and 
transportation are found in both NRC and DOT regulations. The DOT 
regulations contain requirements that govern many aspects of 
UF6 packaging and shipment preparation, including a 
requirement that the UF6 material be packaged in cylinders 
that meet the ANSI N14.1 standard. NRC regulations address fissile 
materials and Type B packaging designs for all materials.
    TS-R-1 contains detailed requirements for UF6 packages 
designed for transport of more than 0.1 kilogram (kg) UF6. 
First, TS-R-1 requires the use of the International Organization for 
Standardization (ISO) 7195, ``Packaging of Uranium Hexafluoride for 
Transport.'' Second, TS-R-1 requires that all packages containing more 
than 0.1 kg UF6 must meet the ``normal conditions of 
transport'' drop test, a minimum internal pressure test, and the 
hypothetical accident condition thermal test (para 630). However, TS-R-
1 does allow a competent national authority to waive certain design 
requirements, including the thermal test for packages designed to 
contain greater than 9,000 kg UF6, provided that 
multilateral approval is obtained. Third, TS-R-1 prohibits 
UF6 packages from using pressure relief devices (para 631). 
Fourth, TS-R-1 includes a new exception for UF6 packages 
regarding the evaluation of criticality safety of a single package. 
This new exception (para 677(b)) allows UF6 packages to be 
evaluated for criticality safety without considering the in leakage of 
water into the containment system. Consequently, a single fissile 
UF6 package does not have to be subcritical assuming that 
water leaks into the containment system. This provision only applies 
when there is no contact between the valve body and the cylinder body 
under accident tests, and the valve remains leak-tight, and when there 
are quality controls in the manufacture, maintenance, and repair of 
packages coupled with tests to demonstrate closure of each package 
before each shipment.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC responses for this issue 
follows:
    Comment. Five commenters expressed support for the proposed changes 
to UF6 package rules that continue the current practice of 
moderator exclusion for UF6. One commenter cited the strong 
safety record applying these rules as evidence that the practice is 
adequate. Two commenters objected to the 5 percent enrichment limit 
provision in proposed Sec. 71.55(g), and a third commenter expressed 
concern with the enrichment limit. One commenter noted that the safety 
case for the specific enrichment to use can be a part of the package 
certification application and, therefore, does not need to be specified 
by rule. The same commenter further noted that arguments that water in 
leakage is not a realistic scenario for a UF6 cylinder 
regardless of enrichment and that the 5 percent limit, if imposed for 
transportation, could have very high cost implications in light of 
pending decisions to use higher enrichments in the fuel cycle. One 
commenter suggested that the rule retain the limit of 5 percent for the 
existing ANSI N14.1 Model 30B cylinder, but that the rule also contain 
provisions that permit greater than 5 percent enrichments in an 
``improved UF6 package with special design features'' to 
accommodate future industry plans.
    Response. The NRC's decision to exempt uranium hexafluoride 
cylinders from Sec. 71.55(b) with a limiting condition of 5 weight 
percent enriched uranium was made based on:
    (1) Consistency with the worldwide practice and limits established 
in national and international standards (ANSI N14.1 and IS 7195) and 
current U.S. regulations (49 CFR 173.417(b)(5));
    (2) The history of safe shipment; and
    (3) The essential need to transport the commodity.
    The NRC staff believes that further expansion of the practice of 
authorizing shipment of materials in packages that do not meet Sec. 
71.55(b), without a strong technical safety basis and without full 
understanding of the potential reduction in safety margins, is not 
prudent or necessary at this time. In addition, provisions are 
available to request approval of alternative package designs that could 
be used for the shipment of uranium hexafluoride with uranium 
enrichments greater than 5 weight percent under the provisions of Sec. 
71.55(b) or Sec. 71.55(c). Merits of a new or modified design that 
included special design features could be reviewed and approved under 
the provisions of Sec. 71.55, including Sec. 71.55(c).
    Because package certification is directly tied to the regulations, 
any assessment of the safety of enrichments greater than 5 weight 
percent uranium-235, considering the potential or probability of water 
in leakage, would not be part of the safety case of an application if 
the enrichment limit is not included as part of the regulation.
    Although it is correct that the water in leakage scenario is not 
changed for enrichments less than or greater than 5 weight percent, it 
is not clear that the safety margins against accidental nuclear 
criticality for all enrichments would be the same if water were 
introduced into the containment vessel accidentally. Because these 
margins are undefined at this time, it does not seem prudent or 
necessary to modify the regulatory standard that was based on worldwide 
practice in existence today. Future changes in the fuel cycle that 
could necessitate transport of enrichments greater than 5 weight 
percent uranium-235 could result in new packages designed to meet the 
normal fissile material package standards in Sec. 71.55(b), as are 
required for other commodities, or could include special design 
features that would enhance nuclear criticality safety for transport 
for approval under the provisions of Sec. 71.55(c). Alternatively, a 
safety assessment could be developed for possible transport of 
enrichments greater than 5 weight percent to support some future 
rulemaking to modify Sec. 71.55(g) to increase the enrichment 
limitation.
    For the previously mentioned reasons, the NRC staff has retained 
the 5 percent enrichment limit in the final rule.
    Comment. One commenter stated an opinion that all UF6 
packages should have overpacks and noted that the proposed rule should 
resolve this issue.
    Response. The NRC staff does not agree with the position that all 
UF6 packages be required by rule to

[[Page 3725]]

incorporate an overpack. Design and performance standards for fissile 
UF6 packages are stated in part 71, and design and 
performance standards for nonfissile UF6 packages appear in 
DOT regulations. Use of specific design features (e.g., overpacks) to 
meet regulatory standards is left to designers.
    Comment. One commenter expressed concern that NRC had not provided 
data to back up its proposal to ``relax the current packaging 
requirements'' in Sec. 71.55(b) for UF6. The commenter 
stated that NRC should not adopt this proposal unless it can provide 
justification for doing so. The commenter was also concerned that NRC's 
EA does not address any impacts associated with this proposal.
    Response. The NRC staff disagrees with the commenter's assertion 
that adoption of Sec. 71.55(g) is a relaxation of current packaging 
requirements in Sec. 71.55(b). As noted by the commenter, NRC's 
proposed rule (67 FR 21400) explains that the new Sec. 71.55(g) 
provisions are consistent with existing worldwide practice for UF6 
packages. This worldwide practice has been in use since its development 
in the 1950s, and the functioning of the nuclear fuel cycle in the U.S. 
relies upon transport of this commodity. The exception was limited to 5 
weight percent enriched uranium consistent with the worldwide practice 
and limits established in national and international standards (ANSI 
N14.1 and IS 7195) and current U.S. regulations (49 CFR 173.417(b)(5)). 
The new regulatory text replaces the more general ``special features'' 
allowances with a more explicit provision pertaining to certain 
UF6 packages.
    Comment. Two commenters expressed opposition for the relaxation of 
testing for radioactive transport containers. One commenter stated that 
the drop test, minimum internal pressure test, and the hypothetical 
accident condition test must be accompanied by the thermal test to 
assure public protection in the event of an accident. One commenter 
cited both the Baltimore tunnel fire and the Arkansas bridge incident 
as justifications for not allowing any exemptions.
    Response. The NRC staff reviewed these comments and determined that 
they concern the nonfissile UF6 packaging issues discussed 
in Issue 6 in the DOT's proposed rulemaking (April 30, 2002; 67 FR 
21337), not the fissile UF6 package matters in Issue 4 in 
the related NRC proposed rulemaking. The NRC staff noted that the 
commenter's letter was jointly addressed to NRC and DOT for resolution 
in their final rule.
Issue 5. Introduction of the Criticality Safety Index Requirements
    Summary of NRC Final Rule. The final rule adopts the TS-R-1 
(paragraphs 218 and 530). Paragraph 218 results in NRC incorporating a 
Criticality Safety Index (CSI) in part 71 that is determined in the 
same manner as current part 71 ``Transport Index for criticality 
control purposes,'' but now it must be displayed on shipments of 
fissile material (paragraphs 544-545) using a new ``fissile material'' 
label. NRC's adoption of TS-R-1 (paragraph 530) increases the CSI-per 
package limit from 10 to 50 for fissile material packages in 
nonexclusive use shipments. (The previous Transport Index criticality 
limit was 10.) The TI is determined in the same way as the ``TI for 
radiation control purposes'' and continues to be displayed on the 
traditional ``radioactive material'' label. The basis for these changes 
that makes part 71 compatible with TS-R-1 is that NRC believes the 
differentiation between criticality control and radiation protection 
would better define the hazards associated with a given package and, 
therefore, provide better package hazard information to emergency 
responders. The increase in the per package CSI limit may provide 
additional flexibility to licensees by permitting the increased use of 
less expensive, nonexclusive use shipments. However, licensees will 
still retain the flexibility to ship a larger number of packages of 
fissile material on an exclusive use conveyance. The adoption of the 
CSI values would make part 71 consistent with TS-R-1 and, therefore, 
would enhance regulatory efficiency.
    Affected Sections. Sections 71.4, 71.18, 71.20, 71.59.
    Background. Historically, the IAEA and U.S. regulations (both NRC 
and DOT) have used a term known as the Transport Index (TI) to 
determine appropriate safety requirements during transport. The TI has 
been used to control the accumulation of packages for both radiological 
safety and criticality safety purposes and to specify minimum 
separation distances from persons (radiological safety). The TI has 
been a single number which is the larger of two values: the ``TI for 
criticality control purposes''; and the ``TI for radiation control 
purposes.'' Taking the larger of the two values has ensured 
conservatism in limiting the accumulation of packages in conveyances 
and in-transit storage areas.
    TS-R-1 (paragraph 218) has introduced the concept of a CSI separate 
from the old TI. As a result, the TI was redefined in TS-R-1. The CSI 
is determined in the same way as the ``TI for criticality control 
purposes,'' but now it must be displayed on shipments of fissile 
material (paragraphs 544 and 545) using a new ``fissile material'' 
label. The redefined TI is determined in the same way as the ``TI for 
radiation control purposes'' and continues to be displayed on the 
traditional ``radioactive material'' label.
    TS-R-1 (paragraph 530) also increased the allowable per package TI 
limit (for criticality control purposes (new CSI)) from 10 to 50 for 
nonexclusive use shipments. No change was made to the per package 
radiation TI limit of 10 for nonexclusive use shipments. As noted 
above, a consolidated radiation safety and CSI existed in the past. In 
this consolidated index, the per package TI limit of 10 was 
historically based on concerns regarding the fogging of photographic 
film in transit, because film might also be present on a nonexclusive 
use conveyance. Consequently, when the single radiation and criticality 
safety indexes were split into the TI and CSI indexes, the IAEA 
determined that the CSI per package limit, for fissile material 
packages that are shipped on a nonexclusive use conveyance, could be 
raised from 10 to 50. The IAEA believed that limiting the total CSI to 
less than or equal to 50 in a nonexclusive use shipment provided 
sufficient safety margin, whether the shipment contains a single 
package or multiple packages. Therefore, the per package CSI limit, for 
nonexclusive use shipments, can be safely raised from 10 to 50, thereby 
providing additional flexibility to shippers. Additionally, no change 
was made to the per package CSI limit of 100 for exclusive use 
shipments.
    The NRC believes the differentiation between criticality control 
and radiation protection would better define the hazards associated 
with a given package and, therefore, provide better package hazard 
information to emergency responders. The increase in the per package 
CSI limit may provide additional flexibility to licensees by permitting 
the increased use of less expensive, nonexclusive use shipments. 
However, licensees will still retain the flexibility to ship a larger 
number of packages of fissile material on an exclusive use conveyance.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment 1. One commenter requested a basic explanation of the CSI

[[Page 3726]]

and TI. The commenter questioned if the proposed changes would increase 
public risk. Another commenter asked for clarification on how NRC would 
calculate CSI for radiological shipments to ensure that a shipment is 
under limits.
    Response. The requested explanation was provided during the June 4, 
2001, public meeting at which the first comment was made (see NRC 
rulemaking interactive Web site at http://ruleforum.llnl.gov. In 
addition, the proposed rule contains background on the CSI; regarding 
increased public risk. The draft RA concluded the change is appropriate 
from a safety perspective. Also, see Background discussion for this 
issue.
    Comment. One commenter expressed opposition to the text that would 
restrict accumulations of fissile material to a total CSI of 50 in 
situations where radioactive materials are stored incident to 
transport. The commenter added that this would effectively remove the 
ability to transport internationally and/or by multiple modes under 
exclusive use conditions and would negatively impact the international 
movement of fissile materials under nonproliferation programs. The 
commenter further noted that this provision would apply only to 
shipments to or from the U.S., thus creating a disadvantage for 
American businesses in the international market.
    Response. The NRC agrees with these comments. The intent of the 
storage phrase was to permit segregation of groups of stored packages, 
consistent with IAEA and DOT requirements, but the NRC staff believes 
that the proposed text did not accommodate that practice. DOT 
requirements restrict accumulation of packages during transport, based 
on summing the packages' CSI or TI, including during storage incident 
to transport. In light of the division of regulatory responsibilities 
explained in the NRC-DOT Memorandum of Understanding (44 FR 38690; July 
2, 1979), the NRC exemptions for carriers-in-transit in 10 CFR 70.12, 
and DOT's proposed 49 CFR 173.457 (67 FR 21384; April 30, 2002), the 
NRC staff believes that storage in transit provisions proposed in 
Sec.Sec. 71.59(c)(1), 71.22(d)(3), and 71.23(d)(3) are unwarranted. The 
NRC has deleted the phrase ``or stored incident to transport'' from 
these sections.
    Comment. One commenter stated that in proposed Sec.Sec. 71.59( 
c)(1), (2) and (3), and 71.55(f)(3), the values of 50.0 and 100.0 
should be changed to 50 and 100 to be consistent with the application 
of the CSI.
    Response. The NRC staff did not intend nor does it believe that 
there is a substantive difference between ``50'' and ``50.0'' as used 
in part 71. In proposing to use the decimal place, the NRC staff was 
attempting to increase precision when the CSI is exactly 50.0 and 
promote consistency as the CSI is by definition rounded to the nearest 
tenth. However, the NRC staff noted that both DOT's proposed rule and 
IAEA TS-R-1 use ``50'' without a decimal place. The NRC staff agrees 
that consistency amongst the three rules is desirable unless a reason 
exists for differentiating. Accordingly, conforming changes have been 
made to the part 71 final rule.
    Comment. One commenter expressed opposition to the rounding of the 
CSI provision in the proposed rule, because it is inconsistent with TS-
R-1 and places additional limits on the array size of shipments.
    Response. The commenter correctly observes that Sec. 71.59(b) 
requires all nonzero CSIs to be rounded up to the first decimal place 
and that the corresponding TS-R-1 requirement (paragraph 528) does not 
require such rounding. Rounding up the CSI is necessary to ensure that 
an unanalyzed number of packages are not transported together; rounding 
a CSI down would permit such situations. The NRC staff notes that this 
U.S. provision predates the currently contemplated changes for 
compatibility with TS-R-1 (viz., the existing U.S. domestic regulations 
are also different than the 1985 IAEA transport regulations in this 
respect).
    Consistent with the NRC proposal, the IAEA's implementing guidance 
for TS-R-1 (i.e., TS-G-1.1 at para. 528.3) states, ``The CSI for a 
package * * * should be rounded up to the first decimal place'' and 
``the CSI should not be rounded down.'' The NRC staff noted that the 
IAEA's guidance, however, does observe that use of the exact CSI value 
may be appropriate in cases when rounding results in less than the 
analyzed number of packages to be shipped.
    The NRC staff believes that the rule is compatible with IAEA TS-R-
1. Furthermore, because the domestic convention on rounding predates 
this rulemaking for compatibility with 1996 TS-R-1, and because the 
statements of consideration did not explicitly discuss the rounding 
practice, the potential elimination of the rounding practice is beyond 
the scope of the current rulemaking action.
    Comment. Three commenters expressed agreement with NRC's proposed 
position. One of the three commenters expressed support for the NRC's 
CSI proposal, reasoning that it provides more accurate communication 
regarding radioactive material in transport, especially in conjunction 
with the TI for radiation exposure. The commenter noted that the CSI is 
important to ensure consistency between domestic and international 
movements of fissile material. Another commenter stated that use of the 
CSI would ``remove a source of confusion with the old TI values. The 
resulting enhancement of the safety of shipments makes the extra 
efforts necessary to implement these proposals worthwhile.''
    Response. No response is necessary.
    Comment. One commenter stated that the CSI ``should be set so as to 
maximize protective benefit for workers and the public without regard 
for added costs to licensees and users.'' The commenter added that 
there doesn't seem to be a ``strong argument against adoption'' of the 
IAEA CSI but then stated that the increase from 10 to 50 per package 
does not have adequate justification. Further, the commenter stated 
that if cost reduction for licensees is the only reason for this 
change, then the proposal is unacceptable.
    Response. The CSI is derived to prevent nuclear criticality for 
single packages and arrays of packages, both in incident-free and 
accident conditions of transport. Therefore, the NRC staff has 
determined that the application of the CSI does support protection of 
workers and the public. The basis for increasing the accumulation of 
packages from 10 TI under the old system to 50 CSI in the new system is 
given in the proposed rule (at 67 FR 21401), and it is not a solely 
economic basis. Specifically, the limit of 10 TI was based on radiation 
damage to film, so when the TI and CSI were split in 1996, a separate 
limit on package accumulation based on criticality prevention, of 50 
CSI, became warranted.
Issue 6. Type C Packages and Low Dispersible Material
    Summary of NRC Final Rule. The final rule does not adopt the Type C 
or Low dispersible material (LDM) requirements for plutonium air 
transport as introduced in the IAEA TS-R-1. NRC decided not to adopt 
Type C or LDM requirements because the U.S. regulations in Sec.Sec. 
71.64 and 71.71 governing plutonium air transportation to, within, or 
over the United States contains more rigorous packaging standards than 
those in the IAEA TS-R-1. Furthermore, the NRC's perception is that 
there is a lack of current or anticipated need for such packages, and 
NRC acknowledges that the DOT import/export provisions permit use of 
IAEA regulations.

[[Page 3727]]

    Affected Sections. None (not adopted).
    Background. TS-R-1 introduced two new concepts: the Type C package 
(paragraphs 230, 667-670, 730, 734-737) and the LDM. The Type C 
packages are designed to withstand severe accident conditions in air 
transport without loss of containment or significant increase in 
external radiation levels. The LDM has limited radiation hazard and low 
dispersibility; as such, it could continue to be transported by 
aircraft in Type B packages (i.e., LDM is excepted from the TS-R-1 Type 
C package requirements). United States regulations do not contain a 
Type C package or LDM category but do have specific requirements for 
the air transport of plutonium (Sec.Sec. 71.64 and 71.74). These 
specific NRC requirements for air transport of plutonium would continue 
to apply.
    The Type C requirements apply to all radionuclides packaged for air 
transport that contain a total activity value above 3,000 A1 
or 100,000 A2, whichever is less, for special form material, 
or above 3,000 A2 for all other radioactive material. Below 
these thresholds, Type B packages would be permitted to be used in air 
transport. The Type C package performance requirements are 
significantly more stringent than those for Type B packages. For 
example, a 90-meter per second (m/s) impact test is required instead of 
the 9-meter drop test. A 60-minute fire test is required instead of the 
30-minute requirement for Type B packages. There are other additional 
tests, such as a puncture/tearing test, imposed for Type C packages. 
These stringent tests are expected to result in package designs that 
would survive more severe aircraft accidents than Type B package 
designs.
    The LDM specification was added in TS-R-1 to account for 
radioactive materials (package contents) that have inherently limited 
dispersibility, solubility, and external radiation levels. The test 
requirements for LDM to demonstrate limited dispersibility and 
leachability are a subset of the Type C package requirements (90-m/s 
impact and 60-minute thermal test) with an added solubility test, and 
must be performed on the material without packaging for nonplutonium 
materials. The LDM must also have an external radiation level below 10 
mSv/hr (1 rem/hr) at 3 meters. Specific acceptance criteria are 
established for evaluating the performance of the material during and 
after the tests (less than 100 A2 in gaseous or particulate 
form of less than 100-micrometer aerodynamic equivalent diameter and 
less than 100 A2 in solution). These stringent performance 
and acceptance requirements are intended to ensure that these materials 
can continue to be transported safely in Type B packages aboard 
aircraft.
    In 1996, the NRC communicated to the IAEA that the NRC did not 
oppose the IAEA adoption of the newly created Type C packaging 
standards (letter dated May 31, 1996, from James M. Taylor, EDO, NRC, 
to A. Bishop, President, Atomic Energy Control Board, Ottawa, Canada). 
However, Mr. Taylor stated in the letter that to be consistent with 
U.S. law, any plutonium air transport to, within, or over the U.S. will 
be subject to the more rigorous U.S. packaging standards. Industry 
needs to be aware of changes or potential changes based on new IAEA 
standards.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Four commenters expressed support for NRC's proposal to 
not adopt the requirements for Type C packages and LDM. One commenter 
also expressed support for the NRC's decision to ensure that there is a 
mechanism for reviewing validations of foreign approvals. One commenter 
stated that the IAEA specification is too broad and that NRC and DOT 
should work with IAEA to reduce the scope to a few packages containing 
fissile oxides of plutonium, but there is no need for this package to 
transport Class 7 materials.
    Two commenters stated that the benefits did not justify the costs 
of the proposed changes and strongly supported the NRC position not to 
adopt the Type C requirements. One commenter stated that many parties 
are asking IAEA to modify the Type C requirements. The commenter urged 
NRC to see how these change proposals will affect the Type C 
requirements before adopting them into the U.S. regulations. 
Additionally, the commenter stated that the need for Type C packages 
for all radioactive material has not been demonstrated.
    Response. The NRC staff acknowledges these comments that endorse 
the position to not adopt Type C package requirements at this time, for 
the reasons specified in the proposed rule (67 FR 21402). The NRC staff 
agrees that Type C issues will likely receive further consideration in 
future IAEA rule cycles. No further response is necessary.
    Comment. Two commenters stated that the threat of terrorism should 
be taken into account when exempting radionuclides from transport 
regulations and changing container regulations. One commenter stated 
that the fact of the September 11, 2001, attacks needs to be accounted 
for with upgraded Types B and C testing, which are currently believed 
to be insufficient. The commenter added that these tests should 
``assure the highest probability that packages will survive 
unbreached.''
    Response. The NRC acknowledges the concern expressed regarding the 
threat of terrorism. However, the NRC does not propose adopting Type C 
and LDM requirements at this time. The NRC staff notes that the IAEA is 
conducting further evaluations on Type C package requirements, which 
may result in other changes for safety and security purposes. Also, see 
Section II, above, for general comments on terrorism.
    Comment. One commenter asked if workers will be protected and 
notified when handling Type C packages and plutonium, and whether they 
will be notified that there will be increased hazards once the proposed 
rule is effective.
    Response. The requested information on worker protection was 
provided at the public meeting at which the comment was made. 
Application of DOT's regulations, including hazardous materials 
training requirements, package radiation limits, and contamination 
limits, will protect workers for Type C packages just as for other 
shipments. In addition, the robustness of the packaging would provide 
protection in accidents. Thus, changes to the probability or 
consequences of releases in accidents do not result from proposed 
changes to Type C packages. The NRC does not propose adopting IAEA Type 
C or LDM standards at this time, and domestic regulations were not 
revised.
    Comment. One commenter recommended that the NRC ``adopt these 
provisions in order to better the goal of compatibility with IAEA 
regulations.'' This commenter continued by stating that ``industry 
would then have a basis for developing such a package if desirable.''
    Response. These comments recommend adoption of Type C standards in 
the interest of the goal of IAEA compatibility and speculate that a 
domestic Type C package regulation and certification might be desirable 
in the future. The NRC staff does not believe that deferring domestic 
rules on Type C packages makes U.S. regulations incompatible with IAEA 
regulations (viz., the U.S. and IAEA rules are not identical but they 
are compatible). The NRC staff believes there is not a need to adopt 
Type C standards at this time because of the reasons specified in the 
proposed rule (67 FR 21402) and

[[Page 3728]]

    (a) The perception of a lack of a current or anticipated need,
    (b) The DOT import/export provisions that permit use of IAEA 
regulations, and
    (c) The existing U.S. regulations and laws covering plutonium air 
transport.
    This can be reevaluated during future periodic rulemakings for IAEA 
compatibility, as necessary. In addition, the proposed rule stated that 
upon request from DOT, NRC would perform a technical review of Type C 
packages against IAEA TS-R-1 standards. The comments do not indicate a 
current need; therefore, the NRC staff has decided to retain the 
position explained in its proposed rule to not adopt Type C or LDM 
requirements.
    Comment. One commenter said that air transport of plutonium and 
other radionuclides should be prohibited under all circumstances. The 
commenter stated that ``low dispersible materials'' is a faulty concept 
regarding air transport and urged NRC to abandon this concept.
    Response. The NRC staff disagrees with the comments that air 
transport of plutonium and other radionuclides should be prohibited 
under all circumstances. These practices are recognized in multiple 
U.S. laws and regulations, and have been carried out with an excellent 
safety record. Consistent with the position expressed in the proposed 
rule, the NRC decided not to adopt the low dispersible material 
provisions at this time.
Issue 7. Deep Immersion Test
    Summary of NRC Final Rule. The final rule adopts the requirement 
for an enhanced water immersion test (deep immersion test) which is 
applicable to any Type B or C packages containing activity greater than 
105A2. The purpose of the deep immersion test is 
to ensure package recoverability. The basis for expanding the scope of 
the deep immersion test to include additional Type B or C packages 
containing activity greater that 105A2 was due to 
the fact that radioactive materials, such as plutonium and high-level 
radioactive waste, are increasingly being transported by sea in large 
quantities. The threshold defining a large quantity as a multiple of 
A2 is considered to be a more appropriate criterion to cover 
all radioactive materials and is based on a consideration of potential 
radioactive exposure resulting from an accident. Also, the NRC is 
retaining the current test requirements in Sec. 71.61 of ``one hour w/o 
collapse, buckling or leakage of water.'' The NRC is retaining this 
acceptance criterion of ``w/o collapse, buckling, or leakage'' as 
opposed to the acceptance criterion specified in TS-R-1 of only ``no 
rupture'' of the containment. NRC has determined that the term 
``rupture'' cannot be determined by engineering analysis and the term 
``w/o collapse, buckling or leakage of water'' is a more precise 
definition for acceptance criterion.
    Affected Sections. Sections 71.41, 71.51, 71.61.
    Background. TS-R-1 expanded the performance requirement for the 
deep water immersion test (paragraphs 657 and 730) from the 
requirements in the IAEA Safety Series No. 6, 1985 edition. Previously, 
the deep immersion test was only required for packages of irradiated 
fuel exceeding 37 PBq (1,000,000 Ci). The deep immersion test 
requirement is found in Safety Series No. 6, paragraphs 550 and 630, 
and basically stated that the test specimen be immersed under a head of 
water of at least 200 meters (660 ft) for a period of not less than 1 
hour, and that an external gauge pressure of at least 2 MPa (290 psi) 
shall be considered to meet these conditions. The TS-R-1 expanded 
immersion test requirement (now called enhanced immersion test) now 
applies to all Type B(U) (unilateral) and B(M) (multilateral) packages 
containing more than 10 \5\ A2, as well as Type C packages.
    In its September 28, 1995 (60 FR 50248), rulemaking for part 71 
compatibility with the 1985 edition of Safety Series No. 6, the NRC 
addressed the new Safety Series No. 6 requirement for spent fuel 
packages by adding Sec. 71.61, ``Special requirements for irradiated 
nuclear fuel shipments.'' Currently, Sec. 71.61 is more conservative 
than Safety Series No. 6 with respect to irradiated fuel package design 
requirements. It requires that a package for irradiated nuclear fuel 
with activity greater than 37 PBq (10 \6\ Ci) must be designed so that 
its undamaged containment system can withstand an external water 
pressure of 2 MPa (290 psi) for a period of not less than 1 hour 
without collapse, buckling, or inleakage of water. The conservatism 
lies in the test criteria of no collapse, buckling, or inleakage as 
compared to the ``no rupture'' criteria found in Safety Series No. 6 
and TS-R-1. The draft advisory document for TS-R-1 (TS-G-1.1, 
paragraphs 657.1 to 657.7) recognizes that leakage into the package and 
subsequent leakage from the package are possible while still meeting 
the IAEA requirement.
    The Safety Series No. 6 test requirements were based on risk 
assessment studies that considered the possibility of a ship carrying 
packages of radioactive material sinking at various locations. The 
studies found that, in most cases, there would be negligible harm to 
the environment if a package were not recovered. However, should a 
large irradiated fuel package (or packages) be lost on the continental 
shelf, the studies indicated there could be some long-term exposure to 
man through the food chain. The 200-meter (660-ft) depth specified in 
Safety Series No. 6 is equivalent to a pressure of 2 MPa (290 psi), and 
roughly corresponds to the continental shelf and to depths that the 
studies indicated radiological impacts could be important. Also, 200 
meters (660 ft) was a depth at which recovery of a package would be 
possible, and salvage would be facilitated if the containment system 
did not rupture. (Reference Safety Series No. 7, paragraphs E-550.1 
through E-550.3.)
    The expansion in scope of the deep immersion test was due to the 
fact that radioactive materials, such as plutonium and high-level 
radioactive wastes, are increasingly being transported by sea in large 
quantities. The threshold defining a large quantity as a multiple of 
A2 is considered to be a more appropriate criterion to cover 
all radioactive materials and is based on a consideration of potential 
radiation exposure resulting from an accident.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter stated that a 1-hour test is ``wholly 
inadequate as a risk basis, given that as many as 100,000 shipments of 
highly irradiated `spent' fuel are anticipated to being moved 
transcontinentally on highways and railroads.'' The commenter added 
that ``barge shipments should be prohibited outright.'' Finally, the 
commenter recommended more stringent immersion testing for shipping 
canisters.
    Response. The NRC acknowledges the comment. However, the NRC 
believes it is already moving towards more stringent standards with 
this rule. The 1-hour test is sufficient to demonstrate structural 
integrity and prevent inleakage. Most hydrostatic testing of components 
are for durations much less than 1 hour. A test duration of 1 hour is 
reflective of a practical requirement that will ensure the desired 
package performance. While a longer duration test may appear to be more 
reflective of the actual immersion times that might exist following an 
accident, the duration of the test must be considered in conjunction 
with the purpose of the test and the acceptance criteria specified for 
successfully passing the test.

[[Page 3729]]

    The purpose of the deep immersion test, as described in IAEA TS-G-
1.1, paragraphs 657.1 to 657.7, is to ensure package recoverability. 
The acceptance criterion specified in TS-R-1 is that there be no 
``rupture'' of the containment system. As described in the rule, NRC 
believes that a more precisely defined acceptance criterion of no 
``collapse, buckling, or inleakage of water'' is preferable. Type B 
package designs that are capable of withstanding a 1-hour test without 
``collapse, buckling, or inleakage of water'' are likely to be 
sufficiently robust that a longer duration test would not produce 
significantly greater structural damage.
    Comment. One commenter suggested that the deep immersion test 
should consider the possibility that the cask could already be damaged 
or ruptured at the time of immersion. The commenter asked if there has 
been an analysis of the dissemination of radionuclides at high 
pressures for partially or completely ruptured casks. The commenter 
stated that this issue is relevant due to the frequent transportation 
of radioactive waste across the Great Lakes and between the U.S. and 
other nations, such as Russia.
    Response. The acceptance criterion for the deep immersion test is 
no ``collapse, buckling, or inleakage of water.'' If a cask is already 
damaged or ruptured at the time of immersion, then the immersion test 
becomes a moot point because the acceptance criterion cannot be met. 
Studies have been performed, including the IAEA-sponsored Coordinated 
Research Project on ``Severity, probability and risk of accidents 
during the maritime transport of radioactive material,'' that examined 
the potential radiological consequences of such accidents. The report 
of the Coordinated Research Project, IAEA-TECDOC-1231, is available 
online at: http://www.iaea.org/ns/rasanet/programme/radiationsafety/
transportsafety/Downloads/Files2001/t1231.pdf.
    Comment. One commenter stated that if older, previously certified 
packages can no longer be ``grandfathered,'' it will take significant 
effort to show that these packages meet the deep immersion test and 
will result in little safety benefit for the shipments.
    Response. The commenter's connection between immersion testing and 
grandfathering (see Issue 8) of existing certified packages is not 
obvious. Under current NRC regulations (Sec. 71.61), a package for 
irradiated nuclear fuel with activity greater than 37 PBq 
(106 Ci) must meet the immersion test requirement. Under the 
revised requirement, these same packages could be used for shipment of 
irradiated nuclear fuel containing activity greater than 105 
A2 and would not require additional immersion testing 
(because the packages must already comply with the test requirement).
    Comment. Three commenters expressed support for NRC's position on 
this issue. One commenter stated that the proposed rule's deep 
immersion test provisions would increase cask safety.
    Response. No response is required.
    Comment. One commenter urged the NRC to require more stringent 
testing procedures for both old and new shipping containers (including 
longer drops; greater crash impacts; longer and higher pressure water 
submersion; leakage resistance; higher, longer, more intense fire 
temperatures; and much greater explosive forces). Another commenter 
requested that NRC change its standards so that casks damaged in 
sequential tests would be required to survive immersion at depths 
greater than those in the proposed rule.
    Response. The NRC acknowledges this comment but believes that it 
has adequate package testing requirements in the rule.
    Comment. One commenter asked if containers that were not currently 
certified to carry over one million curies would become authorized to 
carry over one million curies under the proposed rule.
    Response. If a package design is not currently certified to carry 
over one million curies, its status will not be changed by this 
rulemaking. Any restrictions on a package design imposed through the 
NRC-issued CoC remain unaffected.
    Comment. One commenter stated that the cost of compliance was 
grossly underestimated, particularly for demonstrating cask integrity 
at 200 meters.
    Response. NRC staff appreciates the comment and fully understands 
the importance of accurate cost data. As part of the proposed 
rulemaking, the NRC specifically requested cost-benefit information on 
this issue as well as a number of other issues. To the extent NRC 
received data from public comments, these data were considered in 
developing its final decision.
    Comment. One commenter asked if the deep immersion test would apply 
to all packages shipped across Lake Michigan.
    Response. Under the proposed rule, the deep immersion test would be 
applied to any Type B or C package that contains greater than 
105 A2, regardless of the transport mode. 
Therefore, the immersion test requirement would be applicable to all 
shipments involving a package with an activity exceeding 105 
A2, including any across Lake Michigan.
    Comment. One commenter asked if the deep immersion test actually 
requires a physical test. If the deep immersion test did not actually 
require a physical test, the commenter asked NRC to clarify what it 
means by ``test.'' The commenter also wanted NRC to clarify to what the 
test specifically applies.
    Response. As cited in the IAEA advisory document TS-G-1.1, 
paragraph 730.2: ``The water immersion test may be satisfied by 
immersion of the package, a pressure test of at least 2 MPa, a pressure 
test on critical components combined with calculations, or by 
calculations for the whole package.'' In answer to the commenter's 
specific question, a physical test is not required, and calculational 
techniques may be used. Regarding what the test specifically applies 
to, ST-2, Section 730.3, states that: ``The entire package does not 
have to be subjected to a pressure test. Critical components such as 
the lid area may be subjected to an external gauge pressure of at least 
2 MPa and the balance of the structure may be evaluated by 
calculation.'' Thus, testing may be performed physically, by analysis, 
or by a combination of the two.
    Comment. One commenter stated that industry supports the NRC 
position on deep immersion testing.
    Response. The comment is acknowledged.
    Comment. One commenter expressed concern that the deep immersion 
test only requires that packages be submerged for 1 hour. The concern 
is based on the belief that it is unlikely a package could be recovered 
within an hour following a real accident.
    Response. The 1-hour time limit only applies to the immersion test 
and is the minimum time that the package shall be subjected to the test 
conditions. It is not expected that a package could be recovered within 
1 hour of an accident involving submergence of the package. In fact, in 
the IAEA advisory document TS-G-1.1, paragraph 657.7 states: 
``Degradation of the total containment system could occur with 
prolonged immersion and the recommendations made in the above 
paragraphs (657.1 through 657.6) should be considered as being 
applicable, conservatively, for immersion periods of about 1 year, 
during which recovery should readily be completed.''
    Comment. One commenter asked NRC to clarify its assertion that the 
immersion test is stricter than the IAEA's test because the NRC's 
language

[[Page 3730]]

does not allow collapse, buckling, or any leakage of water.
    Response. TS-R-1, paragraph 657, states, in part, that for a 
package subjected to the enhanced water immersion test (NRC uses the 
term deep immersion test), there would be no ``rupture of the 
containment system.'' The term rupture is not a defined engineering 
term in the IAEA literature related to TS-R-1. Further, the IAEA 
advisory document TS-G-1.1, paragraph 730.3, states, in part, that some 
degree of buckling or deformation is acceptable during the enhanced 
water immersion test. Lacking specificity to the term rupture, the NRC 
imposed specific, and it believes conservative, requirements that do 
not allow collapse, buckling, or inleakage of water for a package 
undergoing the deep immersion test.
Issue 8. Grandfathering Previously Approved Packages
    Summary of NRC Final Rule. The final rule adopts the following 
grandfathering provisions for previously approved packages in section 
71.13:
    (1) Packages approved under NRC standards that are compatible with 
the provisions of the 1967 edition of Safety Series No. 6 may no longer 
be fabricated, but may be used for a 4-year-period after adoption of a 
final rule;
    (2) Packages approved under NRC standards that are compatible with 
the provisions of the 1973 or 1973 (as amended) editions of Safety 
Series No. 6 may no longer be fabricated; however, may still be used;
    (3) Packages approved under NRC standards that are compatible with 
the provisions of the 1985 or 1985 (as amended 1990) editions of Safety 
Series No. 6, and designated as ``-85'' in the identification number, 
may not be fabricated after December 31, 2006, but may be continued to 
be used; and
    (4) Package designs approved under any pre-1996 IAEA standards 
(i.e., packages with an ``-85'' or earlier identification number) may 
be resubmitted to the NRC for review against the current standards. If 
the package design described in the resubmitted application meets the 
current standards, the NRC may issue a new CoC for that package design 
with a ``-96'' designation.
    Thus, the final rule adopts, in part, the provisions for 
grandfathering contained in TS-R-1. The NRC believes that packages 
previously approved under the 1967 edition of Safety Series No. 6 lack 
the enhanced safety enrichments which have been incorporated in the 
packages approved under the provisions of the 1973, 1973 (as amended), 
1985 and 1985 (as amended) editions of Safety Series No. 6. For 
example, later designs demonstrate a greater degree of leakage 
resistance and are subject to quality assurance requirements in subpart 
H of part 71. Furthermore, NRC believes that by discontinuing the use 
of package designs that have been approved to Safety Series No. 6, 
1967, for both domestic and international transport of radioactive 
material, it will ensure safety during transportation and thus will 
increase public confidence. However, NRC has not adopted the immediate 
phase out of 1967-approved packages as the IAEA has, Instead, NRC 
implemented a 4-year transition period for the grandfathering provision 
on packages approved under the provisions of the 1967 edition of Safety 
Series No. 6. This period provides industry the opportunity to phase 
out old packages and phase in new ones, or demonstrate that current 
requirements are met. NRC recognizes that when the regulations change 
there is not necessarily an immediate need to discontinue use of 
packages that were approved under previous revisions of the 
regulations. The final rule includes provisions that would allow 
previously-approved designs to be upgraded and to be evaluated to the 
newer regulatory standards. Note that in 1996, IAEA first published 
that the 1967-approved packages would be eliminated from use. Thus, 
with the final rule 4-year phase out of these older packages, industry 
will have had 12 years (i.e., until 2008) to evaluate its package 
designs and prepare for the eventual phase out.
    Affected Sections. Section 71.13.
    Background. Historically, the IAEA, DOT, and NRC regulations have 
included transitional arrangements or ``grandfathering'' provisions 
whenever the regulations have undergone major revision. The purpose of 
grandfathering is to minimize the costs and impacts of implementing 
changes in the regulations on existing package designs and packagings. 
Grandfathering typically includes provisions that allow: (1) Continued 
use of existing package designs and packagings already fabricated, 
although some additional requirements may be imposed; (2) completion of 
packagings that are in the process of being fabricated or that may be 
fabricated within a given time period after the regulatory change; and 
(3) limited modifications to package designs and packagings without the 
need to demonstrate full compliance with the revised regulations, 
provided that the modifications do not significantly affect the safety 
of the package.
    Each transition from one edition of the IAEA regulations to another 
(and the corresponding revisions of the NRC and DOT regulations) has 
included grandfathering provisions. The 1985 and 1985 (as amended 1990) 
editions of Safety Series No. 6 contained provisions applicable to 
packages approved under the provisions of the 1967, 1973, and 1973 (as 
amended) editions of Safety Series No. 6. TS-R-1 includes provisions 
which apply to packages and special form radioactive material approved 
under the provisions of the 1973, 1973 (as amended), 1985, and 1985 (as 
amended 1990) editions of Safety Series No. 6.
    TS-R-1 grandfathering provisions (see TS-R-1, paragraphs 816 and 
817) are more restrictive than those previously in place in the 1985 
and 1985 (as amended 1990) editions of Safety Series No. 6. The primary 
impact of these two paragraphs is that packagings approved under the 
1967 edition of Safety Series No. 6 are no longer grandfathered; i.e., 
cannot be used. The second impact is that fabrication of packagings 
designed and approved under Safety Series No. 6 1985 (as amended 1990) 
must be completed by a specified date. Regarding special form 
radioactive material, TS-R-1 paragraph 818 does not include provisions 
for special form radioactive material that was approved under the 1967 
edition of Safety Series No. 6. Special form radioactive material that 
was shown to meet the provisions of the 1973, 1973 (as amended), 1985, 
and 1985 (as amended 1990) editions of Safety Series No. 6 may continue 
to be used. However, special form radioactive material manufactured 
after December 31, 2003, must meet the requirements of TS-R-1. Within 
current NRC regulations, the provisions for approval of special form 
radioactive material are already consistent with TS-R-1.
    In TS-R-1, packages approved under Safety Series No. 6, 1973 and 
1973 (as amended) can continue to be used through their design life, 
provided the following conditions are satisfied: (1) Multilateral 
approval is obtained for international shipment; (2) applicable TS-R-1 
quality assurance (QA) requirements and A1 and A2 activity limits are 
met; and (3) if applicable, the additional requirements for air 
transport of fissile material are met. While existing packagings are 
still authorized for use, no new packagings may be fabricated to this 
design standard. Changes in the packaging design or content that 
significantly affect safety require that the package meet current 
requirements of TS-R-1.

[[Page 3731]]

    TS-R-1 further states that those packages approved for use based on 
the 1985 or 1985 (as amended 1990) editions of Safety Series No. 6 may 
continue to be used with unilateral approval until December 31, 2003, 
provided the following conditions are satisfied: (1) TS-R-1 QA 
requirements and A1 and A2 activity limits are 
met; and (2) if applicable, the additional requirements for air 
transport of fissile material are met. After December 31, 2003, use of 
these packages for foreign shipments may continue under the additional 
requirement of multilateral approval. Changes in the packaging design 
or content that significantly affect safety require that the package 
meet current requirements of TS-R-1. Additionally, new fabrication of 
this type of packaging must not be started after December 31, 2006. 
After this date, subsequent package designs must meet TS-R-1 package 
approval requirements.
Analysis of Public Comments on the Proposed Rule
    The NRC notes that although there were a significant number of 
comments reflecting opposition to the proposed grandfathering change to 
the regulation, the majority of these comments were received from two 
commenters representing the same company. The remaining comments 
reflected opinions ranging from strong opposition to any grandfathering 
of designs to full support for the proposed rule change. Accordingly, 
following discussions with the DOT, NRC changed the transition period 
from 3 years in the proposed rule to 4 years in the final rule. With 
the effective date of this final rule being October 1, 2004, the 
transition period is almost 5 years. A review of the specific comments 
and the NRC staff's responses for this issue follows.
    Comment. One commenter stated that the IAEA standards are consensus 
based and that NRC must recognize they do not necessarily consider the 
risk-informed, performance-based aspects of regulations that are 
developed in the United States. The commenter added that NRC 
regulations should also provide allowance for domestic-only 
applications, which would include, for example, the grandfathering 
provision. While the IAEA provisions must apply to international 
shipments, for domestic-only shipments the grandfathering provision 
would allow the continued use of existing packages manufactured to the 
1967 standard, but prohibit the manufacture of any new packages.
    Response. The NRC staff finding is to phase out those packages 
approved to Safety Series No. 6, 1967 Edition, over a 4-year period 
after October 1, 2004. The NRC believes this time period allows 
industry adequate time to phase out old packages, phase in new ones, or 
resubmit a package design for review against the current standards. NRC 
considers it undesirable to be incompatible with IAEA with respect to 
this provision. In eliminating the grandfathering of these older 
designs, the IAEA concluded and NRC agrees that the continuance of 
packages that could not be shown to meet updated standards was no 
longer justified. As described, certain packages approved under the 
1967 edition of the regulations may lack safety enhancements that later 
designs have incorporated. The NRC acknowledges the comment about risk-
informed, performance-based regulations but notes that the 
applicability of this change was not justified.
    Comment. One commenter suggested that NRC require far more 
stringent testing procedures for both old and new shipping containers 
(longer drops; greater crash impacts; longer and higher pressure water 
submersion; leakage resistance; higher, longer, more intense fire 
temperatures; and much greater explosive forces). Another commenter 
stated that ``packages and containers should be subject to upgraded 
safety testing and more rigorous standards than have been required in 
the past,'' especially after the events of September 11, 2001.
    Response. The NRC acknowledges these comments and notes that the 
commenters did not provide justification for the proposed changes. 
Packages designed to regulations that are based on the 1973 and later 
editions of Safety Series 6, in general, may include safety 
enhancements, including designs, that demonstrate a greater degree of 
leakage resistance. Major changes in the physical test parameters for 
Type B packages are not being considered at this time, either by NRC or 
the IAEA. NRC is confident that packages designed to meet the current 
Type B standards provide a high degree of safety in transport, even 
under severe transportation accidents.
    Comment. One commenter objected to any grandfathering of casks. The 
commenter stated that ``it will be a number of years before appreciable 
amounts of `spent' fuel can be transported for more permanent 
disposition'' and that this ``gives a substantial window of time for 
design, development, and proof testing of new, better shipping casks.''
    Response. The NRC and DOT have in place comprehensive regulations 
that will support the safety of a large scale shipping campaign to a 
central geologic repository should one ever be built. Such safety is 
reliant upon the use of certified casks with robust design and 
regulations that address training of staff dealing with shipments and 
use of routes that minimize potential dose to the public. The safety 
record of shipments of spent fuel both here and overseas has been 
excellent. NRC regulations are compatible with IAEA regulations with 
respect to grandfathering previously approved designs. These provisions 
allow continued use of designs approved to earlier regulatory 
standards; however, the provisions include certain restrictions with 
respect to package modifications and fabrication. These provisions have 
been adopted to allow a transition to newer regulations while 
maintaining a high level of safety in transport. Packages that were 
approved to the 1967 IAEA standards are being phased out because they 
may not include safety enhancements of later designs.
    Comment. One commenter stated that accurate data are not currently 
available to forecast cost-benefit impacts. The commenter urged NRC to 
work with those who hold Type B packages to determine whether they want 
to maintain these packages. A second commenter stated that the costs of 
requiring the replacement of 1967-specification packages are 
substantial and that the benefits of requiring the replacements for 
domestic use are zero. The commenter also stated that the NRC should 
allow usage periods to be extended long enough to ensure that the 
``money's worth'' has been obtained. The commenters added that NRC 
should not propose changes when no harm or hazard has been 
demonstrated.
    Response. The NRC has made the decision to begin a 4-year phase out 
of packages that have been approved to Safety Series No. 6, 1967. 
However, NRC will allow package designs to be submitted for review 
against the current requirements (TS-R-1). Based on this pathway, over 
the 4-year period (after effective date of the final rule), industry 
can determine which Type B packages they choose to submit for review to 
the current requirements or have them phased out of use for shipping. 
NRC has no current plans to contact individual design holders of 
affected package designs to suggest an action on their part.
    In evaluating the cost and benefits associated with the proposed 
phasing out of the 1967-based packages, the NRC staff considered that 
these designs may fall into one of the following five categories:

[[Page 3732]]

    (1) Package designs that may meet current safety standards with no 
modifications but have not been submitted for recertification. This 
category includes package designs for which there is probably 
sufficient supporting technical safety basis to support certification 
under current requirements. For example, test data and engineering 
analyses probably exist and are still relevant to the current safety 
standards.
    Costs associated with these package designs include the following:
    (a) Development of an application ($10-$50K); and
    (b) Review costs for NRC certification ($20K for 135 hours--
nonspent fuel amendment).
    The total costs might be expected to be in the range of $30-$70K 
per package design.
    (2) Package designs that can be shown to meet current safety 
standards with probably relatively minor design changes.
    Costs associated with these package designs include the following:
    (a) Design analysis and physical testing for modifications ($10K-
$100K);
    (b) Development of revised package application ($10K-$50K--based on 
approximately 200 staff hours of work);
    (c) Review costs for NRC certification ($20K--based on 135 staff 
hours for review of nonspent fuel amendment requests); and
    (d) Packaging modifications to fleet of packagings (minor--$200 per 
packaging, major--$5K per packaging).
    The total cost would be expected to be in the range of $40K to 
$170K depending on the modifications in the design or testing 
information. This does not include the costs for making the physical 
changes in the packagings, which could vary significantly for different 
package types and different design modifications, in addition to the 
number of packagings that needed to be modified.
    For packages in Categories 1 and 2, NRC staff believe that the 
expense of recertifying the design should be reasonable and is small 
when considering the length of time these package designs have already 
been in service (longer than 20 years). There is additional financial 
incentive for upgrading these designs, because upgrading would allow 
additional packagings to be fabricated and allow certificate holders to 
request a wide range of modifications, both to the package design and 
the authorized contents.
    (3) Package designs that may meet current safety standards but are 
impractical to recertify.
    This category is intended to capture the special nature of spent 
fuel casks that were certified to the 1967 IAEA standards. These 
package designs may be considered separately for several reasons, 
including:
    (a) Domestic regulatory design standards for spent fuel casks 
existed before standards for other package types;
    (b) QA requirements were applied to this type of package, whereas 
other package types were not subjected to the same level of QA either 
for design or fabrication; and
    (c) These packages normally have a limited specific use and are, 
therefore, not present in large numbers in general commerce.
    For packages in this category, NRC staff will be willing to review 
an application under the exemption provisions of Sec. 71.8 that 
requests an exemption to specific performance requirements for which 
demonstration is not practical. The applicant would be free to propose, 
for example, additional operational controls that would provide 
equivalent safety. The exemption request could use risk information in 
justifying the continued use of these existing packagings.
    Costs associated with these package designs include the following:
    (a) Development of application, including risk information ($150K); 
and
    (b) NRC review costs ($40,000--based on 270 staff hours for a 
``non-standard'' spent fuel package amendment request).
    (4) Package designs that cannot be shown to meet current safety 
standards.
    Costs associated with these package designs include the following:
    (a) Development of new designs ($100-150K);
    (b) Analysis and physical tests ($50K for prototype + 100K);
    (c) Development of package application;
    (e) NRC review costs ($40,000--based on 270 staff hours for review 
of new designs for nonspent fuel); and
    (f) Fabrication costs ($50K per package).
    The cost information for development of new designs and the 
analysis and testing of these newly designed packages (Category 4) were 
provided to NRC by industry commenters during the public comment 
period.
    (5) Packages for which the safety performance of the package design 
under the current safety standards is not known. This is due primarily 
to a lack of documentation available regarding the package design and 
performance.
    NRC staff believes it is appropriate to phase out the use of 
designs that fall into Categories 4 and 5. NRC staff believes that 
there are package designers that may be willing and able to develop new 
designs provided there is a financial incentive. With the continued use 
of packages that cannot be shown to meet current standards, there will 
be no financial incentive to upgrade designs. In addition, most 
packagings certified to the 1967 design standards are more than 20 
years old. Although proper maintenance of transportation packagings is 
required, it is not clear that the service life of many types of 
packagings would justify continued use.
    The cost estimates associated with NRC review are based on 
historical information gathered over years of performing technical 
reviews of transportation package designs. There are many factors that 
significantly influence the review time associated with performing 
staff technical reviews for new package designs and amendments. Some of 
the most important factors are: quality of the application, design 
margins in the package, and a clear and unambiguous demonstration that 
the regulatory acceptance criteria have been met. The costs previously 
cited are not considered maximum or minimum but are representative and 
conservative averages based on receipt of a complete and high-quality 
package application.
    The estimates of costs associated with development of designs, 
testing, and preparation of application are extrapolated from 
information provided by commenters to the proposed rule.
    Comment. One commenter stated that packages that were manufactured 
to the 1967 safety standard should be allowed to continue in domestic 
service, unless a safety problem is identified. This commenter provided 
monetized data to show how expensive our proposed position could be.
    Response. In the final rule published September 28, 1995 (60 FR 
50254), NRC wrote: ``NRC believes that the international package 
standards should be used by the United States for both domestic and 
international shipments, to the extent practicable. However, based on a 
history of safe use under earlier safety standards, and the absence of 
unfavorable operational data, NRC will allow the continued use of 
existing packages in domestic transport until the end of their useful 
lives. NRC will not allow, however, the continued fabrication of 
packages to the old designs. This action permits use of existing 
packages. It does not perpetuate package designs that can be discarded 
or upgraded to satisfy the new standards.''
    Further, in the April 30, 2002 (67 FR 21405), proposed rule, NRC 
wrote ``The NRC recognizes that when the

[[Page 3733]]

regulations change there is not an immediate need to discontinue use of 
packages that were approved under previous revisions of the 
regulations. Part 71 has included provisions that would allow 
previously-approved designs to be upgraded and to be evaluated to the 
newer regulatory standards. NRC believes that packages approved under 
the provisions of the 1967 edition of Safety Series No. 6, and which 
have not been updated to later editions, may lack safety enhancements 
which have been included in the packages approved under the provision 
of the 1973, 1973 (as amended), 1985 and 1985 (as amended 1990) 
editions of Safety Series No. 6. Therefore, the NRC believes that it is 
appropriate to begin a phased discontinuance of these earlier packages 
(1967-approved) to further improve transport safety.''
    NRC adopted the 1985 IAEA standards on April 1, 1996 (60 FR 50248), 
which allowed continued use of 1967 packages. In 1996, however, IAEA 
published new regulations in TS-R-1 which discontinued grandfathering 
these older designs. NRC agrees with IAEA's position that continuance 
of these older designs is no longer justified. Therefore, to be 
compatible with IAEA, NRC will begin a phased discontinuance of the 
packages approved to Safety Series No. 6, 1967 after adoption of a 
final rule.
    The NRC has justified phasing out these designs based on the 
following:
    Safety standards have been upgraded three times since these designs 
were initially evaluated and approved. In some cases, the documented 
safety basis for these designs is substantially incomplete. Although 
NRC knows of no imminent safety hazards posed by use of these packages, 
it is judged to be prudent to be consistent with IAEA in phasing out 
these designs. In addition, the performance of the package in a 
transportation accident may not be known until a challenging accident 
occurs.
    Opportunity was provided to upgrade these designs to later 
regulatory standards; however, applicants chose not to provide an 
application to show that the designs met later safety standards. That 
opportunity still exists and should be used by package owners that rely 
on these packages for transporting their products.
    Although there is a financial impact for phasing out these designs, 
it is judged that there will also be a financial benefit to package 
designers that choose to develop replacement packages that meet current 
domestic and international safety standards.
    Comment. One commenter stated that the proposed rule has no 
discernible safety benefit to adopting TS-R-1 on this issue, there is 
no direct economic information on the effect of implementing this 
proposal, and NRC has requested cost-benefit information from the 
regulated community.
    Response. The NRC does not agree that there is no safety benefit in 
adopting TS-R-1 provisions on grandfathering. The NRC believes that 
packages approved to later safety standards (after 1967) may include 
important safety enhancements. The grandfathering provision allows a 4-
year phase out period. Based on this pathway, over the impending 4-year 
period (after effective date of the final rule), certificate holders 
can determine which Type B packages they choose to have phased out or 
reviewed to the current requirements. The commenter accurately notes 
that NRC has solicited cost information regarding this proposal.
    Comment. Three commenters stated that the proposed rule's effort to 
phase out 1967-specification packages would negatively impact their own 
business. One commenter argued that phasing out these packages would 
have such a high cost that it would drive many small nuclear-shipping 
businesses out of business with no ready successors. Another commenter 
stated that phasing out these packages would cost about $20-$25 million 
and could force some entities out of business, which could create an 
unintended side-effect of orphaning over 1,000 radioactive sources of 
considerable size. Another commenter discussed his business of 
designing, manufacturing, servicing, shipping and disposing of devices 
(principally calibrators and irradiators) that use Type B quantities of 
Cobalt-60 or Cesium-137 sources, and the process of shipping 
radioactive sources and how it relates to his business. The commenter 
discussed the impact of phasing out 1967-specification packages. The 
commenter argued that phasing out these packages for domestic shipments 
would impose substantial economic, safety, and environmental costs 
without any benefits.
    Response. The NRC believes that packages approved under the 
provisions of the 1967 edition of Safety Series No. 6, and which have 
not been upgraded to later editions, may lack safety enhancements which 
have been included in packages developed to later standards. NRC is 
seeking to be compatible with the IAEA on the issue of grandfathering 
and is not seeking to put shipping companies out of business. 
Therefore, this final rule will phase out, 4 years after the rule 
effective date, those packages that have been approved to Safety Series 
No. 6, 1967. The NRC believes that many of the suggested orphaned 
sources would qualify as Type A quantities and would not be negatively 
impacted by the phase out of the 1967-approved packages.
    Comment. One commenter opposed NRC's proposal on this issue because 
it will have detrimental effects on his business. The commenter 
explained that his company has 1,200 new packages built to the 1967 
Safety Series No. 6 specifications that will be used in a contract that 
runs through 2006. The company estimates that replacing these packages 
would cost $5,000-$10,000 per package, which overall would devastate 
the contract and be ruinous to the business. The commenter believes 
that packages should be removed from service when they no longer meet 
the safety requirements they were designed to meet or if a new safety 
issue with the package is identified which would prevent the package 
from meeting its intended safety function; neither of these conditions 
have been identified for the package.
    Response. With the adoption of the final rule, the opportunity 
exists to have packages that were built to the 1967 Safety Series No. 6 
specifications reevaluated to the current standards. Since August 1986, 
fabrication of new packages to the old (1967) specifications has not 
been authorized by NRC. The comment supports NRC's pre-1995 position 
that, based on satisfactory performance, the 1967-type packages could 
continue to be used. The new packages suggested in the comment are 
assumed to have been fabricated in accordance with DOT regulations. 
However, NRC's and DOT's current position, which is consistent with the 
IAEA's on grandfathering, is to phase out the packages with these old 
designs over a 4-year period. This time period will allow certificate 
holders to determine which packages they will phase out or resubmit to 
NRC for evaluation to the current standards. Industry needs to be aware 
of changes or potential changes based on IAEA rules. Note in 1996, IAEA 
first published that the 1967-approved packages would be eliminated, 
and 5 years later (i.e., 2001) the international regulations were 
implemented. Thus, with the 4-year phase out of the 1967-approved 
packages, industry will have had 12 years (i.e., until 2008) to 
evaluate their package designs, evaluate those designs that will not 
meet the new standards, and prepare for the eventual phase out.

[[Page 3734]]

    Comment. One commenter stated that eliminating 1967-specification 
packages would cause severe harm. The commenter argued that many 
businesses would have to requalify, relicense, and rebuild virtually 
all of their current shipping containers at a very high cost. The 
commenter noted that the RA did not take these costs into account. The 
commenter argued that prohibiting the use of 1967-specification 
packages would create thousands of orphan sources, creating a public 
health risk, and that these sources could only be moved at very high 
costs.
    Response. The NRC notes that businesses may choose to requalify, 
relicense, or rebuild their packages. Based on the long history 
associated with grandfathering various packages, NRC believes that a 4-
year time period will allow certificate holders adequate opportunity to 
make a responsible business decision as to which pathway to proceed--
phasing a package design out or resubmitting it for evaluation to the 
current standards.
    Comment. One commenter stated that certain containers excluded by 
the proposed legislation couldn't be easily replaced because no 
alternative packaging currently exists at comparable prices. The 
commenter explained that designing, testing, and licensing a new 
package is expensive (approximately $500,000) and usually takes over a 
year to accomplish.
    Response. The NRC acknowledges the comment about the cost and time 
to design a new package. The staff notes that from the time TS-R-1 
became effective to the date when NRC's grandfathering phase out 
becomes effective will have been a significant and sufficient amount of 
time for designers to learn about the new requirements, and to adopt 
design and fabrication effort accordingly. As such new and conforming 
packages would be available for use when needed by shippers.
    Comment. One commenter stated that the RA lacks consideration of 
costs to industry and health and safety benefits of the proposed 
changes. The commenter believes that there were no arguments to be made 
and that the only rationale would be harmonization with the IAEA, which 
is not binding under U.S. law.
    Response. The NRC disagrees that the only rationale for this 
rulemaking is harmonization with the IAEA. NRC continues to believe 
that harmonizing NRC's and DOT's regulations, when appropriate, will 
prove beneficial to NRC, industry, and the general public. NRC believes 
that packages approved to the 1967 standards lack safety enhancements 
that were included in packages approved to later editions of Safety 
Series No. 6 (i.e., 1973 and 1985).
    Comment. One commenter stated that numerous participants in this 
market sector are small entities within the meaning of the Regulatory 
Flexibility Act and would be adversely affected by the proposed rule, 
and neither agency's draft RA accounts for this fact.
    Response. The NRC disagrees with this comment. The Commission 
certified in Section XI of this notice that this rule will not have a 
significant economic impact on a substantial number of small entities. 
This rule affects NRC licensees, including operators of nuclear power 
plants, who transport or deliver to a carrier for transport, relatively 
large quantities of radioactive material in a single package. These 
companies do not generally fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
size standards adopted by the NRC (10 CFR 2.810).
    Only one small entity commented on the proposed changes suggesting 
that small entities would be negatively affected by the rule. Reviewing 
records of licensed QA programs, NRC found that only 15 of the 127 NRC 
licensed QA progams were small entities. Furthermore, of these 15 
companies, NRC staff expects that only 2 or 3 would be negatively 
affected by the final rule, given these companies' lines of business 
and day-to-day operations. Based on this data, it is believed there 
will not be significant economic impacts for a substantial number of 
small entities.
    Comment. One commenter asked how important this issue is to the 
future success of small businesses that routinely transport Type B 
quantities of radioactive materials domestically. The commenter found 
it difficult to understand why some packages with proven safety records 
would ``unjustly'' be phased out for domestic shipments in as little as 
2 years after the proposed rule is issued.
    Response. To be compatible with the IAEA on grandfathering, NRC has 
made a decision to phase out those packages that may lack safety 
enhancements found in other packages. This phase out will impact 
packages approved to Safety Series No. 6, 1967, and will be completed 4 
years after adoption of a final rule. This phase out is consistent with 
NRC's belief that packages approved to the 1967 edition of Safety 
Series No. 6 may lack safety enhancements that are included in packages 
approved to later editions.
    Comment. One commenter supported grandfathering casks made for the 
1967 standards for domestic shipping and urged NRC to retain the 
A2 value for molybdenum-99 and the A1 and 
A2 values for californium-252, also for domestic shipping.
    Response. NRC will retain the current A2 value for 
molybdenum-99 (7.4E-1 TBq; 2.0E1 Ci) and the A2 value for 
californium-252 (0.1 TBq; 2.7 Ci) (see Table A-1). The NRC is not 
adopting the A1 value for californium-252 because the IAEA 
is considering changing the value that appears in TS-R-1 back to what 
presently appears in part 71. For reasons stated in the previous 
response to comments, NRC will not allow grandfathering of packages 
certified to the 1967 standard.
    Comment. Because IAEA does not necessarily consider the risk-
informed, performance-based aspects of regulations that the NRC has 
developed in the United States, a commenter suggested that the NRC 
should consider the unique aspects of U.S.-only applications. The 
commenter also suggested that the package identification number should 
be revised to the appropriate identification number prefix together 
with a suffix of ``-96'' provided that such packages shall be for 
domestic use only and no additional packages be fabricated.
    Response. The NRC does not agree with this suggestion because it 
would allow continued use of B( ) packages for domestic use. NRC has 
determined that only those packages that have enhanced safety features 
(i.e., post-1967 package designs) will be allowed to be used and 
manufactured beyond the 4-year phase-out period for all use (domestic 
and international). When a package design designated as B( ) (i.e., 
approved to Safety Series No. 6, 1967) is submitted to NRC for review 
to the current standards, the NRC may revise the package identification 
number to designate the package design as a B, BF, B(U), B(M), etc, and 
may assign the ``-96'' suffix to indicate that the design has met the 
requirements of part 71. Those submitted package designs that do not 
meet the current standard will not be assigned the ``-96'' suffix.
    Comment. One commenter stated that adopting the revised 
``grandfathering'' provision rule would have a significant impact on 
the commenter's operations. The commenter highlighted how their 
operational need to store fuel would cause unnecessary handling of 
fuel, especially in light of design parameters to which their existing 
containers must adhere. Replacement of certified containers with 
satisfactory safety records is believed unnecessary by the commenter.

[[Page 3735]]

    Furthermore, the commenter added that, if adopted, this proposal 
would eliminate the flexibility to use M-130 containers on an ``as 
needed'' basis. The commenter stated that these containers are safe and 
asked that NRC consider allowing certified containers with satisfactory 
safety records to continue to be ``grandfathered.''
    Response. The NRC acknowledges the comment but notes that the 
certificate holder could choose to request a recertification before use 
beyond the 4-year phase-out period.
    Comment. One commenter was concerned that, in departing from IAEA 
grandfathering standards, NRC is placing the burden entirely on the 
regulated industry to develop the justification for such a departure. 
The commenter asserted that this is a problem because there was no 
basis for having adopted the IAEA grandfathering standards in the first 
place.
    Response. In the interest of maintaining compatibility with the 
IAEA regarding approved package designs to support the NRC's decision 
to be consistent with IAEA on the grandfathering issue (i.e., phasing 
out the Safety Series No. 6, 1967 package designs), and to allow only 
those package designs with enhanced safety features to continue to be 
used as viable packages, NRC will phase out the 1967-approved B( ) 
packages over a 4-year period after adoption of the final rule. Thus, 
NRC does not agree with the comment ``departing from IAEA 
grandfathering standards'' because NRC is making an effort to adopt the 
IAEA grandfathering standards. The primary difference between the IAEA 
and the NRC on this issue, however, is that IAEA has made an immediate 
phase out of the 1967-approved packages, while NRC will phase out the 
same packages over a 4-year period.
    Comment. One commenter requested specific information on the types 
and numbers of packages that would be affected and the timetable under 
which packages would be excluded.
    Response. The response to this comment is found at 67 FR 21406; 
April 30, 2002. NRC does not require certificate holders or licensees 
to submit information concerning the number of packages made to a 
particular CoC.
    Comment. One commenter stated that a regular 2-year reconsideration 
of package design regulations will lead to a situation where package 
designers and users will constantly be trying to keep up with ever-
changing regulations.
    Response. NRC is aware of this concern and does not anticipate 
major changes to the IAEA packaging standards every 2 years. 
Additionally, NRC participates in the 2-year IAEA revision process and 
will work with the IAEA and other member nations to assure that 
proposed changes include appropriate justification with respect to cost 
and safety.
    Comment. One commenter disagreed with the proposed grandfathering 
rule, stating that 1967-specification packages have operated 
successfully for years and that there is no health or safety reason for 
phasing them out. The commenter stated that extending the transition 
period beyond 3 years would delay the negative economic impacts of 
excluding these packages. The commenter did agree with the stricter 
standards for new packages in the proposed legislation. The commenter 
also agreed with the phase out of 1967-specification packages from 
international sources.
    Response. NRC agrees that the 1967-approved packages have appeared 
to provide adequate performance in the past. However, these packages 
lack the safety enhancements that other similar packages currently have 
in place (i.e., post-1967 approved packages). Therefore, NRC believes 
the time has come to phase out those package designs before a safety 
issue occurs and to capitalize on those packages that have incorporated 
the safety enhancements described in the proposed rule (67 FR 21406; 
April 30, 2002). This phase out of the 1967 approved package designs is 
consistent with the NRC's decision to be compatible with the IAEA on 
the grandfathering issue.
    Comment. One commenter expressed concern about the backfitting 
issue and indicated that NRC should demonstrate that the basis for 
IAEA's position is tenable in the U.S., or develop an independent 
satisfactory basis for their position. The commenter stated that this 
is particularly important with regard to grandfathering packages when 
there may be different environments for international and domestic 
shipments.
    Response. The NRC does not support allowing the continued use of 
the 1967-approved packages for domestic-use only. The NRC will continue 
to phase out those package designs that currently meet Safety Series 
No. 6, 1967, over a 4-year period after adoption of a final rule. This 
approach is consistent with the NRC's desire to be compatible with the 
IAEA on the grandfathering issue.
    Comment. One commenter said that the proposed 3-year transition 
period is too long.
    Response. NRC has used the 3-year time line in previous rulemakings 
and believes that this time period adequately supports those steps that 
could be taken regarding grandfathering. However, NRC has worked with 
the DOT and determined that a 4-year transition period would allow 
certificate holders an additional year to determine the most effective 
pathway for a particular design; namely, phase out old package designs, 
phase in new package designs, or submit an existing package design for 
review against the current standard.
    Comment. One commenter was concerned that the proposed rule would 
essentially remove from service any and all containers that could be 
used to transport isotopes from DOE's Advanced Test Reactor for medical 
or industrial use.
    Response. As with other package designs approved to the 1967 
standards, it is expected that certificate holders may request review 
of these designs to the current regulatory standards.
    Comment. Two commenters asserted that there is no safety benefit to 
phasing out the 1967-specification packages. One of these commenters 
noted that packages built to the 1967-specifications have an excellent 
safety record and that NRC and DOT agree that the level of safety of 
the 1967-specification is satisfactory. The commenter stated that the 
phase out may be required for international shipping but not for 
domestic shipping. The other commenter provided information on the high 
cost of recertification and stated that these costs would likely drive 
companies out of business.
    Response. NRC is aware of the safety record of those packages 
approved to Safety Series No. 6, 1967. However, NRC has made a decision 
based on safety to be compatible with the IAEA on the issue of 
grandfathering previously approved packages. Therefore, NRC will impose 
a 4-year phase out of those package designs approved to the 1967 
standards. While the IAEA has immediately terminated the use of 1967-
approved packages, the NRC has elected to terminate their use over a 4-
year period after adoption of a final rule. Any package design impacted 
by the phase out may be submitted to NRC for review against the current 
standards. While this review may be costly, it ensures package safety 
during transport and is compatible with the IAEA.
    Comment. One commenter asserted that the 1967-specification 
packages may be impossible to replace at any cost because these devices 
lack the ``QA Paper'' required under the NRC's regulations at 10 CFR 
part 71. The commenter stated that these packages serve unique 
functions and that phasing them out would leave thousands of Type B 
sources stranded, and the cost of moving them would be prohibitive. The 
commenter raised concerns about

[[Page 3736]]

exposure to these immovable packages and terrorism threats.
    Response. NRC is aware that packages built to the 1967 standards 
were not subject to QA requirements and that fabrication documents may 
not be available. This is one reason why the NRC decided to incorporate 
new standards in NRC regulations and discontinue use of the packages 
certified to the 1967 standards.
    Comment. One commenter said that currently approved DOT 
specification packages should continue to be approved for domestic 
shipments. The commenter based this suggestion on the fact that 
packages that are currently accepted for use and proven to be safe 
should continue to be used until they reach the end of their useful 
life. The commenter did not believe that the costs that would be 
associated with phasing out safely used transportation packages could 
be justified on the basis of harmonization of regulations with TS-R-1.
    Response. NRC has made a decision based on safety to phase out the 
package designs that do not include the safety enhancements that other 
packages currently maintain. Thus, the package designs that were 
approved to Safety Series No. 6, 1967, will be phased out over a 4-year 
period after adoption of the final rule. This approach is consistent 
with the NRC decision to eliminate these types of packages for 
transportation of radioactive materials. The safety enhancements for 
post-1967 package designs can be found in the proposed rule (67 FR 
21406; April 30, 2002).
    Comment. One commenter urged the NRC to accept Competent Authority 
Certificates for foreign-made Type B packages without requiring 
revalidation by a U.S. Competent Authority. The commenter stated that 
revalidation of foreign-made packages for which a country has issued a 
Competent Authority Certificate other than the United States in 
accordance with TS-R-1 is a redundancy that provides no additional 
benefit.
    Response. General license provisions in part 71 authorized use of 
foreign-approved designs for import or export shipments provided that 
DOT has revalidated the certificate. DOT may choose to request NRC 
technical review of those designs. NRC experience has been that review 
of those designs has been useful in identifying possible safety issues.
    Comment. One commenter stated that there needs to be an effective 
date applied to some or all of the proposed rule changes to grandfather 
existing approved transport cask designs. Without that, all part 71 CoC 
holders will be subject to backfit for compliance with no commensurate 
safety benefit. The commenter urged NRC to perform a comprehensive 
evaluation of what impact the proposed changes will have on existing 
dual-purpose certificate holders if a grandfather clause is not 
included in the rule.
    Response. NRC is committed to working with DOT and the IAEA to 
assure that future changes in package performance standards are limited 
to those that are justified and are shown to be significant with 
respect to safety.
    Comment. One commenter urged NRC to provide a flexible CoC design 
concept, which would permit internal packages whose dimensions and 
weight fell within defined ranges (rather than being unique), to be 
linked with one outerpack design of specific dimensions for shipment, 
thus minimizing the number of separate CoCs to be obtained.
    Response. Grandfathering provisions in Sec. 71.13 include certain 
restrictions with respect to changes to previously approved designs. 
However, for designs approved under the current regulations, a CoC can 
be issued to show ranges for dimensions and weights at the request of a 
certificate holder. The application for such a provision should include 
an evaluation that shows that the ranges of weights and dimensions 
would not negatively affect the performance of the package and its 
ability to meet the requirements of part 71.
    Comment. One commenter requested specification of the means by 
which existing packages that were built before required compliance with 
NRC QA standards can be qualified under the new regulations, without 
requiring full, unobtainable ``QA Paper'' compliance.
    Response. Packagings constructed to designs approved under the 1967 
regulations were, in general, not subject to QA requirements in part 
71. This was a consideration in NRC's decision to discontinue the use 
of packages certified to the 1967 standards and to remain compatible 
with IAEA on the grandfathering provisions. QA requirements in subpart 
H of part 71 include provisions for existing packagings with respect to 
QA.
    Comment. One commenter suggested that NRC change the ``timely 
renewal'' principle so as to enable holders of 1967-specification 
packages that submit substantially complete applications for new or 
requalified packages at least 1 year ahead of the ultimate phase-out 
date to continue shipments past the phase-out deadline, pending NRC's 
action on their request for certification or recertification.
    Response. NRC does not agree with this comment or the suggested 
approach. In 1996, IAEA rules indicated that package designs approved 
to Safety Series No. 6, 1967, would be eliminated. The NRC is revising 
its rules to maintain compatibility with these IAEA rules. Therefore, 
the idea of phasing out these packages has been public knowledge for 7 
years. IAEA rules regarding the elimination of the 1967-approved 
packages were implemented in 2001 (5 years after being published). NRC 
has posed a phase out of these package designs 4 years after adoption 
of a final rule (i.e., in 2008). Thus, the overall timeframe already 
encompasses 12 years, which is more than ample time to submit design 
upgrades and have them approved by the NRC.
    Comment. Two commenters expressed support for the proposed rule on 
this issue. One commenter encouraged NRC to accept the IAEA 
transitional requirements including the phase out of Type B 
specification packages and the termination of authorization of Safety 
Series 6 (1967) packages. The commenter said that these packages were 
not designed and constructed according to standards where their 
continued use would be consistent with the intent of the regulations.
    Response. NRC acknowledges these comments. NRC will phase out the 
packages designed to Safety Series No. 6, 1967, 4 years after adoption 
of the final rule.
    Comment. One commenter expressed support for NRC's proposal to 
allow continued safe use of existing packaging through incorporation of 
the TS-R-1 transitional arrangement provisions.
    Response. NRC acknowledges this comment.
    Comment. One commenter suggested that changes to A1 and 
A2 exemption values were relevant to grandfathering 
transport casks. The commenter believed that the NRC grandfathering 
proposal could adversely impact currently certified casks by not 
guaranteeing that casks certified under previous revisions ``will still 
be usable without modification or analysis in the future.''
    Response. The A1 and A2 values were last 
changed in part 71 in 1995 (see 60 FR 50248; September 28, 1995) to 
make the NRC regulations compatible with Safety Series No. 6, 1985. 
With those changes and the adoption of new LSA definitions came the 
awareness that a licensee, when using a CoC-controlled transport 
container, had to apply the new A1 or A2 value 
for a given radionuclide, determine the appropriate LSA limit, yet not 
exceed the activity

[[Page 3737]]

limit for which the transport package was tested, and which was based 
on the old (pre-September 28, 1995) A values. A very similar scenario 
also exists regarding the new A1 and A2 values 
and the existing transport containers. In other words, the new 
A1 and A2 values would be used as the limits for 
a shipment by a licensee, but the transport container's activity limit 
would still be based on the pre-September 28, 1995, A values. Should a 
package design be submitted for review to the current part 71, that 
design would be subject to the current (i.e., TS-R-1) A1 and 
A2 values that are part of this final rule. Thus, while NRC 
is aware of the commenter's concern, industry has already had to 
respond to a similar situation after April 1, 1996, when the September 
28, 1995, final rule became effective.
    Comment. One commenter expressed support for the phase out of the 
1967-specification containers for international shipping to comply with 
IAEA regulations. However, the commenter opposed the phase out for 
domestic shipping, arguing that as long as these packages are 
performing their function safely, then there is no benefit to the phase 
out and extremely high economic costs. The commenter stated that there 
would be huge environmental costs to the creation of hundreds or 
thousands of new orphan sources. The commenter stated that there would 
be large economic costs of these orphan sources because they will have 
to be kept secure. The commenter noted that no facility in possession 
of one of these devices will ever be able to terminate its license or 
perform a close-out radiation survey, and sale or shutdown will be 
impossible.
    Response. The NRC has made a decision to phase out those package 
designs that have been approved to Safety Series No. 6, 1967, for both 
domestic and international transport of radioactive material. NRC 
believes that package designs that include the safety enhancements (see 
67 FR 21406; April 30, 2002) better suit the goals of the NRC and its 
desire to ensure safe transport of all radioactive materials. NRC will 
work closely with those licensees who may have sources that cannot be 
easily transported as a direct result of this rule to provide a 
suitable resolution. This could result in economic incentives for 
package designers to develop new packages to retrieve orphan sources. 
This could also result in the development and certification of a new 
generation of Type B packages that could meet current safety standards 
and fulfill that need for transport of certain radiation sources.
    Comment. One commenter discussed the economic impacts of phasing 
out 1967-specification packages on the entire nuclear waste-shipping 
industry, estimating the total costs to the sector at over $1 billion. 
The commenter argued that these estimates refuted the projection in 
both NRC's and DOT's rulemaking notices, and the NRC's draft RA that 
did not expect any significant costs to be associated with the 
implementation of the rule. To arrive at this estimate, the commenter 
predicted three possible outcomes and discussed these scenarios in the 
comment letter. In two scenarios, the customers would have to design 
and construct new containers and ship them at high costs. The commenter 
discussed these costs in detail. In the third scenario, large amounts 
of radioactive sources would be orphaned and would remain immovable 
indefinitely.
    Response. Based on the information provided by this commenter and 
others regarding the costs of replacement packages, the NRC developed 
an estimated cost of impacts, as previously described. The estimate is 
based on either showing that the old designs meet current standards or 
replacing older designs. The NRC does not have sufficient information 
to substantiate the large costs estimated in this comment, partly 
because NRC does not collect information regarding the number of 
individual packagings fabricated to each design. However, based on 
staff's knowledge, the following financial impacts specified in the 
comment may not be reasonable:
    1. The commenter claims that the cost of design, testing, and 
licensing of new designs is estimated as $12 to $98 million. Based on 
the assessment provided, even assuming that about half of the current 
1967-based designs do not meet current safety standards and would need 
to be phased out, the total costs to industry would not approach these 
values. The derivation of these values cannot be substantiated by 
information available to the NRC.
    2. Cost of construction of new overpacks is stated as $7 to $13 
million. These costs do not seem consistent with NRC knowledge of the 
number of overpack designs currently in use.
    3. Loss of existing overpacks and the loss of value of existing 
devices are estimated from $500 to over $1,000 million. The derivation 
of this value cannot be substantiated by information available to the 
NRC.
    Comment. One commenter stated that phasing out 1967-specification 
containers would cause many nuclear-shipping firms to go out of 
business, which would create thousands of orphan sources that are 
unshippable and unmovable. The commenter stated that NRC would be 
responsible for storing and securing these sources indefinitely and 
protecting worker and public safety. The commenter noted that this 
could create national security concerns with the potential for theft by 
terrorists. The commenter stated that as long as these sources are 
immovable, an entity could not conduct a final radiation survey and 
terminate its license, forcing the entity to remain indefinitely on NRC 
or Agreement State rolls.
    Response. The commenter provided no justification for the opinion 
that shipping firms would be forced to go out of business. The NRC 
believes that if this situation occurs, package designers would be 
motivated to develop new packages to retrieve orphan sources. This 
could result in the development and certification of a new generation 
of Type B packages (that would incorporate the current package 
standards) that could fulfill that need.
    Comment. One commenter stated that new containers would be 
adequate, if they could be feasibly built. The commenter also stated 
that the existing containers are adequate. The commenter stated that 
orphan sources created by ``sunset'' on use of existing 1967-
specification containers decrease protection of public health and 
safety protection.
    Response. Regarding transport of radioactive material, NRC believes 
that phasing out those package designs approved to Safety Series No. 6, 
1967, will assure transport safety due to the fact that the package 
designs will have enhanced safety features that the 1967-approved 
packages lack. Furthermore, NRC is aware that packagings built to the 
1967 standards were not subject to QA requirements, and that 
fabrication documents may not be available. NRC does not agree that 
this fact (lack of QA paperwork) enhances public confidence. Public 
confidence may be increased by removal of such packages from use in 
shipping. NRC will work closely with licensees who may have a source 
that has been impacted by the elimination of its package to ensure 
that, on a case-by-case basis, a suitable resolution is determined.
    Comment. One commenter stated that orphan sources should be 
considered in risk assessments and in assessing the costs and benefits 
of the proposed ban on 1967-specification containers. The commenter 
believes that when these factors are taken into consideration, they 
argue overwhelmingly against the proposed change.

[[Page 3738]]

    Response. The comment is acknowledged. The phase out of the Safety 
Series No. 6, 1967, packages will occur 4 years after adoption of the 
final rule. Thus, should orphan sources result as consequence of this 
rule, industry will have a minimum of 4 years to establish a program 
and a means to eliminate them from its inventory.
    Comment. One commenter stated that any modification of current 
requirements must not operate to prevent a device built to be 
transported in DOT Specification 20WC containers, and which has 
integral shielding and housing that is part of its ``packaging'' for 
regulatory purposes, from being shippable merely because it was not 
constructed fully under the part 71 QA rubric. The commenter warns that 
the device would become, overnight, an ``orphan source.''
    Response. Applicability of NRC QA requirements is specified in 
subpart H of part 71, including provisions for fabrication of 
packagings approved for use before January 1, 1979. Substantive 
technical changes to the QA provisions in part 71 are not being made as 
part of this rulemaking. Transport of packages that were built for the 
DOT Specification 20WC overpacks would require that the package, which 
includes the device within the overpack, be evaluated and certified to 
the new regulations after the 4-year phase-out period.
    Comment. One commenter stated that the U.S. is not bound to IAEA 
requirements for domestic shipping. The commenter notes that NRC and 
DOT have already deviated from the IAEA standards on other domestic-
only issues.
    Response. NRC acknowledges these comments and adds that the NRC has 
made a decision based on safety considerations not to deviate from the 
IAEA on the grandfathering issue for packages. Thus, the NRC will move 
forward to phase out those packages approved to Safety Series No. 6, 
1967.
    Comment. One commenter stated that both NRC and DOT have 
misassessed the impact of their proposals on small entities protected 
by the Regulatory Flexibility Act, 5 U.S.C. 601 et seq. The commenter 
stated that NRC fails to consider the many small entities that would be 
adversely impacted by phasing out the 1967-specification packages. The 
commenter also disagreed with DOT's argument that international 
uniformity will help small entities by the discarding of dual systems 
of regulation. The commenter noted that in the U.S., unlike in Europe, 
many firms do not have to deal with international shipping at all. The 
commenter disagreed with DOT's argument that the proposed phase-in 
period of 2 years would provide a smooth transition to the NRC approval 
process. The commenter believes that the 2-year window was not 
adequate.
    Response. The NRC acknowledges these comments. This commenter was 
the only small entity that made comments on this issue. Therefore, it 
is not clear to the NRC that many small entities would be adversely 
affected by this phase out. Further, NRC has made a decision based on 
safety considerations not to deviate from the IAEA on the 
grandfathering issue for packages. The NRC will move forward to phase 
out those packages over a 4-year period after adoption of the final 
rule. This time period should allow all businesses to assess their 
particular packages and either have them phased out or resubmit them to 
the NRC for review to the current standards. (The NRC staff notes that 
DOT has also decided to adopt a 4-year transition period for DOT 
specification packages.)
    Comment. One commenter stated that there is no reason to compel 
removal of properly inspected, properly maintained 1967-specification 
packages from service for U.S. domestic shipments of special form Type 
B quantities of radioactive material. The commenter argued that 
requiring owners and users to inspect and maintain older packages, or 
to convert to newer packages, would ensure safety. The commenter 
concurred that it is reasonable to ban further construction of 1967-
specification packages.
    Response. The packages approved to Safety Series No. 6, 1967, may 
lack the safety enhancements possessed by post-1967 approved packages. 
Thus, NRC will phase out these packages over a 4-year period including 
production of new packages to these old standards. Alternatively, 
owners and users of older packages have the opportunity to submit an 
application showing that the design, or a modified design, meets the 
current regulations. Recertification of these designs then would allow 
continued fabrication of additional packagings.
    Comment. One commenter stated that NRC and DOT should not subscribe 
to the useful lifetime limitations for shipping packages implicit in 
the IAEA's intended biennial review of its regulations. The commenter 
stated that the cost of such forced obsolescence on an ongoing basis 
would raise the cost of transportation unwarrantedly.
    Response. NRC believes that those packages approved to Safety 
Series No. 6, 1967, do not reflect the current safety standards. Thus, 
these packages will be eliminated over a 4-year period after adoption 
of a final rule. NRC does not anticipate that the future biennial 
changes within IAEA standards will be as significant as the changes 
found in the 1996 TS-R-1 standards. Therefore, based on the summary of 
the impact that will occur on various packages (see 67 FR 21406; April 
30, 2002), NRC will move forward with the elimination of certain 
packages for radioactive material transport.
    Comment. One commenter noted that there is a potential for 
substantial delay in approving new designs or recertifying existing 
designs. The commenter stated that any ``sunset'' deadline on the use 
of any package design being phased out under this proposal should 
permit its continued use pending an ultimate decision by the NRC on 
either recertification of the existing design or approval of a new 
design, as long as (1) a good-faith, substantially complete application 
for approval or recertification, as the case may be, has been filed 
with the NRC at least 12 months before the nominal ``sunset date'' on 
use of the existing design; and (2) the application for approval or 
certification is clearly related in the application to a design which 
is subject to the ``sunset'' provision.
    Response. The NRC has published guidance for applicants to use 
regarding package approval. The purpose of the guidance is to document 
practices used by NRC staff to review applications for package 
approval. This guidance is available in NUREG-1609, ``Standard Review 
Plan for Transportation Packages for Radioactive Material,'' and NUREG-
1617, ``Standard Review Plan for Transportation Packages for Spent 
Nuclear Fuel.'' Using this guidance will assist applicants to prepare a 
suitable application which will facilitate NRC review and ensure that 
such a review is concluded in a timely fashion. Note that these NUREG 
documents are available full-text on the NRC Web site (www.nrc.gov/NRC/
NUREGS/indexnum.html). Regarding the ``sunset'' issue, note that 
eliminating the 1967 packages was first published by IAEA in 1996 
(i.e., 7 years ago) and that the international regulations were 
implemented 5 years later in 2001. Industry should be aware of pending 
changes or possible changes based on IAEA rules. Therefore, including 
an additional 4-year implementation period (i.e., to 2008 (at least)) 
makes at least 12 years that industry has had the opportunity to 
evaluate its package designs, identify designs that may not meet the 
new standards, and prepare for the eventual phase out. The commenter is 
essentially requesting another year of

[[Page 3739]]

use while the paperwork is in review. NRC does not agree with this 
approach.
    Comment. One commenter asserted that if a specific ``sunset'' date 
is chosen, it should be significantly longer than the ones proposed by 
either NRC or DOT to date. The commenter also requested that NRC and 
DOT should agree on a common ``sunset'' date.
    Response. The NRC and DOT have adopted a suitable transition date 
for eliminating packages approved to Safety Series No. 6, 1967. Both 
agencies believe that a 4-year phase-out period is adequate.
    Comment. One commenter urged that the NRC allow for a substantially 
longer transitional time than now proposed. The commenter argued that 
the time necessary to design, fabricate, test, and complete NRC's 
review of a new CoC design would be much greater than the 2-year 
transition period proposed by DOT. The commenter stated that this would 
cause a shipping hiatus.
    Response. The NRC published the issues paper at 65 FR 44360; July 
17, 2000, which indicated the position on the issues associated with 
compatibility with the IAEA on many different issues, including 
grandfathering of those packages approved to Safety Series No. 6, 1967 
(see Issue 8). Thus, as a minimum, industry has been aware of the 
overall proposed impact of phasing out the 1967-approved packages for 
quite some time. Both NRC and DOT believe that a 4-year phase out 
period provides adequate time for industry to phase out old packages, 
phase in new packages, or demonstrate that current requirements are 
met. The 4-year phase out will commence with the adoption of the final 
rule.
    Comment. One commenter supported grandfathering casks made for the 
1967 standards for domestic shipping and urged NRC to retain the 
A2 value for molybdenum-99 and the A1 and 
A2 values for californium-252. The commenter also stated 
that the package identification number should be revised to the 
appropriate identification number prefix together with a suffix of ``-
96'' provided that such packages shall be for domestic use only and no 
additional packages shall be fabricated.
    Response. The NRC acknowledges the comments about grandfathering 
and A1 and A2 values for domestic shipping. For 
the comment about the package identification number, the NRC does not 
agree with this comment (see earlier response and response below).
    Comment. One commenter stated that the unique 1967-packages that 
cannot be easily replaced should not be replaced. The commenter 
supported the general concept of phasing out older packages and agreed 
that use of most 1967-certified packages should be discontinued. The 
commenter discussed the high costs of requalifying packages as ruinous 
for some businesses. The commenter argued that this would result in 
many orphan sources.
    Response. The NRC will move forward to phase out the Safety Series 
No. 6, 1967, packages that may not have the built-in safety 
enhancements that other (post-1967) packages maintain. The NRC will 
work in the future on a case-by-case basis with licensees who may have 
orphaned sources in their inventory as a result of this final rule.
    Comment. One commenter stated that if packages can be shown to meet 
the proposed regulations, the package identification number should be 
revised to the appropriate identification number prefix together with a 
suffix of ``-96'' provided that such packages shall be for domestic use 
only and no additional packages be fabricated.
    Response. The NRC staff disagrees with this comment. Inasmuch as 
this would allow continued use of B( ) packages for domestic use, NRC 
has determined that only those packages that have enhanced safety 
features (i.e., post-1967 package designs) will be allowed to be used 
and manufactured beyond the 4-year phase-out period for all use 
(domestic and international). When a package design is designated as B( 
) (i.e., approved to Safety Series No. 6, 1967) and is submitted to NRC 
for review to the current standards, the NRC may revise the package 
identification number to designate the package design as B, B(U), B(M), 
etc, and may assign the ``-96'' suffix.
Issue 9. Changes to Various Definitions
    Summary of NRC Final Rule. The final rule adopts the TS-R-1 
definition of Criticality Safety Index (CSI). NRC believes this 
provides internal consistency and compatibility with TS-R-1. 
Additionally, the following definitions have been revised to improve 
their clarity and maintain consistency with DOT: A1, 
A2, Consignment, LSA-I, LSA-II, LSA-III, and Unirradiated 
uranium. NRC believes that terms must be clearly defined so that they 
can be used to accurately communicate requirements to licensees. By 
modifying existing definitions and adding new definitions, the licensee 
would benefit through more effective understanding of the requirements 
of part 71.
    Affected Sections. Section 71.4.
    Background. The changes implemented by NRC in this rulemaking 
require changes to various definitions in Sec. 71.4 to provide internal 
consistency and compatibility with TS-R-1. These terms must be clearly 
defined so that they can be used to accurately communicate requirements 
to licensees. By modifying existing definitions and adding new 
definitions, the licensee benefits from a more effective understanding 
of the requirements of part 71.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Four commenters generally supported the proposal. One 
commenter specifically asked that NRC and DOT agree on the definition 
of ``common terms'' before issuance of the final rules.
    Response. The DOT and the NRC continue to coordinate rulemaking 
efforts to ensure regulatory consistency.
    Comment. One commenter stated that `` `Radioactive materials' and 
`contamination' should not be redefined as presented in the draft rule; 
the new definitions would expand exemptions and the deregulation and 
recycling of more nuclear materials and wastes.'' Another commenter 
expressed concern over the omission of a definition for 
``contamination.'' See response to comment on non-fixed contamination 
below.
    Response. The comments appear to be addressing a DOT concern, as 
NRC has not proposed to adopt a definition for ``contamination'' in 
this rulemaking. Currently, NRC regulations in Sec. 71.87(i) refer to 
the contamination levels found in DOT regulations. The NRC notes that 
contamination levels/concerns are not criteria for packaging approval 
within part 71. Rather, they are a factor in safe transport of an 
actual package of radioactive material.
    Comment. One commenter stated that the definition of ``person'' as 
stated in Sec. 70.4 should be included under Sec. 71.4 so it is clear 
that entities such as DOE are not a person under proposed Sec. 71.0(e).
    Response. The NRC does not agree with this comment. ``Person'' is 
defined within each part of Title 10. It is only these entities who 
would make shipments of radioactive material under part 71. Therefore, 
the NRC will rely on the existing definitions to support the 
transportation activities found in part 71.
    Comment. Three commenters stated that the definition of LSA-I and 
LSA-II should agree with the proposed DOT definition. One commenter 
provided specific information in objection to the proposed definitions 
of LSA-I and LSA-II.

[[Page 3740]]

    Response. NRC agrees that the definitions for LSA-I and LSA-II 
should be consistent between the NRC and DOT regulations. Therefore, 
NRC modified its regulations appropriately in Sec. 71.4 and changed the 
definitions for LSA-I and LSA-II to agree with the definitions found in 
DOT's final rule. Additionally, NRC noted that DOT adopted the TS-R-1 
definition for LSA-III material. To maintain consistency between these 
regulations, NRC also adopted DOT's definition for LSA-III.
    Comment. One commenter stated that defining only the containment 
system is broad enough to include the confinement system, because 
defining them differently will be confusing.
    Response. NRC acknowledges the comment.
    Comment. Three commenters were concerned about the omission of a 
definition for ``consignment.'' One commenter suggested that NRC use 
the definition provided in the DOT proposed rule.
    Response. NRC is adding a definition for ``consignment'' in Sec. 
71.4 that is consistent with DOT.
    Comment. Two commenters were concerned about the omission of a 
definition for ``unirradiated uranium.''
    Response. NRC is adding a definition for ``unirradiated uranium'' 
to Sec. 71.4 that is consistent with DOT.
    Comment. Two commenters stressed the importance of including the 
definition of ``non-fixed contamination.''
    Response. NRC disagrees. Section 71.87(i) refers to the nonfixed 
(removable) contamination regarding the contamination levels found in 
DOT regulations in 49 CFR 173.443, Table 11. NRC notes that the 
definition of ``nonfixed contamination'' has been removed from Sec. 
173.403 in DOT's rule. Furthermore, the definition of contamination 
from TS-R-1, including the definitions for fixed and nonfixed 
contamination, have also been added to Sec. 173.403 in DOT's proposed 
rule. Contamination controls are not a function of NRC package approval 
as much as they are a factor in safe transport of a package. Thus, it 
is appropriate to define contamination in DOT's regulations, but not in 
the NRC's.
    Comment. One commenter supported the proposed adoption of the 
specified definitions, and also urged NRC to adopt the TS-R-1 
definitions for confinement system, consignment, contamination, fixed 
contamination, nonfixed contamination, shipment, and transport index. 
The commenter also stated that NRC defined LSA-I differently from DOT, 
and that NRC and DOT should ensure compatibility between the rules.
    Response. See response to the previous comments in this issue. NRC 
agrees that the definition of ``transport index (TI)'' should be 
consistent between NRC and DOT regulations. Therefore, NRC modified 
Sec. 71.4 to include a definition for TI that is consistent with DOT. 
NRC does not agree, however, with the comment to adopt the TS-R-1 
definition of TI, as the definition adopted provides more clarity and 
explanation for the applicability of the TI.
Issue 10. Crush Test for Fissile Material Package Design
    Summary of NRC Final Rule. The final rule adopts, in Sec. 71.73, 
the TS-R-1 requirement for a crush test for fissile material package 
designs and eliminated the 1000 A2 criterion, but maintained 
the current part 71 testing sequence and drop and crush test 
requirements.
    By adopting TS-R-1, the weight and density criteria will apply to 
fissile uranium material packages, and packages that were previously 
exempted because of the 1000 A2 criterion will now require 
crush testing. Adopting crush test requirements and eliminating the 
1000 A2 criterion is appropriate because not adopting the 
TS-R-1 requirements would result in an inconsistency between part 71 
requirements and TS-R-1, which could affect international shipments, 
and fissile material package designs would continue to not be evaluated 
for criticality safety against a potential crush test accident 
condition.
    The NRC did not adopt the TS-R-1 test sequence requirements because 
no new information existed to address concerns from a previous 
rulemaking regarding the difference in test requirements between 
essentially the same IAEA requirements contained in Safety Series No. 6 
and part 71. The NRC chose to remain more conservative than the IAEA by 
requiring both a drop and crush test, rather than one or the other as 
TS-R-1 would permit.
    Affected Sections. Section 71.73.
    Background. The crush test requirements in TS-R-1 were broadened to 
apply to fissile material package designs (regardless of package 
activity). Previously, IAEA Safety Series No. 6 and part 71 required 
the crush test for certain Type B packages. This broadened application 
was created in recognition that the crush environment was a potential 
accident force that should be protected against for both radiological 
safety purposes (packages containing more than 1000 A2 in 
normal form) and criticality safety purposes (fissile material package 
design).
    Under requirements for packages containing fissile material, TS-R-
1, paragraph 682(b), requires tests specified in paragraphs 719-724 
followed by whichever of the following is the more limiting:
    (1) The drop test onto a bar as specified in paragraph 727(b) and 
either the crush test as indicated in paragraph 727(c) for packages 
having a mass not greater than 500 kg (1100 lbs) and an overall density 
not greater than 1000 kg/m3 (62.4 lbs/ft3) based 
on external dimensions, or the 9-meter (30-ft) drop test as defined in 
paragraph 727(a) for all other packages; or
    (2) The water immersion test as specified in paragraph 729.
    Both Safety Series No. 6, paragraph 548, and current Sec. 71.73 
require the crush test for packages having a mass not greater than 500 
kg (1100 lbs), an overall density not greater than 1000 kg/
m3 (62.4 lbs/ft3) based on external dimensions, 
and radioactive contents greater than 1000 A2 not as special 
form radioactive material. Under TS-R-1, the criterion for radioactive 
contents greater than 1000 A2 was eliminated for packages 
containing fissile material. The 1000 A2 criterion still 
applies to Type B packages and is also applied to the IAEA newly 
created Type C package category.
    Full compliance with TS-R-1 requirements for fissile material would 
require changes to the hypothetical accident conditions test sequencing 
of Sec. 71.73 and would require performance of the 9-meter (30-ft) free 
drop test or the crush test, but not both, as presently required by 
Sec. 71.73. The TS-R-1 test requirements are essentially the same as 
those contained in Safety Series No. 6 (1985 edition). NRC addressed 
the difference between Safety Series No. 6 and Sec. 71.73 in a previous 
rulemaking and concluded that the two tests evaluate different features 
of a package, and both tests are necessary to determine whether a 
package response is within applicable limits (final rule, 60 FR 50248; 
Sept. 28, 1995).
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter stated that the additional cost of the crush 
test for fissile material is estimated at about $5,000,000. This cost 
is to design, certify, and manufacture replacement packages currently 
in use for the shipment of uranium oxide. The commenter thought that 
currently three

[[Page 3741]]

to five packages are in use that will need to be modified and 
recertified.
    Response. The information provided by the commenter was considered 
in the development of NRC's rule.
    Comment. One commenter recounted how they were almost crushed under 
``a boulder the width of the highway in the Wyoming Wind River Range 
some years ago'' and stated that ``No vehicle or container could have 
withstood the impact of that boulder's fall from several hundred feet 
above.'' The commenter also stated that based on such probable events, 
crush tests must be mandatory, with the cost borne by licensee or user. 
The commenter added that the NRC needs to implement more rigorous crush 
and drop tests than its current standard so that it can ensure 
container survival in the event of severe accidents. The commenter also 
recommended that because the TS-R-1 document was not readily available, 
it was ``ingenuous, at best, for the NRC to give the references to the 
actual testing requirements in terms of TS-R-1 paragraph citations.''
    Response. The recommendation to implement more rigorous crush and 
drop tests than the current regulatory standards to ensure container 
survival for severe accidents is noted, but was not justified, and is 
outside the scope of the current rulemaking. Further, it should be 
noted that TS-R-1 is readily available online at: http://
www.pub.iaea.org/MTCD/publications/pdf/Pub1098_scr.pdf.
    Comment: Three commenters advocated more stringent testing 
procedures. Specifically, one commenter stated support for NRC's effort 
to adopt crush tests for all fissile material packages regardless of 
size or activity (while rejecting the IAEA's option of choosing to 
perform either a drop or a crush test on a container). The commenter 
also urged the NRC to use a physical (as opposed to a simulating test 
using computer modeling) crush test with a full-size package to provide 
a realistic testing environment. The commenter suggested that the NRC's 
proposal should include all containers, including the DT-22 (which 
failed the dynamic crush test) and the 9975 container (which failed the 
30-foot drop test). Further, it was noted that the redesigned 9975 
container has not yet been ``crush tested to show the results of high-
speed impact against an unyielding surface.'' For this unit, the 
commenter urged NRC to require a physical, as opposed to a simulated, 
crush test with a full-size package to provide a realistic testing 
environment. The commenter also stated that the NRC needs to require 
other testing and noted that ``neither the DT-22 nor the 9975 have been 
sufficiently tested against fire.'' Also, the commenter contended that 
the current test (i.e., burn at 1475 degrees Fahrenheit for 30 minutes) 
ignores the fact of ``more than 20 materials routinely transported on 
highways that burn at more than twice this temperature.'' Two 
commenters suggested that this heat test be made more stringent and 
realistic. NRC also needs to test these two containers for ``durability 
to terrorist attack with a variety of weapons, such as mortars or anti-
tank missiles, under a variety of conditions.'' Furthermore, ``all Type 
B containers should be subject to rigorous testing for terrorist 
resistance.''
    Another commenter expressed concern that the proposed rule would 
allow the DP-22 package to be licensed and approved, despite the fact 
that it does not meet either the drop or crush test requirements.
    Another commenter expressed concern that crush testing is not 
required for packages having a mass greater than 500kg, which includes 
rail SNF waste packages. The commenter suggested that the NRC ``require 
rail transportation casks be subject to crush testing (scaled up to 
produce impact energies of the magnitude expected in a railway 
accident).'' The commenter cited a 1995 report entitled ``Rail 
Transportation of Spent Nuclear Fuel--A Risk Review'' that argued small 
packages are shipped in large numbers and ``as a result demonstrate a 
higher possibility of experiencing crush loads than large packages 
would.'' In addition, the commenter cited how packages transported by 
North American rail would have a high probability of experiencing 
dynamic crushing in an accident.
    Response. The comment regarding more rigorous testing for all Type 
B packages for terrorist resistance is noted. Please refer to the 
second comment in Section II, under the heading: Terrorism Concerns. 
The comment regarding stringency of heat tests is noted but is outside 
the scope of the current rulemaking. With respect to comments regarding 
the DT-22 and 9975 container, NRC staff is not familiar with these 
designs as they are used within the DOE program and are authorized 
under DOE's package approval authority. These containers do not 
currently have an NRC CoC. The NRC staff also is not familiar with the 
DP-22 design that the commenter alludes to as it does not currently 
have an NRC CoC. To receive an NRC CoC, it would have to meet the NRC's 
testing requirements, including drop and crush test if required.
    The comment regarding crush testing for packages greater than 500 
kg (1100 lb) is acknowledged. The NRC has already gone beyond the IAEA 
testing requirements in requiring that all Type B packages subject to 
the crush test must also be subjected to the free drop test. Extending 
the crush test to other Type B packages (i.e., those exceeding 500 kg 
(1100 lbs)) is beyond the scope of the current rulemaking.
    Regarding the comment on requiring physical crush testing, rather 
than simulated tests, and the use of full scale packages for physical 
testing, the NRC staff believes that the use of computer code analysis 
of finite element models and the use of scale models for physical 
testing are valid methods for demonstrating compliance with the NRC's 
package testing requirements. It should be noted that these methods 
should be NRC approved.
    Comment. Three commenters questioned the requirements for both a 
drop test and a crush test. One commenter requested that if both a 
crush test and a drop test are required on packages that meet the 
requirements for the crush test, the rules should specify that this 
could be carried out on two different packages. The commenter explained 
that it does not make sense to require both tests for the same package, 
because in an accident scenario, a single package would not experience 
both conditions.
    Two commenters stated that packages should either pass a drop test 
or the crush test, but not both. The first commenter said that the rule 
should state that separate packages should be used for each test, and 
that the same package should not be used to pass both tests in 
sequence. The second commenter said that, ``A line for deciding which 
test a package should undergo could be based on the gross weight of the 
package.''
    Response. The current requirements under Sec. 71.73(a) state that: 
``Evaluation for hypothetical accident conditions is to be based on 
sequential application of the tests specified in this section, in the 
order indicated, to determine their cumulative effect on a package or 
array of packages.'' However, Sec. 71.73(a) does specifically allow for 
an undamaged specimen to be used for the immersion test of Sec. 
71.73(c)(6). NRC staff is aware that IAEA regulations do not require 
both the free drop and crush test on a single specimen, but has chosen 
to remain more conservative in this regard. In the NRC rulemaking for 
compatibility with IAEA Safety Series No. 6 (September 28, 1995; 60 FR 
50248), NRC staff stated the position that: ``NRC is requiring both the 
crush test and drop

[[Page 3742]]

test for lightweight packages to ensure that the package response to 
both crush test and drop forces is within applicable limits.'' NRC 
staff is not aware of any new information that would cause NRC to 
deviate from that position.
    NRC staff does not agree with the commenter's assertion that 
performing a drop and crush test is a double drop test. In the drop 
test from 9 meters (30 feet), the specimen itself is dropped onto an 
unyielding surface; in the crush test (if required by both the package 
weight and density criteria), a 500-kg (1100-lb) weight is dropped from 
9 meters (30 feet) onto the specimen. These are two independent tests 
that may have different outcomes depending on the package and the 
location where maximum damage is expected to occur for each test.
    Comment. Two commenters supported NRC's proposal regarding crush 
test requirements. One commenter expressed support for the NRC's 
proposal to accept the part of IAEA's rule change under TS-R-1 which 
requires a crush test for fissile material packages regardless of size 
or activity while rejecting the IAEA's option of performing either 
crush or drop tests of containers.
    Response. No response is necessary.
Issue 11. Fissile Material Package Design for Transport by Aircraft
    Summary of NRC Final Rule. The final rule adopts TS-R-1, paragraph 
680, Criticality evaluation, in a new Sec. 71.55(f) that only applies 
to fissile material package designs that are intended to be transported 
aboard aircraft. Section 71.55 specifies the general package 
requirements for fissile materials, and the existing paragraphs of Sec. 
71.55 are unchanged. Among other requirements, TS-R-1, paragraph 680, 
requires that packages must remain subcritical when subjected to the 
tests for Type C packages, because:
    (1) The NRC has deferred adoption of the Type C packaging tests 
(see Issue 6);
    (2) TS-R-1, paragraph 680 requires Type C tests; and
    (3) Paragraph 680 applies to more than Type C packages; only the 
salient text of paragraph 680 was inserted into Sec. 71.55(f) and 
applies to domestic shipments.
    Adopting this change will provide regulatory consistency. Shippers 
would have been required to meet the TS-R-1 air transport requirements 
even if the NRC did not adopt them, because the International Civil 
Aviation Organization had adopted regulations consistent with TS-R-1 on 
July 1, 2001. U.S. domestic air carriers require compliance with the 
ICAO regulations even for domestic shipments. Therefore, these changes 
are expected to benefit industry by eliminating the need for two 
different package designs.
    Affected Sections. Section 71.55.
    Background. TS-R-1 introduced new requirements for fissile material 
package designs that are intended to be transported aboard aircraft. 
TS-R-1 requires that shipped-by-air fissile material packages with 
quantities greater than excepted amounts (which would include all NRC-
certified fissile packages) be subjected to an additional criticality 
evaluation.
    In TS-R-1, paragraph 680, requirements for packages to be 
transported by air are in addition to the normal condition and accident 
tests that the package must already meet. Thus:
    Type A fissile package by air must:
    (1) Withstand normal conditions of transport with respect to 
release, shielding, and maintaining subcriticality (single package and 
5xN array; \1\
---------------------------------------------------------------------------

    \1\ N represents the maximum number of fissile material packages 
that can be shipped on a single conveyance.
---------------------------------------------------------------------------

    (2) Withstand accident condition tests with respect to maintaining 
subcriticality single package and 2xN array); and
    (3) Comply with TS-R-1, paragraph 680, with respect to maintaining 
subcriticality (single package);
    Type B fissile package by air must:
    (1) Withstand normal conditions of transport and Type B tests with 
respect to release, shielding, and maintaining subcriticality (single 
package and 5xN array/normal and 2xN array/accident); and
    (2) Comply with TS-R-1, paragraph 680, with respect to maintaining 
subcriticality.
    TS-R-1, paragraphs 816 and 817, state that fissile package designs 
intended to be transported by aircraft are not allowed to be 
grandfathered. Consequently, all of these fissile package designs will 
be evaluated before their use.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Four commenters supported the NRC's position on this 
issue. One commenter supported NRC's proposal to ensure consistent 
review of package designs affected by the requirements of the 
International Civil Aviation Organization. Another commenter said 
adoption of Type C packages should be scheduled for future 
harmonization with IAEA regulations.
    Response. The NRC believes the changes create a uniform regulatory 
framework for the review of package designs for both national and 
international air shipments.

B. NRC-Initiated Issues

Issue 12. Special Package Authorizations
    Summary of NRC Final Rule. The final rule adopts, in Sec. 71.41, 
special package authorizations that will apply only in limited 
circumstances and only to one-time shipments of large components. 
Special package authorization regulations are necessary because there 
are no regulatory provisions in part 71 for dealing with nonstandard 
packages, other than the exemption provisions and Sec. 71.41(c). The 
NRC processing of one-time exemptions for nonstandard packages, such as 
the Trojan reactor vessel, has required the expenditure of considerable 
NRC resources. Further, the NRC's policy is to avoid the use of 
exemptions for recurring licensing actions. Special package 
authorization requirements will result in enhanced regulatory 
efficiency by standardizing the requirements to provide greater 
regulatory certainty and clarity, and will ensure consistent treatment 
among licensees requesting authorization for shipment of special 
packages.
    Any special package authorization will be issued on a case-by-case 
basis, and requires the applicant to demonstrate that the proposed 
shipment would not endanger life or property nor the common defense and 
security, following the basic process used by applicants to obtain a 
CoC for nonspecial packages from NRC.
    The applicant will be required to provide reasonable assurance that 
the special package, considering operational procedures and 
administrative controls employed during the shipment, would not 
encounter conditions beyond those for which it had been analyzed and 
demonstrated to provide protection. The NRC will review applications 
for special package authorizations. Approval will be based on NRC staff 
determination that the applicant will meet the requirements of subpart 
D of 10 CFR part 71. If approved, the NRC will issue a CoC or other 
approval (i.e., special package authorization letter).
    NRC will consult with DOT on making the determinations required to 
issue an NRC special package authorization.
    Affected Sections. Section 71.41.
    Background. The basic concept for radioactive material 
transportation is that radioactive contents are placed in

[[Page 3743]]

an authorized container, or packaging, and then shipped. The packaging, 
together with its contents, is called the package. In general, the 
transportation regulations in TS-R-1, 10 CFR part 71, and 49 CFR are 
based on the shipment of radioactive contents in a separate, authorized 
packaging. There are a few exceptions. In cases involving larger 
quantities of radioactive material, the content to be shipped may 
itself be a container. A storage tank containing a radioactive residue 
is an example. It is not necessary for the shipper to place the tank 
within an authorized packaging if the shipper demonstrates that the 
tank satisfies the requirements for the packaging. DOT and NRC have 
jointly provided guidance on such shipments (see ``Categorizing and 
Transporting Low Specific Activity Materials and Surface Contaminated 
Objects,'' NUREG-1608, RAMREG-003, July 1998).
    As older nuclear facilities are decommissioned, DOT and NRC are 
being asked to approve the shipment of large components, including 
reactor vessels and steam generators. These components may contain 
significant quantities of radioactive material, but they are so large 
that it may not be practical to fabricate authorized packagings for 
them. Because the potential shipment of these components was not 
contemplated when the NRC transportation regulations were developed, 
the regulations do not specifically address them.
    Large components can be shipped under DOT regulations if the 
components meet the definition of Surface Contaminated Object (SCO) or 
Low Specific Activity (LSA) material (see 49 CFR 173.403 for SCO and 
LSA definitions). For example, steam generators that meet the DOT SCO 
definition are exempt from part 71 and are shipped under 49 CFR, 
following guidance provided in NRC Generic Letter 96-07 dated December 
5, 1996. This method has been applied to several shipments of steam 
generators and small reactor vessels to the low level waste disposal 
facility at Barnwell, SC. NRC and DOT intend to continue employing this 
approach and method for steam generators and similar components that 
can be shipped under DOT regulations.
    Large components that exceed the SCO and LSA definitions are 
subject to part 71. An example is the Trojan reactor vessel which was 
transported to the disposal facility on the Hanford Nuclear Reservation 
near Richland, Washington. The Trojan Reactor Pressure Vessel (TRPV) 
contained approximately 74 PBq (2 million Ci) in the form of activated 
metal and 5.7 TBq (155 Ci) in the form of internal surface 
contamination, and was filled with low-density concrete, and weighed 
approximately 900 metric tons (1,000 tons). Normally, large curie 
contents are required to be shipped in a Type B packaging, but the TRPV 
was too large and massive to be shipped within another packaging.
    Section 71.8 provides that NRC may grant any exemption from the 
requirements of the regulations in part 71 that it determines is 
authorized by law and will not endanger life or property nor the common 
defense and security.
    Currently, no regulatory provisions exist in part 71 for dealing 
with nonstandard packages, other than the exemption provisions and Sec. 
71.41(c). The NRC's practice is to avoid the use of exemptions for 
recurring licensing actions. The new rule language will support this 
practice.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter stated that relaxation of requirements 
applicable to large packages could potentially reduce the cost of these 
shipments for parties who must routinely demonstrate that all 
shipments, including reactor vessels and larger reactor compartments, 
are made in compliance with part 71. However, the commenter asked that 
the NRC relax the restriction that a special package authorization may 
be approved only for ``one-time shipments'' and allow a limited number 
of shipments to be approved if they are of the same design to avoid 
repetitious certification requests.
    Response. The NRC believes that standardizing the special package 
authorization process will increase efficiency during the review of 
large shipment components. These special packages were not provided for 
specifically in earlier regulations. Establishing a standard process 
for authorization also will reduce the regulatory burden associated 
with shipping these packages. The NRC envisions the process for special 
package authorization to be similar to authorization for Type B 
packages, with specific criteria for approval judged on a case-by-case 
basis. The special package authorization is not intended for repeat or 
routine shipments of components. It is reserved for those unique 
instances where traditional packaging and approval methods are 
impractical. Therefore, NRC is not extending special package 
authorizations to multiple shipments of the same component.
    Comment. One commenter opposed NRC's proposal to allow special 
package exemptions stating that it would not be a responsible action by 
NRC and could lead to further requests to loosen regulatory 
restrictions in the future. The commenter cited the precedent of 
Shippingport, Trojan, and Yankee Rowe as reason for the concern. The 
commenter further stated that post-September 11, 2001, NRC ``should not 
assume the legality or safety of any exemptions from full packaging 
container requirements.'' The commenter added that the TS-R-1, 
paragraph 312, ``is not in the public interest and should be changed'' 
and NRC should not allow this decision to remain with DOT. The 
commenter stated that NRC itself admits that DOT uses altered 
definitions to justify transporting special (large) components without 
the amount of protection demanded of lesser components; this is 
unacceptable and a failure by NRC to exercise its mandated 
responsibility. The commenter also requested the NRC to provide a 
definition of ``reasonable assurance.''
    This commenter further stated that the ``shortcoming of dual 
regulation is evident in the handoff of regulatory control from one 
agency to another'' and added that it is unacceptable ``for NRC to wash 
its hands of its responsibility for packaging and containers by handing 
over authority to another agency.'' The commenter then asked if NRC 
planned this as ``merely a cost reduction for licensees,'' and stated 
that NRC needed to provide a justification for this proposal. The 
commenter also questioned the safety of these shipments.
    The commenter also stated that the NRC's focus on high-level waste 
transport would result in the NRC ignoring allowances for exemptions 
for lower activity materials and wastes. This would result in these 
materials and wastes passing from a ``regulated status to exemption and 
release into commerce or unregulated `disposal' and would `increase 
risks to the public that NRC ignores.' The commenter ended by stating 
that this ``is not an acceptable deregulation, is a capricious failure 
to protect the general welfare, and is therefore contrary to law'' and 
reiterated the ``objection to NRC's reliance on `performance-based risk 
informed' regulation that permits less stringent requirements for 
containment and for transportation.''
    Response. The special package authorization does not reduce the 
protection of public health and safety;

[[Page 3744]]

rather, it affects the process used to approve nonstandard packages. 
The special package authorization requirement clearly states that the 
overall safety in transport for shipments approved under special 
package authorization will be at least (emphasis added) equivalent to 
that which would be provided if all applicable requirements had been 
met. The NRC is not adding a definition for the term ``reasonable 
assurance'' because it is not used in a regulatory requirement.
    It is important to repeat that NRC approval will be required for 
special package authorizations. In addition, DOT regulations will be 
modified to recognize NRC's special package authorizations. The process 
efficiencies offered by special package authorizations result in more 
effective and efficient regulation.
    The special package authorization will reduce the need for 
exemptions in the package approval process and will not result in the 
disposal of radioactive material.
    Comment. One commenter stated that the Trojan reactor shipment 
should not be used as a precedent for special package approval. The 
commenter reasoned that the Trojan reactor shipment was an easy 
shipment due to its origin and destination.
    Response. The NRC believes the Trojan reactor vessel shipment 
indicates there is a need for special package approvals because it 
represents a class of contents that, due to their size, mass, or other 
unique factors, are impractical to transport within standard 
radioactive material packaging. The origin and destination of the 
Trojan shipment has no bearing on this rule.
    Comment. One commenter requested more information about how the NRC 
is going to approve special packages. The commenter stated that a 
better explanation of this process would aid regulated bodies in 
acquiring special package authorization.
    Another commenter indicated that with the current proposal, ``the 
special package authorization is not bounded and applicants do not have 
a common basis for preparation of an application'' and requested that 
the NRC staff establish general criteria against which special packages 
can be evaluated.
    One commenter suggested that NRC establish general criteria for the 
special package authorization process.
    One commenter stated that the ``special package'' designator should 
be clearly defined in terms of package size or other appropriate 
feature to ensure that the rule is applied correctly.
    Response. The purpose of this change is to establish general 
criteria for the authorization of special package designs without the 
need for the licensee to request an exemption from the current 
regulations. The NRC agrees that additional information on special 
package approvals is needed. NRC intends to develop regulatory guidance 
in this area before this rule is implemented. In the interim, any 
applications for special package approvals will be considered on a 
case-by-case basis.
    Comment. One commenter requested the NRC to view every shipment of 
a reactor vessel as a significant process requiring National 
Environmental Policy Act (NEPA) review. The commenter argued that a 
NEPA process would allow for public input in the process of 
decommissioning a reactor vessel.
    Response. A NEPA review will not be required for the new special 
package authorizations. Package approvals authorized by our regulations 
are specifically excluded from the requirement to prepare an EA 
pursuant to NEPA (Sec. 51.22(c)(13)). In contrast, an EA for the Trojan 
reactor vessel was thought to be necessary because the NRC did not rely 
on specific package approval regulations, but rather relied on an 
exemption from those requirements.
    Comment. One commenter suggested that shipping retired reactor 
vessels should be a separate issue from the exception process.
    Response. The NRC disagrees that reactor vessels should be excluded 
from special package authorization. The NRC believes reactor vessels 
are an example of the type of shipment that would benefit from special 
package authorization, because the authorization would follow a more 
standardized and efficient design review process. NRC's package design 
review process has been shown to provide adequate protection of public 
health and safety.
    Comment. One commenter stated that no additional limitations should 
be applied to the conditions under which one could apply for a package 
authorization. The commenter noted that the few packages that have been 
authorized have moved without incident and without undue risk to the 
public, workers, or the environment.
    Response. Comment noted. No response necessary.
    Comment. Five commenters supported the proposed provisions in Sec. 
71.41(d) for special package authorizations. Two of these commenters 
stated that this revision provides a consistent approach to dealing 
with the transport of large pieces of equipment and nonstandard items, 
and that the revision would improve the safety and cost effectiveness 
of onsite and offsite transfers of large equipment items. Two other 
commenters supported corresponding with DOT to eliminate duplicitous 
exemptions, but urged the NRC to work closely to ensure the clear 
implementation of this proposal.
    Response. No response necessary.
Issue 13. Expansion of Part 71 Quality Assurance (QA) Requirements to 
Certificate of Compliance (CoC) Holders
    Summary of NRC Final Rule. The final rule adds the terms 
``certificate holder'' and ``applicant for a CoC'' to subpart H, part 
71 and adds a new section, Sec. 71.9, on employee protection. Adopting 
these requirements will ensure that the regulatory scheme of part 71 
will remain more consistent with other NRC regulations in that 
certificate holders and applicants for a CoC will be responsible for 
the behavior of their contractors and subcontractors.
    This expansion is necessary to enhance NRC's ability to enforce 
nonconformance by the certificate holders and applicants for a CoC. 
Although CoC's are legally binding documents, certificate holders and/
or applicants and their contractors and subcontractors have not clearly 
been brought into the scope of part 71 requirements. This is because 
the terms ``certificate holder'' and ``applicant for a certificate of 
compliance'' do not appear in part 71, subpart H; rather, subpart H 
only mentions ``licensee'' in these regulations. Consequently, the NRC 
has not had a clear basis to cite applicants for, and holders of CoC's 
for violations of part 71 requirements in the same way it has 
licensees.
    The NRC also added a new section (Sec. 71.9) on employee protection 
to part 71. The NRC believes that employee protection regulations 
should be added to cover the employees of certificate holders and 
applicants for a CoC to provide greater regulatory equivalency between 
part 71 licensees and certificate holders.
    Affected Sections. Sections 71.0, 71.1, 71.6, 71.7, 71.8 , 71.9, 
71.91, 71.93, 71.100, and 71.101 through 71.137.
    Background. On October 15, 1999 (64 FR 56114), the Commission 
issued a final rule to expand the QA provisions of part 72, subpart G, 
to specifically include certificate holders and applicants for a CoC. 
In a Staff Requirements Memorandum (SRM) to SECY-97-214, the Commission 
directed the staff to consider whether conforming changes to the QA 
regulations in part 71 would be necessary because of the existence of 
dual-purpose cask designs.

[[Page 3745]]

    The 1999 rule requires that Part 72 licensees, certificate holders, 
and applicants for a CoC are responsible for assuring that their 
contractors and subcontractors (e.g., fabricators) are implementing 
adequate QA programs. Similarly, by this final rule, part 71 licensees, 
certificate holders, and applicants for a CoC are responsible under 
Sec. 71.115 for assuring that their contractors and subcontractors 
(e.g., fabricators) are implementing adequate QA programs.
    Under part 71, the NRC reviews and approves applications for Type B 
and fissile material packages for the transport of radioactive 
material. The NRC's approval of a package is documented in a CoC. 
Applicants for a CoC are currently required by Sec. 71.37 to describe 
their QA program for the design, fabrication, assembly, testing, 
maintenance, repair, modification, and use of the proposed package. 
Further, existing Sec. 71.101(a) describes QA requirements that apply 
to design, purchase, fabrication, handling, shipping, storing, 
cleaning, assembly, inspection, testing, operation, maintenance, 
repair, and modification of components of packagings that are important 
to safety. Type B packages are intended to transport radioactive 
material that contains quantities of radionuclides greater than the 
A1 or A2 limits for each radionuclide (see 
Appendix A to part 71 for examples of A1 or A2 
limits). Fissile material packages are intended to transport fissile 
material in quantities greater than the part 71, subpart C, general 
license limits for fissile material (e.g., existing Sec.Sec. 71.18, 
71.20, 71.22, and 71.24).
    Although CoCs are legally binding documents, certificate holders or 
applicants for a CoC and their contractors and subcontractors have not 
clearly been brought into the scope of part 71 requirements. This is 
because the terms ``certificate holder'' and ``applicant for a 
certificate of compliance'' do not appear in part 71, subpart H; 
rather, subpart H only mentions ``licensee'' in these regulations. 
Consequently, the NRC has not had a clear basis to cite certificate 
holders and applicants for a CoC for violations of part 71 requirements 
in the same way it has licensees.
    When the NRC has identified a failure to comply with part 71 QA 
requirements by certificate holders or applicants for a CoC, it has 
issued a Notice of Nonconformance (NON) rather than a Notice of 
Violation (NOV). Although an NON and an NOV appear to be similar, the 
Commission prefers the issuance of an NOV because:
    (1) The issuance of an NOV effectively conveys to both the person 
violating the requirement and the public that a violation of a legally 
binding requirement has occurred;
    (2) The use of graduated severity levels associated with an NOV 
allows the NRC to effectively convey to both the person violating the 
requirement and the public a clearer perspective on the safety and 
regulatory significance of the violation; and
    (3) Violation of a regulation reflects the NRC's conclusion that 
potential risk to public health and safety could exist. Therefore, the 
NRC believes that limiting the available enforcement sanctions to 
administrative actions is insufficient to address the performance 
problems observed in industry.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Five commenters supported the NRC's proposed position on 
this issue. One commenter recommended that NRC establish and apply a 
uniform set of QA requirements. Another commenter added that it would 
like to see the consistent application of QA requirements throughout 
the regulations.
    Response. Expansion of the QA provisions enhances NRC's ability to 
enforce noncompliance and will ensure broader, uniform application of 
QA requirements. However, extension of the requirement beyond part 71 
is outside the bounds of this rulemaking.
Issue 14. Adoption of the American Society of Mechanical Engineers 
(ASME) Code
    Summary of NRC Final Rule. The NRC has decided not to incorporate 
the ASME Code, section III, division 3 requirements into part 71. 
Public Law 104-113 requires that Federal agencies use consensus 
standards in lieu of government-unique standards, if this use is 
practical or inconsistent with other existing laws. Because a major 
revision to the ASME Code is forthcoming and because the changes in 
that revision are not yet available for staff and stakeholder review, 
the NRC staff considered it an imprudent use of NRC and stakeholder 
resources to initiate rulemaking on the current ASME Code revision only 
to have the ASME Code requirements change during the part 71 
rulemaking.
    Affected Sections. None (not adopted).
    Background. Currently, no ASME Code requirements exist in part 71 
for fabrication/construction of spent fuel transportation packages. The 
NRC considered the adoption of the ASME Boiler and Pressure Vessel 
(B&PV) Code, section III, division 3, for two reasons. First, previous 
NRC inspections at vendor and fabricator shops (for fabrication of 
spent fuel storage canisters and transportation casks) identified 
quality control (QC) and QA problems. Some of these problems would have 
been prevented with improved QA programs, and may have been prevented 
had fabrication occurred under more prescriptive requirements such as 
the ASME Code requirements. Second, Public Law 104-113, ``National 
Technology Transfer and Advancement Act,'' enacted in 1996, requires 
that Federal agencies use, as appropriate, consensus standards (e.g., 
the ASME B&PV Code), except when there are justified reasons for not 
doing so.
    With respect to conformance to Public Law 104-113, the ASME issued 
a consensus standard in May 1997, entitled: ``Containment Systems and 
Transport Packages for Spent Fuel and High Level Radioactive Waste,'' 
ASME B&PV Code, section III, division 3. The ASME Code requires the 
presence of an Authorized Nuclear Inspector during construction to 
ensure that the ASME Code requirements are met and the stamping of 
components (i.e., the transportation cask's containment) constructed to 
the ASME Code. NRC staff participated, and continues to participate, in 
the ASME subcommittee that developed the ASME Code requirements. It is 
the NRC staff's understanding, through participation in the 
subcommittee, that the ASME Code document is undergoing extensive 
review and modification and that a major revision will be issued. 
Therefore, NRC staff believes that inclusion of the ASME Code in part 
71 is not appropriate at this time.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Four commenters expressed support for the decision not to 
adopt the ASME code. One commenter said that these are voluntary 
standards and should not be made into requirements.
    Response. No response is required.
Issue 15. Change Authority for Dual-Purpose Package Certificate Holders
    Summary of NRC Final Rule. The Commission does not reach a final 
decision on the issue of change authority for dual-purpose package

[[Page 3746]]

certificate holders in this final rule. The NRC has determined that 
implementation of this change would result in new regulatory burdens 
and costs which could be significant. The Commission believes it needs 
further input from stakeholders on the values and impacts of this 
change before deciding whether to adopt a final rule providing change 
authority for dual-purpose package certificate holders. The NRC staff 
plans to conduct public meetings with appropriate stakeholders to 
develop a final regulatory solution which it will propose to the 
Commission. At that time, the Commission will either issue a final rule 
resolving this issue, taking into account the comments received on the 
proposed rule and in any future public meetings, or will withdraw 10 
CFR part 71 subpart I of the proposed rule.
    Affected Sections. None.
    Background. The Commission approved a final rule to expand the 
provisions of Sec. 72.48, ``Changes, Tests, and Experiments,'' to 
include part 72 certificate holders and licensees (64 FR 53582; October 
4, 1999). Part 72 certificate holders and licensees are allowed, under 
Sec. 72.48, to make certain changes to a spent fuel storage cask's 
design or procedures used with the storage cask and to conduct tests 
and experiments without prior NRC review and approval. Part 71 does not 
contain any similar provisions to permit a CoC holder to change the 
design of a part 71 transportation package, without prior NRC review 
and approval. The NRC has issued separate CoC's under parts 71 and 72 
for dual-purpose spent fuel storage casks and transportation packages. 
This has created a situation where an entity holding both a part 71 and 
a part 72 CoC would be allowed under part 72 to make certain changes to 
the design of a dual-purpose cask (i.e., changes that affected a 
component or design feature that has a storage function) without 
obtaining prior NRC approval. However, the entity would not be allowed 
under part 71 to make changes to the design of this same dual-purpose 
cask (package) if that component or feature also has a transportation 
function without obtaining prior NRC approval, even when the same 
physical component and change are involved (i.e., the change involves a 
component that has both storage and transportation functions).
    NRC staff recognized a need to consider making both part 72 and 
part 71 more consistent in dealing with design changes of a minor 
nature. Thus, in SECY-99-054,\2\ NRC staff recommended that an 
authority similar to Sec. 72.48 be created for dual-purpose spent fuel 
storage casks and transportation packages intended for domestic use 
only. NRC staff also recommended that this authority be limited to the 
part 71 CoC holder.
---------------------------------------------------------------------------

    \2\ SECY-99-054; February 22, 1999, ``Plans for Final Rule-
Revisions to Requirements of 10 CFR parts 50, 52, and 72 Concerning 
Changes, Tests, and Experiments.''
---------------------------------------------------------------------------

    Since the proposed rule was published, the NRC has evaluated 
comments received from the public and has conducted a detailed analysis 
of the implementation of the change authority, as proposed. Based on 
this analysis, the NRC has determined not to finalize subpart I, Type 
B(DP) Package Approval, as proposed. Instead, the NRC will seek further 
input on the values and impacts of this change and then decide whether 
to proceed with a final rule.
    Proposed Sec. 71.153 stated that the application for a Type B(DP) 
package shall include an analysis of potential accidents, package 
response to these potential accidents, and any consequences to the 
public. Currently, under part 71, an applicant has to demonstrate, 
either by test or analysis, that a package design can withstand the 
cumulative effects of the Hypothetical Accident Conditions of a 30-foot 
drop test, a 40-inch puncture test, a thermal test, and immersion tests 
as described in Sec. 71.73 and Sec. 71.61, and meet Subpart E--Package 
Approval Standards. Applicants are not required to perform an 
independent analysis of potential transportation accidents specific to 
that design and plans for use, project package responses to ``real 
world'' transportation accidents, or determine the consequences to the 
public from such accidents.
    The NRC reviewed and considered the comments that were received 
about this proposed change. The new process included the need to 
establish a design specific accident assessment for the cask design 
response to potential ``real world'' transportation accidents. Such an 
accident analysis has not been required for a transportation cask 
application before. Which accidents would be appropriate, for which 
routes, under what conditions, for what duration, and with what 
combinations of forces and assumptions, all would be questions that 
would need to be answered by CoC applicants who have not been required 
to perform such analysis for cask designs applications.
    To provide new guidance for the development of an acceptable 
accident analysis for a transportation cask, the NRC staff would need 
to perform significant research on what types of accidents would be 
required to be included. The NRC believes that such an analysis can be 
performed; however, the NRC does not believe that it had fully 
considered in the proposed rule the rigor, resources, and time that 
such a requirement would require. The detailed associated cost 
estimates had not been included in the RA for this part of the rule 
change. The RA has been revised, and the costs of implementation for 
CoC holders could be significantly higher than that reflected in the 
proposed rulemaking. This additional regulatory burden had not been 
accurately reflected in the draft RA. The Safety Analysis Report (SAR) 
for part 71 applications is based, in part, on demonstrating compliance 
with the Hypothetical Accident Conditions of part 71. Thus, there is 
not a clear linkage between the SAR and regulatory conditions for 
making changes to a design without NRC approval, such as a minimal 
increase in the probability of an accident sequence or the creation of 
accidents of a different type. Given these revised cost estimates, the 
NRC is uncertain whether the benefits to be gained from this change 
outweigh the costs. The NRC intends to explore this issue further 
before deciding whether to proceed to a final rule.
    The proposed Sec. 71.175, ``Changes,'' establishes methods to 
determine if a proposed change to a Type B(DP) package can be made 
without prior NRC approval. As stated in a public comment, the language 
in this section mirrors that in Sec. 72.48. It should be noted that the 
design and application process under part 72 does require that an 
applicant perform an accident analysis as part of its application for 
approval, but such a requirement has never been incorporated into part 
71 as noted above.
    The intent of subpart I was to allow a certificate holder 
flexibility to make minor changes to the design of the package to be 
consistent with the change authority provided under Sec. 72.48 for 
spent fuel storage casks in a cost and time effective manner. The NRC 
notes that transportation CoCs issued under part 71 do allow for many 
changes to be made to package designs without NRC approval, provided 
the changes do not impact upon compliance with part 71 standards. For 
example, changes in the SAR for a transportation package, in general, 
do not require NRC approval provided the changes do not affect the 
conditions listed in the CoC or the ability of the package to meet the 
requirements of part 71. Additionally,

[[Page 3747]]

packaging design drawings that are included as conditions in the CoC do 
not need to specify fabrication details that are not important to 
safety. In this way, changes may be made to nonsafety features without 
modifying the drawings and without NRC review and approval. This is in 
contrast to the approaches for part 72 CoCs. It is therefore important 
that applications for package approval, including packaging design 
drawings, are developed to focus on the safety features of the design. 
The NRC notes that the current regulatory process for evaluating and 
approving CoC amendments for transportation packaging may be more 
efficient than developing a new regulatory infrastructure. To aid in 
receiving high quality transportation applications, the NRC staff is 
preparing an amended standard format and content regulatory guide.
    The NRC has determined that implementation of the proposed change 
process would result in new regulatory burdens and costs which could be 
significant. The NRC also recognizes the concerns of public commenters 
related to the potential benefits of allowing changes to the design of 
a Type B(DP) package without prior NRC approval. The NRC staff will 
work with appropriate stakeholders to determine whether a final rule is 
the preferred method for resolving the need for a change process in 
part 71 or whether there may be other regulatory solutions that meet 
this need. The NRC staff will then propose a final regulatory solution 
to the Commission. The Commission will then determine if subpart I 
should be issued as a final rule or if other regulatory solutions to 
this issue obviate the need for going forward with a final rule. If a 
final rule is not needed, then proposed subpart I will be withdrawn and 
the comments received on this issue will be addressed at that time.
Issue 16. Fissile Material Exemptions and General License Provisions
    Summary of NRC Final Rule. The final rule adopts various revisions 
to the fissile material exemptions and the general license provisions 
in part 71 to facilitate effective and efficient regulation of the 
transport of small quantities of fissile material. The fissile 
exemptions (Sec. 71.15) have been revised to include controls on 
fissile package mass limit combined with package fissile-to-nonfissile 
mass ratio. The general license for fissile material (Sec. 71.22) has 
been revised to consolidate and simplify current fissile general 
license provisions from Sec.Sec. 71.18, 71.20, 71.22, and 71.24. Under 
the final rule, the general license is based on mass-based limits and 
the CSI. In light of comments and applicable DOT requirements, the 
final rule removes proposed rule language references to ``storage 
incident to transportation.'' Also, the exemptions for low level 
materials in Sec. 71.14 were revised to apply only to nonfissile and 
fissile-exempt materials.
    Affected Sections. Sections 71.4, 71.10, 71.11, 71.18, 71.20, 
71.22, 71.24, 71.53, 71.59, and 71.100. (Currently effective Sec. 71.10 
was relocated to Sec. 71.14 with additional language. Currently 
effective Sec.Sec. 71.18, 71.20, 71.22, 71.24, and 71.53 are replaced 
by new Sec.Sec. 71.15 and 71.22.)
    Background. The NRC published an emergency final rule amending its 
regulations on shipments of small quantities of fissile material (62 FR 
5907; February 10, 1997). This rule revised the regulations on fissile 
exemptions in Sec. 71.53 and the fissile general licenses in Sec.Sec. 
71.18 and 71.22. The NRC determined that good cause existed, under 
section 553(b)(B) of the Administrative Procedure Act (APA) (5 U.S.C. 
553(b)(B)), to publish this final rule without notice and opportunity 
for public comment. Further, the NRC also determined that good cause 
existed, under section 553(d)(3) of the APA (5 U.S.C. 553(d)(3)), to 
make this final rule immediately effective. Notwithstanding the final 
status of the rule, the NRC provided for a 30-day public comment 
period. The NRC subsequently published in the Federal Register (64 FR 
57769; October 27, 1999) a response to the comments received on the 
emergency final rule and a request for information on any unintended 
economic impacts caused by the emergency final rule.
    The NRC issued this emergency final rule in response to a 
regulatory defect in the fissile exemption regulation in Sec. 71.53 
which was identified by an NRC licensee. The licensee was evaluating a 
proposed shipment of a special fissile material and moderator mixture 
(beryllium oxide mixed with a low concentration of high-enriched 
uranium). The licensee concluded that while Sec. 71.53 was applicable 
to the proposed shipment, applying the requirements of Sec. 71.53 
could, in certain circumstances, result in an inadequate level of 
criticality safety (i.e., an accidental nuclear criticality was 
possible in certain unique circumstances).\3\
---------------------------------------------------------------------------

    \3\ For transportation purposes, ``nuclear criticality'' means a 
condition in which an uncontrolled, self-sustaining, and neutron-
multiplying fission chain reaction occurs. ``Nuclear criticality'' 
is generally a concern when sufficient concentrations and masses of 
fissile material and neutron moderating material exist together in a 
favorable configuration. Neutron moderating material cannot achieve 
criticality by itself in any concentration or configuration. 
However, it can enhance the ability of fissile material to achieve 
criticality by slowing down neutrons or reflecting neutrons.
---------------------------------------------------------------------------

    The NRC staff confirmed the licensee's analysis that this beryllium 
oxide and high-enriched uranium mixture created the potential for 
inadequate criticality safety during transportation. An added factor in 
the urgency of the situation was that under the NRC regulations in 
Sec.Sec. 71.18, 71.20, 71.22, 71.24, and 71.53, these types of fissile 
material shipments could be made without prior approval of NRC. For 
many years, NRC allowed these shipments of small quantities of fissile 
material based on NRC's understanding of the level of risk involved 
with these shipments, as well as industry's historic transportation 
practices. This experience base had led NRC (and its predecessor, the 
Atomic Energy Commission (AEC)) to conclude that shipments made under 
the fissile exemption provisions of part 71 typically required minimal 
regulatory oversight (i.e., NRC considered these types of shipments to 
be inherently safe).\4\
---------------------------------------------------------------------------

    \4\ The NRC's regulations in part 71 ensure protection of public 
health and safety by requiring that Type AF, B, or BF packages used 
for transportation of large quantities of radioactive materials be 
approved by the NRC. This approval is based upon the NRC's review of 
applications which contain an evaluation of the package's response 
to a specific set of rigorous tests to simulate both normal 
conditions of transport (NCT) and hypothetical accident conditions 
(HAC). However, certain types of packages are exempted from the 
testing and NRC prior approval; these are fissile material packages 
that either contain exempt quantities (Sec. 71.53), or are shipped 
under the general license provisions of Sec.Sec. 71.18, 71.20, 
71.22, or 71.24.
---------------------------------------------------------------------------

    All public comments on the emergency final rule supported the need 
for limits on special moderators (i.e., moderators with low neutron-
absorption properties such as beryllium, graphite, and deuterium). 
However, the commenters stated that the restrictions were far too 
limiting (to the point that some inherently safe packages were excluded 
from the fissile exemption) and could lead to undue cost burdens with 
no benefit to safety. In addition, the commenters believed that the 
consignment mass limits set to deter undue accumulation of fissile mass 
would be extremely costly. Therefore, the commenters recommended that 
further rulemaking was necessary to resolve these excessive 
restrictions. Based on the public comments on the emergency final rule, 
NRC staff contracted with Oak Ridge National Laboratory (ORNL) to 
review the fissile

[[Page 3748]]

material exemptions and general license provisions, study the 
regulatory and technical bases associated with these regulations, and 
perform criticality model calculations for different mixtures of 
fissile materials and moderators. The results of the ORNL study were 
documented in NUREG/CR-5342,\5\ and NRC published a notice of the 
availability of this document in the Federal Register (63 FR 44477; 
August 19, 1998). The ORNL study confirmed that the emergency final 
rule was needed to provide safe transportation of packages with special 
moderators that are shipped under the general license and fissile 
material exemptions, but the regulations may be excessive for shipments 
where water moderation is the only concern. The ORNL study recommended 
that NRC revise part 71.
---------------------------------------------------------------------------

    \5\ NUREG/CR-5342, ``Assessment and Recommendations for Fissile-
Material Packaging Exemptions and General Licenses Within 10 CFR 
Part 71,'' July 1998.
---------------------------------------------------------------------------

    In the October 27, 1999 (64 FR 57769) final rule, the Commission 
requested additional information on the cost impact of the emergency 
final rule from the public, industry, and DOE because the NRC staff was 
not successful in obtaining this information. Specifically, NRC 
requested information on the cost of shipments made under the fissile 
material exemptions and general license provisions of part 71, before 
the publication of the emergency final rule, and those costs and/or 
changes in costs resulting from implementation of the emergency rule. 
One commenter agreed with the NRC approach but stated that, ``the 
limits for those materials containing no special moderators can and 
should be increased, hopefully back to their pre-emergency rule 
levels.''
    As part of NUREG/CR-5342, ORNL performed computer model 
calculations of keff (k-effective) for various combinations 
of fissile material and moderating material, including beryllium, 
carbon, deuterium, silicon-dioxide, and water, to verify the accuracy 
of current minimum critical mass values. These minimum critical mass 
values were then applied to the regulatory structure contained in part 
71, and revised mass limits for both the general license and exemption 
provisions to part 71 were determined. Also, ORNL researched the 
historical bases for the fissile material exemption and general license 
regulations in part 71 and discussed the impact of the emergency final 
rule's restrictions on NRC licensees. ORNL concluded that the 
restrictions imposed by the emergency final rule were necessary to 
address concerns relative to uncontrolled accumulation of exempt 
packages (and thus fissile mass) in a shipment and the potential for 
inadequate safety margin for exempt packages with large quantities of 
special moderators.
    Based on its new keff calculations, ORNL suggested that: 
(1) The mass limits in the general license and exemption provisions 
could be safely increased and thereby provide greater flexibility to 
licensees shipping fissile radioactive material; and (2) additional 
revisions to part 71 were appropriate to provide increased 
clarification and simplification of the regulations. Copies of NUREG/
CR-5342 may be obtained by writing to the Superintendent of Documents, 
U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-
9328. A copy is also available for inspection and copying, for a fee, 
at the NRC Public Document Room in the NRC Headquarters at One White 
Flint North, Room O-1F21, 11555 Rockville Pike, Rockville, MD 20852-
2738.
    The current restrictions on fissile exempt and general license 
shipments under Sec.Sec. 71.53, and 71.18 through 71.24, respectively, 
are burdensome for a large number of shipments that actually contain no 
special moderating materials (i.e., packages that are shipped with 
water considered as the potential moderating material). This problem 
was clearly expressed in public comments on the emergency final rule. 
Another regulatory problem is that the current fissile exempt and 
general license provisions are cumbersome and outdated; this was one of 
the main conclusions of the ORNL study.
    The NRC proposed changes (67 FR 21417) were made on the basis of 17 
recommendations contained in NUREG/CR-5342. These changes included: (1) 
Revising Sec. 71.10, ``Exemption for low level materials,'' to exclude 
fissile material, also redesignate Sec. 71.10 as Sec. 71.14; (2) 
redesignating Sec. 71.53 as Sec. 71.15, ``Exemption from classification 
as fissile material,'' and revise the fissile exemptions; (3) 
consolidation of the existing four general licenses in existing 
Sec.Sec. 71.18, 71.20, 71.22, and 71.24 into one general license in new 
Sec. 71.22, revise the mass limits, and add Type A package, CSI, and QA 
requirements; and (4) consolidation of the existing general license 
requirements for plutonium-beryllium sealed sources, which are 
contained in existing Sec.Sec. 71.18 and 71.22 into one general license 
in new Sec. 71.23 and revise the mass limits. Additionally, changes 
were proposed to be made to Sec. 71.4, ``Definitions,'' and Sec. 
71.100, ``Criminal penalties.''
    The NRC also proposed: (1) To adopt the use of the CSI for general 
licensed fissile packages; and (2) to retain the current per package 
(CSI) limit of 10, rather than raising the per package limit to 50 (see 
Issue 5). TS-R-1 does not address the issue of fissile general 
licenses, so no compatibility issues arise with retention of the 
current NRC per package limit of 10. NRC staff believes that because 
reduced regulatory oversight is imposed on fissile general license 
shipments (e.g., the package standards of Sec.Sec. 71.71 and 71.73, 
fissile package standards of Sec. 71.55, and fissile array standards of 
Sec. 71.59 are not imposed for fissile general license shipments), 
retention of the current per package limit of 10 is appropriate. 
Furthermore, retention of the current per package limit of 10 would not 
impose a new burden on licensees; rather, licensees shipping fissile 
material under the general license provisions of Sec.Sec. 71.22 and 
71.23 would not be permitted to take advantage of the relaxation of the 
per package CSI limit from 10 to 50 that would be permitted for Types 
AF and B(F) package shipments.
    As a result of stakeholder meetings and public comments, the NRC 
has incorporated the following changes to the proposed language for 
Sec.Sec. 71.15 and 71.22 in the final rule:
    (1) Small quantities of fissile materials such as environmental 
samples shipped for testing are judged to be of sufficient low quantity 
that, if individually packaged, the risk (probability and consequence) 
of accumulating the number and type of packages needed to present a 
potential criticality hazard is judged to be inconsequential. 
Therefore, a new Sec. 71.15(a) has been added to exempt packages 
containing 2 grams or less fissile material.
    (2) Proposed Sec. 71.15(a) (Sec. 71.15(b) in the final rule) 
specifically referred to iron as the nonfissile material for 
calculating limiting ratio of 200:1. Commenters suggested that this 
would require a new definition (of iron) and would complicate 
implementation. There is no technical reason to require that iron be 
identified as the nonfissile materials to be included with a mass ratio 
of 200:1. Other nonspecial moderating materials such as stainless 
steel, concrete, etc., are appropriate. The mass ratio wording has been 
modified. The modification maintains the need for the mass ratio of 
200:1, but the required nonfissile material is required to be a solid. 
As worded, the nonfissile mass can include the packaging mass. It is 
judged that sufficient distribution of fissile material in small 
quantities (i.e., 1 g of fissile material per 200 g of solid nonfissile 
material) will provide adequate protection against nuclear

[[Page 3749]]

criticality. This specification ensures that large numbers of packages, 
containing 15 g of fissile material per package, will remain safely 
subcritical because of the fissile material dilution and density 
reduction by nonfissile materials which are not special moderators 
(e.g., beryllium, graphite, etc.). For example, 1 g of optimally 
moderated uranium-235 in a mixture at about 0.05 g Uranium-235/cm\3\ 
occupies a volume of about 20 cm\3\. Two hundred grams of aluminum 
metal at about 2.7 g of aluminum/cm\3\ occupies a volume of about 74 
cm\3\. As specified, the 15 g of uranium-235 per package will have a 
diluted volume of about 1,410 cm\3\ at a density of about 0.01 g 
uranium-235/cm\3\ and a density reduction by a factor of 5. Though 
aluminum is a minor absorber of low-energy neutrons, most other common 
materials of packaging have moderate neutron-absorbing properties that 
further ensure safely subcritical accumulations of such packages. The 
increase in the subcritical mass of 620 g of optimally moderated 
uranium-235, permitted by the reduction of fissile material density, is 
related to the ratio of the densities to the power of 1.8 (see Ref. 1 , 
pp. 19-22). Given the density reduction of 5 in the above example, the 
adjusted subcritical mass becomes 11,125 g of uranium-235, requiring in 
excess of about 741 packages (containing 15 g of uranium-235 per 
package) to exceed the determined equivalent quantity of material.
    (3) Proposed Sec. 71.15(b) (Sec. 71.15(c) in the final rule), was 
modified by referring to fissile and nonfissile materials as solid 
materials instead of using ``noncombustible'' and ``insoluble-in-
water.'' The modification was a pragmatic consideration and was made to 
avoid reference to the undefined/specified word, ``noncombustible,'' 
and the phrase, ``insoluble-in-water,'' while addressing the need to 
avoid fissile and nonfissile liquids/gases that easily could be 
consolidated or lost (thereby decreasing nuclear criticality safety) in 
normal and hypothetical accident transportation circumstances. An 
additional modification, Sec. 71.15(c)(2) in the final rule, also 
removes the limit of 350 g in a package and instead specifies criteria 
for commingling of the material such that, within any selected 360 kg 
of nonfissile solid material, there can be no more than 180 g of 
fissile material. Thus, a large rail car with a homogenized 
distribution of fissile material within a nonfissile waste matrix might 
exceed the 180 g limit but would be effectively mixed at low enough 
concentration to enable safe shipment.
    (4) The basis for Sec. 71.15(c)(1) is that a 2000:1 mass ratio of 
nonfissile to fissile material is [sim]60% of the minimum critical 
fissile material concentration of 1.33 g uranium-235/L in a 1,600 g 
SiO2/L matrix. The 60-percent value is judged to be a 
reasonably conservative decrease in g uranium-235/g nonfissile material 
(e.g., SiO2) to accommodate other nonfissile materials. The 
minimum critical fissile material concentration in SiO2 was 
derived from studies to compare ``special'' and ``natural'' neutron 
moderators with fissile materials. In those studies various systems 
were examined that had different species of fissile material (i.e., 
uranium-235, uranium-233, or plutonium-239) combined with water and 
other nonfissile neutron scatterers/moderators (e.g., polyethylene, 
beryllium, carbon, deuterium, and SiO2). SiO2 was 
selected for consideration in the transport exemptions because it is 
judged to be the most representative, arbitrary, and nonspecial 
moderator matrix for commingling with fissile material. SiO2 
has a very low probability for absorbing neutrons and has a large 
abundance in nature (i.e., 33 weight percent, second only to oxygen at 
49 weight percent). An independent study compared the relative 
importance of other elements to silicon with dilute fissile materials. 
Except for the category of special moderators (i.e., deuterium, 
beryllium, and graphite) and pure forms of magnesium (i.e., magnesium 
carbonate, magnesium fluoride, magnesium oxalate, magnesium oxide, 
magnesium peroxide, magnesium silicates) and bismuth (i.e., bismuth 
basic carbonate, bismuth tri-or penta-fluorides, bismuth oxide), 
silicon or silicon dioxide is the most neutronically reactive diluent 
for fissile materials. The 1.6-g SiO2/L is representative of 
dry bulk mean world soil density.
    (5) Section 71.15(d) (Sec. 71.15(c) in proposed rule) has been 
revised to reflect ``mass of beryllium, graphite, and hydrogenous 
material enriched in deuterium constitute less than 5 percent of the 
uranium mass'' (less than 0.1 percent of the fissile mass being the 
proposed phrase). This change was made in response to a comment about 
the difficulty that shippers would experience based on the proposed 
rule language. The staff reviewed the 0.1 percent of fissile mass 
language and determined that limiting the low-neutron-absorbing 
materials to the proposed ratio would be impractical to implement. The 
final language reflecting 5 percent of the uranium mass assures 
subcriticality for all moderators of concern and is less burdensome to 
measure and implement as a requirement.
    (6) Section 71.15(e) (Sec. 71.15(d) in the proposed rule) states 
``total plutonium and uranium-233 content not exceeding 0.002 percent 
of the mass of uranium'' while the proposed language stated ``does not 
exceed 0.1 percent of the mass of uranium-235.'' This change was made 
in response to a public comment that the proposed rule changes should 
be consistent with the international regulations. The final language 
for this section has been revised to be consistent with the 1996 IAEA 
standards.
    (7) Section 71.15(f) (proposed Sec. 71.15(e)) was reworded for 
clarity but reflects the same requirements and guidance as in the 
proposed language.
    (8) Proposed Sec. 71.22 (e)(5)(iii), Exemption from classification 
as fissile material, was revised to read `` * * * The uranium is of 
unknown Uranium-235 enrichment or greater than 24 weight percent 
enrichment; or * * * '' The reason for the Sec. 71.22(e)(5)(iii) 
modification was that enrichments of U-235 greater than 24 weight 
percent were not accommodated in the proposed text. Because the minimum 
critical mass transition between 24 and 100 weight percent enrichments 
of 235U vary slightly, the text was changed to require the 
use of Table 71-1 values for all enrichments greater than 24 weight 
percent as well as materials of unknown enrichments. The values in 
Table 71-1 were developed for 100 weight percent uranium-235 enriched 
uranium and are conservatively applied down to 24 weight percent 
uranium-235.
    (9) Proposed Sec. 71.22, Table 71-1, was modified in the final rule 
to replace uranium-235 (Y) with uranium-233 (Y)--change to uranium-233 
(Y). The reason is to correct a typographical error in the table.
    In the final rule, the NRC has deleted the phrase ``or stored 
incident to transport'' from proposed Sec.Sec. 71.22(d)(3) and 
71.23(d)(3). The intent of the storage phrase was to permit segregation 
of groups of stored packages, consistent with IAEA and DOT 
requirements, but the NRC staff believes that the proposed text did not 
accommodate that practice because it did not accommodate storage and 
segregation of groups of packages. DOT requirements properly restrict 
accumulation of packages during transport, based on summing the 
packages' CSI or TI, including during storage incident to transport. In 
light of the division of regulatory responsibilities explained in the 
NRC-DOT Memorandum of Understanding (44 FR 38690; July 2, 1979), the 
NRC

[[Page 3750]]

exemptions for carriers-in-transit in Sec. 70.12, and DOT's revision to 
49 CFR 173.457 (67 FR 21384), the NRC staff believes that storage in 
transit provisions as proposed in Sec.Sec. 71.22(d)(3) and 71.23(d)(3) 
are unnecessary.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter noted that this is a significant deviation 
from the TS-R-1 requirement, which now has a 15-g uranium-235 limit as 
well as a mass consignment limit.
    Response. On February 10, 1997 (62 FR 5907), the NRC published a 
final rule on fissile exemptions. That final rule essentially adopted 
the 1996 TS-R-1 requirements, including the 15-g per package limit and 
400-g consignment mass limit. Both the consignment mass limit (400 g ) 
and the package mass limit (15 g) were used to control package 
accumulations. In consideration of comments received on the 1997 rule, 
the NRC has proposed changes to the fissile exemptions; one of the 
principal concerns with the 1997 rule was the practicability of the 
350-g consignment mass limit (see 67 FR 21418; April 30, 2002). The 
proposed rule suggested a mass ratio system together with the per 
package limit to eliminate this consignment mass limit. The IAEA is 
currently considering changes to the current international regulations 
in the area of the fissile material exemptions.
    Comment. Three commenters indicated that this provision would 
overly complicate the shipping of fissile material and negatively 
impact intermodal and international shipping. One commenter noted that 
the three-tiered system would dramatically complicate the shipping of 
fissile material because the mass ratio requirement makes it difficult 
to determine how to classify UF6 into the three tiers. This 
same commenter stated that companies that ship internationally will 
have a difficult time complying with the proposed system as well as the 
international system and suggested that NRC simplify compliance for 
these companies. The other commenter stated that if NRC's proposal is 
adopted as written, shippers would need to have detailed information 
available regarding the materials in each packaging. The commenter 
reasoned that this approach assumes that the detailed information would 
be readily available and disseminated to shippers, and further, 
shippers making international shipments would likely need to meet both 
NRC's domestic requirements for determining fissile exempt quantities 
and the international mass consignment limits, thus further 
complicating the evaluation of criticality controls for a shipment.
    Response. The NRC staff believes that the changes are warranted to 
alleviate the unnecessary regulatory burden created by the 1997 
emergency final rule, including the consignment mass limit. The changes 
implemented by the 1997 rule are essentially the same as TS-R-1. These 
amendments permit greater flexibility for domestic transport, in 
consideration of the comments received when the U.S. adopted the TS-R-1 
approach in 1997. However, NRC recognizes that international transport 
will also need to comply with IAEA TS-R-1, and the burden has been 
unchanged. The IAEA is currently considering changes to the current 
international regulations in the area of the fissile material 
exemptions. The NRC staff did review the proposed language for the 
proposed Sec. 71.15(c) and determined that the 0.1 percent ratio of the 
mass of beryllium, graphite, and hydrogenous material enriched in 
deuterium to the total fissile mass was a requirement that was 
difficult to implement and therefore the language has been changed as 
noted above in the rule language description.
    Comment. Several commenters expressed concern about material 
definitions, with one commenter noting that the definition of iron is 
unclear. One commenter requested clarification of what constitutes iron 
with regard to Tier 1 or fissile exempt quantities and specifically 
asked if steel is considered iron. Another stated that it is difficult 
to obtain information on materials to carry out the calculations under 
the proposed regulations.
    Response. Many materials have the neutronic properties that would 
permit them to be considered as the nonfissile material mass to be 
mixed with up to 15 g of fissile material in a ratio of 200:1. Iron, 
generic steels, stainless steels, and concrete are good examples of 
materials for use. Only lead, beryllium, graphite, and hydrogenous 
material enriched in deuterium should be excluded as noted in the 
revised text. The wording has been modified and clarified in the final 
rule.
    Comment. One commenter requested that the NRC explain why NRC 
proposes changing the total shipment CSI in cases where there is 
storage incident to transport, effectively doing away with an exclusive 
use condition. The commenter considered this proposal a significant 
change in the method of calculating the CSI per consignment and wanted 
to remind us that the proposed rule maintains segregation and storage 
requirements.
    Response. The ``storage incident to transport'' language has been 
deleted. See the comment responses under Issue 5.
    Comment. Two commenters said that NRC should clarify how the mass 
limits for general license packages (found in Sec. 71.22 (a)(3), Tables 
71-1 and 71-2) are used for uranium enriched greater than 24 percent. 
Both commenters stated that highly enriched uranium does not meet the 
criteria under Sec. 71.22(e)(5). Moreover, if uranium enriched greater 
than 24 percent cannot be shipped in a DOT 7A, this provision would 
have significant cost and operational impacts on the DOE.
    Response. Uranium enriched to greater than 24 percent can be 
shipped provided the appropriate X value from Table 71-1 is used in the 
equation to determine the CSI. The proposed rule had intended Sec. 
71.22(e)(3) to guide the reader to using Table 71-1 for uranium-235 
enrichments greater than 24 percent. However, the text for Sec. 
71.22(e)(5)(iii) has been revised to clarify the use of Table 71-1 for 
uranium-235 enrichments greater than 24 percent.
    Comment. Several commenters discussed the economic impact of the 
proposed regulation. Two commenters asserted that the regulation will 
cause an increase in the number of shipments required with an 
associated increase in costs, with one predicting required transports 
to increase two-to three-fold. Another warned of significant negative 
economic consequences if NRC did not retain the current provision for 
15 g per package, at least until it is demonstrated unsafe.
    Response. These comments appear to be concerned with the rule's 
restrictions on package accumulation based on CSI due to the ``storage 
incident to transport'' language in the proposed rule. The ``storage 
incident to transport'' language has been deleted. Also see the 
response to second comment under Issue 5.
    Comment. One commenter stated that ``under no circumstances should 
the NRC issue general licenses for shipments of radioactive materials 
and wastes (or, for that matter, for other purposes).'' The commenter 
then added that NRC shouldn't allow fissile materials to be exempted 
from packaging and transportation regulations nor should NRC allow 
``transport subject to even remotely possible criticality accidents 
during shipment'' under any circumstances. The commenter added that it 
is ``an outrage, furthermore, that the NRC had

[[Page 3751]]

approved an ``emergency final rule'' allowing shipments of fissile 
materials in 1997 without affording the public full opportunity for 
comment * * *'' The commenter cited NRC's footnote (see 67 FR 21418; 
April 30, 2002) and stated doubts regarding NRC's process for requiring 
NRC's approval for ``all Type AF, B, or BF packages.'' The commenter 
concluded by stating that ``NRC approval is virtually guaranteed in 
almost all cases, whether or not the decision contributes to public 
health and safety, not to mention the environment.''
    Response. The NRC staff believes that current regulations and 
programs for transporting fissile materials, and in particular the 
general licensing approach in part 71, result in a high degree of 
safety as evidenced by a long record of safe transport of these 
materials. The staff believes that a graded series of requirements for 
hazardous materials, including the fissile exemptions and general 
licenses, remains appropriate.
    Comment. Two commenters expressed concern about the use of the part 
110 definitions of ``deuterium'' and ``graphite'' in the proposed rule. 
The commenters suggested that NRC reconsider these definitions because 
they are inappropriate for the purpose of nuclear criticality safety.
    Response. The final rule stipulates that ``Lead, beryllium, 
graphite, and hydrogenous material enriched in deuterium may be present 
in the package, but must not be included in determining the required 
mass of solid nonfissile material.'' Materials enriched in deuterium 
and graphite are often termed special moderators because their very low 
neutron absorption properties give rise to special consideration for 
large systems with low concentration of fissile material and, 
therefore, warrant consideration in the criticality control approach. 
In the interests of consistency within NRC regulations, the NRC staff 
believes that the definitions of graphite and deuterium are sufficient 
for purposes of defining the materials that cannot be used in the Sec. 
71.15 determination.
    Comment. One commenter opposed the fissile material exemptions.
    Response. No response is necessary.
    Comment. Two commenters expressed general support for the fissile 
material exemptions. One of whom expressed support for the graduated 
exemptions for fissile material shipments because they would allow 
increasing quantities in shipments, provided that the packages also 
contained a corresponding increase in the ratio of non-fissile to 
fissile material. They also appreciated NRC consolidating four fissile 
material general licenses into one and consolidating existing general 
license requirements for PuBe sources into one section and updating the 
mass limits.
    Response. The comments are acknowledged. No further response is 
necessary.
    Comment. Several commenters requested that NRC include and/or 
improve various definitions in the proposed rule. One commenter stated 
that improved definitions were necessary to categorize the ratio 
calculations.
    Three commenters added that NRC should not exclude the definition 
of ``shipment'' from the rule. Another suggested that the proposed rule 
was ambiguous as to whether iron in the packaging (e.g. internal 
structure) can be used to meet the 200:1 ratio requirement in the 15-g 
exception.
    Two commenters noted that the proposed rule did not include a 
definition for ``insoluble in water,'' one of whom stated that the 
proposed rule fails to clarify the issue in part because of the 
rulemaking's lack of clarity. This same commenter questioned NRC's 
decision to omit definitions for ``consignment'' and ``shipment'' and 
urged NRC to adopt the TS-R-1 definition for these terms.
    Response. The NRC staff believes the terms ``ratio'' and 
``calculations'' are sufficiently clear without corresponding 
definitions. The terms ``iron in the packaging'' and ``insoluble in 
water'' have been deleted from the rule. Because of its bearing upon 
the fissile exemptions rule, a definition of ``consignment'' that is 
consistent with the definition in DOT's corresponding rulemaking has 
been added to the final rule language. The NRC staff does not believe a 
definition of the common-usage term shipment is warranted.
    Comment. One commenter noted that Sec. 71.15(b) does not identify 
what standard is to be used in applying either the term 
``noncombustible'' or the term ``insoluble-in-water.'' The commenter 
stated that if this section is kept as proposed, there is a need to 
clarify the terms and specify an appropriate standard.
    Response. The text from the proposed rule has changed. Rather than 
clarify the words ``noncombustible'' and ``insoluble-in-water,'' the 
new text indicates only the need for the nonfissile material to be a 
``solid.'' The NRC believes that new definitions are not necessary.
    Comment 13. One commenter requested that NRC delete the proposed 
exemptions for plutonium-244 in proposed Sec. 71.14(b)(1) because there 
are no special form plutonium-244 sources available.
    Response: Section 71.14(b)(1) was changed to provide clarification 
and simplification of the language that existed in the current 
regulation (Sec. 71.10), while retaining the substance of the 
exemption. The current Sec. 71.10 (b)(1) exempts shipments that contain 
no more than a Type A quantity of radioactive material from all of the 
requirements of part 71, except for Sec.Sec. 71.5 and 71.88. Similarly, 
Sec. 71.10(b)(3) exempts domestic shipments that contain less than an 
aggregate 20 Curies (Ci) of special form americium or plutonium from 
all of the requirements of part 71, except for Sec.Sec. 71.5 and 71.88. 
The current Type A (A1) limit for plutonium-244 is 8 Ci. The 
rule raises the A1 limit for plutonium-244 to 11 Ci--still 
less than the 20-Ci exemption of the current Sec. 71.10(b)(3). 
Consequently, for plutonium-244, the two exemption criteria of the 
current Sec. 71.10(b)(1) and (b)(3) were in conflict. The NRC's 
proposed rule resolved that conflict. The commenter's proposed solution 
would retain that conflict. Accordingly, absent a substantive basis for 
changing the proposed rule, the NRC is retaining the existing 20-Ci 
exemption for domestic shipments of special form americium or plutonium 
in Sec. 71.14(b)(1) in this final rule. Furthermore, because the 
A1 limits for all other nuclides of plutonium are greater 
than 20 Ci, only plutonium-244 is mentioned in paragraph (b)(1).
    Comment. Two commenters asserted that the regulations are overly 
complex and inconsistent with international regulations. One commenter 
agreed with NRC's proposal to change the requirements for fissile 
material shipments, but did have several objections. The three primary 
objections were that NRC hadn't adequately defined the terms to 
categorize the ratio calculations; information on the materials, 
necessary to perform calculations, is difficult to obtain; and the 
proposal is overly complex and inconsistent with international 
regulations. This same commenter stated that the proposed rule does not 
adequately account for both packages of large volume and packages of 
small volume. The proposed changes do not provide for the ability to 
ship large volumes of decommissioning waste in an effective manner and 
will complicate international trade of fissile exempt materials. 
Furthermore, the proposed ratio control is inadequate, and NRC should 
define ``insoluble in water.'' The commenter recommended inclusion of 
the TS-R-1 provisions for fissile exempt

[[Page 3752]]

materials. Lastly, the commenter stated that, while NRC should go 
forward with the rulemaking, it should work with industry to determine 
operational limits that will assure that the mass or concentration 
limit is maintained under accident conditions.
    Response. The staff has reviewed the proposed rule language and has 
determined that section Sec. 71.15(d) was not consistent with the 
language in TS-R-1 and has been revised. The commenter should note, 
that the intent for this rule change is to provide greater flexibility 
in transportation with a concomitant improvement of a shipper's 
knowledge about the contents of materials in the package. The rule has 
been revised to address the concerns about shipments of very small 
quantities of fissile material in small packages and shipment of low 
concentrations of fissile material where the large volume of the 
container and mass of nonfissile material might enable one to exceed 
the fissile limit in the proposed rule. The IAEA is currently 
considering changes to the current international regulations in the 
area of the fissile material exemptions. The concept put forward in the 
current rule is one of those under consideration. The other option 
proposed to the IAEA to provide safety in the event of uncontrolled 
accumulation of fissile exempt packages is to implement a CSI for all 
packages containing fissile material. The NRC considered both options 
and chose to implement the option that did not require a CSI on fissile 
exempt packages.
    Comment. One commenter expressed concern that NRC's proposal to add 
atomic ratio criteria to the previously used 15-g \235\U mass criterion 
may restrict exemption of fissile materials, not containing special 
moderators, that are currently acceptable. Another commenter expressed 
support for the concept of exemptions for fissile material shipments 
under specific conditions. However, the commenter said that NRC's 
proposal in Sec. 71.15 was overly conservative and resulted in a 
reduction in the limits of fissile material content without 
justification.
    Response. The NRC staff agrees, in part, with these comments. 
Proposed Sec. 71.15(c)(1) has been modified by removing the limit of 
350 g in a package and instead specifies criteria for commingling of 
the material such that, within any selected 360 kg of nonfissile solid 
material, there can be no more than 180 g of fissile material. Thus, a 
large rail car with a homogenized distribution of fissile material 
within a nonfissile waste matrix might exceed the 180-g limit but would 
be effectively mixed at low enough concentration to enable safe 
shipment. In the case of small sample shipments, a limit of 2 g per 
package has been added to Sec. 71.15(a) and applies without regard to 
any mass ratios.
    Comment. One commenter stated that the proposed fissile material 
exemptions do not agree with the TS-R-1 exemptions and appear to 
contain requirements that are not necessary for nuclear criticality 
safety. This commenter also expressed concern about the discontinuance 
of the exemption for material containing less than 5 grams of uranium-
235 per 10-liter volume and its impact on shipments related to 
decommissioning activities. The commenter also voiced support for the 
proposed new limit of 350 g of fissile material with a 2000:1 ratio to 
noncombustible and insoluble-in-water material.
    Response. The NRC staff acknowledges the comment of support for one 
of the proposed changes. Regarding the comment about the exemption 
discontinuance, the commenter did not provide any detailed 
justification for this concern; thus, no change has been made to the 
rule language. As stated above, the NRC has determined for a number of 
issues that it does not harmonize completely with all changes made in 
the IAEA guidance documents based on safety and other technical 
reasons.
Issue 17. Decision on Petition for Rulemaking on Double Containment of 
Plutonium (PRM-71-12)
    Summary of Decision on PRM-71-12. Currently in 10 CFR 71.63(b), 
plutonium in excess of 0.74 TBq (20 Ci) must be packaged in a separate 
inner container placed within an outer packaging. This is referred to 
as double containment. It is the combination of the inner container and 
the outer packaging that is subjected to the normal conditions of 
transport (Sec. 71.71) and the hypothetical accident conditions (Sec. 
71.73). Upon application of the normal conditions of transport and 
hypothetical accident conditions, the acceptance criteria for 
shielding, containment, and sub-criticality in Sec. 71.51 must be also 
met for the total package (inner container and outer packaging), but 
the containment dispersal acceptance (10-6 A2/
hour or 1 A2/week) are applied to each boundary (i.e., the 
inner container and the outer packaging). Note however, as a point of 
clarification, double containment does not mean two Type B containers 
nested into one.
    The final rule grants the petitioner's request to remove the double 
containment requirement of Sec. 71.63(b). However, the requirement of 
Sec. 71.63(a) that shipments whose contents contain greater than 0.74 
TBq (20 Ci) of plutonium must be made with the contents in solid form 
is retained. Thus, the petitioner's alternative proposal is denied. 
This completes action on PRM-71-12.
    The NRC has decided to remove the double containment requirement 
because this regulation is neither risk-informed nor performance-based. 
There are many nuclides with A2 values the same or lower 
than plutonium's for which double containment has never been required. 
Thus, requiring double containment for plutonium alone is not 
consistent with the relative hazard rankings in Table A-1. The Type B 
packaging standards, which the outer containment of plutonium shipments 
must meet, in and of themselves, provide reasonable assurance that 
public health and safety and the environment are protected during the 
transportation of radioactive material. This position is supported by 
an excellent safety record in which no fatalities or injuries have been 
attributed to material transported in a Type B package. The imposition 
of an additional packaging requirement (in the form of a separate inner 
container) is fundamentally inconsistent with this position and is 
technically unnecessary to assure safe transport. Further, removal of 
this requirement will reduce an unnecessary regulatory burden on 
licensees, will likely result in reduced risk to radiation workers, and 
will serve to harmonize part 71 with TS-R-1.
    On the other hand, the imposition of the requirement that plutonium 
in excess of 0.74 TBq (20 Ci) per package be shipped as a solid does 
not create a regulatory inconsistency with the Type B package 
standards. The NRC considers the contents of a package when it is 
evaluating the adequacy of a packaging's design. The approved content 
limits and the approved packaging design together define the CoC for a 
package. However, other than criticality controls and the solid form 
requirement of Sec. 71.63(a), subparts E and F do not contain any 
restrictions on the contents of a package. Thus, while the inner 
containment requirement in Sec. 71.63(b) can be seen as conflicting 
with the Type B package standard because the inner containment affects 
the packaging design, the solid form requirement of Sec. 71.63(a) does 
not conflict with the packaging requirements of the Type B package 
standard because the solid form requirement affects only the contents 
of the package, not the packaging itself.
    Affected Sections. Section 71.63.

[[Page 3753]]

    Discussion of PRM-71-12: The NRC received a petition for rulemaking 
from International Energy Consultants, Inc. (IEC), dated September 25, 
1997. The petition was docketed as PRM-71-12 and was published for 
public comment (63 FR 8362; February 19, 1998). Based on a request from 
General Atomic, the comment period was extended to July 31, 1998 (see 
63 FR 34335; June 24, 1998). Nine public comments were received on the 
petition. Four commenters supported the petition, and five commenters 
opposed the petition.
    The petitioner requested that Sec. 71.63(b) be removed. The 
petitioner argued that the double containment provisions of Sec. 
71.63(b) cannot be supported technically or logically. The petitioner 
stated that based on the ``Q-system for the Calculation of 
A1 and A2 Values,'' an A2 quantity of 
any radionuclide has the same potential for damaging the environment 
and the human species as an A2 quantity of any other 
radionuclide.
    The NRC believes that the Q-values are based upon radiological 
exposure hazard models which calculate the allowable quantity limit 
(the A1 or A2 value) necessary to produce a known 
exposure (i.e., one A2 of plutonium-239 or one A2 
of cobalt-60 will both yield the same radiation dose under the Q-system 
models, even though the A2 values for these nuclides are 
different (e.g., one A2 of plutonium-239 = 2 x 
10-4 TBq, and one A2 of cobalt-60 = 1 TBq). The 
Q-system models take into account the exposure pathways of the various 
radionuclides, typical chemical forms of the radionuclide, methods for 
uptake into the body, methods for removal from the body, the type of 
radiation the radionuclide emits, and the bodily organs the 
radionuclide preferentially affects. The specific A1 and 
A2 values for each nuclide are developed using radiation 
dosimetry approaches recommended by the World Health Organization and 
the ICRP. The models are periodically reviewed by international health 
physics experts (including representatives from the United States), and 
the A1 and A2 values are updated during the IAEA 
revision process, based upon the best available data. (Note that 
changes to the A1 and A2 values as a result of 
changes to the models in TS-R-1 are also discussed in Issue 3 of this 
rule.) These values are then issued by the IAEA in safety standards 
such as TS-R-1. When the IAEA has revised the A1 and 
A2 values in previous revisions of its transport 
regulations, these revised values have been adopted by the NRC and DOT 
into the transportation regulations in 10 CFR part 71 and 49 CFR part 
173, respectively.
    NRC's review of the current A1 and A2 values 
in Appendix A to part 71, Table A-1, reveals that 5 radionuclides have 
an A2 value lower than plutonium (i.e., plutonium-239), and 
11 radionuclides have an A2 value that is equal to 
plutonium-239. Because the models used to determine the A1 
and A2 values all result in the same radiation exposure 
(i.e., hazard), a smaller A1 and A2 value for one 
radionuclide would indicate a greater potential hazard to humans than a 
radionuclide with a larger A1 and A2 value. Thus, 
overall, Table A-1 can also be viewed as a relative hazard ranking (for 
transportation purposes) of the listed radionuclides. In that light, 
requiring double containment for plutonium alone is not consistent with 
the relative hazard rankings in Table A-1.
    The petitioner also argued that the Type B package requirements 
should be applied consistently for any radionuclide, whenever a 
package's contents exceed an A2 limit. However, part 71 is 
not consistent by imposing the double containment requirement for 
plutonium. The petitioner believes that if Type B package standards are 
sufficient for a quantity of a particular radionuclide which exceeds 
the A2 limit, then Type B package standards should also be 
sufficient for any other radionuclide which also exceeds the 
A2 limit. The petitioner stated that:
    While, for the most part, part 71 regulations embrace this simple 
logical congruence, the congruence fails under 10 CFR 71.63(b) wherein 
packages containing plutonium must include a separate inner container 
for quantities of plutonium having a radioactivity exceeding 20 curies 
(0.74 TBq) (with certain exceptions).
The petitioner further stated that:
    If the NRC allows this failure of congruence to persist, the 
regulations will be vulnerable to the following challenges: (1) The 
logical foundation of the adequacy of A2 values as a proper 
measure of the potential for damaging the environment and the human 
species, as set forth under the Q-System, is compromised; (2) the 
absence of a limit for every other radionuclide which, if exceeded, 
would require a separate inner container, is an inherently inconsistent 
safety practice; and (3) the performance requirements for Type B 
packages, as called for by 10 CFR part 71, establish containment 
conditions under different levels of package trauma. The satisfaction 
of these Type B package standards should be a matter of proper design 
work by the package designer and proper evaluation of the design 
through regulatory review. The imposition of any specific package 
design feature such as that contained in 10 CFR 71.63(b) is gratuitous. 
The regulations are not formulated as package design specifications, 
nor should they be.
    The NRC agrees that the part 71 regulations are not formulated as 
package design specifications; rather, the part 71 regulations 
establish performance standards for a package's design. The NRC reviews 
the application to evaluate whether the package's design meets the 
performance requirements of part 71. Consequently, the NRC can then 
conclude that the design of the package provides reasonable assurance 
that public health and safety and the environment are adequately 
protected.
    The petitioner also believes that the continuing presence of Sec. 
71.63(b) engenders excessively high costs in the transport of some 
radioactive materials without a clearly measurable net safety benefit. 
The petitioner stated that this is so, in part, because the ultimate 
release limits allowed under part 71 package performance requirements 
are identical with or without a ``separate inner container,'' and 
because the presence of a ``separate inner container'' promotes 
additional exposures to radiation through the additional handling 
required for the ``separate inner container.'' Consequently, the 
petitioner asserted that the presence or absence of a separate inner 
container barrier does not affect the standard to which the outer 
container barrier must perform in protecting public health and safety 
and the environment. Therefore, the petitioner concluded that given 
that the outer containment barrier provides an acceptable level of 
safety, the separate inner container is superfluous and results in 
unnecessary cost and radiation exposure. According to the petitioner, 
these unnecessary costs involve both the design, review, and 
fabrication of a package, as well as the costs of transporting the 
package. And the unnecessary radiation exposure involves workers having 
to handle (i.e., seal, inspect, or move) the ``separate inner 
container.''
    As an alternative to the primary petition, the petitioner believes 
that an option to eliminate both Sec. 71.63(a) and (b) should also be 
considered. Section 71.63(a) requires that plutonium in quantities 
greater than 0.74 TBq (20 Ci) be shipped in solid form. This option 
would have the effect of removing Sec. 71.63 entirely. The petitioner 
believes that the arguments set forth to support the elimination of 
Sec. 71.63(b) also support the elimination of Sec. 71.63(a).

[[Page 3754]]

The petitioner did not provide a separate regulatory or cost analysis 
supporting the request to remove Sec. 71.63(a).
    History of the Double Containment Requirement: On June 17, 1974 (39 
FR 20960), the AEC issued a final rule which imposed special 
requirements on the shipment of plutonium. These requirements are 
located in Sec. 71.63 and apply to shipments of radioactive material 
containing quantities of plutonium in excess of 0.74 TBq (20 curies). 
Section 71.63 contains two principal requirements. First, the plutonium 
contents of the package must be in solid form [Sec. 71.63(a)]. Second, 
the packaging containing the plutonium must provide a separate inner 
containment (i.e., the ``double containment'' requirement) [Sec. 
71.63(b)]. In addition, the AEC specifically excluded from the double 
containment requirement of Sec. 71.63(b) plutonium in the form of 
reactor fuel elements, metal or metal alloys, and other plutonium-
bearing solids that the Commission (AEC or NRC) may determine, on a 
case-by-case basis, do not require double containment. This regulation 
remained essentially unchanged from 1974 until 1998, when vitrified 
high-level waste in sealed canisters was added to the list of exempt 
forms of plutonium in Sec. 71.63(b) (63 FR 32600; June 15, 1998). The 
double containment requirement is in addition to the existing 10 CFR 
part 71 subparts E and F requirements imposed on Type B packagings 
(e.g., the normal conditions of transport and hypothetical accident 
conditions of Sec.Sec. 71.71 and 71.73, respectively, and the fissile 
package requirements of Sec.Sec. 71.55 and 71.59). Part 71 does not 
impose a double containment requirement for any radionuclide other than 
plutonium. Additionally, IAEA standard TS-R-1 does not provide for a 
double containment requirement (in lieu of the single containment Type 
B package standards) for any radionuclide.
    The AEC issued this regulation at a time when AEC staff anticipated 
widespread reprocessing of commercial spent fuel, and existing 
shipments of plutonium were made in the form of liquid plutonium 
nitrate. Because of physical changes to the plutonium that was expected 
to be reprocessed (i.e., higher levels of burnup in commercial reactors 
for spent fuel, which would then be reprocessed), and regulatory 
concerns with the possibility of package leakage, the AEC issued a 
regulation that imposed the double containment requirement when the 
package contained more than 0.74 TBq (20 Ci) of plutonium. This double 
containment was in addition to the existing Type B package standards on 
packages intended for the shipment of greater than an A1 or 
A2 quantity of plutonium.
    The NRC staff has reviewed the available regulatory history for 
Sec. 71.63, and has provided a recapitulation of the supporting 
information which led to the issuance of this regulation. The NRC staff 
has extracted the following information from several SECY papers the 
AEC staff submitted to the Commission on this regulation. The NRC staff 
believes this information is relevant and will provide stakeholders 
with perspective in understanding the bases for this regulation, and 
thereby assist stakeholders in evaluating the staff's proposed changes 
to this regulation.
    In SECY-R-702,\6\ the AEC staff identified two considerations that 
were the genesis of the rulemaking that led to Sec. 71.63. AEC staff 
stated:
---------------------------------------------------------------------------

    \6\ SECY-R-702, ``Consideration of Form for Shipping 
Plutonium,'' June 1, 1973.
---------------------------------------------------------------------------

    First, increasingly larger quantities of plutonium will be 
recovered from power reactor spent fuel. Second, the specific activity 
of the plutonium will increase with higher reactor fuel burnup 
resulting in greater pressure generation potential from plutonium 
nitrate solutions in shipping containers, greater heat generation, and 
higher gamma and neutron radiation levels. These changes will make the 
present nitrate packages obsolete. Thus, from both safety and economic 
considerations, the transportation of plutonium as [liquid] nitrate 
will soon require substantial redesign of packages to handle larger 
quantities as well as to deal with the higher levels of gas evolution 
(pressurization), heat generation, and gamma and neutron radiation.
    There is little doubt that larger plutonium nitrate packages could 
be designed to meet regulatory standards. The increased potential for 
human error and the consequences of such error in the shipment of 
plutonium nitrate are not so easily controlled by regulation. Even 
though such packages may be adequately designed, their loading and 
closure requires high operation performance by personnel on a 
continuing basis. As the number of packages to be shipped increases, 
the probability of leakage through improperly assembled and closed 
packages also increases. * * * More refined or stringent regulatory 
requirements, such as double containment, would not sufficiently lessen 
this concern because of the necessary dependence on people to affect 
engineered safeguards.
    In SECY-R-74-5,\7\ AEC staff summarized the factors relevant to 
consideration of a proposed rule following a June 14, 1973, meeting to 
discuss SECY-R-702, between the Regulatory and General Manager's staffs 
(i.e., the rulemaking and operational sides of the AEC). The AEC 
stated:
---------------------------------------------------------------------------

    \7\ SECY-R-74-5, ``Consideration of Form for Shipping 
Plutonium,'' dated July 6, 1973.
---------------------------------------------------------------------------

    As a result of this meeting (on June 14, 1973), the (Regulatory and 
General Manager's) staffs have agreed that the basic factors pertinent 
to the consideration of form for shipment of plutonium are:
    1. The experience with shipping plutonium as an aqueous nitrate 
solution in packages meeting current regulatory criteria has been 
satisfactory to date.
    2. The changing characteristic of plutonium recovered from power 
reactors will make the existing packaging obsolete for plutonium 
nitrate solutions and possibly for solid form. Economic factors will 
probably dictate considerably larger shipments (and larger packages) 
than currently used.
    3. It is expected that packages can be designed to meet regulatory 
standards for either aqueous solutions or solid plutonium compounds. 
Just as in any situation involving the packaging of radioactive 
materials, a high level of human performance is necessary to assure 
against leakage caused by human error in packaging. As the number of 
plutonium shipments increases, as it will, and packages become larger 
and more complex in design, the probability of such human error 
increases.
    4. The probability of human error with the packaging for liquid, 
anticipated to be more complex in design, is probably greater than with 
the packaging for solid. Furthermore, should a human error occur in 
package preparation or closure, the probability of liquid escaping from 
the improperly prepared package is greater than for most solids and 
particularly for solid plutonium materials expected to be shipped.
    5. Staff studies reported in SECY-R-62 and SECY-R-509 \8\ conclude 
that the consequences of release of solid or aqueous solutions do not 
differ appreciably. Therefore, this paper (SECY-R-702) does not deal 
with the consequences of releases.
---------------------------------------------------------------------------

    \8\ SECY-R-62, ``Shipment of Plutonium,'' and SECY-R-509, 
``Plutonium Handling and Storage,'' dated October 16, 1970. These 
papers concluded that there is no scientific or technical reason to 
prohibit shipment of plutonium nitrate and recommended that 
Commission (AEC) efforts be directed toward providing improved 
safety criteria for shipping containers.

---------------------------------------------------------------------------

[[Page 3755]]

    6. It is, therefore, concluded that safety would be enhanced if 
plutonium were shipped as a solid rather than in solution.
    The arguments for requiring a solid form of plutonium for shipment 
are largely subjective, in that there is no hard evidence on which to 
base statistical probabilities or to assess quantitatively the 
incremental increase in safety which is expected. The discussion in the 
regulatory paper, SECY-R-702, is not intended to be a technical 
argument which incontrovertibly leads to a conclusion. It is, rather, a 
presentation of the rationale which has led the Regulatory staff to its 
conclusion that a possible problem may develop and that the proposed 
action is a step towards increased assurance against the problem 
developing. In SECY-R-74-172,\9\ AEC staff submitted a final rule to 
the Commission for approval.
---------------------------------------------------------------------------

    \9\ SECY-R-74-172, ``Consideration of Form for Shipping 
Plutonium,'' April 18, 1974.
---------------------------------------------------------------------------

    The proposed rule had contained a requirement that the plutonium be 
contained in a special form capsule. However, in response to comments 
from the AEC General Manager, the final rule changed this requirement 
to a separate inner container (i.e., the double containment 
requirement). The AEC staff indicated in a response to a public comment 
in Enclosure B (to SECY-R-74-172) that ``[t]he need for the inner 
containment is based on the desire to provide a substitute for not 
requiring the plutonium to be in a `nonrespirable' form.''
    The regulatory history of Sec. 71.63 indicates that the AEC's 
decision to require a separate inner container for shipments of 
plutonium in excess of 0.74 TBq (20 Ci) was based on existing policy 
and regulatory concerns (i.e., ``that a possible problem may develop 
and that the proposed action [in SECY-R-702] is a step towards 
increased assurance against the problem developing''). Because of the 
expectation of a significant increase in the number of liquid plutonium 
nitrate shipments, the AEC used a defense-in-depth philosophy (i.e., 
the double containment and solid form requirements), to ensure that 
respirable plutonium would not be released to the environment during a 
transportation accident. However, the regulatory history does indicate 
that the AEC's concerns did not involve the adequacy of existing liquid 
plutonium nitrate packages. Rather, the AEC's regulatory concern was on 
the increased possibility of human error combined with an expected 
increase in the number of shipments that would yield an increased 
probability of leakage during shipment. The AEC's policy concern was 
based on an economic decision on whether the AEC should require the 
reprocessing industry to build new, larger liquid plutonium-nitrate 
shipping containers, capable of handling higher burnup reactor spent 
fuel, or to build new, dry, powdered plutonium-dioxide shipping 
containers. The regulatory history indicates that the AEC staff judged 
that new, larger, higher burnup-capacity liquid plutonium-nitrate 
packages could be designed, approved, built, and safely used. However, 
one of the AEC's principal underlying assumptions for this rule was 
obviated in 1979 when the Carter administration decided that 
reprocessing of civilian spent fuel and reuse of plutonium was not 
desirable. Consequently, the expected plutonium reprocessing economy 
and widespread shipments of liquid plutonium nitrate within the U.S. 
never materialized.
    On June 15, 1998 (63 FR 32600), in response to a petition for 
rulemaking submitted by DOE (PRM-71-11) (February 18, 1994; 59 FR 
8143), the Commission issued a final rule revising Sec. 71.63(b) to add 
vitrified high-level waste (HLW) contained in a sealed canister to the 
list of forms of plutonium exempt from the double containment 
requirement (June 15, 1998; 63 FR 32600). In its original response to 
PRM-71-11, NRC proposed in SECY-96-215 \10\ to make a ``determination'' 
under Sec. 71.63(b)(3) that vitrified HLW contained in a sealed 
canister did not require double containment. However, the Commission in 
an SRM on SECY-96-215, dated October 31, 1996, disapproved the staff's 
approach and directed that resolution of this petition be addressed 
through rulemaking (the June 15, 1998, final rule was the culmination 
of this effort). In addition to disapproving the use of a 
``determination'' process, the Commission also directed the staff to 
``* * * also address whether the technical basis for 10 CFR 71.63 
remains valid, or whether a revision or elimination of portions of 10 
CFR 71.63 is needed to provide flexibility for current and future 
technologies.'' In SECY-97-218,\11\ NRC responded to the SRM's 
direction and stated ``[t]he technical basis remains valid and the 
provisions provide adequate flexibility for current and future 
technologies.''
---------------------------------------------------------------------------

    \10\ SECY-96-215, ``Requirements for Shipping Packages Used to 
Transport Vitrified Waste Containing Plutonium,'' dated October 8, 
1996.
    \11\ SECY-97-218, ``Special Provisions for Transport of Large 
Quantities of Plutonium (Response to Staff Requirements Memorandum--
SECY-96-215),'' dated September 29, 1997.
---------------------------------------------------------------------------

    Summary of Comments Received on the Petition (PRM-71-12): Nine 
public comments were received on the petition (petition was published 
for public comment in 63 FR 8362; February 19, 1998). Four commenters 
supported the petition, and five commenters opposed the petition. The 
four commenters supporting the petition essentially stated that the 
IAEA's Q-system accurately reflects the dangers of radionuclides, 
including plutonium, and that elimination of Sec. 71.63(a) and (b) 
would make the regulations more performance based, reduce costs and 
personnel exposures, and be consistent with the IAEA standards.
    The five commenters opposing the petition essentially stated that: 
(1) Plutonium is very dangerous, especially in liquid form, and 
therefore additional regulatory requirements are warranted; (2) 
existing regulations are not overly burdensome, especially in light of 
the total expected transportation cost; (3) TRUPACT-II packages meet 
current Sec. 71.63(b) requirements (TRUPACT-II is a package developed 
by DOE to transport transuranic wastes (including plutonium) to the 
Waste Isolation Pilot Plant (WIPP) and has been issued a part 71 CoC, 
No. 9218); (4) a commenter (the Western Governors' Association) has 
worked for over 10 years to ensure a safe transportation system for 
WIPP, including educating the public about the TRUPACT-II package; (5) 
any change now would erode public confidence and be detrimental to the 
entire transportation system for WIPP shipments; and (6) additional 
personnel exposure due to double containment is insignificant.
    Analysis of Public Comments on the Issues Paper: The NRC has 
received 48 public comments on this issue in response to the issue 
paper, in subsequent public meetings, and the workshop (the issues 
paper was published at 65 FR 44360; July 17, 2000). Industry 
representatives and some members of the public support the petition. 
Public interest organizations, Agreement States and State 
representatives, and the Western Governors' Association, and other 
members of the public oppose the petition. Several commenters expressed 
their belief that Congress, in approving the Waste Isolation Pilot 
Plant Land Withdrawal Act (the Act), Pub. L. 102-579 (106 Stat. 4777), 
section 16(a), which mandates that the NRC certify the design of 
packages used to transport transuranic waste to WIPP, expected those 
packages to have a double containment. The NRC researched this

[[Page 3756]]

issue and found that section 16(a) of the Act does not contain any 
explicit provisions mandating the use of a double containment in 
packages transporting transuranic waste to or from WIPP. Section 16(a) 
of the Act states, in part, ``[n]o transuranic waste may be transported 
by or for the Secretary [of the DOE] to or from WIPP, except in 
packages the design of which has been certified by the Nuclear 
Regulatory Commission * * *'' Furthermore, the NRC has reviewed the 
legislative history \12\ associated with the Act and has not identified 
any discussions on the use of double containment for the shipment of 
transuranic waste. The legislative history does mention that the design 
of these packages will be certified by the NRC; however, this language 
is identical to that contained in the Act itself. Therefore, the NRC 
believes the absence of specific language in section 16(a) of the Act 
requiring double containment should be interpreted as requiring the NRC 
to apply its independent technical judgment in establishing standards 
for package designs and in evaluating applications for certification of 
package designs, to ensure that such packages would provide reasonable 
assurance that public health and safety and the environment would be 
adequately protected. In carrying out its mission, the courts have 
found that the NRC has broad latitude in establishing, maintaining, and 
revising technical performance criteria necessary to provide reasonable 
assurance that public health and safety and the environment are 
adequately protected. An example of these technical performance 
criteria is the Type B package design standards. Accordingly, the NRC 
believes that the proposed revision of a technical package standard 
(i.e., removal of the double containment requirement for plutonium from 
the Type B package standards) is not restricted by the mandate of 
section 16(a) of the Act for the NRC to certify the design of packages 
intended to transport transuranic material to and from WIPP.
---------------------------------------------------------------------------

    \12\ See Congressional Record Vol. 137, November 5, 1991, pages 
S15984-15997 (Senate approval of S. 1671); Cong. Rec. Vol. 138, July 
21, 1992, pages H6301-6333 (House approval of H.R. 2637); Cong. Rec. 
Vol. 138, October 5, 1992, pages H11868-11870 (House approval of 
Conference Report on S. 1671); Cong. Rec. Vol. 138, October 8, 1992 
(Senate approval of Conference Report on S. 1671); and Cong. Rec. 
Vol. 138, October 5, 1992, pages H12221-12226 (Conference Report on 
S. 1671-H. Rpt. 102-1037).
---------------------------------------------------------------------------

    Other commenters stated that stakeholders' expectations were that 
packages intended to transport transuranic material to and from WIPP 
would include a double containment provision. Consequently, the 
commenters expressed a belief that removal of the double containment 
requirement would decrease public confidence in the NRC's 
accomplishment of its mission in the approval of the design of packages 
for the transportation of transuranic waste to and from WIPP. The 
commenters stated that the public would view elimination of the double 
containment requirement as a relaxation in safety. The presence of a 
separate inner container provides defense-in-depth through an 
additional barrier to the release of plutonium during a transportation 
accident, according to commenters. In addition, the commenters stated 
that plutonium is so inherently deadly, that defense-in-depth is 
appropriate. The NRC agrees that a double containment does provide an 
additional barrier. However, the NRC believes that, for the reasons 
discussed below, double containment is unnecessary to protect public 
health and safety. The NRC and AEC have not required an additional 
containment barrier for Type B packages transporting any radionuclides 
other than plutonium and, before 1974, the AEC did not require double 
containment for plutonium.
    In response to some of the comments opposed to the petition, the 
NRC believes that removal of Sec. 71.63(b) would not invalidate the 
design of existing packages intended for the shipment of plutonium. 
These packages could continue to be used with a separate inner 
container. The NRC agrees with the commenters that a quantitative cost 
analysis was not provided by the petitioner.
    The NRC has issued part 71 CoC No. 9218 to DOE for the TRUPACT-II 
package (Docket No. 71-9218), for the transportation of transuranic 
waste (including plutonium) to and from the WIPP. The TRUPACT-II 
package complies with the current Sec. 71.63(b) requirements and has a 
separate inner container. The TRUPACT-II SAR indicates that the weight 
of the inner container and its lid is approximately 2,620 lbs. 
Hypothetically, elimination of the separate inner container would 
increase the available payload for the TRUPACT-II package from the 
current 7,265 to 9,885 lbs. Thus, removal of the double containment 
requirement would potentially increase the TRUPACT-II's available 
payload by 36 percent. Further, the removal of the inner container from 
the TRUPACT-II would also potentially increase the available volume. 
The NRC believes that the final rule would not invalidate the existing 
TRUPACT-II design (i.e., it would still meet all remaining applicable 
requirements of part 71). Thus, DOE could continue to use the TRUPACT-
II to ship transuranic waste to and from WIPP, or DOE could consider an 
alternate Type B package.
    Additionally, based on comments received in the public meetings, 
the NRC believes that a misperception exists with respect to TRUPACT-II 
shipments; removal of the Sec. 71.63(b) double containment requirement 
would not result in loose plutonium waste being placed inside a 
TRUPACT-II package. Based upon information contained in the SAR, 
plutonium wastes (i.e., used gloves, anti-Cs, rags, etc.) are placed in 
plastic bags, and these bags are sealed inside lined 55-gallon steel 
drums. Plutonium residues are placed inside cans which are then sealed 
inside a pipe overpack (a 6-inch or 12-inch stainless steel cylinder 
with a bolted lid), and the pipe overpack is then sealed inside a lined 
55-gallon steel drum. The 55-gallon drums are then sealed inside the 
TRUPACT-II inner containment vessel, and finally the inner containment 
vessel is sealed inside the TRUPACT-II package. Consequently, the 
TRUPACT-II shipping practices employ multiple barriers and would 
continue to do so. Removal of the inner containment vessel would not be 
expected to produce a significant incremental increase in the 
possibility of leakage during normal transportation. The NRC notes that 
some NRC regulations have established additional requirements for 
plutonium (e.g., the special nuclear material license application 
provisions of Sec. 70.22(f)).
    The NRC believes that the Type B packaging standards, in and of 
themselves, provide reasonable assurance that public health and safety 
and the environment would be adequately protected during the 
transportation of radioactive material. This belief is supported by an 
excellent safety record in which no fatalities or injuries have been 
attributed to material transported in a Type B package. Type B 
packaging standards have been in existence for approximately 40 years 
and have been incorporated into the part 71 regulations by both the NRC 
and its predecessor, the AEC. The NRC's Type B package standards are 
based on IAEA's Type B package standards. Moreover, IAEA's Type B 
package standards have never required a separate inner container for 
packages intended to transport plutonium, nor for any other 
radionuclide.
    Therefore, the NRC believes that imposition of an additional 
packaging

[[Page 3757]]

requirement (in the form of a separate inner container) is 
fundamentally inconsistent with the position that Type B packaging 
standards, in and of themselves, provide reasonable assurance that 
public health and safety and the environment would be adequately 
protected during the transportation of (any type of) radioactive 
material. Thus, the NRC believes that maintaining Sec. 71.63(b) is not 
consistent with the other existing Type B packaging standards contained 
in part 71.
    The NRC also believes that the regulatory history of Sec. 71.63 
demonstrates that the AEC's decision to add this section was based on 
policy and regulatory concerns. However, the NRC also agrees that the 
use of a double containment does provide defense-in-depth and does 
decrease the absolute risk of the release of respirable plutonium to 
the environment during a transportation accident. Consequently, while 
the defense-in-depth afforded by a double containment does reduce risk, 
the NRC believes the question which should be focused on is whether the 
double containment requirement is risk-informed. The NRC is unaware of 
any risk studies that would provide a quantitative indication of the 
risk reduction associated with the use of an NRC-certified double 
containment packaging in transportation of plutonium. Rather, the NRC 
would look to the demonstrated performance record of existing Type B 
package standards to conclude that double containment is not necessary.
    In summary, the AEC indicated (in SECY-R-702 and SECY-R-74-5) that 
liquid plutonium nitrate packages were safe, and new, larger packages 
to handle higher burnup reactor spent fuel could also be designed. NRC 
believes that the AEC's assumption for initiating this requirement was 
that large scale reprocessing of civilian reactor spent fuel and reuse 
of plutonium would occur. The decision of former President Carter's 
administration to forgo the reprocessing of civilian reactor spent fuel 
and reuse of plutonium obviated the AEC's assumption. Consequently, the 
AEC's supposition that a human error occurring while sealing a package 
of liquid plutonium nitrate was more likely to occur with the expected 
increase in shipments of plutonium nitrate was also obviated by the 
Government's decision to forgo the reprocessing of civilian reactor 
spent fuel. In SECY-97-218, NRC staff indicated that the separate inner 
container provided an additional barrier to the release of plutonium in 
an accident. NRC continues to believe that a separate inner container 
provides an additional barrier to the release of plutonium in an 
accident, just as a package with triple containment would provide an 
even greater barrier to the release of plutonium in an accident. 
However, this type of approach is neither risk informed nor performance 
based. Consequently, based upon review of the petition, comments on the 
petition, and research into the regulatory history of the double 
containment requirement, the NRC agrees that a separate inner container 
is not necessary for Type B packages containing solid plutonium. NRC 
believes that the worldwide performance record over 40 years of Type B 
packages demonstrates that a single containment barrier is adequate. 
Therefore, the NRC agrees with the petitioner and believes that Sec. 
71.63(b) is not technically necessary to provide a reasonable assurance 
that public health and safety and the environment will be adequately 
protected during the transportation of plutonium.
    While the NRC believes a case can be made for elimination of the 
separate inner container requirement in Sec. 71.63(b), elimination of 
the solid form requirement in Sec. 71.63(a) is not as clear. While the 
same arguments can be made on the obviation of the AEC's basis for 
originally issuing Sec. 71.63(a) (i.e., the elimination of reprocessing 
of plutonium), the same regulatory inconsistency between Type B package 
standards and the inner containment requirement does not exist for the 
liquid versus solid form argument. The NRC considers the contents of a 
package when it is evaluating the adequacy of a packaging's design. The 
approved content limits and the approved packaging design together 
define the CoC for a package. However, other than criticality controls 
and the liquid form requirement of Sec. 71.63(a), 10 CFR part 71 
subparts E and F do not contain any restrictions on the contents of a 
package. Thus, while the inner containment requirement in Sec. 71.63(b) 
can be seen as conflicting with the Type B package standard because the 
inner containment affects the packaging's design, the solid form 
requirement of Sec. 71.63(a) does not conflict with the packaging 
requirements of the Type B package standard because the solid form 
requirement affects only the contents of the package, not the packaging 
itself.
    The NRC expects that cost and dose savings would accrue from the 
removal of Sec. 71.63(b). However, because no shipments of liquid 
plutonium nitrate are contemplated in the U.S., NRC would not expect 
cost or dose savings to accrue from the removal of Sec. 71.63(a), if 
that section were to be also removed. Further, the AEC's original bases 
have been obviated by former President Carter's administration's 
decision to not pursue a commercial fuel cycle involving the 
reprocessing of plutonium.
    After weighing this information, the NRC continues to believe that 
the Type B package standards, when evaluated against 40 years of use 
worldwide, and millions of safe shipments of Type B packages, together 
provide reasonable assurance that public health and safety and the 
environment would be adequately protected during the transportation of 
radioactive material. The NRC believes that, in this case, the 
reasonable assurance standard, provided by the Type B package 
requirements, provides an adequate basis for the public's confidence in 
the NRC's actions.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Several commenters suggested that all radioactive 
materials should require double packaging. Two of these commenters 
stated double containment is a security and safety precaution. A third 
stated that existing container requirements are the minimum standards 
necessary for safety, security, and public acceptance. Another 
commenter simply objected to the removal of the requirement for double 
containment of plutonium.
    Response. The NRC disagrees with these comments. The NRC has made a 
finding that single containment of radioactive material provides an 
adequate level of safety for all radioactive materials. The 
A1 and A2 value summary found at 67 FR 21422; 
April 30, 2002, under the heading Issue 3, provides information that 
supports the NRC's basis for this decision. The comments provided no 
justification for the double containment requirement for shipment of 
all nuclear materials.
    Comment. Several commenters were concerned with NRC's proposal to 
eliminate double containment. The first of these commenters asked if 
there is any basis to eliminate the double containment requirement 
other than to harmonize our rules with the IAEA regulations. The second 
commenter expressed concern that the ``only benefits from eliminating 
double containment * * * would accrue to the DOE, to contractors, 
licensees, and shippers in the form of cost savings.'' Furthermore, the 
commenter stated that the cost of maintaining transportation

[[Page 3758]]

safety standards should be borne by those in the industry and that 
costs should not be ``used as an excuse for deregulation or 
exemptions.'' A similar argument was made by another commenter who 
urged NRC not to remove Sec. 71.63(b) reasoning that, as noted in the 
proposed rulemaking, the petitioner did not provide a quantitative cost 
analysis; therefore, the contention that ``presence of Sec. 71.63(b) 
engenders excessively high costs'' is unsubstantiated. Another 
commenter stated that while an 8-13 percent volume reduction due to 
weight restrictions caused by double containment is not trivial, the 
benefits from reducing this weight penalty needs to be balanced against 
the resulting increase in radiation doses, the increased likelihood of 
a release in the event of a severe accident, and the increased cost of 
certifying a new package.
    Response. The primary reason for removing the double containment 
requirement is that the NRC has no technical justification or basis for 
maintaining double containment for plutonium or any other radionuclide. 
The NRC believes the arguments for removing double containment have 
been adequately addressed earlier in this notice and in the proposed 
rule under this issue.
    While NRC acknowledges that there may be monetary benefits 
associated with removing double containment, there are other reasons as 
well, including reduction in personnel exposure for those individuals 
involved in loading packages for transport. Further, while double 
containment does provide an additional barrier against release, the NRC 
believes that, for reasons previously explained, double containment is 
unnecessary to protect public health and safety. Moreover, NRC has been 
and remains committed to providing regulations that are not only risk 
informed, but also reduce unnecessary regulatory burden.
    Comment. One commenter stated that removing the double containment 
requirement would reduce costs of packaging and associated hardware. 
The commenter asserted that double containment increases costs without 
measurable benefit. The commenter then provided cost information and 
discussed the design, certification, and fabrication of future 
packaging (e.g., TRUPACT III or the DPP-1 and DPP-2) needed to complete 
DOE's Accelerated Cleanup strategy for resolution of the legacy wastes 
and materials from the Cold War.
    Response. NRC acknowledges the comment.
    Comment. Many commenters opposed the elimination of the double 
containment requirement because of possible public health and safety 
consequences.
    Response. The commenters provided no basis for their assertions 
that removing the double-containment requirement would increase public 
exposure risks. The NRC staff believes that the current Type B package 
requirements, as applied to all radionuclides, are adequate to protect 
public health and safety.
    Comment. One commenter stated that the principal benefit of 
removing the double containment requirement would be a reduction in 
exposure to the workers. The commenter added that it would also result 
in lower costs.
    Response. NRC acknowledges the comment.
    Comment. One commenter expressed concern that the A1 and 
A2 values have been used as a justification for single-shell 
containers for plutonium.
    Response: The NRC does not agree with this unsubstantiated 
statement that the A1 and A2 values have been 
used as justification for the elimination of the double containment 
requirement for plutonium. The justifications for elimination of the 
double containment requirement were detailed in the proposed rule on 
April 30, 2002 (67 FR 21421 through 21425), and focus more on the fact 
that the original AEC requirement for double containment of plutonium 
was based on existing policy and regulatory concerns and was not risk 
informed. While the A1 and A2 values are 
referenced in the discussion, they are referenced from the standpoint 
that there are other radionuclides with the same or lower A1 
and A2 values than plutonium. Because these radionuclides 
have never required double containment, it cannot be argued from a risk 
standpoint that the shipment of plutonium should be treated any 
differently.
    Comment. Three commenters expressed support for the proposed 
removal of the requirement for ``double containment'' of plutonium from 
Sec. 71.63. One commenter asserted that a single containment barrier is 
adequate for Type B packages containing more than 20 curies of solid 
form plutonium. The commenter further stated that the former AEC's 
rationale for requiring the double containment provision is now moot 
because the expectation for liquid plutonium nitrate shipments has 
never materialized. The commenter also expressed opposition to the 
double containment requirement because it presents continuing costs 
without commensurate benefits. The commenter stated that removing the 
double containment requirement would result in a small and acceptable 
increase in public risk. Furthermore, the requirement removes 
flexibility in package designs that might be needed to meet DOE's 
mission.
    Another commenter expressed concern that the double containment 
requirement was implemented in the 1970s without adequate 
justification.
    The third commenter said that using double containment causes 
unnecessary worker radiation exposure. This commenter said this 
unnecessary worker radiation is estimated to be 1200 to 1700 person-rem 
over a 10-year period. The commenter also said the conditions that 
justified double containment during the early 1970s have disappeared. 
These include large numbers of shipments of nitrate solutions or other 
forms from reprocessing, compounded by crude containment requirements, 
and the absence of quality assurance requirements. This position was 
justified because France, Germany, and the United Kingdom, as well as 
other IAEA Member Nations, no longer require double containment for 
plutonium. The commenter believed that harmonization of part 71 with 
IAEA TS-R-1 was an important goal of this rulemaking because to do so 
would allow for consistent regulation among the principal nations 
shipping nuclear materials. Furthermore, it was recommended that NRC 
eliminate the special requirements for plutonium shipments in Sec. 
71.63 for consistency with the use of prescriptive, performance-based 
safety standards.
    Response. The comments are generally in line with statements in the 
proposed rule on April 30, 2002 (67 FR 21421 through 21425), that 
described the NRC's bases for elimination of the double containment 
requirement.
    Comment. Several commenters stated that double containment provides 
more protection to the public than single containment. One of these 
commenters stated the belief that the commenter and a majority of the 
Western Governors are concerned with the proposal to eliminate the 
double containment requirement for plutonium shipments. The commenter 
stated that ``the regulatory analysis is defective in its failure to 
recognize likely impacts on the agreement among the Western Governors' 
Association, the individual Western States, and DOE for a system of 
extra regulatory transportation safeguards, which we believe are at the 
heart of both government and public acceptance of the WIPP 
transportation program.'' One commenter stated that if

[[Page 3759]]

Sec. 71.63(b) is deleted, there will very likely be some use of single-
contained packages for future WIPP shipments.
    Response. With respect to the last commenter's statement, the use 
of single containment packages for future shipments is one possible 
outcome of the change. NRC acknowledges that agreements between DOE and 
States may be impacted by the elimination of the double containment 
regulatory requirement. However, any change to NRC regulations that 
impact how DOE conducts its transportation operations is a DOE 
decision. As such, DOE and the States may need to negotiate and resolve 
issues related to DOE's operations.
    Comment. One commenter stated that the proposed rule is not risk 
informed and does not use a common sense approach. Another commenter 
stated strong agreement with this first commenter. Another commenter 
recommended that both Sec.Sec. 71.63(a) and (b) be retained but that 
the limit be expressed as 0.74 TBq (20 Ci) for the total of all 
actinides with A2 values equal to or less than 1.0 x 
10-3 TBq (2.7 x 10-2 Ci).
    Response. The NRC believes the decision to eliminate double 
containment is risk informed and reduces an unnecessary regulatory 
burden. In this context, there is adequate actual operating experience 
with Type B package shipments to support the Commission's decision to 
remove the double containment requirement for plutonium packages. There 
are many nuclides with A2 values the same or lower than 
plutonium's that have never required double containment.
    Further, current NRC regulations state that, in certain 
circumstances, plutonium in excess of 0.74 TBq (20 Ci) can be shipped 
as a normal form solid without requiring double containment. The 
shipment of reactor fuel elements containing plutonium is one example. 
Using the most conservative A2 value of 0.00541 Ci, 0.74 TBq 
(20 Ci) of plutonium (Pu-238, Pu-239, Pu-240) equates to an 
A2 multiple of roughly 3700. In contrast, using 19 risk-
significant nuclides (including Am-241) from a typical single boiling 
water reactor spent fuel assembly (reference NUREG/CR-6672, 
``Reexamination of Spent Fuel Shipment Risk Estimates,'' page 7-17), 
one can calculate a curie content of 148,346 Ci with a cumulative 
A2 multiple of just under 790,000 (the assembly also would 
contain an A2 multiple of 455,000 of plutonium nuclides). If 
the A2 multiple is viewed as a measure of potential health 
effect, then from a risk-informed standpoint, the shipment of one 
particular nuclide in a Type B package should not be treated 
differently from any other nuclide of comparable A2 in a 
Type B package. It should be noted that for domestic shipments, there 
is a well established and excellent safety record associated with the 
shipment of spent fuel assemblies in single containment spent fuel 
packages.
    Comment. Two commenters stated that removing the double containment 
requirement would provide health benefits for radiation workers. One 
commenter argued that the cost of reducing the exposure to workers to 
the required 1 mrem/yr would be very high. One commenter asserted that 
we need to balance public safety and the safety of radiation workers.
    Response. As discussed in the draft EA, NRC agrees that the removal 
of the double containment requirement would result in reduced risk to 
radiation workers.
    Comment. One commenter stated that worker exposure estimates are 
not supported by data. Another commenter stated that the conclusion 
that single containment will decrease radiation doses is incorrect for 
WIPP shipments. The commenter contends that radiation doses would 
increase to both workers and the general public.
    Response. The first commenter's remark about lack of data on worker 
exposure estimates was true at the time of the public meeting on June 
24, 2002, where the comment was made. However, during the comment 
period, DOE, one of the major entities affected by the current double 
containment rule, submitted the results of a detailed study they 
performed to evaluate the impacts for elimination of the current 
requirement. In that study, they presented quantifiable data that 
indicates that over a 10-year period, they could expect to see a 
reduction of 1200 to 1700 person-rem if the double containment 
provision is eliminated. The second commenter provided qualitative and 
quantitative information (some of which concerned a non-NRC certified 
cask) that comes to a contrary conclusion. While the NRC does not 
endorse or dispute either study's conclusions, the NRC believes worker 
dose would be reduced due to less handling. Further, radiation 
protection of transport workers (e.g., drivers, inspectors) and the 
public is provided through the package maximum radiation levels set 
forth in DOT regulations, which are not a function of double 
containment.
    Comment. One commenter stated that the NRC has not fully evaluated 
the regulatory impact of the proposed change on the use of the TRUPACT 
II design.
    Response. During the development of the proposed rule, NRC staff 
used all available data to evaluate the costs and benefits of the 
proposed change. NRC staff requested specific information on costs and 
benefits as part of the proposed rule, and the information received was 
considered during the development of a final position. NRC received a 
study from the commenter and, while the NRC does not endorse or dispute 
the study's conclusions, the results are in line with the NRC's 
contention that elimination of the double containment requirement will 
likely result in a reduction in worker radiation exposure.
    Comment. One commenter asked if NRC considers powder a solid form.
    Response. Yes, the NRC has always considered powder as a solid form 
when implementing Sec. 71.63(a). However, powders, under the eliciting 
rule, were not considered as a solid form that was exempt from the 
double containment requirements of Sec. 71.63(b).
    Comment. One commenter endorsed NRC's proposal to retain the 
requirement that shipments whose contents exceed 20 curies of plutonium 
must be made in a solid form as provided under Sec. 71.63(a).
    Response. The comment is acknowledged.
    Comment. One commenter expressed support for the NRC position.
    Response. The comment is acknowledged.
    Comment. Several commenters expressed concern that removing the 
double containment requirement would erode public confidence in the 
Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. One of 
the commenters noted that NRC's decision is not supported by any 
studies to demonstrate that the change is minimal and that NRC should 
only relax the double containment provisions when NRC receives 
scientific evidence that demonstrates beyond a reasonable doubt that 
single containment is as safe as double containment for shipments to 
WIPP. Another commenter cited the economic, shipping, and public 
confidence aspects of a severe accident release as the primary 
arguments in support of retaining double containment.
    Response. The comments are acknowledged. With regard to the last 
commenter's citation, as is the case with other nuclides, NRC-certified 
Type B packagings provide for safety in transportation accidents. With 
regard to non-safety focused arguments (economic and public confidence

[[Page 3760]]

issues), as well as the other commenter's concerns, the reader is 
referred to a related discussion earlier on this issue, under the 
heading: Analysis of Public Comments on the Issues Paper.
    Comment. One commenter discussed an incident involving the shipment 
of plutonium-containing transuranic waste to DOE's Waste Isolation 
Pilot Plant in New Mexico. A truck carrying TRU waste was involved in a 
traffic accident. While no radiation was released, the inner container 
was discovered to be contaminated with radiation to the extent that it 
could not be unloaded. The commenter pointed out that the double-walled 
container provided a margin of safety that would not have existed under 
the proposed rule. The commenter stated that the incident underscores 
the importance of maintaining the double containment requirement, as it 
has been a crucial element in the success of the WIPP TRU waste 
shipping campaign to date.
    Response. In the cited case, NRC staff understands that neither 
containment was compromised due to the accident.
    Comment. One commenter stated that all shipping requirement 
revisions should be more, rather than less, protective of public 
health. Two other commenters stated that the AEC's original 1974 
reasoning for imposing the double containment requirements was still 
valid, including the possibility for human error and expected increases 
in the number of shipments. The commenter also responded to the claim 
that adopting a single containment requirement would be safer for 
personnel who handle the inner container by stating that this may 
simply be a shifting of risk from personnel to the public.
    Response. The comment that shipping requirement revisions should 
all be more, rather than less, protective of public health, is 
acknowledged. The NRC's transportation regulations are designed to 
provide adequate protection to the public health and safety from 
radioactive material transportation activities. In doing so, NRC seeks 
to balance its regulations by ensuring public health and safety while 
at the same time not creating unnecessary regulatory burden.
    Regarding the comment that the AEC's original 1974 reasoning for 
imposing double containment is still valid, the NRC notes that the 
AEC's original reasoning was based on the fact of transporting liquids; 
that is no longer the case. The justifications for elimination of the 
double containment requirement detailed in the proposed rule on April 
30, 2002 (67 FR 21421 through 21425) is based on technical arguments 
and focus on the confidence in Type B packages. While there is an 
increase in the number of shipments to WIPP, the vast majority of these 
shipments do not involve liquids.
    The NRC disagrees with the comment that while the adoption of a 
single containment requirement would be safer for personnel who handle 
the inner container, this constitutes a shifting of the risk from 
personnel to the public. The NRC believes that the risk of shipping 
plutonium in a single containment Type B package is no different than 
that of shipping other radionuclides with the same or lower 
A1 and A2 values than plutonium.
    Comment. One commenter stated that although spent fuel that is 
damaged to the extent that the rod cladding's integrity is in question 
may be subject to the requirements of Sec. 71.63, it is not clear that 
all damaged fuel will require double containment.
    Response. NRC has previously published guidance (ISG-1, Rev. 1, 
dated October 25, 2002) on when the double containment provision is 
required for damaged spent fuel. Basically, canning (double 
containment) is required if the spent fuel contains known or suspected 
cladding defects greater than a pinhole leak or hairline crack that 
have the potential for release of significant amounts of fuel into the 
cask.
    Comment. One commenter stated that additional procedures (e.g., 
closures and testing) are required to implement Sec. 71.63, which leads 
to added worker exposures. The commenter provided quantitative and 
monetized data detailing the extra time and amount of money that the 
double containment requirement imposes on TRU Waste, Plutonium Oxides, 
and Damaged Spent Nuclear Fuel Operations.
    Response. NRC acknowledges this comment.
    Comment. One commenter stated that additional containment systems 
reduce cask capacities and consequently require more shipments to move 
the same material. This commenter also said that the double containment 
represents extra weight that must be moved and then provided estimates 
of the cost for moving the extra weight in the double-containment 
structure in the cases of TRU Waste, Plutonium Oxides, and Damaged 
Spent Nuclear Fuel operations.
    Response. The comment is acknowledged.
    Comment. One commenter stated that design costs and costs for NRC 
certification services are incurred by increased design complexity 
relating to the provision of the double-containment barrier. The 
commenter noted that the alternative to the design and certification 
cost penalty is to petition for an exemption under Sec. 71.63(b)(4); 
however, preparing this petition is time-consuming and probably similar 
in cost to getting a separate containment boundary designed and 
certified. The commenter estimated certification and capital cost 
penalties for the cases of CH-TRU and RH-TRU Wastes, Plutonium Oxides, 
DHLW Glass Exemption, and Damaged Spent Nuclear Fuel.
    Response. The comment is acknowledged.
    Comment. One commenter stated that while the restrictions of Sec. 
71.63 remain in effect, it must continue to expend funds unnecessarily 
for double-containment packaging. This commenter provided tables of 
monetized breakdowns of these estimates. The commenter estimated that 
the net result from all three areas (TRU wastes, plutonium oxides and 
residues, and damaged spent nuclear fuel) is that double-containment 
requirements will produce an avoidable cost of approximately $12 
million in capital cost, $20 million in operational cost, and $26 
million to $40 million in shipping and receiving costs. In addition, 
the commenter estimated that the double containment requirement will 
result in additional worker radiation exposure amounting to 1250 to 
1770 person-rem.
    Response. The commenter has provided information that appears to 
support the NRC's contention that removal of double containment would 
provide for cost savings and decreased personnel exposure.
    Comment. One commenter stated that double containment provides some 
additional protection to the public in both normal and accident 
situations. The commenter stated that most of this additional 
protection relates to a potential reduction in population exposure. 
However, the commenter estimated that the total radiation exposure 
reduction in most cases amounts to a maximum of about 30 person-rem/
year distributed among a potentially exposed population of tens of 
millions of persons. The commenter stated that such an effect would not 
be perceptible.
    Response. NRC acknowledges the comment.
    Comment. One commenter stated that, although double containment 
reduces the risk incurred by the public of exposure to radiation from 
the package in incident-free transport, the reduction is likely to be 
relatively small. The dose rate is already small enough at distances

[[Page 3761]]

where the public is likely to be exposed that the impact of single-or 
double-contained material will not be consequential. This commenter 
also noted that one effective containment boundary is sufficient to 
meet containment requirements implicit in Type B design approvals, but 
the materials shipped are already within one or more inner containers. 
The commenter believes the presence of these redundant containers 
effectively rules out any problems that might result from human errors 
in achieving a required level of leak-tightness for single contained 
Type B packages.
    Response. NRC acknowledges the comment.
    Comment. One commenter stated that doubly contained packages pose 
lower risks and is not, by itself, sufficient justification for using 
doubly contained packages. The commenter stated that, in general, the 
likelihood of achieving an accident sufficient to compromise 
containment of a singly contained Type B package has been estimated to 
be fewer than 1 in 200 in the event of a severe accident. Achieving 
damage to two redundant containments could be expected to be as much as 
a factor of 10 lower risk relative to the single containment case. The 
commenter stated that this is not as large a benefit as it may seem; 
the decrease in absolute risk will be very small because the risk of 
shipping singly contained plutonium is exceedingly small to start. The 
commenter provided monetized and quantified estimates of the cost/risk 
tradeoffs associated with double-containment versus single-containment 
for the handling of Contact-Handled TRU Waste, Plutonium Oxide and 
Plutonium-Bearing Wastes, Remote-Handled TRU Waste, and Failed Fuel.
    Response. NRC acknowledges the comment.
    Comment. Two commenters stated that if the NRC continues to pursue 
the proposal to relax the plutonium shipment double containment 
standards, then it should conduct a series of hearings on the 
rulemaking, with at least one of those hearings held in the western 
U.S. Another commenter objected to the lack of public education 
regarding the ``numerous, confusing, and complicated'' proposed rule 
changes, which, when presented as they were, encourage nonengagement. 
The commenter requested that an extension be placed on the comment 
period and that ``ordinary'' language be used to explain the actual 
proposals, how they will impact public health, what agencies and rules 
are involved, and how one can easily reply to all agencies involved in 
these proposals by mail, email, or fax.
    Response. The rulemaking process does not include the opportunity 
for formal hearings because the proposed rulemaking is not a licensing 
action, which does require hearings. The NRC staff thinks that the 
commenter meant holding public meetings to discuss the issue. Hearings 
were held in this rulemaking in the form of public meetings. Two 
meetings were held in June 2002, in Chicago, IL, and the NRC TWFN 
Auditorium, and 3 meetings were held in NRC Headquarters, Atlanta, GA, 
and Oakland, CA, during August and September 2000. The NRC did not 
extend the 90-day public comment period, because the public had ample 
opportunity to comment on this rule during the 1-year period following 
March 2001, when the proposed rule was posted on the Secretary of the 
Commission Web site.
Issue 18. Contamination Limits as Applied to Spent Fuel and High-Level 
Waste (HLW) Packages
    Summary of NRC Final Rule. The final rule does not adopt any 
changes to part 71 for this issue because experience with regulations 
requiring that licensees monitor the external surfaces of labeled 
radioactive material packages for contamination upon receipt and 
opening indicates the rate of packages exceeding allowable levels en 
route is low, and therefore, in transit decontamination of packages is 
not warranted. Further, requiring such decontamination of packages 
could result in a significant increase in worker doses without a 
commensurate increase in public health and safety.
    Affected Sections. None (not adopted).
    Background. In the period of December 1997 through April 1998, the 
French Nuclear Installations Safety Directorate inspected a French 
nuclear power plant and railway terminal used by La Hague reprocessing 
plant. The inspectors noticed that, since the beginning of the 1990's, 
a high percentage of spent fuel packages and/or railcars had a level of 
removable surface contamination that exceeded IAEA regulatory limits by 
as much as a factor of 1000. Subsequent investigations found that the 
contamination incidents involved shipments from other European 
countries, and the French transport authorities notified their 
counterparts of their findings. Subsequently, French, German, Swiss, 
Belgian, and Dutch spent fuel shipments were temporarily suspended.
    After estimating the occupational and public doses from the 
contamination incidents, the European transport authorities concluded 
that these incidents did not have any radiological consequence. The 
contamination was believed to be caused by contact of the spent fuel 
package surface with contaminated water from the spent fuel storage 
pool during package handling operations. The authorities concluded that 
there were deficiencies in the contamination measurement procedures and 
the distribution of that information.
    Media reports on these incidents focused attention on IAEA's 
regulations for removable contamination on package surfaces. TS-R-1 
contains contamination limits for all packages of 4.0 Bq/cm2 
for beta and gamma and low toxicity alpha emitting radionuclides, and 
0.4 Bq/cm2 for all other alpha emitting radionuclides. 
Although TS-R-1 uses the term ``limit,'' IAEA considers these 
``limits'' to be guidance values, or derived values, above which 
appropriate action should be considered. In cases of contamination 
above the limit, that action is to decontaminate to below the limits.
    TS-R-1 further provides that in transport, ``* * * the magnitude of 
individual doses, the number of persons exposed, and the likelihood of 
incurring exposure shall be kept as low as reasonable, economic and 
social factors being taken into account * * *'' The IAEA contamination 
regulations have been applied to radioactive material packages in 
international commerce for almost 40 years, and practical experience 
demonstrates that the regulations can be applied successfully. With 
respect to contamination limits, TS-R-1 contains no changes from 
previous versions of IAEA's regulations.
    Part 71 does not contain contamination limits, but Sec. 71.87(i) 
requires that licensees determine that the level of removable 
contamination on the external surface of each package offered for 
transport is as low as is reasonably achievable, and within the limits 
specified in DOT regulations in 49 CFR 173.443.
    The IAEA established a Coordinated Research Project (CRP) to review 
contamination models, approaches to reduce package contamination, 
strategies to address cask-weeping, and possible recommendations for 
revisions to the contamination standard that consider risks, costs, and 
practical experience. The IAEA CRP facilitates the investigation of 
radioactive material transportation issues by key IAEA Member States. 
IAEA is considering the CRP report, and any further actions or remedies 
that may be warranted are being addressed by the IAEA Transportation 
Safety Standards Committee (TRANSSC). NRC supported

[[Page 3762]]

the IAEA initiative to establish the CRP, and NRC would participate in 
the IAEA review of surface contamination standards.
Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. One commenter expressed support of the NRC position not to 
change from current standards.
    Response. The NRC acknowledges these comments. No further response 
necessary.
    Comment. One commenter requested that the NRC keep ``removable 
contamination of external ``spent'' fuel shipping packages'' to the 
``absolute minimum attainable, even if extra cost is incurred in doing 
so.'' The commenter added that ``full data on container surface 
contamination must be kept and submitted to the regulatory agency as 
part of required manifest records.''
    Response. Keeping contamination to an absolute minimum could result 
in a significant increase in worker dose, due to the additional 
exposures required to achieve that low level of contamination, without 
a commensurate increase in public health and safety. Current DOT 
regulations require that shippers be able to provide to inspectors upon 
request documentation that supports the shipper's certification that 
radioactive material shipments were made in compliance with applicable 
requirements, including contamination limits. This practice has worked 
well, and NRC has no basis to change it.
    Comment. One commenter stated that the NRC's measures should allow 
for decontamination of nuclear waste shipments during transport if they 
begin to exceed allowable radiation levels en route. The commenter 
stated that this would reduce exposure to the public and prevent 
shipments from having to return to the point of origin.
    Response. Current NRC regulations require that licensees monitor 
the external surfaces of labeled radioactive material packages for 
contamination upon receipt and opening (see details at Sec. 
20.1906(b)(1)). Based on its experience with these regulations, the 
rate of packages exceeding allowable levels en route is low, and NRC 
does not believe that in transit decontamination of packages is 
warranted.
    Comment. One commenter asserted that there is no reason to seek any 
special dose consideration or reduction in the handling and transport 
of spent fuel or storage casks. The commenter added that industry has 
not attributed any problems with decontamination and dose to the 
handling and transport of spent fuel or storage casks. The commenter 
did note that although industry did experience some of the weeping 
issues in the early 1990's, industry has taken steps to eliminate this 
condition.
    Response. NRC agrees that incidents of cask weeping have subsided 
in recent years. However, NRC notes that considerable occupational dose 
is expended to achieve compliance with current regulatory limits that 
do not appear to be risk-informed, and that occupational and public 
doses associated with spent fuel cask surface contamination limits do 
not appear to be optimized.
    Comment. One commenter requested that the NRC not relax ``radiation 
protection in any shipments, especially high-level wastes and intensely 
irradiated ``spent'' fuel,'' the reason being that, in the near future, 
shipments of high-level wastes and spent fuel may increase in number, 
and this would justify NRC staff's maintaining ``maximum control * * * 
as a principal goal of the NRC.'' The commenter also stated that while 
``Europeans may dismiss contamination ``incidents'' as having no 
radiological consequences * * * that is not convincing, in view of 
recent research findings concerning adverse impacts of low-level 
radiation at the cellular and molecular levels.''
    Response. No change to the contamination limit is being adopted in 
the final rule, and no relaxation of radiation protection has been 
proposed.
    Comment. Two commenters expressed opposition to allowing greater 
contamination on surfaces of irradiated fuel and high-level radioactive 
waste containers and supported NRC's decision to refuse this. Two other 
commenters supported the NRC's proposal to make no changes in the 
contamination levels for these packages.
    Response. No response is necessary.
    Comment. One commenter expressed opposition to allowing greater 
contamination on surfaces of irradiated fuel and high level radioactive 
waste containers.
    Response: The NRC acknowledges these comments. No response is 
necessary.
Issue 19. Modifications of Event Reporting Requirements
    Summary of NRC Final Rule. The final rule revises, in Sec. 71.95, 
the event reporting submission period to provide a written report from 
30 to 60 days. Other regulatory requirements to orally notify the NRC 
Operations Center promptly of an event and for licensees to report 
instances of failure to follow the conditions of the CoC while 
packaging was in use remain unchanged. The revision lengthening the 
time for submission of the written report is consistent with changes to 
similar requirements in Part 50.
    Affected Sections. Section 71.95.
    Background. The Commission recently issued a final rule to revise 
the event reporting requirements in Part 50 (see 65 FR 63769; October 
20, 2000). This final rule revised the verbal and written event 
notification requirements for power reactor licensees in Sec.Sec. 50.72 
and 50.73. In SECY-99-181,\13\ NRC staff informed the Commission that 
public comments on the proposed part 50 rule had suggested that 
conforming changes also be made to the event notification requirements 
in part 72 (Licensing Requirements for the Independent Storage of Spent 
Fuel) and part 73 (Physical Protection of Plants and Materials). In 
response, the Commission directed the NRC staff to study whether 
conforming changes should be made to parts 72 and 73. During this 
study, the NRC also reviewed the part 71 event reporting requirements 
in Sec. 71.95 and concluded that similar changes could be made to the 
part 71 event reporting requirements.
---------------------------------------------------------------------------

    \13\ SECY-99-181, ``Proposed Plans and Schedules to Modify 
Reporting Requirements Other than 10 CFR 50.72 and 50.73 for Power 
Reactors and Material Licensees,'' dated July 9, 1999.
---------------------------------------------------------------------------

Analysis of Public Comments on the Proposed Rule
    A review of the comments and the NRC staff's responses for this 
issue follows:
    Comment. Two commenters expressed support for the proposed 
modifications. One commenter stated that the proposed modifications to 
event reporting requirements will enhance safety. The other commenter 
noted that many States respond to incidents involving radioactive 
materials on a regular basis and would not want to wait until the full 
60 days for reporting purposes.
    Response. The NRC acknowledges the comments supporting the change 
to require a 60-day report instead of a 30-day report for a 
transportation event. The comment that States would need to respond to 
incidents and would need reports sooner than 60 days is not consistent 
with the fact that prompt reporting to the National Response Center, 
NRC Operations Center, and appropriate State Authorities occurs after 
an event. The written report to the NRC will not affect this practice. 
Therefore, the change in the time to

[[Page 3763]]

provide a written report would have no effect on the emergency response 
and information exchange actions that would still be performed by 
licensees or the DOT National Response Center. Therefore, no changes in 
the proposed rule language are being made.
    Comment. One commenter asked how this proposed change affects other 
parts of the proposed rulemaking and urged the NRC to ensure that it 
conforms with the rest of the proposed rulemaking.
    Response. There are no other impacts on the regulations associated 
with adopting this specific change.
    Comment. Two commenters opposed the proposed event reporting 
requirements. The first commenter stated that there should never be a 
30-or 60-day ``delay in filing a report on any event involving 
malperformance of a package or container,'' but that a report should be 
filed immediately with the NRC when a problem occurs. The second 
commenter suggested that ``reporting should serve the needs of the 
(NRC) staff-and public safety,'' rather than the licensee. This 
commenter also claimed that an extra 30 days may be too long an 
extension if there is a serious safety problem.
    Response. The NRC notes that if a serious safety problem resulted 
from an incident, it would be reported promptly to the NRC Operations 
Center. The NRC staff notes that a review of the regulatory analysis 
included in the proposed rule stated that: ``In new paragraph (a)(3), 
[of section 71.95] the NRC would retain the existing requirement for 
licensees to report instances of failure to follow the conditions of 
the CoC while a packaging was in use.'' This section was inadvertently 
left out of the proposed rule language and was added to the final rule.
    Comment. One commenter indicated concern about the lack of data to 
support NRC's position on extending the reporting period from 30 to 60 
days.
    Response. There is sufficient rationale as reflected in other 
regulations for reducing the regulatory burden related to the time for 
submitting written reports. See the discussion in the proposed rule 
(April 30, 2002; 67 FR 21427) for additional detail on the 
justification for the change. Therefore, no change to the rule is 
proposed.
    Comment. One commenter was concerned about difficulties in 
compiling a jointly written report by the certificate holder and the 
shipper if they are in different countries.
    Response. The commenter's concern about coordination of a jointly 
written event report is valid; however, the longer time being proposed 
for submitting an event report should accommodate delays in the 
communication interface and help ensure completion within the 60-day 
reporting period. Therefore, no changes have been made to the proposed 
rule language.
    Comment. One commenter found the event reporting requirements 
unclear in two places. The proposed rule would direct the licensee to 
request information from certificate holders; however, neither the 
supporting discussion nor regulatory text addresses a situation in 
which a certificate holder declines to provide comments. The commenter 
asked whether the licensee's obligation would be satisfied at the point 
that a request is made to CoC holders. The commenter also found it 
unclear whether NRC intended to exempt DOT specification and foreign 
package designs holding U.S. validations from the reporting 
requirements. The commenter asserted that if NRC intends to make a 
distinction between NRC-approved packages and other authorized 
packages, it may be necessary to develop separate QA procedures and 
related instructions. The impacts on resources associated with such 
development may require further investigation.
    Response. Regarding the first question about what would happen if a 
licensee did not receive supporting information in its process to issue 
an event report to the NRC to comply with the requirements of Sec. 
71.95, the NRC notes that the licensee should make an earnest attempt 
to obtain relevant information from the CoC holder. In the case where 
the CoC holder refused to provide input to the report, the licensee 
would still need to submit the report to the NRC within the 60-day time 
period. NRC technical staff would determine if CoC staff input should 
have been included in the report and would obtain it directly from the 
CoC holder as necessary. Further, if the NRC determined that the CoC 
holder's lack of support resulted in a report that was incorrect or 
incomplete, then the NRC would pursue appropriate regulatory action 
against the CoC holder.
    Regarding the second question about the reporting requirement being 
applicable to DOT specification and foreign package designs with U.S. 
validation, the NRC notes that its regulations only apply directly to 
its licensees or CoC holders. NRC will, however, forward this comment 
to DOT for appropriate consideration. No change to NRC rule language is 
being made.
    Comment. One commenter stated that the requirement of the CoC 
holder to rely on other licensees or registered users, over whom the 
holder has no authority or control, to identify problems or package 
deficiencies, is inappropriate and must be modified. Another commenter 
stated that the authorized package user should be making the required 
report.
    Response. Both comments deal with the original language in the 
existing Sec. 71.95 which states that licensees are responsible for 
providing event reports to the NRC.

IV. Section-by-Section Analysis

    Several sections in part 71 are redesignated in this rulemaking to 
improve consistency and ease of use. For some sections, only the 
section number is changed. However, for other sections, revisions are 
being made to the regulatory language. The following table is provided 
to aid the public in understanding the numerical changes to sections of 
part 71.

                           Redesignation Table
------------------------------------------------------------------------
            New section number                 Existing section number
------------------------------------------------------------------------
Sec. 71.8.................................  Sec. 71.11.
Sec. 71.9.................................  New section.
Sec. 71.10................................  New section.
Sec. 71.11 (Reserved).....................  NA.
Sec. 71.12................................  Sec. 71.8.
Sec. 71.13................................  Sec. 71.9.
Sec. 71.14................................  Sec. 71.10.
Sec. 71.15................................  Sec. 71. 53.
Sec. 71.16 (Reserved).....................  NA.
Sec. 71.17................................  Sec. 71.12.
Sec. 71.18 (Reserved).....................  NA.
Sec. 71.19................................  Sec. 71.13.
Sec. 71.20................................  Sec. 71.14.
Sec. 71.21................................  Sec. 71.16.
Sec. 71.22................................  Sec. 71.18.
Sec. 71.23................................  Sec. 71.20.
Sec. 71.24 (Reserved).....................  Sec. 71.22 (Section
                                             removed).
Sec. 71.25 (Reserved).....................  Sec. 71.24 (Section
                                             removed).
Sec. 71.53 (Reserved).....................  Sec. 71.53 (Section
                                             redesignated).
------------------------------------------------------------------------

Subpart A--General Provisions

Section 71.0 Purpose and scope
    Paragraph (d) has been reformatted into three paragraphs to 
simplify this regulation and to better use plain language. Paragraph 
(d)(1) indicates that general licenses, for which no NRC package 
approval is required, are issued in new Sec.Sec. 71.20 through 71.23. 
This change reflects the removal of existing Sec.Sec. 71.22 and 71.24 
(redesignated Sec.Sec. 71.24 and 71.25 (Reserved)). Paragraph (d)(2) 
indicates that an application for package approval must be completed in 
accordance with subpart D. Paragraph (d)(3) continues to

[[Page 3764]]

require a licensee transporting, or delivering material to a carrier 
for transport, to meet the requirements of the applicable portions of 
subparts A, G, and H.
    New paragraph (e) has been added to indicate that persons who hold, 
or apply for, a part 71 CoC for Type AF, Type B, Type BF, Type B(U)F, 
or Type B(M)F packages are within the scope of part 71 regulations.
    Existing paragraphs (e) and (f) have been redesignated as new 
paragraphs (f) and (g), respectively. The rule text in new paragraph 
(f) is the same as existing paragraph (e) text. New paragraph (g) has 
been revised to reflect the redesignation of existing Sec. 71.11 as new 
Sec. 71.8.
Section 71.1 Communications and Records
    In Sec. 71.1, paragraph (a) has been revised to indicate that 
documents submitted to the NRC should be addressed to the attention of 
the ``Document Control Desk,'' not the ``Director of the Office of 
Nuclear Material Safety and Safeguards.'' Provisions have also been 
added to provide requirements when a due date for a document falls on a 
Saturday, Sunday, or Federal holiday. In that case, the document would 
be due the next Federal workday. This change is identical to a change 
made to Sec. 72.4 in a recent part 72 final rule (see 64 FR 33178; June 
22, 1999).
Section 71.2 Interpretations
    No changes were made to the text of this section; however, it has 
been retained in the revision of this subpart for completeness.
Section 71.3 Requirement for License
    No changes were made to the text of this section; however, it has 
been retained in the revision of this subpart for completeness.
Section 71.4 Definitions
    The existing definitions for ``A1,'' ``Fissile 
material,'' ``Low Specific Activity (LSA) material,'' ``Package,'' and 
``Transport index (TI)'' are revised as conforming changes. New 
definitions for ``A2,'' ``Certificate of Compliance,'' 
``Consignment,'' ``Criticality Safety Index (CSI),'' ``Deuterium,'' 
``U.S. Department of Transportation (DOT),'' ``Graphite,'' ``Spent 
fuel,'' and ``unirradiated uranium'' have been added as conforming 
changes.
    The definition of ``A1'' has been revised to split the 
previous combined definition for ``A1'' and 
``A2'' into two individual definitions. This approach is 
consistent with the standard in TS-R-1. Furthermore, no change has been 
made to the current technical content of the definition for 
``A1''; however, the text is revised to improve readability.
    A definition for ``A2'' has been added, because the 
previous joint definition for ``A1'' and ``A2'' 
has been split into two definitions. (See also definition for 
``A1.'')
    A definition for ``Certificate of Compliance'' has been added. This 
definition is similar to the definition for the same term found in Sec. 
72.3.
    A definition for ``Consignment'' has been added.
    A definition of ``Criticality Safety Index (CSI)'' has been added.
    A definition of ``Deuterium'' has been added that applies to new 
Sec.Sec. 71.15 and 71.22.
    A definition of ``U.S. Department of Transportation (DOT)'' has 
been added.
    The definition of ``Fissile material'' has been revised by removing 
238Pu from the list of fissile nuclides; clarifying that 
``fissile material'' means the fissile nuclides themselves, not 
materials containing fissile nuclides; and redesignating the reference 
to exclusions from fissile material controls from Sec. 71.53 to new 
Sec. 71.15.
    A definition of ``Graphite'' has been added that applies to new 
Sec.Sec. 71.15 and 71.22.
    The definition of ``Low Specific Activity (LSA)'' material (LSA-I, 
LSA-II, and LSA-III) has been revised to be consistent with DOT, and to 
reflect the existence of Sec. 71.77 (Sec. 71.77 provides requirements 
on the qualification of LSA-III material).
    A definition for ``Optimum interspersed hydrogenous moderation'' 
has been added (the definition itself was included in the proposed rule 
Sec. 71.4, but, inadvertently, no mention of that fact was made in this 
Section).
    The definition of ``Package'' has been revised by clarifying in 
paragraph (1) that Fissile material package also means a Type AF, Type 
BF, Type B(U)F, or Type B(M)F package. New paragraph (2) has been added 
defining Type A packages in accordance with DOT regulations contained 
in 49 CFR Part 173. Existing paragraph (2) defining Type B packages has 
been redesignated as subparagraph (3). No changes have been made to the 
redesignated text.
    A definition of ``Spent nuclear fuel'' or ``Spent fuel'' has been 
added. This definition is the same as that currently found in Sec. 
72.3.
    The definition for ``Transport index (TI)'' has been revised to 
reflect the new definition of Criticality Safety Index; however, the 
method for determining the TI of a package, based on the package's 
radiation dose rate, remains unchanged.
    A definition for ``unirradiated uranium'' has been added as it is 
part of the LSA-I definition.
Section 71.5 Transportation of Licensed Material
    No changes were made to the text of this section; however, it has 
been included in the revision of this subpart for completeness.
Section 71.6 Information Collection Requirements: OMB Approval
    This section has been redesignated from subpart B, Exemptions, to 
subpart A, General Provisions. Paragraph (b) of this section has been 
revised as a conforming change to reflect the addition of new 
information collection requirements. Additionally, the existing 
information collection requirement in Appendix A to part 71, paragraph 
II, was inadvertently omitted from the list of approved information 
collection requirements in a previous rulemaking; consequently, NRC 
staff has added Appendix A, paragraph II, to paragraph (b) to correct 
this error. Furthermore, the reference to Sec. 71.6a has been removed, 
because no such section currently exists in part 71.
Section 71.7 Completeness and Accuracy of Information
    This section has been redesignated from subpart B, Exemptions, to 
subpart A, General Provisions. Further, paragraphs (a) and (b) have 
been revised by adding the terms ``certificate holder'' and ``applicant 
for a CoC.''
Section 71.8 Deliberate Misconduct
    This section has been redesignated from subpart B, Exemptions, to 
subpart A, General Provisions. Further, in subpart A, Sec. 71.11 has 
been redesignated as Sec. 71.8. However, the current text of Sec. 71.11 
has not changed in the redesignated Sec. 71.8.
Section 71.9 Employee Protection
    New Sec. 71.9 has been added to provide requirements on employee 
protection. Currently, requirements relating to the protection of 
employees against firing or other discrimination when the employee 
engages in certain ``protected activities'' are provided under the 
parts of title 10 for which a specific license was issued to possess 
radioactive material. However, no provisions were provided in part 71 
relating to the protection of employees against firing or other 
discrimination when employees engage in certain ``protected 
activities'' when they are the employees of a certificate holder or 
applicant for a CoC.

[[Page 3765]]

The NRC believes these employees should also be afforded the same 
rights and protection as are currently afforded employees of licensees. 
The new section is identical to the existing Sec. 72.10, ``Employee 
protection.'' In including licensees in the new Sec. 71.9, the NRC 
recognizes that the potential for duplication occurs for licensees 
regulated under multiple title 10 parts. However, the NRC believes that 
by including licensees along with certificate holders and applicants 
for a CoC, improved regulatory clarity would be achieved, and any 
potential confusion would be minimized.
Section 71.10 Public Inspection of Application
    A new section has been added indicating that applications and 
documents submitted to the Commission, in connection with an 
application for a package approval, shall be available for public 
review in accordance with the provisions of parts 2 and 9. This new 
section is similar to existing Sec. 72.20. Existing Sec. 71.10 has been 
redesignated Sec. 71.14 with changes to the text as discussed under 
Sec. 71.14, below.
Section 71.11 (Reserved)
    This section has been redesignated from subpart B, Exemptions, to 
subpart A, General Provisions, and is reserved. Existing Sec. 71.11 has 
been redesignated as Sec. 71.8.

Subpart B--Exemptions

Section 71.12 Specific Exemptions
    Existing Sec. 71.8 has been redesignated as Sec. 71.12. No changes 
have been made to the contents of this section. Existing Sec. 71.12 has 
been redesignated as Sec. 71.17, with changes to the text as discussed 
under Sec. 71.17, below.
Section 71.13 Exemption of Physicians
    Existing Sec. 71.9 has been redesignated as Sec. 71.13. No changes 
have been made to the contents of this section. Existing Sec. 71.13 has 
been redesignated as Sec. 71.19, with changes to the text as discussed 
under Sec. 71.19, below.
Section 71.14 Exemption for Low-Level Materials
    Existing Sec. 71.10 has been redesignated as Sec. 71.14. Existing 
Sec. 71.14 has been redesignated as Sec. 71.20, with no changes to the 
text.
    In new Sec. 71.14, paragraph (a) has been revised by removing the 
existing single 70 Bq/g (0.002 [mu]Ci/g) specific activity value. 
Additionally, paragraph (a) has been reformatted by adding two new 
paragraphs. Subparagraph (a)(1) provides an increased exemption for 
natural radioactive materials and ores. Subparagraph (a)(2) provides an 
exemption for radioactive material based on the ``Activity 
Concentration for Exempt Material'' and the ``Activity Limit for Exempt 
Consignment'' found in Table A-2 in Appendix A to part 71.
    Paragraph (b) has been revised to consolidate the exemption 
provisions for LSA and SCO material. The LSA and SCO exemptions 
contained in existing paragraphs (b)(2) and (c) of this section have 
been consolidated into a revised paragraph (b)(3). The reference to 
material exempt from classification as fissile material has been 
revised from Sec. 71.53 to Sec. 71.15, because of the redesignation of 
the section.
    Existing paragraph (b)(3) has been removed. The 0.74-TBq (20-Ci) 
exemption for special form americium and special form plutonium has 
been removed. However, the 0.74-TBq (20-Ci) exemption for special form 
plutonium-244, transported in domestic commerce, has been retained as 
new paragraph (b)(2). For international shipments, the A1 quantity 
limit for special form plutonium-244 continues to apply.
Section 71.15 Exemption From Classification as Fissile Material
    Existing Sec. 71.11 has been redesignated as Sec. 71.8. Existing 
Sec. 71.53 has been redesignated as Sec. 71.15, and relocated to 
subpart B with the other part 71 exemptions. This section has been 
revised by providing mass-ratio based limits in classifying fissile-
exempt material. This approach removes the concentration- and 
consignment-based limits of the current Sec. 71.53 and returns to 
package-based mass limits, with required minimum ratios of nonfissile-
to-fissile mass.
    The title has been changed to ``Exemption from classification as 
fissile material.''
    New paragraph (a) has been added and allows for small samples of 
fissile material to be shipped. In paragraph (b), the fissile mass per 
package is limited to 15 grams with a nonfissile-to-fissile mass ratio 
of 200:1. In paragraph (c), the allowed provided there is less than 150 
g of fissile material per 360 Kg ratio of nonfissile-to-fissile 
material is also raised to 2000:1. The mass of any lead, graphite, 
beryllium, and deuterium in the package cannot be included in 
determining the nonfissile material mass.
    In current Sec. 71.53, paragraph (c) has been redesignated as 
paragraph (e), and has been reformatted and revised to clarify that the 
nitrogen to uranium atomic ratio, for shipments of liquid uranyl 
nitrate, must be greater than or equal to 2.0. A new requirement has 
been added specifying the use of DOT Type A packaging.
    In current Sec. 71.53, paragraph (d) has been redesignated as 
paragraph (e), and has been reformatted and revised to clarify the mass 
limits for plutonium. No substantive changes have been made to this 
paragraph.
Section 71.16 (Reserved)
    This section has been redesignated from subpart C, General 
Licenses, to subpart B, Exemptions, and is reserved. Further, existing 
Sec. 71.16 has been redesignated as Sec. 71.21. However, the current 
text of Sec. 71.16 has not been changed in the redesignated Sec. 71.21.

Subpart C--General Licenses

Section 71.17 General License: NRC-Approved Package
    Existing Sec. 71.12 has been redesignated as Sec. 71.17. The text 
of paragraphs (a) and paragraph (b) has not been changed.
    Paragraph (c)(3) has been revised using plain language and to 
reflect the NRC's requirement to address information submitted to the 
NRC to the attention of the NRC's Document Control Desk, in accordance 
with Sec. 71.1.
    Paragraph (d) has not been changed.
    Paragraph (e) has been revised to reflect the redesignation of Sec. 
71.13 to Sec. 71.19. No other change was made for this paragraph.
Section 71.18 Reserved
Section 71.19 Previously Approved Package
    Existing Sec. 71.13 has been redesignated as Sec. 71.19. Paragraph 
(a) has been revised to reflect the current package designators (e.g., 
B(U)F, B(M)F, AF) and to reflect the redesignation of Sec. 71.12 to 
Sec. 71.17. Additionally, the contents of paragraph (a)(2) have been 
removed to reflect that these packages are no longer recognized 
internationally. Existing paragraph (a)(3) has been redesignated as 
(a)(2) with no change to the contents. Also, an expiration date for 
grandfathering these packages has been established in new paragraph 
(a)(3). Paragraph (b) has been updated to remove the LSA packages, as 
these packages no longer exist, and to reflect the redesignation of 
Sec. 71.12 to Sec. 71.17. No other changes were made. A new paragraph 
(c) has been added to reflect the type B(U) and B(M) packages that have 
met the requirements of IAEA Safety Series 6 1985 (as amended 1990) and 
to correct a typographical error. Additionally, a date by which 
fabrication of these packages must be complete has been added. Existing 
paragraph (c) has been redesignated as

[[Page 3766]]

paragraph (d). Existing paragraph (d) has been redesignated as 
paragraph (e) and updated to reflect the identification number suffix 
of ``-96'' for previously approved package designs that have been 
resubmitted for review by the NRC and have been approved, and to remove 
the package designated as Type A from this paragraph.
Section 71.20 General License: DOT Specification Container
    Existing Sec. 71.14 has been redesignated as Sec. 71.20. No changes 
have been made to the contents of paragraphs (a) through (d). New 
paragraph (e) has been added to indicate that these types of packages 
will be phased out 4 years after the effective date of this final rule.
Section 71.21 General License: Use of Foreign Approved Package
    Existing Sec. 71.16 has been redesignated as Sec. 71.21. No changes 
have been made to the contents of this section.
Section 71.22 General License: Fissile Material
    Existing Sec. 71.18 has been redesignated as Sec. 71.22. The 
current Sec. 71.22 has been removed. This section has been amended by 
consolidating and simplifying the current fissile general license 
provisions contained in existing Sec.Sec. 71.18, 71.20, 71.22, and 
71.24 into a new Sec. 71.22. The new Sec. 71.22, while retaining some 
of the provisions of the existing general licenses, principally uses 
mass-based limits and a Criticality Safety Index (CSI). Concentration-
based limits have been removed. Exceptions relating to plutonium-
beryllium sealed sources in existing Sec.Sec. 71.18 and 71.22 have been 
relocated to new Sec. 71.23. The values contained in new Tables 71-1 
and 71-2 have been revised from the values contained in the table in 
existing Sec. 71.22 and in Table 1 in existing Sec. 71.20, 
respectively; and are based on new minimum critical mass calculations 
described in NUREG/CR-5342. In some instances, the allowable mass limit 
has been increased from the current limits in existing Sec.Sec. 71.18, 
71.20, 71.22, and 71.24; in other instances, the allowable mass limit 
has been reduced. The values contained in new Tables 71-1 and 71-2 are 
used as the variables X, Y, and Z in the equation in paragraph (e).
    The title has been revised to indicate that this general license is 
not restricted to a specific type of fissile material shipment.
    Paragraph (a) has been revised to require that fissile material 
shipped under this general license be contained in a DOT Type A 
package. Additionally, while the existing exception from subparts E and 
F requirements has been maintained, the DOT Type A package regulations 
of 49 CFR part 173 has also been specified.
    Paragraph (b) remains unchanged.
    Paragraph (c) has been revised to remove the specific gram limits 
for uranium and plutonium but retains the existing Type A quantity 
limit. Revised gram limits have been relocated to new Table 71-1, which 
is associated with new paragraphs (d) and (e). A requirement has also 
been added to limit the amount of special moderating materials 
beryllium, graphite, and hydrogenous material enriched in deuterium 
present in a package to less than 500 g.
    Existing paragraph (d) has been removed. Revised gram limits for 
fissile material mixed with material having a hydrogen density greater 
than water (i.e., a moderating effectiveness greater than 
H2O) have been placed in new Table 71-1. A note has been 
added to new Table 71-1 to indicate that reduced mass limits apply when 
more than 15 percent of a mixture of moderating materials contains 
moderating material with a hydrogen density greater than 
H2O.
    New paragraph (d) has been added to require that shipments of 
packages containing fissile material be labeled with a CSI, that the 
CSI per package be less than or equal to 10.0, and that the sum of the 
CSIs in a shipment of multiple fissile material packages be limited to 
less than or equal to 50.0 for a nonexclusive use conveyance, and to 
less than or equal to 100.0 for an exclusive use conveyance.
    Existing Paragraphs (e) and (f) have been removed.
    New paragraph (e) has been added to require that the CSI be 
calculated via a new equation for any of the fissile nuclides. Guidance 
on applying the equation and the mass limit input values of Tables 71-1 
and 71-2 is also contained in this paragraph.
Section 71.23 General License: Plutonium-Beryllium Special Form 
Material
    The existing Sec. 71.20, ``General license: Fissile material, 
limited moderator per package,'' has been removed. A new section on the 
shipment of plutonium-beryllium (Pu-Be) special-form fissile material 
(i.e., sealed sources) has been added as a new Sec. 71.23. New Sec. 
71.23 consolidates regulations on shipment of Pu-Be sealed sources 
contained in existing Sec.Sec. 71.18 and 71.22 into one location in 
part 71. The new Sec. 71.23 reduces the maximum quantity of fissile 
plutonium Pu-Be sealed sources that could be shipped on a single 
conveyance through changes in the mass limits and calculation of the 
CSI. Currently, a Pu-Be sealed source package can contain up to 400 g 
of fissile plutonium with a CSI equal to 10.0. Consequently, the 
current conveyance limits are 4,000 g per shipment for an exclusive-use 
vehicle and 2000 g per shipment for a nonexclusive use vehicle. The new 
Sec. 71.23 increases the maximum CSI per package from 10 to 100; 
however, the maximum quantity of plutonium per conveyance (i.e., 
shipment) would be reduced to 1000 g. The 1000-g per shipment limit and 
240 g of fissile plutonium limit are equivalent to those in new Sec. 
71.22(f) (1000 g per shipment and 200 g of fissile plutonium). The 240 
g versus 200 g of fissile plutonium per package is due to the increased 
confidence that the fissile plutonium, within a sealed source capsule, 
would not escape from the capsule during an accident and reconfigure 
itself into an unfavorable geometry.
    New Sec. 71.23 has been titled: ``General license: Plutonium-
beryllium special form material.'' Paragraph (a) describes the 
applicability of this section, exceptions to the requirements of 
subparts E and F, and the requirement to ship Pu-Be sealed sources in 
DOT Type A packages.
    Paragraph (b) requires that shipments of Pu-Be sealed sources be 
made under an NRC-approved QA program.
    Paragraph (c) requires a 1000 g per package limit. In addition, 
plutonium-239 and plutonium-241 constitute only 240 g of the 1000 g 
limit.
    Paragraph (d) requires that a CSI be calculated per paragraph (e), 
and the CSI must be less than or equal to 100.0. For shipments of 
multiple packages, the sum of the CSIs is limited to less than or equal 
to 50.0 for a nonexclusive use conveyance and to less than or equal to 
100.0 for an exclusive use conveyance.
    Paragraph (e) provides an equation to calculate the CSI for Pu-Be 
sources. This equation is based upon the 240-g mass limit for fissile 
nuclide plutonium-239 and plutonium-241 in paragraph (c).
Section 71.24 (Reserved)
Section 71.25 (Reserved)
    Existing Sec.Sec. 71.22 and 71.24 have been redesignated as 
Sec.Sec. 71.24 and 71.25. New Sec.Sec. 71.24 and 71.25 have been 
removed and reserved.

Subpart D--Application for Package Approval

Section 71.41 Demonstration of Compliance
    Paragraph (a) has been revised to require that a Type B package 
which contains radioactive contents with

[[Page 3767]]

activity greater than 105A2 of any radionuclide 
must meet the enhanced deep immersion test found in Sec. 71.61. A new 
paragraph (d) has been added to provide special package authorizations.
Section 71.51 Additional Requirements for Type B Packages
    Paragraph (a) has been revised to remove the reference to Sec. 
71.52, because the requirements of Sec. 71.52 have expired. Paragraph 
(d) has been added to require that a package which contains radioactive 
contents with activity greater than 105A2 of any 
radionuclide must also meet the enhanced deep immersion test found in 
Sec. 71.61.
Section 71.53 Fissile Material Exemptions (Reserved)
    This section has been removed and reserved; its contents have been 
moved to Sec. 71.15.
Section 71.55 General Requirements for Fissile Material Packages
    New paragraphs (f) and (g) have been added. Paragraph (f) specifies 
design and testing for fissile material package designs for transport 
by aircraft, and paragraph (g) addresses UF6 criticality 
exception from Sec. 71.55(b). Additionally, as a conforming change, 
paragraph (b) has been updated to support new paragraph (g).
Section 71.59 Standards for Arrays of Fissile Material Packages
    Paragraphs (b) and (c) have been revised to use the term CSI 
(criticality safety index).
    Paragraph (b) has been revised to refer to a CSI rather than a TI 
for nuclear criticality control. The method for calculating a CSI is 
the same as the existing method for a TI for nuclear criticality 
control.
    Paragraph (c) has been revised to provide direction to licensees 
when the CSI is exactly equal to 50 and to use plain language. 
Subparagraph (1) has been revised by replacing the term ``(n)ot in 
excess of 10,'' with the term ``(l)ess than or equal to 50.'' New 
paragraph (c)(2) has been added to provide for shipment of packages 
with a CSI of less than 50 on an exclusive use conveyance. The current 
conveyance limit of 100 has been retained. Existing paragraph (c)(2) 
has been redesignated as new paragraph (c)(3) and has been revised by 
replacing the term ``(i)n excess of 10,'' with the term ``(g)reater 
than 50.'' These three changes: (1) Provide greater clarity and 
mathematical consistency among paragraphs (c)(1), (c)(2), and (c)(3); 
(2) clarify the CSI limits for storage incident to transport; and (3) 
increase the CSI limit per package from 10 to 50 for shipments made 
with nonexclusive use conveyances.
Section 71.61 Special Requirements for Type B Packages Containing More 
Than 105A2
    This section has been revised to require an enhanced water 
immersion test for packages used for radioactive contents with activity 
greater than 105A2. The title of this section has 
also been revised to reflect that the scope has been broadened beyond 
irradiated nuclear fuel.
Section 71.63 Special Requirement for Plutonium Shipments
    The title has been revised to reflect only a single ``requirement'' 
rather than multiple requirements.
    Paragraph (b) has been removed.
    The designation of the remaining text as paragraph (a) has been 
removed, because only one paragraph remains. The text of former 
paragraph (a) has been revised to use plain language. The 0.74-TBq (20-
Ci) limit and solid form requirement have been retained.
Section 71.73 Hypothetical Accident Conditions
    A new paragraph (c)(2) has been added to require a crush test for 
fissile material packages.
Section 71.88 Air Transport of Plutonium
    Paragraph (a)(2) has been revised to remove the 70-Bq/g (0.002-
[mgr]Ci/g) specific activity value and substitute activity 
concentration values for plutonium found in Appendix A, Table A-2, of 
this part. This revision is a conforming change to the revision to new 
Sec. 71.14 to ensure consistent treatment of plutonium between these 
two sections.

Subpart G--Operating Controls and Procedures

Section 71.91 Records
    As a conforming change to subpart H, paragraphs (b) and (c) have 
been redesignated as paragraphs (c) and (d), respectively, and are 
revised by adding the terms ``certificate holder'' and ``applicant for 
a CoC.'' New paragraph (b) has been added to require a certificate 
holder to keep records on the model, serial number, and date of 
manufacture of a packaging. These requirements are similar to the 
requirements in paragraph (a), though less information is required. No 
change has been made to paragraph (a).
Section 71.93 Inspection and Tests
    As a conforming change to subpart H, paragraphs (a) and (b) have 
been revised by adding the terms ``certificate holder'' and ``applicant 
for a CoC.'' Paragraph (c) has been revised to require the certificate 
holder to notify the NRC before it begins fabrication of a packaging 
that can contain material having a decay heat load in excess of 5 kW or 
a maximum normal operating pressure of 103 kPa (kilo Pascals) (15 lbf/
in2) gauge. This notification could be for either 
fabricating a single packaging or the beginning of a campaign for 
fabricating multiple packagings. This notification is in accordance 
with the requirements of Sec. 71.1, rather than an NRC Regional 
Administrator. This change in notification location reduces confusion 
in identifying the appropriate Regional Administrator when the 
certificate holder and fabrication location are overseas. Licensees 
have been removed from this paragraph because the NRC believes that 
requiring a licensee, who does not own the packaging, to notify the NRC 
in advance of a packaging fabrication, when the licensee may not use 
the packaging for years, is inappropriate and an unreasonable burden. 
The NRC believes that requiring certificate holders and applicants for 
a CoC to notify the NRC in advance of fabricating a packaging(s) would 
allow the NRC adequate opportunity to inspect these activities. This 
change is similar to the current requirement in Sec. 72.232(d) for part 
72 certificate holders or applicants for a CoC to notify the NRC 45 
days before starting the fabrication of the first storage cask under a 
part 72 CoC. This action improves the harmonization between these two 
regulations in parts 71 and 72.
Section 71.95 Reports
    The existing introductory text and paragraphs (a), (b), and (c) 
have been combined into a new paragraph (a) which requires a licensee, 
after requesting the certificate holder's input, to submit a written 
report to the NRC in certain circumstances. The requirement for the 
licensee to request input from the certificate holder during 
development of the written event report will ensure that design 
deficiency issues have been thoroughly considered. The licensee will 
also be required to provide the certificate holder with a copy of the 
written event report, after the report is submitted to the NRC. This 
will permit the certificate holder to monitor and trend the package 
performance information, arising from package use by multiple 
licensees. Additionally,

[[Page 3768]]

requirements on timing and submission location for the written reports 
have been relocated to new paragraph (c). Furthermore, the 30-day 
reporting requirement has been lengthened to a 60-day reporting 
requirement.
    The existing paragraph (c) has been redesignated as paragraph (b) 
and revised for clarity.
    New paragraphs (c) and (d) have been added to provide requirements 
on the timing, submission location, form, and content of the written 
reports.
Section 71.100 Criminal Penalties
    Section 223 of the Atomic Energy Act of 1954, as amended, (the Act) 
provides for criminal sanctions for willful violation of, attempted 
violation of, or conspiracy to violate, any regulation issued under 
sections 161b, 161i, or 161o of the Act. The Commission stated in a 
final rule on ``Clarification of Statutory Authority for Purposes of 
Criminal Enforcement'' (57 FR 55082; November, 24, 1992), that 
substantive rules under sections 161b, 161i, or 161o of the Act include 
those rules that create ``duties, obligations, conditions, 
restrictions, limitations, and prohibitions.'' For the NRC to consider 
the possibility of criminal sanctions for willful violation of, 
attempted violation of, or conspiracy to violate, any substantive 
regulations, the NRC must have clearly identified to affected parties 
which regulations in part 71 are substantive rules. Accordingly, 
paragraph (b) of this section identifies those part 71 regulations that 
the NRC does not consider as substantive regulations. Thus, willful 
violation of, attempted violation of, or conspiracy to violate any of 
the regulations listed in paragraph (b) is not subject to possible 
criminal sanctions.
    Paragraph (b) of this section has been revised as a conforming 
change. The NRC has reviewed new Sec.Sec. 71.10 and considers that this 
regulation is not a substantive rule. Therefore, new Sec.Sec. 71.10 has 
been added to the list of sections in paragraph (b). The NRC reviewed 
new Sec.Sec. 71.9, 71.18, and 71.23 and considers that these 
regulations are substantive rules. Therefore, these sections have not 
been added to paragraph (b). Additionally, the NRC has reviewed the 
existing Sec.Sec. 71.9, 71.10, and 71.53 and concluded these sections 
should be recharacterized as substantive rules. Therefore, new Sec.Sec. 
71.13, 71.14, and 71.18 have not been included in paragraph (b). 
Additionally, existing Sec.Sec. 71.52 and 71.53 have been removed from 
paragraph (b), because these section numbers have been removed from 
part 71.

Subpart H--Quality Assurance

Section 71.101 Quality Assurance Requirements
    Paragraph (a) has been revised by adding two new sentences to the 
end of the paragraph specifying responsibilities for certificate 
holders and applicants for a CoC.
    Paragraph (b) has been revised to add the terms ``certificate 
holder'' and ``applicant for a CoC.'' The second sentence has been 
revised to provide greater clarity and consistency within subpart H by 
referring to ``the QA requirement's importance to safety.''
    Paragraph (c) has been revised by redesignating the existing text 
as paragraph (c)(1), and new text has been added on submitting QA 
programs in accordance with the requirements of Sec. 71.1. New 
paragraph (c)(2) has been added to provide equivalent requirements on 
the submission of QA programs for certificate holders and applicants 
for a CoC.
    Paragraph (f) has been revised to allow the use of existing NRC-
approved part 71 and part 72 QA programs, in lieu of submitting a new 
QA program. Additionally, the terms ``certificate holder'' and 
``applicant for a CoC'' have been added.
    Paragraph (g) has been revised by making a minor change to clarify 
that Sec. 34.31(b) is located in chapter I of title 10 of the Code of 
Federal Regulations. Additionally, as a conforming change, Sec. 
71.12(b) has been redesignated as Sec. 71.17(b).
Section 71.103 Quality Assurance Organization
    Paragraph (a) has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.105 Quality Assurance Program
    Paragraphs (a) through (d) have been revised by adding the terms 
``certificate holder'' and ``applicant for a CoC.''
Section 71.107 Package Design Control
    Paragraph (a) has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.'' Further, the last sentence has 
been revised to improve clarity and consistency within subpart H by 
referring to ``processes that are essential to the functions of the 
materials, parts, and components that are important to safety.''
    Paragraph (b) has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.'' Additionally, the last sentence 
of paragraph (c) has been revised by replacing the text ``(c)hanges in 
the conditions specified in the package approval require NRC approval * 
* *.'' with ``(c)hanges in the conditions specified in the CoC require 
NRC prior approval * * *.''
Section 71.109 Procurement Document Control
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.111 Instructions, Procedures, and Drawings
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.113 Document Control
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.115 Control of Purchased Material, Equipment, and Services
    Paragraphs (a) through (c) have been revised by adding the terms 
``certificate holder'' and ``applicant for a CoC.''
Section 71.117 Identification and Control of Materials, Parts, and 
Components
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.119 Control of Special Processes
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.121 Internal Inspection
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.123 Test Control
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.125 Control of Measuring and Test Equipment
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.127 Handling, Storage, and Shipping Control
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.129 Inspection, Test, and Operating Status
    Paragraph (a) has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''

[[Page 3769]]

Section 71.131 Nonconforming Materials, Parts, or Components
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.133 Corrective Action
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.135 Quality Assurance Records
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
Section 71.137 Audits
    This section has been revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''

Appendix A to Part 71--Determination of A1 and A2

    No changes have been made in paragraphs I, III, and V; however, 
these paragraphs have been included due to revising Appendix A, in its 
entirety.
    Paragraph II has been revised to use plain language and has been 
redesignated as subparagraph II(a). The intent of existing paragraph II 
has not been changed; however, the reference to existing Table A-2 has 
been revised as a conforming change to the new Table A-3. New paragraph 
II(b) has been added to provide direction on determining exempt 
material activity concentration and exempt consignment activity values 
when a radionuclide has been identified as a constituent of a proposed 
shipment, but the individual radionuclide is not listed in Table A-2. 
Consequently, the structure of paragraphs II(a) and II(b) is the same. 
New paragraph II(c) has been added to provide direction to licensees on 
how to submit requests for Commission prior approval of either 
A1 and A2 values or exempt material activity 
concentration and exempt consignment activity values, for radionuclides 
that are not listed in Tables A-1 and A-2, respectively.
    Paragraph IV has been revised by adding new paragraphs (e) and (f) 
to provide equations to use in determining a consolidated exempt 
material activity concentration and exempt consignment activity value 
when a shipment contains multiple radionuclides. The existing text 
describing an alternative method for calculating the A1 or 
A2 value of a mixture has been redesignated as paragraphs 
(c) and (d). No changes have been made from the existing equations.

Appendix A, Table A-1--A1 and A2 Values for 
Radionuclides

    This Table has been revised to reflect the values from TS-R-1.

Appendix A, Table A-2--Exempt Material Activity Concentrations and 
Exempt Consignment Activity Limits for Radionuclides

    A new Table A-2 has been added to Appendix A of part 71. This table 
contains the values of Exempt Material Activity Concentrations and 
Exempt Consignment Activity Limits for selected radionuclides. Table A-
2 is referenced in new Sec. 71.14(a)(2) and is used in Sec. 71.14 to 
determine when concentrations of material are not considered 
radioactive material, for the purposes of transportation.

Appendix A, Table A-3--General Values for A1 and 
A2

    The existing Table A-2 has been redesignated as new Table A-3, and 
the values have been revised to reflect the changes from TS-R-1.

Appendix A, Table A-4--Activity Mass Relationships for Uranium

    The existing Table A-3 has been redesignated as new Table A-4. No 
changes have been made to the values contained in new Table A-4.

V. Criminal Penalties

    For the purposes of section 223 of the Atomic Energy Act (AEA), the 
Commission is amending 10 CFR part 71 under one or more of sections 
161b, 161i, or 161o of the AEA. Willful violations of the rule will be 
subject to criminal enforcement.
    The following is a list of substantive rule sections being revised 
or added in this rulemaking: Sec.Sec. 71.1, 71.3, 71.5, 71.8, 71.9, 
71.12, 71.13, 71.14, 71.15, 71.17, 71.19, 71.20, 71.21, 71.22, 71.23, 
71.61, 71.63, 71.88, 71.91, 71.93, 71.95, 71.101, 71.103, 71.105, 
71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 
71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137.

VI. Issues of Compatibility for Agreement States

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' which became effective on September 3, 1997 
(62 FR 46517), NRC program elements (including regulations) are placed 
into four compatibility categories. In addition, NRC program elements 
also are identified as having particular health and safety significance 
or as being reserved solely to the NRC. Compatibility Category A are 
those program elements that are basic radiation protection standards 
and scientific terms and definitions that are necessary to understand 
radiation protection concepts. An Agreement State should adopt Category 
A program elements in an essentially identical manner to provide 
uniformity in the regulation of agreement material on a nationwide 
basis. Compatibility Category B are those program elements that apply 
to activities that have direct and significant effects in multiple 
jurisdictions. An Agreement State should adopt Category B program 
elements in an essentially identical manner. Compatibility Category C 
are those program elements that do not meet the criteria of Category A 
or B, but the essential objectives of which an Agreement State should 
adopt to avoid conflict, duplication, gaps, or other conditions that 
would jeopardize an orderly pattern in the regulation of agreement 
material on a nationwide basis. An Agreement State should adopt the 
essential objectives of the Category C program elements. Compatibility 
Category D are those program elements that do not meet any of the 
criteria of Category A, B, or C, and thus do not need to be adopted by 
Agreement States for purposes of compatibility. A bracket around a 
category means that the section may have been adopted elsewhere, and it 
is not necessary to adopt it again. Health and Safety (H&S) are program 
elements that are not required for compatibility (i.e., Category D) but 
are identified as having a particular health and safety role (i.e., 
adequacy) in the regulation of agreement material within the State. 
Although not required for compatibility, the State should adopt program 
elements in this category based on those of NRC that embody the 
essential objectives of the NRC program elements because of particular 
health and safety considerations. Compatibility Category NRC are those 
program elements that address areas of regulation that cannot be 
relinquished to Agreement States pursuant to the Atomic Energy Act, as 
amended, or provisions of title 10 of the Code of Federal Regulations. 
These program elements should not be adopted by Agreement States. The 
following table lists the part 71 revisions and their corresponding 
categorization under the ``Policy Statement on Adequacy and 
Compatibility of Agreement State Programs.'' This table has been 
revised to incorporate comments received from the States of California 
and Wisconsin during the 30-day Agreement States comment period which 
began on June 3, 2003.

[[Page 3770]]



      Part 71--Packaging and Transportation of Radioactive Material
------------------------------------------------------------------------
  Regulation                        Compatibility
    section       Section title       category            Comments
------------------------------------------------------------------------
Sec. 71.0.....  Purpose and       D, except         This requirement is
                 Scope.            paragraph C is    designated
                                   [B].              Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this requirement in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.1.....  Communications    D
                 and Records.
Sec. 71.2.....  Interpretations.  D
Sec. 71.3.....  Requirements for  [B].............  This requirement is
                 license.                            designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions since
                                                     it assures
                                                     authorization for
                                                     the transport of
                                                     licensed material.
                                                     An Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this requirement in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.4.....  Definitions:
                A1..............  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                A2..............  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in mulitiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Carrier.........  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.

[[Page 3771]]

 
                Certificate       D--for those      This term is used in
                 holder.           States which      the sections
                                   have no           concerning quality
                                   licensees that    assurance programs
                                   us Type B         for Type B
                                   packages. or      packages. Those
                                                     States which have
                                                     no licensees that
                                                     use Type B packages
                                                     are not required to
                                                     adopt this
                                                     definition. This
                                                     definition is
                                                     designated
                                                     Compatibility
                                                     Category B for
                                                     those States which
                                                     have licensees that
                                                     us Type B packages
                                                     because it applies
                                                     to activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                                  [B]--for those
                                   States which
                                   have licensees
                                   that use Type B
                                   packages.
                Certificate of    D--for those      This term is used in
                 compliance.       States which      the sections
                                   have no           concerning quality
                                   licensees that    assurance programs
                                   use Type B        for Type B
                                   packages.         packages. Those
                                  [B]--for those     States which have
                                   States which      no licensees that
                                   have licensees    use Type B packages
                                   that use Type B   are not required to
                                   packages.         adopt this
                                                     definition. This
                                                     definition is
                                                     designated
                                                     Compatibility
                                                     Category B for
                                                     those States which
                                                     have licensees that
                                                     use Type B packages
                                                     because it applies
                                                     to activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Close reflection  D...............  This definition is
                 by water.                           not required for
                                                     compatibility since
                                                     it defines a term
                                                     which pertains to
                                                     an area reserved to
                                                     NRC. A State may
                                                     adopt this
                                                     definition for
                                                     purposes of clarity
                                                     or communication.
                                                     This definition can
                                                     be adopted by
                                                     Agreement States
                                                     since it in and of
                                                     itself does not
                                                     convey any
                                                     authority whereby a
                                                     State can regulate
                                                     in an exclusive NRC
                                                     jurisdiction.
                                                     However, if a State
                                                     chooses to define
                                                     the term then the
                                                     definition should
                                                     be essentially
                                                     identical.
                Consignment.....  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Containment       D...............  This term is not
                 System.                             used in any section
                                                     requiring Agreement
                                                     State adoption.
                Conveyance......  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.

[[Page 3772]]

 
                Criticality       B...............  This definition is
                 safety Index.                       designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     In addition, this
                                                     definition is
                                                     needed for a common
                                                     understanding
                                                     beyond a plain
                                                     dictionary meaning
                                                     of the term in
                                                     order to implement
                                                     10 CFR 71.22, 71.23
                                                     and 71.59.
                Deuterium.......  B...............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     In addition, this
                                                     definition is
                                                     needed for a common
                                                     understanding
                                                     beyond a plain
                                                     dictionary meaning
                                                     of the term in
                                                     order to implement
                                                     Sec. 71.15.
                DOT.............  D...............  This term does not
                                                     meet any of the
                                                     criteria of
                                                     Category A, B, C,
                                                     or H&S because it
                                                     is a widely
                                                     accepted
                                                     abbreviation for
                                                     the U. S.
                                                     Department of
                                                     Transportation.
                Exclusive use...  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Fissile material  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Graphite........  B...............  This definition is
                                                     needed for a common
                                                     understanding
                                                     beyond a plain
                                                     dictionary meaning
                                                     of the term in
                                                     order to implement
                                                     Sec. 71.15, which
                                                     has direct and
                                                     significant
                                                     transboundary
                                                     effects.
                Licensed          [D].............  This term does not
                 material.                           meet any of the
                                                     criteria of
                                                     Category A, B, C,
                                                     or H&S because it
                                                     is widely accepted
                                                     and understood.
                                                     This definition
                                                     also appears in 10
                                                     CFR 20.1003. For
                                                     purposes of
                                                     compatibility, the
                                                     language of the
                                                     Part 20 definition
                                                     should be used and
                                                     is assigned to
                                                     Compatibility
                                                     Category D.
                Low Specific      [B].............  This definition is
                 Activity (LSA)                      designated
                 material.                           Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.

[[Page 3773]]

 
                Low toxicity      [B].............  This definition is
                 alpha emitters.                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Maximum normal    D...............  The definition of
                 operating                           the term ``maximum
                 pressure.                           normal operating
                                                     pressure'' was
                                                     changed from a
                                                     compatibility
                                                     category ``B'' to a
                                                     category ``D.''
                                                     This term is not
                                                     used in any section
                                                     requiring Agreement
                                                     State adoption; it
                                                     relates to the heat
                                                     conditions in Sec.
                                                     71.71(c)(1), which
                                                     is designated a
                                                     category ``NRC.''
                                                     This definition is
                                                     not required for
                                                     compatibility since
                                                     it defines a term
                                                     which pertains to
                                                     an area reserved to
                                                     the NRC. A State
                                                     may adopt this
                                                     definition for
                                                     purposes of clarity
                                                     or communication.
                                                     This definition can
                                                     be adopted by
                                                     Agreement States
                                                     since it is and of
                                                     itself does not
                                                     convey any
                                                     authority whereby a
                                                     State can regulate
                                                     in an exclusive NRC
                                                     jurisdiction.
                                                     However, if a State
                                                     chooses to define
                                                     this term, then the
                                                     definition should
                                                     be essentially
                                                     identical.
                Natural thorium.  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Normal form       [B].............  This definition is
                 radioactive                         designated
                 material.                           Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Optimum           D...............  This definition is
                 interspersed                        not required for
                 hydrogenous                         compatibility since
                 moderation.                         it defines a term
                                                     which pertains to
                                                     an area reserved to
                                                     NRC. A State may
                                                     adopt this
                                                     definition for
                                                     purposes of clarity
                                                     or communication.
                                                     This definition can
                                                     be adopted by
                                                     Agreement States
                                                     since it in and of
                                                     itself does not
                                                     convey any
                                                     authority whereby a
                                                     State can regulate
                                                     in an exclusive NRC
                                                     jurisdiction.
                                                     However, if a State
                                                     chooses to define
                                                     the term, then the
                                                     definition should
                                                     be essentially
                                                     identical.
                Package.........  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.

[[Page 3774]]

 
                Fissile material  [B].............  This definition is
                 package or Type                     designated
                 AF package,                         Compatibility
                 Type BF, Type                       Category B because
                 B(U)F package,                      it applies to
                 or Type B(M)F.                      activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Type A package..  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Type B package..  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Packaging.......  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Special form      [B].............  This definition is
                 radioactive                         designated
                 material.                           Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Specific          [B].............  This definition is
                 activity.                           designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Spent Nuclear     D...............  This definition is
                 Fuel or Spent                       not required
                 Fuel.                               compatibility since
                                                     it defines a term
                                                     which pertains to
                                                     an area reserved to
                                                     NRC. A State may
                                                     adopt this
                                                     definition for
                                                     purposes of clarity
                                                     or communication.
                                                     This definition can
                                                     be adopted by
                                                     Agreement States
                                                     since it in and of
                                                     itself does not
                                                     convey any
                                                     authority whereby a
                                                     State can regulate
                                                     in an exclusive NRC
                                                     jurisdiction.
                                                     However, if a State
                                                     chooses to define
                                                     the term, then the
                                                     definition should
                                                     be essentially
                                                     identical.
                State...........  D.                ....................

[[Page 3775]]

 
                Surface           [B].............  This definition is
                 Contaminated                        designated
                 Object (SCO).                       Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Transport Index.  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Type A quantity.  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Type B quantity.  [B].............  This definition is
                                                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Unirradiated      [B].............  This definition is
                 uranium.                            designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
                Uranium--         [B].............  This definition is
                 natural,                            designated
                 depleted and                        Compatibility
                 enriched.                           Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this definition in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this definition
                                                     is not necessary.
Sec. 71.5.....  Transportation    [B].............  This requirement is
                 of Licensed                         designated
                 Material.                           Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
Sec. 71.6.....  Information       D...............
                 collection
                 requirements:
                 OMB approval.

[[Page 3776]]

 
Sec. 71.7.....  Completeness and  D...............
                 accuracy of
                 Information.
Sec. 71.8.....  Deliberate        C...............  The Commission
                 misconduct.                         determined in
                                                     response to SECY-97-
                                                     156 that Agreement
                                                     States should adopt
                                                     the essential
                                                     objectives of this
                                                     provision. The
                                                     essential
                                                     objectives of this
                                                     provision are
                                                     provided in
                                                     paragraphs (a),
                                                     (b), (c), and (d).
                                                     If deliberate
                                                     misconduct and
                                                     wrongdoing issues
                                                     involving Agreement
                                                     State licensees
                                                     were not pursued
                                                     and closed by
                                                     Agreement States,
                                                     then a potential
                                                     gap may be created
                                                     between NRC and
                                                     Agreement State
                                                     programs.
Sec. 71.9.....  Employee          D...............  This provision does
                 Protection.                         not meet any of the
                                                     criteria for
                                                     designations
                                                     Category A, B, C,
                                                     or health and
                                                     safety. Thus, it
                                                     does not need to be
                                                     adopted by
                                                     Agreement States.
Sec. 71.10....  Public            D...............  This provision does
                 Inspection of                       not meet any of the
                 Application.                        criteria for
                                                     designations
                                                     Category A, B, C,
                                                     or health and
                                                     safety. Thus, it
                                                     does not need to be
                                                     adopted by
                                                     Agreement States.
Sec. 71.11....  [RESERVED]......
Sec. 71.12....  Specific          D...............
                 exemptions.
Sec. 71.13....  Exemption for     [B].............  This provision is
                 physicians.                         designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
Sec. 71.14....  Exemptions for    [B]-paragraph     Paragraph (a) is
                 low level         (a).              designated as a
                 material.        NRC--paragraph     Compatibility
                                   (b).              Category B because
                                                     of its significant
                                                     transboundary
                                                     impacts with
                                                     respect to the
                                                     establishment of
                                                     exempt materials in
                                                     the area of
                                                     transportation. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this requirement in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
                                                    Paragraph (b) is
                                                     designated
                                                     Compatibility
                                                     Category ``NRC.''
                                                     This provision is
                                                     reserved to the NRC
                                                     because it
                                                     delineates NRC's
                                                     authority from that
                                                     of DOT's in the
                                                     area of
                                                     transportation of
                                                     radioactive
                                                     materials. These
                                                     provisions
                                                     relinquish to DOT
                                                     the control of
                                                     types of shipment
                                                     that are of low
                                                     risk both from
                                                     radiation and
                                                     criticality
                                                     standpoints.
                                                     Further, to ensure
                                                     that only low
                                                     criticality risk
                                                     shipments are
                                                     included in the
                                                     area of DOT
                                                     authority, these
                                                     provisions restrict
                                                     the exemption to
                                                     Type A and low-
                                                     specific-activity
                                                     (LSA) or surface
                                                     contaminated
                                                     objects (SCOs) that
                                                     either contain no
                                                     fissile material or
                                                     satisfy the fissile
                                                     material exemption
                                                     requirements in
                                                     Sec. 71.11.
                                                     Finally, this
                                                     provision is
                                                     reserved to the NRC
                                                     because this
                                                     exemption does not
                                                     relieve licensees
                                                     from DOT
                                                     requirements by
                                                     reason of NRC's
                                                     authority. Thus,
                                                     Agreement States
                                                     should not adopt
                                                     this provision in
                                                     order to retain
                                                     their ability to
                                                     implement all of 49
                                                     CFR as directed by
                                                     DOT.

[[Page 3777]]

 
Sec. 71.15....  Exemptions from   [B].............  This provision is
                 classification                      designated
                 as fissile                          Compatibility
                 material.                           Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
                                                     Note: This
                                                     provision was
                                                     previously
                                                     designated ``NRC.''
                                                     It was changed to
                                                     ``B'' to ensure
                                                     compatibility
                                                     between NRC and
                                                     Agreement States in
                                                     an area that has
                                                     significant and
                                                     direct
                                                     transboundary
                                                     implications.
                                                     During further
                                                     staff review, it
                                                     was noted that the
                                                     requirements in
                                                     this section
                                                     ``Fissile material
                                                     exemptions'' is the
                                                     same as those of
                                                     DOT in 49 CFR
                                                     173.453, ``Fissile
                                                     materials
                                                     exceptions.'' Staff
                                                     noted that States
                                                     adopt these DOT
                                                     regulations as a
                                                     part of their
                                                     transportation
                                                     regulations. Staff
                                                     also noted that in
                                                     accordance with
                                                     Sec. 150.11, an
                                                     Agreement State can
                                                     regulate the
                                                     following fissile
                                                     materials: U-235 in
                                                     quantities not
                                                     exceeding 350
                                                     grams, U-233 in
                                                     quantities not
                                                     exceeding 200
                                                     grams; plutonium in
                                                     quantities not
                                                     exceeding 200
                                                     grams, or any
                                                     combination of
                                                     these materials
                                                     that would be
                                                     sufficient to form
                                                     a critical mass.
                                                     These requirements
                                                     would apply to the
                                                     materials Agreement
                                                     States regulate.
                                                     Thus, the
                                                     compatibility of
                                                     this requirement
                                                     was changed to a
                                                     ``[B],'' which
                                                     indicates that if a
                                                     State has adopted
                                                     this provision as a
                                                     part of the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.16....  [RESERVED]......
Sec. 71.17....  General license:  [B].............  This provision is
                 NRC--approved                       designated
                 package.                            Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.19....  Previously        NRC.............  This provision is
                 approved                            reserved to the NRC
                 package.                            because it
                                                     addresses packages
                                                     intended for both
                                                     the storage and
                                                     transportation of
                                                     spent fuel.
Sec. 71.20....  General license:  [B].............  This provision is
                 DOT                                 designated
                 specification                       Compatibility
                 container                           Category B because
                 material.                           it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.21....  General license:  [B].............  This provision is
                 Use of foreign                      designated
                 approved                            Compatibility
                 package.                            Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.

[[Page 3778]]

 
Sec. 71.22....  General license:  [B].............  This provision
                 Fissile                             designated
                 material.                           Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
                                                    Note: A similar
                                                     provision was
                                                     previously
                                                     designated ``NRC.''
                                                     It was changed to
                                                     ``B'' to ensure
                                                     compatibility
                                                     between NRC and
                                                     Agreement States in
                                                     an area that has
                                                     significant and
                                                     direct
                                                     transboundary
                                                     implications.
                                                     During further
                                                     staff review, it
                                                     was noted that in
                                                     accordance with 10
                                                     CFR 150.11, an
                                                     Agreement State can
                                                     regulate the
                                                     following fissile
                                                     materials: U-235 in
                                                     quantities not
                                                     exceeding 350
                                                     grams, U-233 in
                                                     quantities not
                                                     exceeding 200
                                                     grams; plutonium in
                                                     quantities not
                                                     exceeding 200
                                                     grams, or any
                                                     combination of
                                                     these materials
                                                     that would be
                                                     sufficient to form
                                                     a critical mass.
                                                     These requirements
                                                     would apply to the
                                                     materials Agreement
                                                     States regulate.
                                                     Thus, the
                                                     compatibility of
                                                     this requirement
                                                     was changed to a
                                                     ``[B],'' which
                                                     indicates that if a
                                                     State has adopted
                                                     this provision as a
                                                     part of the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.23....  General license:  [B].............  This provision is
                 Plutonium-                          designated
                 beryllium                           Compatibility
                 special form                        Category B because
                 material.                           it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
Sec. 71.24....  [RESERVED]......
Sec. 71.25....  [RESERVED]......
Sec. 71.31....  Contents of       NRC.............
                 Application.
Sec. 71.33....  Package           NRC.............
                 description.
Sec. 71.35....  Package           NRC.............
                 evaluation.
Sec. 71.37....  Quality           NRC.............
                 Assurance.
Sec. 71.38....  Renewal of a      NRC.............
                 certificate of
                 compliance or
                 quality
                 assurance
                 program
                 approval.
Sec. 71.39....  Requirements for  NRC.............
                 additional
                 information.
Sec. 71.41....  Demonstration of  NRC.............  This provision is
                 Compliance.                         designated NRC
                                                     because it
                                                     addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.43....  General           NRC.............
                 Standards for
                 all packages.
Sec. 71.45....  Lifting and tie-  NRC.............
                 down Standards
                 for all
                 packages.
Sec. 71.47....  External          [B].............  This requirement was
                 radiation                           changed from a
                 Standards for                       compatibility
                 all packages.                       category ``NRC'' to
                                                     ``[B].'' This
                                                     provision was
                                                     changed because it
                                                     establishes the
                                                     external radiation
                                                     standards for all
                                                     transportation
                                                     packages. It is
                                                     essential that the
                                                     Agreement States
                                                     adopt this
                                                     provision in an
                                                     essentially
                                                     identical manner
                                                     because they have
                                                     direct and
                                                     significant
                                                     transboundary
                                                     effects. The
                                                     bracket,``B,''
                                                     indicates that a
                                                     State should adopt
                                                     this provision in
                                                     an essentially
                                                     identical manner
                                                     because of its
                                                     direct and
                                                     significant
                                                     transboundary
                                                     effects; however,
                                                     if a State has
                                                     adopted this
                                                     provision as a part
                                                     of its DOT
                                                     regulations, then
                                                     the adoption of
                                                     this section is not
                                                     necessary.

[[Page 3779]]

 
Sec. 71.51....  Additional        NRC.............  This provision is
                 Requirements                        designated NRC
                 for Type B                          because it
                 packages.                           addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.53....  [RESERVED]......
Sec. 71.55....  General           NRC.............  This provision is
                 Requirements                        designated NRC
                 for fissile                         because it
                 material                            addresses an area
                 packages.                           reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.57....  [RESERVED]......
Sec. 71.59....  Standards for     NRC.............  This provision is
                 arrays of                           designated NRC
                 fissile                             because it
                 material                            addresses an area
                 packages.                           reserved to NRC's
                                                     regulator
                                                     authority.
Sec. 71.61....  Special           NRC.............  This provision is
                 requirements                        designated NRC
                 for Type B                          because it
                 packages                            addresses an area
                 containing more                     reserved to NRC's
                 than 10\5\A2.                       regulatory
                                                     authority.
Sec. 71.63....  Special           NRC.............  This provision is
                 requirements                        designated NRC
                 for plutonium                       because it
                 shipments.                          addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.64....  Special           NRC.............  This provision is
                 requirements                        designated NRC
                 for plutonium                       because it
                 air shipments.                      addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.65....  Additional        NRC.............  This provision is
                 Requirements.                       designated NRC
                                                     because it
                                                     addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.71....  Normal            NRC.............  This provision is
                 conditions of                       designated NRC
                 transport.                          because it
                                                     addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.73....  Hypothetical      NRC.............  This provision is
                 accident                            designated NRC
                 conditions.                         because it
                                                     addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.74....  Accident          NRC.............  This provision is
                 conditions for                      designated NRC
                 air transport                       because it
                 of plutonium.                       addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.75....  Qualification of  NRC.............  This provision is
                 special form                        designated NRC
                 radioactive                         because it
                 material.                           addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.77....  Qualification of  NRC.............  This provision is
                 LSA-III                             designated NRC
                 material.                           because it
                                                     addresses an area
                                                     reserved to NRC's
                                                     regulatory
                                                     authority.
Sec. 71.81....  Applicability of  D...............  This requirement was
                 operating                           changed from a
                 controls.                           compatibility
                                                     category ``B'' to
                                                     ``D.'' This
                                                     designation was
                                                     changed because it
                                                     does not meet any
                                                     of the criteria for
                                                     designation as
                                                     Category A, B, C or
                                                     Health and Safety
                                                     and is not required
                                                     for the purposes of
                                                     compatibility.
Sec. 71.83....  Assumptions as    [B].............  This requirement was
                 to unknown                          changed from a
                 properties.                         compatibility
                                                     category ``NRC'' to
                                                     ``[B].'' Agreement
                                                     States can regulate
                                                     fissile material
                                                     below 350g. This
                                                     provision is needed
                                                     to address fissile
                                                     material regulated
                                                     by the States and
                                                     to assure that a
                                                     regulatory gap in
                                                     the regulations of
                                                     these materials is
                                                     not created. The
                                                     bracket, ``b,''
                                                     indicates that a
                                                     State should adopt
                                                     this provision in
                                                     an essentially
                                                     identical manner
                                                     because of its
                                                     direct and
                                                     significant
                                                     transboundary
                                                     effects; however,
                                                     if a State has
                                                     adopted this
                                                     provision as a part
                                                     of its DOT
                                                     regulations, then
                                                     the adoption of
                                                     this section is not
                                                     necessary.
Sec. 71.85....  Preliminary       [B].............  This provision is
                 determinations.                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.

[[Page 3780]]

 
Sec. 71.87....  Routine           [B].............  This provision is
                 determinations.                     designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this provision
                                                     is not necessary.
Sec. 71.88....  Air transport of  [B].............  This provision is
                 plutonium.                          designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this regulation
                                                     is not necessary.
Sec. 71.89....  Opening           [B].............  This provision is
                 instructions.                       designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this regulation
                                                     is not necessary.
Sec. 71.91....  Records.........  D...............  This provision does
                                                     not meet any of the
                                                     criteria for
                                                     designations
                                                     Category A, B, C,
                                                     or health and
                                                     safety. Thus, it
                                                     does not need to be
                                                     adopted by
                                                     Agreement States.
Sec. 71.93....  Inspection and    D...............  This provision does
                 tests.                              not meet any of the
                                                     criteria for
                                                     designations
                                                     Category A, B, C,
                                                     or health and
                                                     safety. Thus, it
                                                     does not need to be
                                                     adopted by
                                                     Agreement States.
Sec. 71.95....  Reports.........  D...............  This provision does
                                                     not meet any of the
                                                     criteria for
                                                     designations
                                                     Category A, B, C,
                                                     or health and
                                                     safety. Thus, it
                                                     does not need to be
                                                     adopted by
                                                     Agreement States.
Sec. 71.97....  Advance           B...............  This provision is
                 notification of                     designated
                 shipment of                         Compatibility
                 irradiated                          Category B because
                 reactor fuel                        it applies to
                 and nuclear                         activities that
                 waste.                              have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
Sec. 71.99....  Violations......  D...............
Sec. 71.100...  Criminal          D...............
                 penalties.

[[Page 3781]]

 
Sec. 71.101...  Quality           D--Paragraphs     Paragraphs (a), (b),
                 assurance         (a), (b), and     and (c)(1) are
                 requirements.     (c)(1) are        designated Category
                                   designated D      C and the essential
                                   for those         objectives of these
                                   States which      provisions should
                                   have no users     be adopted by those
                                   of Type B         Agreement States
                                   packages-other    which have
                                   than Industrial   licensees who use
                                   Radiography**.    Type B packages.
                                  C--Paragraphs      These provisions
                                   (a), (b) and      are designated
                                   (c)(1) are        Category C because
                                   designated C      the quality
                                   for those         assurance of Type B
                                   States which      packages is an
                                   have users of     activity that is
                                   Type B packages-  needed in order to
                                   other than        avoid a nationwide
                                   Industrial        gap in the
                                   Radiography.**.   regulation of the
                                  D--paragraph (f)   transportation of
                                  C--paragraph (g)   radioactive
                                   NRC-paragraphs    materials. If these
                                   (c)(2), (d) and   provisions are not
                                   (e).              adopted, this could
                                  **Note: 10 CFR     result in
                                   71.101(g)         undesirable
                                   indicates that    consequences in
                                   QA programs for   multiple
                                   industrial        jurisdictions. The
                                   radiography       essential objective
                                   Type B package    of paragraph (a) is
                                   users are         that each licensee
                                   covered by 10     who uses a Type B
                                   CFR 34.31(b).     package is
                                   It also           responsible for the
                                   indicated that    quality assurance
                                   this section      requirements which
                                   satisfies Sec.    apply to the use of
                                   71.12 (b) and     a package. The
                                   thus would        essential objective
                                   satisfy those     of paragraph (b) is
                                   secitons          that each licensee
                                   referenced in     who uses a Type B
                                   this provision    package shall
                                   (Sec.Sec.         establish,
                                   71.101 through    amintain, and
                                   71.137).          execute a quality
                                                     assurance program.
                                                     The essential
                                                     objective of
                                                     paragraph (c)(1) is
                                                     that each licensee
                                                     who uses a Type B
                                                     package shall,
                                                     prior to the use of
                                                     any package for the
                                                     shipment of any
                                                     material subject to
                                                     this part, obtain
                                                     approval of its
                                                     quality assurance
                                                     program by the
                                                     regulatory agency.
                                                    Paragraph (f) is not
                                                     required for
                                                     compatibility
                                                     because the States
                                                     have the
                                                     felxibility to
                                                     determine whether
                                                     they wish to accept
                                                     a previously
                                                     approved quality
                                                     assurance program.
Sec. 71.103...  Quality           D--for those      For paragraph (a),
                 assurance         States which      those States which
                 organization.     have no users     have licenses that
                                   of Type B         use Type B
                                   packages-other    packages, and have
                                   than Industrial   adopted the
                                   Radiography**.    essential
                                  [C]--Paragraph     objectives of Sec.
                                   (a) is            71.101(a), it is
                                   designated [C]    not necessary for
                                   for those         them to adopt this
                                   States which      provision again.
                                   have users of    Paragraph (b) is
                                   Type B packages-  designated as a
                                   other than        Category C, and the
                                   Industrial        essential
                                   Radiography**.    objectives of these
                                  C--Paragraph (b)   provisions should
                                   is designated C   be adopted by those
                                   for those         Agreement States
                                   States which      which have
                                   have users of     licensees who use
                                   Type B packages-  Type B packages.
                                   other than        This provision is
                                   Industrial        designated Category
                                   Radiography**.    C because the
                                  D--paragraphs      quality assurance
                                   (d), (e), and     of Type B packages
                                   (f).              is an activity that
                                  **Note: Sec.       is needed in order
                                   71.101 (g)        to avoid a
                                   indicates that    nationwide gap in
                                   QA programs for   the regulation of
                                   industrial        the transportation
                                   radiography       of radioactive
                                   Type B package    materials. If these
                                   users are         provisions are not
                                   covered by Sec.   adopted, this could
                                   34.31(b). It      result in
                                   also indicated    undesirable
                                   that this         consequences in
                                   section           multiple
                                   satisfies Sec.    jurisdictions. The
                                   71.12(b) and      essential objective
                                   thus would        of paragraph (b) is
                                   satisfy those     that each licensee
                                   sections          who uses a Type B
                                   referenced in     package should
                                   this provision    verify by
                                   Sec.Sec. 71.101   procedures such as
                                   through 71.137).  checking, auditing,
                                                     and inspection,
                                                     that activities
                                                     affecting the
                                                     safety-related
                                                     functions have been
                                                     performed
                                                     correctly.

[[Page 3782]]

 
Sec. 71.105...  Quality           D--for those      Para. (a) is
                 assurance         States which      designated [C] and
                 program.          have no users     para. (b) is
                                   of Type B         designated C for
                                   packages--other   those Agreement
                                   than Industrial   States with
                                   Radiography.      licensees that use
                                  C--Paragraphs      Type B packages and
                                   (a), (c), and     the essential
                                   (d) and.          objectives of these
                                  [C]--paragraph b   provisions should
                                   for those         be adopted by those
                                   States which      Agreement States.
                                   have users of     These provisions
                                   Type B packages-  are designated
                                   -other than       Category C because
                                   Industrial        the QA of Type B
                                   Radiography**.    packages is an
                                  **Note: 10 CFR     activity that is
                                   71.101(g)         needed in order to
                                   indicates that    avoid a nationwide
                                   QA programs for   regulatory gap in
                                   industrial        the regulation of
                                   radiography       the transportation
                                   Type B package    of radioactive
                                   users are         materials. If these
                                   covered by 10     provisions are not
                                   CFR 34.31(b).     adopted, this could
                                   It also           result in
                                   indicated that    undesirable
                                   this section      consequences in
                                   satisfies Sec.    multiple
                                   71.12(b) and      jurisdictions. The
                                   thus would        essential objective
                                   satisfy those     of para. (a) is
                                   sections          that each licensee
                                   referenced in     who uses a Type B
                                   this provision    package shall
                                   (Sec.Sec.         document the
                                   71.101 through    quality assurance
                                   71.137).          program by written
                                                     procedures or
                                                     instructions and
                                                     shall carry out the
                                                     program in
                                                     accordance with
                                                     those procedures
                                                     throughout the
                                                     period during which
                                                     the packaging is
                                                     used, and shall
                                                     identify the
                                                     material and
                                                     components covered
                                                     by the quality
                                                     assurance program.
                                                     The essential
                                                     objective of para.
                                                     (b) is that each
                                                     licensee who uses a
                                                     Type B package
                                                     shall control
                                                     activities
                                                     affecting the
                                                     safety-related
                                                     functions of the
                                                     Type B package.
                                                     Para. (b) is
                                                     bracketed ``C'',
                                                     because the
                                                     essential objective
                                                     of this provision
                                                     is captured by Sec.
                                                     71.103(b); if an
                                                     Agreement State
                                                     adopts the
                                                     essential
                                                     objectives of Sec.
                                                     71.103(b), it is
                                                     not necessary to
                                                     adopt this
                                                     provision again.
                                                     The essential
                                                     objective of para.
                                                     (c) is that the
                                                     licensee and
                                                     certificate holder
                                                     shall base its QA
                                                     program on items
                                                     listed in (1)
                                                     through (5). The
                                                     essential objective
                                                     of para. (d) is
                                                     that the licensee
                                                     and certificate
                                                     holder shall
                                                     provide training of
                                                     personnel
                                                     performing
                                                     activities
                                                     affecting the
                                                     quality of the
                                                     package to assure
                                                     proficiency in
                                                     their knowledge of
                                                     the QA program;
                                                     review the status
                                                     and adequacy of the
                                                     QA program at
                                                     established
                                                     intervals; and
                                                     regular management
                                                     review of the QA
                                                     program by all
                                                     cognizant
                                                     organizations.
Sec. 71.107...  Package design    NRC.............  This provision is
                 control.                            reserved to the NRC
                                                     because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.109...  Procurement       NRC.............  This provision is
                 document                            reserved to the NRC
                 control.                            because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.111...  Instructions,     NRC.............  This provision is
                 procedures, and                     reserved to the NRC
                 drawings.                           because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.113...  Document control  NRC.............  This provision is
                                                     reserved to the NRC
                                                     because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.115...  Control of        NRC.............  This provision is
                 purchased                           reserved to the NRC
                 material,                           because it
                 equipment, and                      addresses the
                 services.                           design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.117...  Identification    NRC.............  This provision is
                 and control of                      reserved to the NRC
                 materials,                          because it
                 parts, and                          addresses the
                 components.                         design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.119...  Control of        NRC.............  This provision is
                 special                             reserved to the NRC
                 processes.                          because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.121...  Internal          NRC.............  This provision is
                 Inspection.                         reserved to the NRC
                                                     because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.123...  Test control....  NRC.............  This provision is
                                                     reserved to the NRC
                                                     because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.

[[Page 3783]]

 
Sec. 71.125...  Control of        NRC.............  This provision is
                 measuring and                       reserved to the NRC
                 test equipment.                     because it
                                                     addresses the
                                                     design,
                                                     fabrication,
                                                     modification, and
                                                     approval of Type B
                                                     packages.
Sec. 71.127...  Handling,         D--for those      This provision is
                 storage, and      States which      designated Category
                 shipping          have no users     C for those States
                 control.          of Type B         which have
                                   packages--other   licensees that use
                                   than Industrial   Type B packages.
                                   Radiography.      This provision is
                                  [C]--for those     designated Category
                                   States which      C because the
                                   have users of     quality assurance
                                   Type B packages-  of Type B packages
                                   -other than       is an activity that
                                   Industrial        is needed in order
                                   Radiography**.    to avoid nationwide
                                  **Note: 10 CFR     gas in the
                                   71.101 (g)        regulation of the
                                   indicates that    transportation of
                                   QA programs for   radioactive
                                   industrial        materials. If this
                                   radiography       provision is not
                                   Type B package    adopted, this could
                                   users are         result in
                                   covered by Sec.   undesirable
                                   34.31(b). It      consequences in
                                   also indicated    multiple
                                   that this         jurisdications. For
                                   section           those States which
                                   satisfies Sec.    have licensees that
                                   71.12(b) and      use Type B
                                   thus would        packages, and have
                                   satisfy those     adopted the
                                   sections          essential
                                   referenced in     objectives of Sec.
                                   this provision    71.105, it is not
                                   (Sec.Sec.         necessary for them
                                   71.101 through    to adopt this
                                   71.137).          provision again.
Sec. 71.129...  Inspection,       D--for those      This provision is
                 test, and         States which      designated Category
                 operating         have no users     C because the
                 status.           of Type B         quality assurance
                                   packages--other   of Type B packages
                                   than Industrial   is an activity that
                                   Radiography**.    is needed in order
                                  [C]--for those     to avoid a
                                   States which      nationwide gap in
                                   have users of     the regulation of
                                   Type B packages-  the transportation
                                   -other than       of radioactive
                                   Industrial        materials. If this
                                   Radiography**.    provision is not
                                  **Note: 10 CFR     adopted, this could
                                   71.101 (g)        result in
                                   indicates that    undesirable
                                   QA programs for   consequences in
                                   industrial        multiple
                                   radiography       jurisdictions. For
                                   Type B package    those States which
                                   users are         have licensees that
                                   covered by Sec.   use Type B
                                   34.31(b). It      packages, and have
                                   also indicated    adopted the
                                   that this         essential
                                   section           objectives of Sec.
                                   satisfies Sec.    71.105, it is not
                                   71.12(b) and      necessary for them
                                   thus would        to adopt this
                                   satisfy those     provision again.
                                   sections
                                   referenced in
                                   this provision
                                   (Sec.Sec.
                                   71.101 through
                                   71.137).
Sec. 71.131...  Nonconforming     D--for those      This provision is
                 materials,        States which      designated Category
                 parts, or         have no users     C because the
                 components.       of Type B         quality assurance
                                   packages-other    of Type B packages
                                   than Industrial   is an activity that
                                   Radiography**.    is needed in order
                                  [C]--for those     to avoid a
                                   States which      nationwide gap in
                                   have users of     the regulation of
                                   Type B packages-  the transportation
                                   -other than       of radioactive
                                   Industrial        materials. If this
                                   Radiography**.    provision is not
                                  **Note: 10 CFR     adopted, this could
                                   71.101 (g)        result in
                                   indicates that    undesirable
                                   QA programs for   consequences in
                                   industrial        multiple
                                   radiography       jurisdictions. For
                                   Type B package    those States which
                                   users are         have licensees that
                                   covered by Sec.   use Type B
                                   34.31(b). It      packages, and have
                                   also indicated    adopted the
                                   that this         essential
                                   section           objectives of Sec.
                                   satisfies Sec.    71.105, it is not
                                   71.12(b) and      necessary for them
                                   thus would        to adopt this
                                   satisfy those     provision again.
                                   sections
                                   referenced in
                                   this provision
                                   (Sec.Sec.
                                   71.101 through
                                   71.137).
Sec. 71.133...  Corrective        D--for those      This provision is
                 action.           States which      designated Category
                                   have no users     C for those States
                                   of Type B         which have
                                   packages--other   licensees that use
                                   than Industrial   Type B packages.
                                   Radiography**.    This provision is
                                  C--for those       designated Category
                                   States which      C because the
                                   have users of     quality assurance
                                   Type B packages-  of Type B packages
                                   -other than       is an activity that
                                   Industrial        is needed in order
                                   Radiography**.    to avoid a
                                  **Note: 10 CFR     nationwide gap in
                                   71.101 (g)        the regulation of
                                   indicates that    the transportation
                                   QA programs for   of radioactive
                                   industrial        materials. If this
                                   radiography       provision is not
                                   Type B package    adopted, this could
                                   users are         result in
                                   covered by Sec.   undesirable
                                   34.31(b). It      consequences in
                                   also indicated    multiple
                                   that this         jurisdictions. The
                                   section           essential objective
                                   satisfies Sec.    of this provision
                                   71.12(b) and      is that each
                                   thus would        licensee who uses a
                                   satisfy those     Type B package
                                   sections          shall establish
                                   referenced in     measures to assure
                                   this provision    that conditions
                                   (Sec.Sec.         adverse to quality,
                                   71.101 through    such as
                                   71.137).          deficiencies,
                                                     deviations,
                                                     defective material
                                                     and equipment, and
                                                     nonconformances,
                                                     are promptly
                                                     identified and
                                                     corrected.

[[Page 3784]]

 
Sec. 71.135...  Quality           D--for those      This provision is
                 assurance         States which      designated a
                 records.          have no users     Category C for
                                   of Type B         those States which
                                   packages--other   have licensees that
                                   than industrial   use Type B
                                   Radiography**.    packages. This
                                  C--for those       provision is
                                   States which      designated Category
                                   have users of     C because the
                                   Type B packages-  quality assurance
                                   -other than       of Type B packages
                                   industrial        is an activity that
                                   radiography**.    is needed in order
                                  **Note: 10 CFR     to avoid a
                                   71.101(g)         nationwide gap in
                                   indicates that    the regulation of
                                   QA programs for   the transportation
                                   industrial        of radioactive
                                   radiography       materials. If this
                                   Type B package    provision is not
                                   users are         adopted, this could
                                   covered by Sec.   result in
                                   34.31(b). It      undesirable
                                   also indicated    consequences in
                                   that this         multiple
                                   section           jurisdictions. The
                                   satisfies Sec.    essential objective
                                   71.12(b) and      of this provision
                                   thus would        is that each
                                   satisfy those     licensee who uses a
                                   sections          Type B package
                                   referenced in     shall maintain
                                   this provision    sufficient written
                                   (Sec.Sec.         records to
                                   71.101 through    demonstrate
                                   71.137).          compliance with the
                                                     quality assurance
                                                     program.
Sec. 71.137...  Audits..........  D--for those      This provision is
                                   States which      designated a
                                   have no users     Category C for
                                   of Type B         those States which
                                   packages--other   have licensees that
                                   than Industrial   use Type B
                                   Radiography**.    packages. This
                                  C--for those       provision is
                                   States which      designated Category
                                   have users of     C because the
                                   Type B packages-  quality assurance
                                   -other than       of Type B packages
                                   Industrial        is an activity that
                                   Radiography**.    is needed in order
                                  **Note: 10 CFR     to avoid a
                                   71.101(g)         nationwide gap in
                                   indicates that    the regulation of
                                   QA program for    the transportation
                                   industrial        of radioactive
                                   radiography       materials. If this
                                   Type B package    provision is not
                                   users are         adopted, this could
                                   covered by Sec.   result in
                                   34.31(b). It      undesirable
                                   also indicated    consequences in
                                   that this         multiple
                                   section           jurisdictions. The
                                   satisfies Sec.    essential
                                   71.12(b) and      objectives of this
                                   thus would        provision are that
                                   satisfy those     each licensee who
                                   sections          uses a Type B
                                   referenced in     package shall carry
                                   this provision    out a system of
                                   Sec.Sec. 71.101   planned and
                                   through 71.137).  periodic audits to:
                                                     (1) verify
                                                     compliance with all
                                                     aspects of the
                                                     quality assurance
                                                     program, (2)
                                                     determine the
                                                     effectiveness of
                                                     the program, (3)
                                                     verify that the
                                                     audits are
                                                     performed by
                                                     appropriately
                                                     trained personnel,
                                                     (4) audits
                                                     performed in
                                                     accordance with
                                                     procedures; (5)
                                                     audit results
                                                     documented and
                                                     reviewed by
                                                     appropriate
                                                     management; and (6)
                                                     follow-up actions
                                                     are taken as
                                                     necessary.
Appendix A....  Determination of  [B].............  This definition is
                 A1 and A2.                          designated
                                                     Compatibility
                                                     Category B because
                                                     it applies to
                                                     activities that
                                                     have direct and
                                                     significant effects
                                                     in multiple
                                                     jurisdictions. An
                                                     Agreement State
                                                     should adopt
                                                     Category B program
                                                     elements in an
                                                     essentially
                                                     identical manner.
                                                     The bracket, ``B,''
                                                     indicates that if a
                                                     State has adopted
                                                     this provision in
                                                     another portion of
                                                     its regulations,
                                                     such as the State's
                                                     DOT regulations,
                                                     then the adoption
                                                     of this requirement
                                                     is not necessary.
------------------------------------------------------------------------

VII. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 104-113, requires that Federal agencies use technical standards that 
are developed or adopted by voluntary consensus standard bodies unless 
the use of such a standard is inconsistent with applicable law or 
otherwise impractical. In this rule, the NRC considered but decided not 
to adopt the ASME Code, Section III, Division 3, as described in Issue 
14. However, NRC has amended its transportation regulations to make 
them compatible with the IAEA transportation standards. This action 
does not constitute the establishment of a standard that establishes 
generally applicable requirements.

VIII. Environmental Assessment: Finding of No Significant Environmental 
Impact

    The Commission has prepared an environmental assessment entitled 
Final Environmental Assessment (EA) of Major Revision of 10 CFR part 71 
(NUREG/CR-6711, December 2003), on this regulation. The EA is available 
on the NRC rulemaking Web site (http://ruleforum.llnl.gov) and is also 
available for inspection in the NRC Public Document Room, 11555 
Rockville Pike, Room O-1F21, Rockville, MD. The following is a brief 
summary of the EA.
    The EA grouped the proposed action into 19 different changes to 
part 71, which could be adopted either all together as one list or 
independently in a partial list. Of these 19 changes, the following 4 
meet the NRC's categorical exclusion criteria:

     Changes to Various Definitions (Issue 9);
     Expansion of Part 71 Quality Assurance 
Requirements to Certificate of Compliance (CoC) Holders (Issue 13);
     Change Authority for Dual-Purpose Package 
Certificate Holders (Issue 15); and
     Modifications of Event Reporting Requirements 
(Issue 19).
    None of the remaining 15 changes are expected to cause a 
significant impact to human health, safety, or the environment, whether 
issued altogether or individually. In fact, most of the

[[Page 3785]]

changes would have negligible effects or result in slight improvements 
in health, safety, and environmental protection. In particular, the 
following changes are primarily administrative in nature, would not 
cause any new negative impacts, and would result in the beneficial 
effect of simplifying and/or harmonizing the NRC's regulations with TS-
R-1:

     Changing Part 71 to the International System of 
Units (SI) Only (Issue 1);
     Revision of A1 and A2 
(Issue 3);
     A new requirement to display the Criticality 
Safety Index on shipping packages of fissile material (Issue 5);
     A provision to ``grandfather'' older shipping 
packages under the part 71 requirements in existence when their 
Certificates of Compliance were issued (Issue 8); and
     Procedures for approval of special arrangements 
for shipment of special packages (Issue 12).
    The following changes would result in slight net improvements in 
health, safety, and environmental protection:

     Addition of uranium hexafluoride package 
requirements (Issue 4);
     Strengthening the requirements in Sec. 71.61 to 
ensure package containment in deep submersion scenarios (Issue 7);
     Adoption of the crush test for fissile material 
package design (Issue 10);
     Adoption of fissile material package design 
requirements for transport by aircraft (Issue 11); and
     Adoption of the ASME Code for spent fuel 
transportation casks (Issue 14).
    The proposal to change the existing 70-Bq/g (0.002-[mu]Ci/g) level 
to radionuclide-specific activity limits (Issue 2) is expected to have 
mixed, although overall minor, effects. For radionuclides with new 
exemption values that are lower than the current limit, there could be 
a decrease in the number of exempted shipments and a commensurate 
slight increase in the level of protection. For radionuclides with new 
exemption values that are higher than the current limit, there could be 
an increase in the number of exempted shipments and a commensurate 
slight increase in associated radiation exposures. However, IAEA and 
the NRC have determined that this change would not significantly 
increase the risk to individuals.
    The addition of the Type C package and low level dispersible 
material concepts (Issue 6) would result in mixed, although overall 
minor, effects. If the same number of packages are handled, the 
radiation doses to workers loading and unloading Type C packages 
shipped by air will be slightly higher than the doses to workers 
loading and unloading other kinds of packages shipped by other means. 
At the same time, ``incident-free'' doses during the shipping of Type C 
packages are expected to be slightly reduced compared to baseline 
conditions, while the risks associated with accidents during shipping 
could be slightly increased or decreased depending on the shipping 
scenario.
    Changes to transportation regulations for fissile materials 
actually consist of 17 individual recommendations for revisions to part 
71 (Issue 16). Ten of these recommendations are expected to result in 
no impact, as they simply clarify definitions, consolidate related 
requirements into single sections, or streamline the regulations. Four 
of the recommendations will result in small improvements to health, 
safety, and environmental protection by eliminating confusion among 
licensees and/or providing added assurance for critical safety. The 
last two recommendations, which would revise exemptions for low-level 
material and remove or modify provisions related to the shipment of Pu-
Be neutron sources, are expected to significantly improve criticality 
safety.
    Changes to the requirements for plutonium shipments in Sec. 71.63 
(PRM-71-12) could result in a slight increase in the probability and 
consequences of accidental releases, primarily when and if plutonium is 
shipped in liquid form. However, most plutonium shipments are either 
related to the disposition of plutonium wastes or to the production of 
mixed oxides, neither of which involve the shipment of a liquid 
solution of plutonium.
    No changes have been identified for the issue related to surface 
contamination limits as applied to spent fuel and high level waste 
(Issue 18). The issue was included in the proposed rule in response to 
Commission direction in SRM-SECY-00-0117. NRC is seeking input on 
whether the NRC should address this issue in future rulemaking 
activities. As a result, no regulatory options were developed, and 
therefore no environmental assessment conducted.
    The Commission has determined, under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
subpart A of 10 CFR part 51, that this rule is not a major Federal 
action significantly affecting the quality of the human environment, 
and therefore an environmental impact statement (EIS) is not required.
    The Commission's ``Final Environmental Statement on the 
Transportation of Radioactive Material by Air and Other Modes,'' NUREG-
0170 \14\, dated December 1977, is NRC's generic EIS, covering all 
types of radioactive material transportation by all modes (road, rail, 
air, and water). From the Commission's latest survey of radioactive 
material shipments and their characteristics, ``Transport of 
Radioactive Material in the United States,'' SAND 84-7174, April 1985, 
the NRC concluded that current radioactive material shipments are not 
so different from those evaluated in NUREG-0170 as to invalidate the 
results or conclusions of that EIS. The environmental assessment of the 
impacts associated with this rulemaking is evaluated in Final 
Environmental Assessment (EA) of Major Revision of 10 CFR part 71 
(NUREG/CR-6711, December 2003).
---------------------------------------------------------------------------

    \14\ Copies of NUREG-0170 may be purchased from the 
Superintendent of Documents, U.S. Government Printing Office, P.O. 
Box 37082, Washington, DC 20013-7082. Copies are also available from 
the National Technical Information Service, 5285 Port Royal Road, 
Springfield, VA 22161. A copy is also available for inspection and 
copying for a fee in the NRC Public Document Room, 11555 Rockville 
Pike, Room O-1F21, Rockville, MD.
---------------------------------------------------------------------------

    NUREG-0170 established the nonaccident related radiation exposures 
associated with transportation of radioactive material in the United 
States as 98 person-Sv (9800 person-rem) which, based on the 
conservative linear radiation dose hypothesis, resulted in a maximum of 
1.7 genetic effects and 1.2 latent cancer effects per year. More than 
half this impact resulted from shipment of medical-use radioactive 
materials. Accident related impacts were established at a maximum of 
one genetic effect and one latent cancer fatality for 200 years of 
transporting radioactive materials. The principal nonradiological 
impacts were found to be two injuries per year and less than one 
accidental death per 4 years. In contrast, nonaccident related 
radiation exposures and accident related impacts associated with this 
rulemaking would not change from the impact of the current part 71 
requirements (i.e., no increase or decrease). Nonradiological traffic 
injuries and nonradiological traffic deaths would not change. These 
impacts are judged to be insignificant compared with the baseline 
impacts established in NUREG-0170.

IX. Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These

[[Page 3786]]

requirements were approved by the Office of Management and Budget, 
approval number 3150-0008.
    The burden to the public for these information collections is 
estimated to average 19.2 hours per licensee, including the time for 
reviewing instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
information collection. Send comments on any aspect of these 
information collections, including suggestions for reducing the burden, 
to the Records Management Branch (T-5F52), U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, or by Internet electronic mail 
to INFOCOLLECTS@nrc.gov; and to the Desk Officer, Office of Information 
and Regulatory Affairs, NEOB-10202,(3150-0008), Office of Management 
and Budget, Washington, DC 20503.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

X. Regulatory Analysis

    The Commission has prepared a regulatory analysis entitled ``Final 
Regulatory Analysis of Major Revision of 10 CFR part 71--NUREG/CR-6713, 
December 2003. `` To support the discussions of the proposed changes, 
selected material from this regulatory analysis has been included 
earlier under each issue. The analysis examines the costs and benefits 
of the alternatives considered by the Commission. The regulatory 
analysis is available on the NRC rulemaking Web site, and is also 
available for inspection at the NRC Public Document Room, 11555 
Rockville Pike, Room O-1F21, Rockville, MD.

XI. Regulatory Flexibility Act Certification

    In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 
605(b)), the Commission certifies that this rule will not have a 
significant economic impact on a substantial number of small entities. 
This rule affects NRC licensees, including operators of nuclear power 
plants, who transport or deliver to a carrier for transport, relatively 
large quantities of radioactive material in a single package. These 
companies do not generally fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
size standards adopted by the NRC (10 CFR 2.810).
    Only one small entity commented on the proposed changes suggesting 
that small entities would be negatively affected by the rule. Reviewing 
records of licensed QA programs, NRC found that only 15 of the 127 NRC-
licensed QA progams were small entities. Furthermore, of these 15 
companies, NRC staff expects that only two or three would be negatively 
affected by the final rule, given these companies' lines of business 
and day-to-day operations. Based on these data, it is believed there 
will not be significant economic impacts for a substantial number of 
small entities.

XII. Backfit Analysis

    The NRC has determined that the backfit rule does not apply to this 
rule; therefore, a backfit analysis is not required for this rule 
because these amendments do not involve any provisions that would 
require backfits as defined in 10 CFR chapter I.

List of Subjects in 10 CFR Part 71

    Criminal penalties, Hazardous materials transportation, Nuclear 
materials, Packaging and containers, Reporting and recordkeeping 
requirements.

0
For the reasons set out in the preamble and under the authority of the 
Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 
1974, as amended; and 5 U.S.C. 552 and 553, the Commission is adopting 
the following amendments to 10 CFR part 71.

PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

0
1. The authority citation for part 71 continues to read as follows:

    Authority: Secs. 53, 57, 62, 63, 81, 161, 182, 183, 234, 68 
Stat. 930, 932, 933, 935, 948, 953, 954, as amended, sec. 1701, 106 
Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 
2201, 2232, 2233, 2297f); secs. 201, as amended, 202, 206, 88 Stat. 
1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 
1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
    Section 71.97 also issued under sec. 301, Pub. L. 96-295, 94 
Stat. 789-790.
0
2. Subparts A, B, and C to part 71 are revised to read as follows:
Subpart A--General Provisions
Sec.
71.0 Purpose and scope.
71.1 Communications and records.
71.2 Interpretations.
71.3 Requirement for license.
71.4 Definitions.
71.5 Transportation of licensed material.
71.6 Information collection requirements: OMB approval.
71.7 Completeness and accuracy of information.
71.8 Deliberate misconduct.
71.9 Employee protection.
71.10 Public inspection of application.
71.11 [Reserved]
Subpart B--Exemptions
71.12 Specific exemptions.
71.13 Exemption of physicians.
71.14 Exemption for low-level materials.
71.15 Exemption from classification as fissile material.
71.16 [Reserved]
Subpart C--General Licenses
71.17 General license: NRC-approved package.
71.18 [Reserved]
71.19 Previously approved package.
71.20 General license: DOT specification container.
71.21 General license: Use of foreign approved package.
71.22 General license: Fissile material.
71.23 General license: Plutonium-beryllium special form material.
71.24 [Reserved]
71.25 [Reserved]

Subpart A--General Provisions


Sec. 71.0  Purpose and scope.

    (a) This part establishes--
    (1) Requirements for packaging, preparation for shipment, and 
transportation of licensed material; and
    (2) Procedures and standards for NRC approval of packaging and 
shipping procedures for fissile material and for a quantity of other 
licensed material in excess of a Type A quantity.
    (b) The packaging and transport of licensed material are also 
subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30, 
40, 70, and 73) and to the regulations of other agencies (e.g., the 
U.S. Department of Transportation (DOT) and the U.S. Postal Service) 
\1\ having jurisdiction over means of transport. The requirements of 
this part are in addition to, and not in substitution for, other 
requirements.
---------------------------------------------------------------------------

    \1\ Postal Service manual (Domestic Mail Manual), Section 124, 
which is incorporated by reference at 39 CFR 111.1.
---------------------------------------------------------------------------

    (c) The regulations in this part apply to any licensee authorized 
by specific or general license issued by the Commission to receive, 
possess, use, or transfer licensed material, if the licensee delivers 
that material to a carrier for transport, transports the material 
outside the site of usage as specified in the NRC license, or 
transports that material on public highways. No provision of this part 
authorizes possession of licensed material.
    (d)(1) Exemptions from the requirement for license in Sec. 71.3 are 
specified in Sec. 71.14. General licenses for which no NRC package 
approval is

[[Page 3787]]

required are issued in Sec.Sec. 71.20 through 71.23. The general 
license in Sec. 71.17 requires that an NRC certificate of compliance or 
other package approval be issued for the package to be used under this 
general license.
    (2) Application for package approval must be completed in 
accordance with subpart D of this part, demonstrating that the design 
of the package to be used satisfies the package approval standards 
contained in subpart E of this part, as related to the tests of subpart 
F of this part.
    (3) A licensee transporting licensed material, or delivering 
licensed material to a carrier for transport, shall comply with the 
operating control requirements of subpart G of this part; the quality 
assurance requirements of subpart H of this part; and the general 
provisions of subpart A of this part, including DOT regulations 
referenced in Sec. 71.5.
    (e) The regulations of this part apply to any person holding, or 
applying for, a certificate of compliance, issued pursuant to this 
part, for a package intended for the transportation of radioactive 
material, outside the confines of a licensee's facility or authorized 
place of use.
    (f) The regulations in this part apply to any person required to 
obtain a certificate of compliance, or an approved compliance plan, 
pursuant to part 76 of this chapter, if the person delivers radioactive 
material to a common or contract carrier for transport or transports 
the material outside the confines of the person's plant or other 
authorized place of use.
    (g) This part also gives notice to all persons who knowingly 
provide to any licensee, certificate holder, quality assurance program 
approval holder, applicant for a license, certificate, or quality 
assurance program approval, or to a contractor, or subcontractor of any 
of them, components, equipment, materials, or other goods or services, 
that relate to a licensee's, certificate holder's, quality assurance 
program approval holder's, or applicant's activities subject to this 
part, that they may be individually subject to NRC enforcement action 
for violation of Sec. 71.8.


Sec. 71.1  Communications and records.

    (a) Except where otherwise specified, all communications and 
reports concerning the regulations in this part and applications filed 
under them should be sent by mail addressed: ATTN: Document Control 
Desk, Director, Spent Fuel Project Office, Office of Nuclear Material 
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001, by hand delivery to the NRC's offices at 11555 Rockville 
Pike, Rockville, Maryland; or, where practicable, by electronic 
submission, for example, via Electronic Information Exchange, or CD-
ROM. Electronic submissions must be made in a manner that enables the 
NRC to receive, read, authenticate, distribute, and archive the 
submission, and process and retrieve it a single page at a time. 
Detailed guidance on making electronic submissions can be obtained by 
visiting the NRC's Web site at http://www.nrc.gov/site-help/eie.html, 
by calling (301) 415-6030, by e-mail to EIE@nrc.gov, or by writing the 
Office of the Chief Information Officer, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001. The guidance discusses, among 
other topics, the formats the NRC can accept, the use of electronic 
signatures, and the treatment of nonpublic information.
    (b) Each record required by this part must be legible throughout 
the retention period specified by each Commission regulation. The 
record may be the original or a reproduced copy or a microform provided 
that the copy or microform is authenticated by authorized personnel and 
that the microform is capable of producing a clear copy throughout the 
required retention period. The record may also be stored in electronic 
media with the capability for producing legible, accurate, and complete 
records during the required retention period. Records such as letters, 
drawings, and specifications must include all pertinent information 
such as stamps, initials, and signatures. The licensee shall maintain 
adequate safeguards against tampering with and loss of records.


Sec. 71.2  Interpretations.

    Except as specifically authorized by the Commission in writing, no 
interpretation of the meaning of the regulations in this part by any 
officer or employee of the Commission, other than a written 
interpretation by the General Counsel, will be recognized to be binding 
upon the Commission.


Sec. 71.3  Requirement for license.

    Except as authorized in a general license or a specific license 
issued by the Commission, or as exempted in this part, no licensee may-
-
    (a) Deliver licensed material to a carrier for transport; or
    (b) Transport licensed material.


Sec. 71.4  Definitions.

    The following terms are as defined here for the purpose of this 
part. To ensure compatibility with international transportation 
standards, all limits in this part are given in terms of dual units: 
The International System of Units (SI) followed or preceded by U.S. 
standard or customary units. The U.S. customary units are not exact 
equivalents but are rounded to a convenient value, providing a 
functionally equivalent unit. For the purpose of this part, either unit 
may be used.
    A1 means the maximum activity of special form radioactive material 
permitted in a Type A package. This value is either listed in Appendix 
A, Table A-1, of this part, or may be derived in accordance with the 
procedures prescribed in Appendix A of this part.
    A2 means the maximum activity of radioactive material, other than 
special form material, LSA, and SCO material, permitted in a Type A 
package. This value is either listed in Appendix A, Table A-1, of this 
part, or may be derived in accordance with the procedures prescribed in 
Appendix A of this part.
    Carrier means a person engaged in the transportation of passengers 
or property by land or water as a common, contract, or private carrier, 
or by civil aircraft.
    Certificate holder means a person who has been issued a certificate 
of compliance or other package approval by the Commission.
    Certificate of Compliance (CoC) means the certificate issued by the 
Commission under subpart D of this part which approves the design of a 
package for the transportation of radioactive material.
    Close reflection by water means immediate contact by water of 
sufficient thickness for maximum reflection of neutrons.
    Consignment means each shipment of a package or groups of packages 
or load of radioactive material offered by a shipper for transport.
    Containment system means the assembly of components of the 
packaging intended to retain the radioactive material during transport.
    Conveyance means:
    (1) For transport by public highway or rail any transport vehicle 
or large freight container;
    (2) For transport by water any vessel, or any hold, compartment, or 
defined deck area of a vessel including any transport vehicle on board 
the vessel; and
    (3) For transport by any aircraft.
    Criticality Safety Index (CSI) means the dimensionless number 
(rounded up to the next tenth) assigned to and placed on the label of a 
fissile material package, to designate the degree of control of

[[Page 3788]]

accumulation of packages containing fissile material during 
transportation. Determination of the criticality safety index is 
described in Sec.Sec. 71.22, 71.23, and 71.59.
    Deuterium means, for the purposes of Sec.Sec. 71.15 and 71.22, 
deuterium and any deuterium compounds, including heavy water, in which 
the ratio of deuterium atoms to hydrogen atoms exceeds 1:5000.
    DOT means the U.S. Department of Transportation.
    Exclusive use means the sole use by a single consignor of a 
conveyance for which all initial, intermediate, and final loading and 
unloading are carried out in accordance with the direction of the 
consignor or consignee. The consignor and the carrier must ensure that 
any loading or unloading is performed by personnel having radiological 
training and resources appropriate for safe handling of the 
consignment. The consignor must issue specific instructions, in 
writing, for maintenance of exclusive use shipment controls, and 
include them with the shipping paper information provided to the 
carrier by the consignor.
    Fissile material means the radionuclides uranium-233, uranium-235, 
plutonium-239, and plutonium-241, or any combination of these 
radionuclides. Fissile material means the fissile nuclides themselves, 
not material containing fissile nuclides. Unirradiated natural uranium 
and depleted uranium and natural uranium or depleted uranium, that has 
been irradiated in thermal reactors only, are not included in this 
definition. Certain exclusions from fissile material controls are 
provided in Sec. 71.15.
    Graphite means, for the purposes of Sec.Sec. 71.15 and 71.22, 
graphite with a boron equivalent content less than 5 parts per million 
and density greater than 1.5 grams per cubic centimeter.
    Licensed material means byproduct, source, or special nuclear 
material received, possessed, used, or transferred under a general or 
specific license issued by the Commission pursuant to the regulations 
in this chapter.
    Low Specific Activity (LSA) material means radioactive material 
with limited specific activity which is nonfissile or is excepted under 
Sec. 71.15, and which satisfies the descriptions and limits set forth 
below. Shielding materials surrounding the LSA material may not be 
considered in determining the estimated average specific activity of 
the package contents. LSA material must be in one of three groups:
    (1) LSA--I.
    (i) Uranium and thorium ores, concentrates of uranium and thorium 
ores, and other ores containing naturally occurring radioactive 
radionuclides which are not intended to be processed for the use of 
these radionuclides;
    (ii) Solid unirradiated natural uranium or depleted uranium or 
natural thorium or their solid or liquid compounds or mixtures;
    (iii) Radioactive material for which the A2 value is 
unlimited; or
    (iv) Other radioactive material in which the activity is 
distributed throughout and the estimated average specific activity does 
not exceed 30 times the value for exempt material activity 
concentration determined in accordance with Appendix A.
    (2) LSA--II.
    (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
    (ii) Other material in which the activity is distributed throughout 
and the average specific activity does not exceed 10-\4\ 
A2/g for solids and gases, and 
10-\5\A2/g for liquids.
    (3) LSA--III. Solids (e.g., consolidated wastes, activated 
materials), excluding powders, that satisfy the requirements of Sec. 
71.77, in which:
    (i) The radioactive material is distributed throughout a solid or a 
collection of solid objects, or is essentially uniformly distributed in 
a solid compact binding agent (such as concrete, bitumen, ceramic, 
etc.);
    (ii) The radioactive material is relatively insoluble, or it is 
intrinsically contained in a relatively insoluble material, so that 
even under loss of packaging, the loss of radioactive material per 
package by leaching, when placed in water for 7 days, would not exceed 
0.1 A2; and
    (iii) The estimated average specific activity of the solid does not 
exceed 2 x 10-3 A2/g.
    Low toxicity alpha emitters means natural uranium, depleted 
uranium, natural thorium; uranium-235, uranium-238, thorium-232, 
thorium-228 or thorium-230 when contained in ores or physical or 
chemical concentrates or tailings; or alpha emitters with a half-life 
of less than 10 days.
    Maximum normal operating pressure means the maximum gauge pressure 
that would develop in the containment system in a period of 1 year 
under the heat condition specified in Sec. 71.71(c)(1), in the absence 
of venting, external cooling by an ancillary system, or operational 
controls during transport.
    Natural thorium means thorium with the naturally occurring 
distribution of thorium isotopes (essentially 100 weight percent 
thorium-232).
    Normal form radioactive material means radioactive material that 
has not been demonstrated to qualify as ``special form radioactive 
material.''
    Optimum interspersed hydrogenous moderation means the presence of 
hydrogenous material between packages to such an extent that the 
maximum nuclear reactivity results.
    Package means the packaging together with its radioactive contents 
as presented for transport.
    (1) Fissile material package or Type AF package, Type BF package, 
Type B(U)F package, or Type B(M)F package means a fissile material 
packaging together with its fissile material contents.
    (2) Type A package means a Type A packaging together with its 
radioactive contents. A Type A package is defined and must comply with 
the DOT regulations in 49 CFR part 173.
    (3) Type B package means a Type B packaging together with its 
radioactive contents. On approval, a Type B package design is 
designated by NRC as B(U) unless the package has a maximum normal 
operating pressure of more than 700 kPa (100 lbs/in2) gauge 
or a pressure relief device that would allow the release of radioactive 
material to the environment under the tests specified in Sec. 71.73 
(hypothetical accident conditions), in which case it will receive a 
designation B(M). B(U) refers to the need for unilateral approval of 
international shipments; B(M) refers to the need for multilateral 
approval of international shipments. There is no distinction made in 
how packages with these designations may be used in domestic 
transportation. To determine their distinction for international 
transportation, see DOT regulations in 49 CFR Part 173. A Type B 
package approved before September 6, 1983, was designated only as Type 
B. Limitations on its use are specified in Sec. 71.19.
    Packaging means the assembly of components necessary to ensure 
compliance with the packaging requirements of this part. It may consist 
of one or more receptacles, absorbent materials, spacing structures, 
thermal insulation, radiation shielding, and devices for cooling or 
absorbing mechanical shocks. The vehicle, tie-down system, and 
auxiliary equipment may be designated as part of the packaging.
    Special form radioactive material means radioactive material that 
satisfies the following conditions:
    (1) It is either a single solid piece or is contained in a sealed 
capsule that can be opened only by destroying the capsule;
    (2) The piece or capsule has at least one dimension not less than 5 
mm (0.2 in); and

[[Page 3789]]

    (3) It satisfies the requirements of Sec. 71.75. A special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on June 30, 1983 (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before July 1, 1985, and a special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on March 31, 1996 (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before April 1, 1998, may continue to be 
used. Any other special form encapsulation must meet the specifications 
of this definition.
    Specific activity of a radionuclide means the radioactivity of the 
radionuclide per unit mass of that nuclide. The specific activity of a 
material in which the radionuclide is essentially uniformly distributed 
is the radioactivity per unit mass of the material.
    Spent nuclear fuel or Spent fuel means fuel that has been withdrawn 
from a nuclear reactor following irradiation, has undergone at least 1 
year's decay since being used as a source of energy in a power reactor, 
and has not been chemically separated into its constituent elements by 
reprocessing. Spent fuel includes the special nuclear material, 
byproduct material, source material, and other radioactive materials 
associated with fuel assemblies.
    State means a State of the United States, the District of Columbia, 
the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American 
Samoa, and the Commonwealth of the Northern Mariana Islands.
    Surface Contaminated Object (SCO) means a solid object that is not 
itself classed as radioactive material, but which has radioactive 
material distributed on any of its surfaces. SCO must be in one of two 
groups with surface activity not exceeding the following limits:
    (1) SCO-I: A solid object on which:
    (i) The nonfixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
cm2) does not exceed 4 Bq/cm2 (10-4 
microcurie/cm2) for beta and gamma and low toxicity alpha 
emitters, or 0.4 Bq/cm2 (10-5 microcurie/
cm2) for all other alpha emitters;
    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
cm2) does not exceed 4 x 10-4 Bq/cm2 
(1.0 microcurie/cm2) for beta and gamma and low toxicity 
alpha emitters, or 4 x 103 Bq/cm2 (0.1 
microcurie/cm2) for all other alpha emitters; and
    (iii) The nonfixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm2 (or the area 
of the surface if less than 300 cm2) does not exceed 4 x 
104 Bq/cm2 (1 microcurie/cm2) for beta 
and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2 
(0.1 microcurie/cm2) for all other alpha emitters.
    (2) SCO-II: A solid object on which the limits for SCO-I are 
exceeded and on which:
    (i) The nonfixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
2) does not exceed 400 Bq/cm2 (10-2 
microcurie/cm2) for beta and gamma and low toxicity alpha 
emitters or 40 Bq/cm2 (10-3 microcurie/
cm2) for all other alpha emitters;
    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
cm2) does not exceed 8 x 105 Bq/cm2 
(20 microcuries/cm2) for beta and gamma and low toxicity 
alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/
cm2) for all other alpha emitters; and
    (iii) The nonfixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm2 (or the area 
of the surface if less than 300 2) does not exceed 8 x 
105 Bq/cm2 (20 microcuries/cm2) for 
beta and gamma and low toxicity alpha emitters, or 8 x 104 
Bq/cm2 (2 microcuries/cm2) for all other alpha 
emitters.
    Transport index (TI) means the dimensionless number (rounded up to 
the next tenth) placed on the label of a package, to designate the 
degree of control to be exercised by the carrier during transportation. 
The transport index is the number determined by multiplying the maximum 
radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from 
the external surface of the package by 100 (equivalent to the maximum 
radiation level in millirem per hour at 1 meter (3.3 ft)).
    Type A quantity means a quantity of radioactive material, the 
aggregate radioactivity of which does not exceed A1 for 
special form radioactive material, or A2, for normal form 
radioactive material, where A1 and A2 are given 
in Table A-1 of this part, or may be determined by procedures described 
in Appendix A of this part.
    Type B quantity means a quantity of radioactive material greater 
than a Type A quantity.
    Unirradiated uranium means uranium containing not more than 2 x 
103 Bq of plutonium per gram of uranium-235, not more than 9 
x 106 Bq of fission products per gram of uranium-235, and 
not more than 5 x 10-3 g of uranium-236 per gram of uranium-
235.
    Uranium--natural, depleted, enriched:
    (1) Natural uranium means uranium with the naturally occurring 
distribution of uranium isotopes (approximately 0.711 weight percent 
uranium-235, and the remainder by weight essentially uranium-238).
    (2) Depleted uranium means uranium containing less uranium-235 than 
the naturally occurring distribution of uranium isotopes.
    (3) Enriched uranium means uranium containing more uranium-235 than 
the naturally occurring distribution of uranium isotopes.


Sec. 71.5  Transportation of licensed material.

    (a) Each licensee who transports licensed material outside the site 
of usage, as specified in the NRC license, or where transport is on 
public highways, or who delivers licensed material to a carrier for 
transport, shall comply with the applicable requirements of the DOT 
regulations in 49 CFR parts 170 through 189 appropriate to the mode of 
transport.
    (1) The licensee shall particularly note DOT regulations in the 
following areas:
    (i) Packaging--49 CFR part 173: subparts A, B, and I.
    (ii) Marking and labeling--49 CFR part 172: subpart D, Sec.Sec. 
172.400 through 172.407, Sec.Sec. 172.436 through 172.440, and subpart 
E.
    (iii) Placarding--49 CFR part 172: subpart F, especially Sec.Sec. 
172.500 through 172.519, 172.556, and appendices B and C.
    (iv) Accident reporting--49 CFR part 171: Sec.Sec. 171.15 and 
171.16.
    (v) Shipping papers and emergency information--49 CFR part 172: 
subparts C and G.
    (vi) Hazardous material employee training--49 CFR part 172: subpart 
H.
    (vii) Hazardous material shipper/carrier registration--49 CFR part 
107: subpart G.
    (2) The licensee shall also note DOT regulations pertaining to the 
following modes of transportation:
    (i) Rail--49 CFR part 174: subparts A through D and K.
    (ii) Air--49 CFR part 175.
    (iii) Vessel--49 CFR part 176: subparts A through F and M.
    (iv) Public Highway--49 CFR part 177 and parts 390 through 397.
    (b) If DOT regulations are not applicable to a shipment of licensed 
material, the licensee shall conform to the standards and requirements 
of the DOT specified in paragraph (a) of this section to the same 
extent as if the shipment or transportation were subject to DOT 
regulations. A request for

[[Page 3790]]

modification, waiver, or exemption from those requirements, and any 
notification referred to in those requirements, must be filed with, or 
made to, the Director, Office of Nuclear Material Safety and 
Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.


Sec. 71.6  Information collection requirements: OMB approval.

    (a) The Nuclear Regulatory Commission has submitted the information 
collection requirements contained in this part to the Office of 
Management and Budget (OMB) for approval as required by the Paperwork 
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or 
sponsor, and a person is not required to respond to, a collection of 
information unless it displays a currently valid OMB control number. 
OMB has approved the information collection requirements contained in 
this part under control number 3150-0008.
    (b) The approved information collection requirements contained in 
this part appear in Sec.Sec. 71.5, 71.7, 71.9, 71.12, 71.17, 71.19, 
71.20, 71.22, 71.23, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41, 
71.47, 71.85, 71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 
71.105, 71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 
71.123, 71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, and 
Appendix A, Paragraph II.


Sec. 71.7  Completeness and accuracy of information.

    (a) Information provided to the Commission by a licensee, 
certificate holder, or an applicant for a license or CoC; or 
information required by statute or by the Commission's regulations, 
orders, license or CoC conditions, to be maintained by the licensee or 
certificate holder, must be complete and accurate in all material 
respects.
    (b) Each licensee, certificate holder, or applicant for a license 
or CoC must notify the Commission of information identified by the 
licensee, certificate holder, or applicant for a license or CoC as 
having, for the regulated activity, a significant implication for 
public health and safety or common defense and security. A licensee, 
certificate holder, or an applicant for a license or CoC violates this 
paragraph only if the licensee, certificate holder, or applicant for a 
license or CoC fails to notify the Commission of information that the 
licensee, certificate holder, or applicant for a license or CoC has 
identified as having a significant implication for public health and 
safety or common defense and security. Notification must be provided to 
the Administrator of the appropriate Regional Office within 2 working 
days of identifying the information. This requirement is not applicable 
to information which is already required to be provided to the 
Commission by other reporting or updating requirements.


Sec. 71.8  Deliberate misconduct.

    (a) This section applies to any--
    (1) Licensee;
    (2) Certificate holder;
    (3) Quality assurance program approval holder;
    (4) Applicant for a license, certificate, or quality assurance 
program approval;
    (5) Contractor (including a supplier or consultant) or 
subcontractor, to any person identified in paragraph (a)(4) of this 
section; or
    (6) Employees of any person identified in paragraphs (a)(1) through 
(a)(5) of this section.
    (b) A person identified in paragraph (a) of this section who 
knowingly provides to any entity, listed in paragraphs (a)(1) through 
(a)(5) of this section, any components, materials, or other goods or 
services that relate to a licensee's, certificate holder's, quality 
assurance program approval holder's, or applicant's activities subject 
to this part may not:
    (1) Engage in deliberate misconduct that causes or would have 
caused, if not detected, a licensee, certificate holder, quality 
assurance program approval holder, or any applicant to be in violation 
of any rule, regulation, or order; or any term, condition or limitation 
of any license, certificate, or approval issued by the Commission; or
    (2) Deliberately submit to the NRC, a licensee, a certificate 
holder, quality assurance program approval holder, an applicant for a 
license, certificate or quality assurance program approval, or a 
licensee's, applicant's, certificate holder's, or quality assurance 
program approval holder's contractor or subcontractor, information that 
the person submitting the information knows to be incomplete or 
inaccurate in some respect material to the NRC.
    (c) A person who violates paragraph (b)(1) or (b)(2) of this 
section may be subject to enforcement action in accordance with the 
procedures in 10 CFR part 2, subpart B.
    (d) For the purposes of paragraph (b)(1) of this section, 
deliberate misconduct by a person means an intentional act or omission 
that the person knows:
    (1) Would cause a licensee, certificate holder, quality assurance 
program approval holder, or applicant for a license, certificate, or 
quality assurance program approval to be in violation of any rule, 
regulation, or order; or any term, condition, or limitation of any 
license or certificate issued by the Commission; or
    (2) Constitutes a violation of a requirement, procedure, 
instruction, contract, purchase order, or policy of a licensee, 
certificate holder, quality assurance program approval holder, 
applicant, or the contractor or subcontractor of any of them.


Sec. 71.9  Employee protection.

    (a) Discrimination by a Commission licensee, certificate holder, an 
applicant for a Commission license or a CoC, or a contractor or 
subcontractor of any of these, against an employee for engaging in 
certain protected activities, is prohibited. Discrimination includes 
discharge and other actions that relate to compensation, terms, 
conditions, or privileges of employment. The protected activities are 
established in section 211 of the Energy Reorganization Act of 1974, as 
amended, and in general are related to the administration or 
enforcement of a requirement imposed under the Atomic Energy Act of 
1954, as amended, or the Energy Reorganization Act of 1974, as amended.
    (1) The protected activities include, but are not limited to:
    (i) Providing the Commission or his or her employer information 
about alleged violations of either of the statutes named in paragraph 
(a) of this section or possible violations of requirements imposed 
under either of those statutes;
    (ii) Refusing to engage in any practice made unlawful under either 
of the statutes named in paragraph (a) of this section or under these 
requirements if the employee has identified the alleged illegality to 
the employer;
    (iii) Requesting the Commission to institute action against his or 
her employer for the administration or enforcement of these 
requirements;
    (iv) Testifying in any Commission proceeding, or before Congress, 
or at any Federal or State proceeding regarding any provision (or 
proposed provision) of either of the statutes named in paragraph (a) of 
this section; and
    (v) Assisting or participating in, or is about to assist or 
participate in, these activities.
    (2) These activities are protected even if no formal proceeding is 
actually initiated as a result of the employee's assistance or 
participation.
    (3) This section has no application to any employee alleging 
discrimination prohibited by this section who, acting without direction 
from his or her employer (or the employer's agent),

[[Page 3791]]

deliberately causes a violation of any requirement of the Energy 
Reorganization Act of 1974, as amended, or the Atomic Energy Act of 
1954, as amended.
    (b) Any employee who believes that he or she has been discharged or 
otherwise discriminated against by any person for engaging in protected 
activities specified in paragraph (a)(1) of this section may seek a 
remedy for the discharge or discrimination through an administrative 
proceeding in the Department of Labor. The administrative proceeding 
must be initiated within 180 days after an alleged violation occurs. 
The employee may do this by filing a complaint alleging the violation 
with the Department of Labor, Employment Standards Administration, Wage 
and Hour Division. The Department of Labor may order reinstatement, 
back pay, and compensatory damages.
    (c) A violation of paragraph (a), (e), or (f) of this section by a 
Commission licensee, certificate holder, applicant for a Commission 
license or a CoC, or a contractor or subcontractor of any of these may 
be grounds for:
    (1) Denial, revocation, or suspension of the license or the CoC;
    (2) Imposition of a civil penalty on the licensee or applicant; or
    (3) Other enforcement action.
    (d) Actions taken by an employer, or others, which adversely affect 
an employee may be predicated upon nondiscriminatory grounds. The 
prohibition applies when the adverse action occurs because the employee 
has engaged in protected activities. An employee's engagement in 
protected activities does not automatically render him or her immune 
from discharge or discipline for legitimate reasons or from adverse 
action dictated by nonprohibited considerations.
    (e)(1) Each licensee, certificate holder, and applicant for a 
license or CoC must prominently post the current revision of NRC Form 
3, ``Notice to Employees,'' referenced in Sec. 19.11(c) of this 
chapter. This form must be posted at locations sufficient to permit 
employees protected by this section to observe a copy on the way to or 
from their place of work. The premises must be posted not later than 30 
days after an application is docketed and remain posted while the 
application is pending before the Commission, during the term of the 
license or CoC, and for 30 days following license or CoC termination.
    (2) Copies of NRC Form 3 may be obtained by writing to the Regional 
Administrator of the appropriate U.S. Nuclear Regulatory Commission 
Regional Office listed in Appendix D to part 20 of this chapter or by 
calling the NRC Publishing Services Branch at 301-415-5877.
    (f) No agreement affecting the compensation, terms, conditions, or 
privileges of employment, including an agreement to settle a complaint 
filed by an employee with the Department of Labor pursuant to section 
211 of the Energy Reorganization Act of 1974, as amended, may contain 
any provision which would prohibit, restrict, or otherwise discourage 
an employee from participating in a protected activity as defined in 
paragraph (a)(1) of this section including, but not limited to, 
providing information to the NRC or to his or her employer on potential 
violations or other matters within NRC's regulatory responsibilities.


Sec. 71.10  Public inspection of application.

    Applications for approval of a package design under this part, 
which are submitted to the Commission, may be made available for public 
inspection, in accordance with provisions of parts 2 and 9 of this 
chapter. This includes an application to amend or revise an existing 
package design, any associated documents and drawings submitted with 
the application, and any responses to NRC requests for additional 
information.


Sec. 71.11  [Reserved]

Subpart B--Exemptions


Sec. 71.12  Specific exemptions.

    On application of any interested person or on its own initiative, 
the Commission may grant any exemption from the requirements of the 
regulations in this part that it determines is authorized by law and 
will not endanger life or property nor the common defense and security.


Sec. 71.13  Exemption of physicians.

    Any physician licensed by a State to dispense drugs in the practice 
of medicine is exempt from Sec. 71.5 with respect to transport by the 
physician of licensed material for use in the practice of medicine. 
However, any physician operating under this exemption must be licensed 
under 10 CFR part 35 or the equivalent Agreement State regulations.


Sec. 71.14  Exemption for low-level materials.

    (a) A licensee is exempt from all the requirements of this part 
with respect to shipment or carriage of the following low-level 
materials:
    (1) Natural material and ores containing naturally occurring 
radionuclides that are not intended to be processed for use of these 
radionuclides, provided the activity concentration of the material does 
not exceed 10 times the values specified in Appendix A, Table A-2, of 
this part.
    (2) Materials for which the activity concentration is not greater 
than the activity concentration values specified in Appendix A, Table 
A-2 of this part, or for which the consignment activity is not greater 
than the limit for an exempt consignment found in Appendix A, Table A-
2, of this part.
    (b) A licensee is exempt from all the requirements of this part, 
other than Sec.Sec. 71.5 and 71.88, with respect to shipment or 
carriage of the following packages, provided the packages do not 
contain any fissile material, or the material is exempt from 
classification as fissile material under Sec. 71.15:
    (1) A package that contains no more than a Type A quantity of 
radioactive material;
    (2) A package transported within the United States that contains no 
more than 0.74 TBq (20 Ci) of special form plutonium-244; or
    (3) The package contains only LSA or SCO radioactive material, 
provided--
    (i) That the LSA or SCO material has an external radiation dose of 
less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the 
unshielded material; or
    (ii) That the package contains only LSA-I or SCO-I material.


Sec. 71.15  Exemption from classification as fissile material.

    Fissile material meeting the requirements of at least one of the 
paragraphs (a) through (f) of this section are exempt from 
classification as fissile material and from the fissile material 
package standards of Sec.Sec. 71.55 and 71.59, but are subject to all 
other requirements of this part, except as noted.
    (a) Individual package containing 2 grams or less fissile material.
    (b) Individual or bulk packaging containing 15 grams or less of 
fissile material provided the package has at least 200 grams of solid 
nonfissile material for every gram of fissile material. Lead, 
beryllium, graphite, and hydrogenous material enriched in deuterium may 
be present in the package but must not be included in determining the 
required mass for solid nonfissile material.
    (c)(1) Low concentrations of solid fissile material commingled with 
solid nonfissile material, provided that:
    (i) There is at least 2000 grams of solid nonfissile material for 
every gram of fissile material, and
    (ii) There is no more than 180 grams of fissile material 
distributed within 360 kg of contiguous nonfissile material.
    (2) Lead, beryllium, graphite, and hydrogenous material enriched in

[[Page 3792]]

deuterium may be present in the package but must not be included in 
determining the required mass of solid nonfissile material.
    (d) Uranium enriched in uranium-235 to a maximum of 1 percent by 
weight, and with total plutonium and uranium-233 content of up to 1 
percent of the mass of uranium-235, provided that the mass of any 
beryllium, graphite, and hydrogenous material enriched in deuterium 
constitutes less than 5 percent of the uranium mass.
    (e) Liquid solutions of uranyl nitrate enriched in uranium-235 to a 
maximum of 2 percent by mass, with a total plutonium and uranium-233 
content not exceeding 0.002 percent of the mass of uranium, and with a 
minimum nitrogen to uranium atomic ratio (N/U) of 2. The material must 
be contained in at least a DOT Type A package.
    (f) Packages containing, individually, a total plutonium mass of 
not more than 1000 grams, of which not more than 20 percent by mass may 
consist of plutonium-239, plutonium-241, or any combination of these 
radionuclides.


Sec. 71.16  [Reserved]

Subpart C--General Licenses


Sec. 71.17  General license: NRC-approved package.

    (a) A general license is issued to any licensee of the Commission 
to transport, or to deliver to a carrier for transport, licensed 
material in a package for which a license, certificate of compliance 
(CoC), or other approval has been issued by the NRC.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the CoC, or other approval of the package, and 
has the drawings and other documents referenced in the approval 
relating to the use and maintenance of the packaging and to the actions 
to be taken before shipment;
    (2) Complies with the terms and conditions of the license, 
certificate, or other approval, as applicable, and the applicable 
requirements of subparts A, G, and H of this part; and
    (3) Before the licensee's first use of the package, submits in 
writing to: ATTN: Document Control Desk, Director, Spent Fuel Project 
Office, Office of Nuclear Material Safety and Safeguards, using an 
appropriate method listed in Sec. 71.1(a), the licensee's name and 
license number and the package identification number specified in the 
package approval.
    (d) This general license applies only when the package approval 
authorizes use of the package under this general license.
    (e) For a Type B or fissile material package, the design of which 
was approved by NRC before April 1, 1996, the general license is 
subject to the additional restrictions of Sec. 71.19.


Sec. 71.18  [Reserved]


Sec. 71.19  Previously approved package.

    (a) A Type B package previously approved by NRC, but not designated 
as B(U), B(M), B(U)F, or B(M)F in the identification number of the NRC 
CoC, or Type AF packages approved by the NRC prior to September 6, 
1983, may be used under the general license of Sec. 71.17 with the 
following additional conditions:
    (1) Fabrication of the packaging was satisfactorily completed by 
August 31, 1986, as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A serial number that uniquely identifies each packaging which 
conforms to the approved design is assigned to, and legibly and durably 
marked on, the outside of each packaging; and
    (3) Paragraph (a) of this section expires (insert date 4 years 
after the effective date of this final rule). The effective date of 
this final rule is October 1, 2004.
    (b) A Type B(U) package, a Type B(M) package, or a fissile material 
package, previously approved by the NRC but without the designation ``-
85'' in the identification number of the NRC CoC, may be used under the 
general license of Sec. 71.17 with the following additional conditions:
    (1) Fabrication of the package is satisfactorily completed by April 
1, 1999, as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A package used for a shipment to a location outside the United 
States is subject to multilateral approval as defined in DOT 
regulations at 49 CFR 173.403; and
    (3) A serial number which uniquely identifies each packaging which 
conforms to the approved design is assigned to and legibly and durably 
marked on the outside of each packaging.
    (c) A Type B(U) package, a Type B(M) package, or a fissile material 
package previously approved by the NRC with the designation ``-85'' in 
the identification number of the NRC CoC, may be used under the general 
license of Sec. 71.17 with the following additional conditions:
    (1) Fabrication of the package must be satisfactorily completed by 
December 31, 2006, as demonstrated by application of its model number 
in accordance with Sec. 71.85(c); and
    (2) After December 31, 2003, a package used for a shipment to a 
location outside the United States is subject to multilateral approval 
as defined in DOT regulations at 49 CFR 173.403.
    (d) NRC will approve modifications to the design and authorized 
contents of a Type B package, or a fissile material package, previously 
approved by NRC, provided--
    (1) The modifications of a Type B package are not significant with 
respect to the design, operating characteristics, or safe performance 
of the containment system, when the package is subjected to the tests 
specified in Sec.Sec. 71.71 and 71.73;
    (2) The modifications of a fissile material package are not 
significant, with respect to the prevention of criticality, when the 
package is subjected to the tests specified in Sec.Sec. 71.71 and 
71.73; and
    (3) The modifications to the package satisfy the requirements of 
this part.
    (e) NRC will revise the package identification number to designate 
previously approved package designs as B, BF, AF, B(U), B(M), B(U)F, 
B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate, 
and with the identification number suffix ``-96'' after receipt of an 
application demonstrating that the design meets the requirements of 
this part.


Sec. 71.20  General license: DOT specification container.

    (a) A general license is issued to any licensee of the Commission 
to transport, or to deliver to a carrier for transport, licensed 
material in a specification container for fissile material or for a 
Type B quantity of radioactive material as specified in DOT regulations 
at 49 CFR parts 173 and 178.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the specification; and
    (2) Complies with the terms and conditions of the specification and 
the applicable requirements of subparts A, G, and H of this part.
    (d) This general license is subject to the limitation that the 
specification

[[Page 3793]]

container may not be used for a shipment to a location outside the 
United States, except by multilateral approval, as defined in DOT 
regulations at 49 CFR 173.403.
    (e) This section expires October 1, 2008.


Sec. 71.21  General license: Use of foreign approved package.

    (a) A general license is issued to any licensee of the Commission 
to transport, or to deliver to a carrier for transport, licensed 
material in a package, the design of which has been approved in a 
foreign national competent authority certificate, that has been 
revalidated by DOT as meeting the applicable requirements of 49 CFR 
171.12.
    (b) Except as otherwise provided in this section, the general 
license applies only to a licensee who has a quality assurance program 
approved by the Commission as satisfying the applicable provisions of 
subpart H of this part.
    (c) This general license applies only to shipments made to or from 
locations outside the United States.
    (d) This general license applies only to a licensee who--
    (1) Has a copy of the applicable certificate, the revalidation, and 
the drawings and other documents referenced in the certificate, 
relating to the use and maintenance of the packaging and to the actions 
to be taken before shipment; and
    (2) Complies with the terms and conditions of the certificate and 
revalidation, and with the applicable requirements of subparts A, G, 
and H of this part. With respect to the quality assurance provisions of 
subpart H of this part, the licensee is exempt from design, 
construction, and fabrication considerations.


Sec. 71.22  General license: Fissile material.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, if the material is shipped in accordance with 
this section. The fissile material need not be contained in a package 
which meets the standards of subparts E and F of this part; however, 
the material must be contained in a Type A package. The Type A package 
must also meet the DOT requirements of 49 CFR 173.417(a).
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) The general license applies only when a package's contents:
    (1) Contain less than a Type A quantity of fissile material; and
    (2) Contain less than 500 total grams of beryllium, graphite, or 
hydrogenous material enriched in deuterium.
    (d) The general license applies only to packages containing fissile 
material that are labeled with a CSI which:
    (1) Has been determined in accordance with paragraph (e) of this 
section;
    (2) Has a value less than or equal to 10; and
    (3) For a shipment of multiple packages containing fissile 
material, the sum of the CSIs must be less than or equal to 50 (for 
shipment on a nonexclusive use conveyance) and less than or equal to 
100 (for shipment on an exclusive use conveyance).
    (e)(1) The value for the CSI must be greater than or equal to the 
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TR26JA04.012

    (2) The calculated CSI must be rounded up to the first decimal 
place;
    (3) The values of X, Y, and Z used in the CSI equation must be 
taken from Tables 71-1 or 71-2, as appropriate;
    (4) If Table 71-2 is used to obtain the value of X, then the values 
for the terms in the equation for uranium-233 and plutonium must be 
assumed to be zero; and
    (5) Table 71-1 values for X, Y, and Z must be used to determine the 
CSI if:
    (i) Uranium-233 is present in the package;
    (ii) The mass of plutonium exceeds 1 percent of the mass of 
uranium-235;
    (iii) The uranium is of unknown uranium-235 enrichment or greater 
than 24 weight percent enrichment; or
    (iv) Substances having a moderating effectiveness (i.e., an average 
hydrogen density greater than H2O) (e.g., certain 
hydrocarbon oils or plastics) are present in any form, except as 
polyethylene used for packing or wrapping.

 Table 71-1.--Mass Limits for General License Packages Containing Mixed
 Quantities of Fissile Material or Uranium-235 of Unknown Enrichment per
                              Sec. 71.22(e)
------------------------------------------------------------------------
                                   Fissile material    Fissile material
                                    mass mixed with     mass mixed with
                                      moderating          moderating
                                   substances having   substances having
        Fissile material              an average          an average
                                   hydrogen density    hydrogen density
                                  less than or equal   greater than H2Oa
                                    to H2O (grams)          (grams)
------------------------------------------------------------------------
\235\ U (X).....................                 60                  38
\233\ U (Y).....................                 43                  27
\239\ Pu or \241\ Pu (Z)........                 37                  24
------------------------------------------------------------------------
a When mixtures of moderating substances are present, the lower mass
  limits shall be used if more than 15 percent of the moderating
  substance has an average hydrogen density greater than H2O.


[[Page 3794]]


Table 71-2.--Mass Limits for General License Packages Containing Uranium-
                235 of Known Enrichment per Sec. 71.22(e)
------------------------------------------------------------------------
                                                               Fissile
                                                              material
    Uranium enrichment in weight percent of \235\ U not        mass of
                         exceeding                           \235\ U (X)
                                                               (grams)
------------------------------------------------------------------------
24........................................................           60
20........................................................           63
15........................................................           67
11........................................................           72
10........................................................           76
9.5.......................................................           78
9.........................................................           81
8.5.......................................................           82
8.........................................................           85
7.5.......................................................           88
7.........................................................           90
6.5.......................................................           93
6.........................................................           97
5.5.......................................................          102
5.........................................................          108
4.5.......................................................          114
4.........................................................          120
3.5.......................................................          132
3.........................................................          150
2.5.......................................................          180
2.........................................................          246
1.5.......................................................          408
1.35......................................................          480
1.........................................................        1,020
0.92......................................................        1,800
------------------------------------------------------------------------

Sec. 71.23  General license: Plutonium-beryllium special form material.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material in the form of plutonium-beryllium (Pu-
Be) special form sealed sources, or to deliver Pu-Be sealed sources to 
a carrier for transport, if the material is shipped in accordance with 
this section. This material need not be contained in a package which 
meets the standards of subparts E and F of this part; however, the 
material must be contained in a Type A package. The Type A package must 
also meet the DOT requirements of 49 CFR 173.417(a).
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) The general license applies only when a package's contents:
    (1) Contain less than a Type A quantity of material; and
    (2) Contain less than 1000 g of plutonium, provided that: 
plutonium-239, plutonium-241, or any combination of these 
radionuclides, constitutes less than 240 g of the total quantity of 
plutonium in the package.
    (d) The general license applies only to packages labeled with a CSI 
which:
    (1) Has been determined in accordance with paragraph (e) of this 
section;
    (2) Has a value less than or equal to 100; and
    (3) For a shipment of multiple packages containing Pu-Be sealed 
sources, the sum of the CSIs must be less than or equal to 50 (for 
shipment on a nonexclusive use conveyance) and less than or equal to 
100 (for shipment on an exclusive use conveyance).
    (e)(1) The value for the CSI must be greater than or equal to the 
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TR26JA04.013

    (2) The calculated CSI must be rounded up to the first decimal 
place.


Sec. 71.24  [Reserved]


Sec. 71.25  [Reserved]

0
3. In Sec. 71.41, paragraph (a) is revised, and a new paragraph (d) is 
added to read as follows:


Sec. 71.41  Demonstration of compliance.

    (a) The effects on a package of the tests specified in Sec. 71.71 
(``Normal conditions of transport''), and the tests specified in Sec. 
71.73 (``Hypothetical accident conditions''), and Sec. 71.61 (``Special 
requirements for Type B packages containing more than 105 
A2''), must be evaluated by subjecting a specimen or scale 
model to a specific test, or by another method of demonstration 
acceptable to the Commission, as appropriate for the particular feature 
being considered.
* * * * *
    (d) Packages for which compliance with the other provisions of 
these regulations is impracticable shall not be transported except 
under special package authorization. Provided the applicant 
demonstrates that compliance with the other provisions of the 
regulations is impracticable and that the requisite standards of safety 
established by these regulations have been demonstrated through means 
alternative to the other provisions, a special package authorization 
may be approved for one-time shipments. The applicant shall demonstrate 
that the overall level of safety in transport for these shipments is at 
least equivalent to that which would be provided if all the applicable 
requirements had been met.

0
4. In Sec. 71.51, the introductory text of paragraph (a) is revised, 
and a new paragraph (d) is added to read as follows:


Sec. 71.51  Additional requirements for Type B packages.

    (a) A Type B package, in addition to satisfying the requirements of 
Sec.Sec. 71.41 through 71.47, must be designed, constructed, and 
prepared for shipment so that under the tests specified in:
* * * * *
    (d) For packages which contain radioactive contents with activity 
greater than 105 A2, the requirements of Sec. 71.61 must be 
met.


Sec. 71.53  [Reserved)

0
5. Section 71.53 is removed and reserved.
0
6. In Sec. 71.55, the introductory text of paragraph (b) is revised, 
and new paragraphs (f) and (g) are added to read as follows:


Sec. 71.55  General requirements for fissile material packages.

* * * * *

[[Page 3795]]

    (b) Except as provided in paragraph (c) or (g) of this section, a 
package used for the shipment of fissile material must be so designed 
and constructed and its contents so limited that it would be 
subcritical if water were to leak into the containment system, or 
liquid contents were to leak out of the containment system so that, 
under the following conditions, maximum reactivity of the fissile 
material would be attained:
* * * * *
    (f) For fissile material package designs to be transported by air:
    (1) The package must be designed and constructed, and its contents 
limited so that it would be subcritical, assuming reflection by 20 cm 
(7.9 in) of water but no water inleakage, when subjected to sequential 
application of:
    (i) The free drop test in Sec. 71.73(c)(1);
    (ii) The crush test in Sec. 71.73(c)(2);
    (iii) A puncture test, for packages of 250 kg or more, consisting 
of a free drop of the specimen through a distance of 3 m (120 in) in a 
position for which maximum damage is expected at the conclusion of the 
test sequence, onto the upper end of a solid, vertical, cylindrical, 
mild steel probe mounted on an essentially unyielding, horizontal 
surface. The probe must be 20 cm (7.9 in) in diameter, with the 
striking end forming the frustum of a right circular cone with the 
dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded 
to a radius of not more than 6 mm (0.25 in). For packages less than 250 
kg, the puncture test must be the same, except that a 250 kg probe must 
be dropped onto the specimen which must be placed on the surface; and
    (iv) The thermal test in Sec. 71.73(c)(4), except that the duration 
of the test must be 60 minutes.
    (2) The package must be designed and constructed, and its contents 
limited, so that it would be subcritical, assuming reflection by 20 cm 
(7.9 in) of water but no water inleakage, when subjected to an impact 
on an unyielding surface at a velocity of 90 m/s normal to the surface, 
at such orientation so as to result in maximum damage. A separate, 
undamaged specimen can be used for this evaluation.
    (3) Allowance may not be made for the special design features in 
paragraph (c) of this section, unless water leakage into or out of void 
spaces is prevented following application of the tests in paragraphs 
(f)(1) and (f)(2) of this section, and subsequent application of the 
immersion test in Sec. 71.73(c)(5).
    (g) Packages containing uranium hexafluoride only are excepted from 
the requirements of paragraph (b) of this section provided that:
    (1) Following the tests specified in Sec. 71.73 (``Hypothetical 
accident conditions''), there is no physical contact between the valve 
body and any other component of the packaging, other than at its 
original point of attachment, and the valve remains leak tight;
    (2) There is an adequate quality control in the manufacture, 
maintenance, and repair of packagings;
    (3) Each package is tested to demonstrate closure before each 
shipment; and
    (4) The uranium is enriched to not more than 5 weight percent 
uranium-235.

0
7. In Sec. 71.59, paragraphs (b) and (c) are revised to read as 
follows:


Sec. 71.59  Standards for arrays of fissile material packages.

* * * * *
    (b) The CSI must be determined by dividing the number 50 by the 
value of ``N'' derived using the procedures specified in paragraph (a) 
of this section. The value of the CSI may be zero provided that an 
unlimited number of packages are subcritical, such that the value of 
``N'' is effectively equal to infinity under the procedures specified 
in paragraph (a) of this section. Any CSI greater than zero must be 
rounded up to the first decimal place.
    (c) For a fissile material package which is assigned a CSI value--
    (1) Less than or equal to 50, that package may be shipped by a 
carrier in a nonexclusive use conveyance, provided the sum of the CSIs 
is limited to less than or equal to 50.
    (2) Less than or equal to 50, that package may be shipped by a 
carrier in an exclusive use conveyance, provided the sum of the CSIs is 
limited to less than or equal to 100.
    (3) Greater than 50, that package must be shipped by a carrier in 
an exclusive use conveyance, provided the sum of the CSIs is limited to 
less than or equal to 100.

0
8. Section 71.61 is revised to read as follows:


Sec. 71.61  Special requirements for Type B packages containing more 
than 10\5\A[bdi2].

    A Type B package containing more than 10\5\A2 must be 
designed so that its undamaged containment system can withstand an 
external water pressure of 2 MPa (290 psi) for a period of not less 
than 1 hour without collapse, buckling, or inleakage of water.

0
9. Section 71.63 is revised to read as follows:


Sec. 71.63  Special requirement for plutonium shipments.

    Shipments containing plutonium must be made with the contents in 
solid form, if the contents contain greater than 0.74 TBq (20 Ci) of 
plutonium.

0
10. In Sec. 71.73, paragraph (c)(2) is revised to read as follows:


Sec. 71.73  Hypothetical accident conditions.

* * * * *
    (c) * * *
    (2) Crush. Subjection of the specimen to a dynamic crush test by 
positioning the specimen on a flat, essentially unyielding horizontal 
surface so as to suffer maximum damage by the drop of a 500-kg (1100-
lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a 
solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a 
horizontal attitude. The crush test is required only when the specimen 
has a mass not greater than 500 kg (1100 lb), an overall density not 
greater than 1000 kg/m \3\ (62.4 lb/ft \3\) based on external 
dimension, and radioactive contents greater than 1000 A2 not 
as special form radioactive material. For packages containing fissile 
material, the radioactive contents greater than 1000 A2 
criterion does not apply.
* * * * *

0
11. In Sec. 71.88, paragraph (a)(2) is revised to read as follows:


Sec. 71.88  Air transport of plutonium.

    (a) * * *
    (2) The plutonium is contained in a material in which the specific 
activity is less than or equal to the activity concentration values for 
plutonium specified in Appendix A, Table A-2, of this part, and in 
which the radioactivity is essentially uniformly distributed; or
* * * * *

0
12. In Sec. 71.91, paragraphs (b) and (c) are revised, and a new 
paragraph (d) is added to read as follows:


Sec. 71.91  Records.

* * * * *
    (b) Each certificate holder shall maintain, for a period of 3 years 
after the life of the packaging to which they apply, records 
identifying the packaging by model number, serial number, and date of 
manufacture.
    (c) The licensee, certificate holder, and an applicant for a CoC, 
shall make available to the Commission for inspection, upon reasonable 
notice, all records required by this part. Records are only valid if 
stamped, initialed, or signed and dated by authorized personnel, or 
otherwise authenticated.
    (d) The licensee, certificate holder, and an applicant for a CoC 
shall maintain sufficient written records to furnish evidence of the 
quality of packaging. The records to be maintained include results of 
the determinations required by Sec. 71.85; design, fabrication,

[[Page 3796]]

and assembly records; results of reviews, inspections, tests, and 
audits; results of monitoring work performance and materials analyses; 
and results of maintenance, modification, and repair activities. 
Inspection, test, and audit records must identify the inspector or data 
recorder, the type of observation, the results, the acceptability, and 
the action taken in connection with any deficiencies noted. These 
records must be retained for 3 years after the life of the packaging to 
which they apply.

0
13. Section 71.93 is revised to read as follows:


Sec. 71.93  Inspection and tests.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
permit the Commission, at all reasonable times, to inspect the licensed 
material, packaging, premises, and facilities in which the licensed 
material or packaging is used, provided, constructed, fabricated, 
tested, stored, or shipped.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
perform, and permit the Commission to perform, any tests the Commission 
deems necessary or appropriate for the administration of the 
regulations in this chapter.
    (c) The certificate holder and applicant for a CoC shall notify the 
NRC, in accordance with Sec. 71.1, 45 days in advance of starting 
fabrication of the first packaging under a CoC. This paragraph applies 
to any packaging used for the shipment of licensed material which has 
either--
    (1) A decay heat load in excess of 5 kW; or
    (2) A maximum normal operating pressure in excess of 103 kPa (15 
lbf/in \2\) gauge.

0
14. Section 71.95 is revised to read as follows:


Sec. 71.95  Reports.

    (a) The licensee, after requesting the certificate holder's input, 
shall submit a written report to the Commission of--
    (1) Instances in which there is a significant reduction in the 
effectiveness of any NRC-approved Type B or Type AF packaging during 
use; or
    (2) Details of any defects with safety significance in any NRC-
approved Type B or fissile material packaging, after first use.
    (3) Instances in which the conditions of approval in the 
Certificate of Compliance were not observed in making a shipment.
    (b) The licensee shall submit a written report to the Commission of 
instances in which the conditions in the certificate of compliance were 
not followed during a shipment.
    (c) Each licensee shall submit, in accordance with Sec. 71.1, a 
written report required by paragraph (a) or (b) of this section within 
60 days of the event or discovery of the event. The licensee shall also 
provide a copy of each report submitted to the NRC to the applicable 
certificate holder. Written reports prepared under other regulations 
may be submitted to fulfill this requirement if the reports contain all 
the necessary information, and the appropriate distribution is made. 
Using an appropriate method listed in Sec. 71.1(a), the licensee shall 
report to: ATTN: Document Control Desk, Director, Spent Fuel Project 
Office, Office of Nuclear Material Safety and Safeguards. These written 
reports must include the following:
    (1) A brief abstract describing the major occurrences during the 
event, including all component or system failures that contributed to 
the event and significant corrective action taken or planned to prevent 
recurrence.
    (2) A clear, specific, narrative description of the event that 
occurred so that knowledgeable readers conversant with the requirements 
of part 71, but not familiar with the design of the packaging, can 
understand the complete event. The narrative description must include 
the following specific information as appropriate for the particular 
event.
    (i) Status of components or systems that were inoperable at the 
start of the event and that contributed to the event;
    (ii) Dates and approximate times of occurrences;
    (iii) The cause of each component or system failure or personnel 
error, if known;
    (iv) The failure mode, mechanism, and effect of each failed 
component, if known;
    (v) A list of systems or secondary functions that were also 
affected for failures of components with multiple functions;
    (vi) The method of discovery of each component or system failure or 
procedural error;
    (vii) For each human performance-related root cause, a discussion 
of the cause(s) and circumstances;
    (viii) The manufacturer and model number (or other identification) 
of each component that failed during the event; and
    (ix) For events occurring during use of a packaging, the quantities 
and chemical and physical form(s) of the package contents.
    (3) An assessment of the safety consequences and implications of 
the event. This assessment must include the availability of other 
systems or components that could have performed the same function as 
the components and systems that failed during the event.
    (4) A description of any corrective actions planned as a result of 
the event, including the means employed to repair any defects, and 
actions taken to reduce the probability of similar events occurring in 
the future.
    (5) Reference to any previous similar events involving the same 
packaging that are known to the licensee or certificate holder.
    (6) The name and telephone number of a person within the licensee's 
organization who is knowledgeable about the event and can provide 
additional information.
    (7) The extent of exposure of individuals to radiation or to 
radioactive materials without identification of individuals by name.
    (d) Report legibility. The reports submitted by licensees and/or 
certificate holders under this section must be of sufficient quality to 
permit reproduction and micrographic processing.

0
15. In Sec. 71.100, paragraph (b) is revised to read as follows:


Sec. 71.100  Criminal penalties.

* * * * *
    (b) The regulations in part 71 that are not issued under sections 
161b, 161i, or 161o for the purposes of section 223 are as follows: 
Sec.Sec. 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35, 
71.37, 71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55, 
71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 71.100.

0
16. Subpart H to part 71 is revised to read as follows:
Subpart H--Quality Assurance
Sec.
71.101 Quality assurance requirements.
71.103 Quality assurance organization.
71.105 Quality assurance program.
71.107 Package design control.
71.109 Procurement document control.
71.111 Instructions, procedures, and drawings.
71.113 Document control.
71.115 Control of purchased material, equipment, and services.
71.117 Identification and control of materials, parts, and 
components.
71.119 Control of special processes.
71.121 Internal inspection.
71.123 Test control.
71.125 Control of measuring and test equipment.
71.127 Handling, storage, and shipping control.
71.129 Inspection, test, and operating status.
71.131 Nonconforming materials, parts, or components.
71.133 Corrective action.
71.135 Quality assurance records.
71.137 Audits.

[[Page 3797]]

Subpart H--Quality Assurance


Sec. 71.101  Quality assurance requirements.

    (a) Purpose. This subpart describes quality assurance requirements 
applying to design, purchase, fabrication, handling, shipping, storing, 
cleaning, assembly, inspection, testing, operation, maintenance, 
repair, and modification of components of packaging that are important 
to safety. As used in this subpart, ``quality assurance'' comprises all 
those planned and systematic actions necessary to provide adequate 
confidence that a system or component will perform satisfactorily in 
service. Quality assurance includes quality control, which comprises 
those quality assurance actions related to control of the physical 
characteristics and quality of the material or component to 
predetermined requirements. The licensee, certificate holder, and 
applicant for a CoC are responsible for the quality assurance 
requirements as they apply to design, fabrication, testing, and 
modification of packaging. Each licensee is responsible for the quality 
assurance provision which applies to its use of a packaging for the 
shipment of licensed material subject to this subpart.
    (b) Establishment of program. Each licensee, certificate holder, 
and applicant for a CoC shall establish, maintain, and execute a 
quality assurance program satisfying each of the applicable criteria of 
Sec.Sec. 71.101 through 71.137 and satisfying any specific provisions 
that are applicable to the licensee's activities including procurement 
of packaging. The licensee, certificate holder, and applicant for a CoC 
shall execute the applicable criteria in a graded approach to an extent 
that is commensurate with the quality assurance requirement's 
importance to safety.
    (c) Approval of program. (1) Before the use of any package for the 
shipment of licensed material subject to this subpart, each licensee 
shall obtain Commission approval of its quality assurance program. 
Using an appropriate method listed in Sec. 71.1(a), each licensee shall 
file a description of its quality assurance program, including a 
discussion of which requirements of this subpart are applicable and how 
they will be satisfied, by submitting the description to: ATTN: 
Document Control Desk, Director, Spent Fuel Project Office, Office of 
Nuclear Material Safety and Safeguards.
    (2) Before the fabrication, testing, or modification of any package 
for the shipment of licensed material subject to this subpart, each 
licensee, certificate holder, or applicant for a CoC shall obtain 
Commission approval of its quality assurance program. Each certificate 
holder or applicant for a CoC shall, in accordance with Sec. 71.1, file 
a description of its quality assurance program, including a discussion 
of which requirements of this subpart are applicable and how they will 
be satisfied.
    (d) Existing package designs. The provisions of this paragraph deal 
with packages that have been approved for use in accordance with this 
part before January 1, 1979, and which have been designed in accordance 
with the provisions of this part in effect at the time of application 
for package approval. Those packages will be accepted as having been 
designed in accordance with a quality assurance program that satisfies 
the provisions of paragraph (b) of this section.
    (e) Existing packages. The provisions of this paragraph deal with 
packages that have been approved for use in accordance with this part 
before January 1, 1979, have been at least partially fabricated before 
that date, and for which the fabrication is in accordance with the 
provisions of this part in effect at the time of application for 
approval of package design. These packages will be accepted as having 
been fabricated and assembled in accordance with a quality assurance 
program that satisfies the provisions of paragraph (b) of this section.
    (f) Previously approved programs. A Commission-approved quality 
assurance program that satisfies the applicable criteria of subpart H 
of this part, Appendix B of part 50 of this chapter, or subpart G of 
part 72 of this chapter, and that is established, maintained, and 
executed regarding transport packages, will be accepted as satisfying 
the requirements of paragraph (b) of this section. Before first use, 
the licensee, certificate holder, and applicant for a CoC shall notify 
the NRC, in accordance with Sec. 71.1, of its intent to apply its 
previously approved subpart H, Appendix B, or subpart G quality 
assurance program to transportation activities. The licensee, 
certificate holder, and applicant for a CoC shall identify the program 
by date of submittal to the Commission, Docket Number, and date of 
Commission approval.
    (g) Radiography containers. A program for transport container 
inspection and maintenance limited to radiographic exposure devices, 
source changers, or packages transporting these devices and meeting the 
requirements of Sec. 34.31(b) of this chapter or equivalent Agreement 
State requirement, is deemed to satisfy the requirements of Sec.Sec. 
71.17(b) and 71.101(b).


Sec. 71.103  Quality assurance organization.

    (a) The licensee,\2\ certificate holder, and applicant for a CoC 
shall be responsible for the establishment and execution of the quality 
assurance program. The licensee, certificate holder, and applicant for 
a CoC may delegate to others, such as contractors, agents, or 
consultants, the work of establishing and executing the quality 
assurance program, or any part of the quality assurance program, but 
shall retain responsibility for the program. These activities include 
performing the functions associated with attaining quality objectives 
and the quality assurance functions.
---------------------------------------------------------------------------

    \2\ While the term ``licensee'' is used in these criteria, the 
requirements are applicable to whatever design, fabrication, 
assembly, and testing of the package is accomplished with respect to 
a package before the time a package approval is issued.
---------------------------------------------------------------------------

    (b) The quality assurance functions are--
    (1) Assuring that an appropriate quality assurance program is 
established and effectively executed; and
    (2) Verifying, by procedures such as checking, auditing, and 
inspection, that activities affecting the functions that are important 
to safety have been correctly performed.
    (c) The persons and organizations performing quality assurance 
functions must have sufficient authority and organizational freedom to-
-
    (1) Identify quality problems;
    (2) Initiate, recommend, or provide solutions; and
    (3) Verify implementation of solutions.
    (d) The persons and organizations performing quality assurance 
functions shall report to a management level that assures that the 
required authority and organizational freedom, including sufficient 
independence from cost and schedule, when opposed to safety 
considerations, are provided.
    (e) Because of the many variables involved, such as the number of 
personnel, the type of activity being performed, and the location or 
locations where activities are performed, the organizational structure 
for executing the quality assurance program may take various forms, 
provided that the persons and organizations assigned the quality 
assurance functions have the required authority and organizational 
freedom.
    (f) Irrespective of the organizational structure, the individual(s) 
assigned the responsibility for assuring effective execution of any 
portion of the quality assurance program, at any location where 
activities subject to this section

[[Page 3798]]

are being performed, must have direct access to the levels of 
management necessary to perform this function.


Sec. 71.105  Quality assurance program.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish, at the earliest practicable time consistent with the 
schedule for accomplishing the activities, a quality assurance program 
that complies with the requirements of Sec.Sec. 71.101 through 71.137. 
The licensee, certificate holder, and applicant for a CoC shall 
document the quality assurance program by written procedures or 
instructions and shall carry out the program in accordance with those 
procedures throughout the period during which the packaging is used. 
The licensee, certificate holder, and applicant for a CoC shall 
identify the material and components to be covered by the quality 
assurance program, the major organizations participating in the 
program, and the designated functions of these organizations.
    (b) The licensee, certificate holder, and applicant for a CoC, 
through its quality assurance program, shall provide control over 
activities affecting the quality of the identified materials and 
components to an extent consistent with their importance to safety, and 
as necessary to assure conformance to the approved design of each 
individual package used for the shipment of radioactive material. The 
licensee, certificate holder, and applicant for a CoC shall assure that 
activities affecting quality are accomplished under suitably controlled 
conditions. Controlled conditions include the use of appropriate 
equipment; suitable environmental conditions for accomplishing the 
activity, such as adequate cleanliness; and assurance that all 
prerequisites for the given activity have been satisfied. The licensee, 
certificate holder, and applicant for a CoC shall take into account the 
need for special controls, processes, test equipment, tools, and skills 
to attain the required quality, and the need for verification of 
quality by inspection and test.
    (c) The licensee, certificate holder, and applicant for a CoC shall 
base the requirements and procedures of its quality assurance program 
on the following considerations concerning the complexity and proposed 
use of the package and its components:
    (1) The impact of malfunction or failure of the item to safety;
    (2) The design and fabrication complexity or uniqueness of the 
item;
    (3) The need for special controls and surveillance over processes 
and equipment;
    (4) The degree to which functional compliance can be demonstrated 
by inspection or test; and
    (5) The quality history and degree of standardization of the item.
    (d) The licensee, certificate holder, and applicant for a CoC shall 
provide for indoctrination and training of personnel performing 
activities affecting quality, as necessary to assure that suitable 
proficiency is achieved and maintained. The licensee, certificate 
holder, and applicant for a CoC shall review the status and adequacy of 
the quality assurance program at established intervals. Management of 
other organizations participating in the quality assurance program 
shall review regularly the status and adequacy of that part of the 
quality assurance program they are executing.


Sec. 71.107  Package design control.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that applicable regulatory requirements 
and the package design, as specified in the license or CoC for those 
materials and components to which this section applies, are correctly 
translated into specifications, drawings, procedures, and instructions. 
These measures must include provisions to assure that appropriate 
quality standards are specified and included in design documents and 
that deviations from standards are controlled. Measures must be 
established for the selection and review for suitability of application 
of materials, parts, equipment, and processes that are essential to the 
functions of the materials, parts, and components of the packaging that 
are important to safety.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
establish measures for the identification and control of design 
interfaces and for coordination among participating design 
organizations. These measures must include the establishment of written 
procedures, among participating design organizations, for the review, 
approval, release, distribution, and revision of documents involving 
design interfaces. The design control measures must provide for 
verifying or checking the adequacy of design, by methods such as design 
reviews, alternate or simplified calculational methods, or by a 
suitable testing program. For the verifying or checking process, the 
licensee shall designate individuals or groups other than those who 
were responsible for the original design, but who may be from the same 
organization. Where a test program is used to verify the adequacy of a 
specific design feature in lieu of other verifying or checking 
processes, the licensee, certificate holder, and applicant for a CoC 
shall include suitable qualification testing of a prototype or sample 
unit under the most adverse design conditions. The licensee, 
certificate holder, and applicant for a CoC shall apply design control 
measures to the following:

    (1) Criticality physics, radiation shielding, stress, thermal, 
hydraulic, and accident analyses;
    (2) Compatibility of materials;
    (3) Accessibility for inservice inspection, maintenance, and 
repair;
    (4) Features to facilitate decontamination; and
    (5) Delineation of acceptance criteria for inspections and 
tests.

    (c) The licensee, certificate holder, and applicant for a CoC shall 
subject design changes, including field changes, to design control 
measures commensurate with those applied to the original design. 
Changes in the conditions specified in the CoC require prior NRC 
approval.


Sec. 71.109  Procurement document control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that adequate quality is required in the 
documents for procurement of material, equipment, and services, whether 
purchased by the licensee, certificate holder, and applicant for a CoC 
or by its contractors or subcontractors. To the extent necessary, the 
licensee, certificate holder, and applicant for a CoC shall require 
contractors or subcontractors to provide a quality assurance program 
consistent with the applicable provisions of this part.


Sec. 71.111  Instructions, procedures, and drawings.

    The licensee, certificate holder, and applicant for a CoC shall 
prescribe activities affecting quality by documented instructions, 
procedures, or drawings of a type appropriate to the circumstances and 
shall require that these instructions, procedures, and drawings be 
followed. The instructions, procedures, and drawings must include 
appropriate quantitative or qualitative acceptance criteria for 
determining that important activities have been satisfactorily 
accomplished.


Sec. 71.113  Document control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control the issuance of documents such as 
instructions, procedures, and drawings, including

[[Page 3799]]

changes, that prescribe all activities affecting quality. These 
measures must assure that documents, including changes, are reviewed 
for adequacy, approved for release by authorized personnel, and 
distributed and used at the location where the prescribed activity is 
performed.


Sec. 71.115  Control of purchased material, equipment, and services.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that purchased material, equipment, and 
services, whether purchased directly or through contractors and 
subcontractors, conform to the procurement documents. These measures 
must include provisions, as appropriate, for source evaluation and 
selection, objective evidence of quality furnished by the contractor or 
subcontractor, inspection at the contractor or subcontractor source, 
and examination of products on delivery.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
have available documentary evidence that material and equipment conform 
to the procurement specifications before installation or use of the 
material and equipment. The licensee, certificate holder, and applicant 
for a CoC shall retain, or have available, this documentary evidence 
for the life of the package to which it applies. The licensee, 
certificate holder, and applicant for a CoC shall assure that the 
evidence is sufficient to identify the specific requirements met by the 
purchased material and equipment.
    (c) The licensee, certificate holder, and applicant for a CoC shall 
assess the effectiveness of the control of quality by contractors and 
subcontractors at intervals consistent with the importance, complexity, 
and quantity of the product or services.


Sec. 71.117  Identification and control of materials, parts, and 
components.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures for the identification and control of materials, 
parts, and components. These measures must assure that identification 
of the item is maintained by heat number, part number, or other 
appropriate means, either on the item or on records traceable to the 
item, as required throughout fabrication, installation, and use of the 
item. These identification and control measures must be designed to 
prevent the use of incorrect or defective materials, parts, and 
components.


Sec. 71.119  Control of special processes.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that special processes, including welding, 
heat treating, and nondestructive testing are controlled and 
accomplished by qualified personnel using qualified procedures in 
accordance with applicable codes, standards, specifications, criteria, 
and other special requirements.


Sec. 71.121  Internal inspection.

    The licensee, certificate holder, and applicant for a CoC shall 
establish and execute a program for inspection of activities affecting 
quality by or for the organization performing the activity, to verify 
conformance with the documented instructions, procedures, and drawings 
for accomplishing the activity. The inspection must be performed by 
individuals other than those who performed the activity being 
inspected. Examination, measurements, or tests of material or products 
processed must be performed for each work operation where necessary to 
assure quality. If direct inspection of processed material or products 
is not carried out, indirect control by monitoring processing methods, 
equipment, and personnel must be provided. Both inspection and process 
monitoring must be provided when quality control is inadequate without 
both. If mandatory inspection hold points, which require witnessing or 
inspecting by the licensee's designated representative and beyond which 
work should not proceed without the consent of its designated 
representative, are required, the specific hold points must be 
indicated in appropriate documents.


Sec. 71.123  Test control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish a test program to assure that all testing required to 
demonstrate that the packaging components will perform satisfactorily 
in service is identified and performed in accordance with written test 
procedures that incorporate the requirements of this part and the 
requirements and acceptance limits contained in the package approval. 
The test procedures must include provisions for assuring that all 
prerequisites for the given test are met, that adequate test 
instrumentation is available and used, and that the test is performed 
under suitable environmental conditions. The licensee, certificate 
holder, and applicant for a CoC shall document and evaluate the test 
results to assure that test requirements have been satisfied.


Sec. 71.125  Control of measuring and test equipment.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that tools, gauges, instruments, and other 
measuring and testing devices used in activities affecting quality are 
properly controlled, calibrated, and adjusted at specified times to 
maintain accuracy within necessary limits.


Sec. 71.127  Handling, storage, and shipping control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control, in accordance with instructions, the 
handling, storage, shipping, cleaning, and preservation of materials 
and equipment to be used in packaging to prevent damage or 
deterioration. When necessary for particular products, special 
protective environments, such as inert gas atmosphere, and specific 
moisture content and temperature levels must be specified and provided.


Sec. 71.129  Inspection, test, and operating status.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to indicate, by the use of markings such as stamps, 
tags, labels, routing cards, or other suitable means, the status of 
inspections and tests performed upon individual items of the packaging. 
These measures must provide for the identification of items that have 
satisfactorily passed required inspections and tests, where necessary 
to preclude inadvertent bypassing of the inspections and tests.
    (b) The licensee shall establish measures to identify the operating 
status of components of the packaging, such as tagging valves and 
switches, to prevent inadvertent operation.


Sec. 71.131  Nonconforming materials, parts, or components.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control materials, parts, or components that do 
not conform to the licensee's requirements to prevent their inadvertent 
use or installation. These measures must include, as appropriate, 
procedures for identification, documentation, segregation, disposition, 
and notification to affected organizations. Nonconforming items must be 
reviewed and accepted, rejected, repaired, or reworked in accordance 
with documented procedures.


Sec. 71.133  Corrective action.

    The licensee, certificate holder, and applicant for a CoC shall 
establish

[[Page 3800]]

measures to assure that conditions adverse to quality, such as 
deficiencies, deviations, defective material and equipment, and 
nonconformances, are promptly identified and corrected. In the case of 
a significant condition adverse to quality, the measures must assure 
that the cause of the condition is determined and corrective action 
taken to preclude repetition. The identification of the significant 
condition adverse to quality, the cause of the condition, and the 
corrective action taken must be documented and reported to appropriate 
levels of management.


Sec. 71.135  Quality assurance records.

    The licensee, certificate holder, and applicant for a CoC shall 
maintain sufficient written records to describe the activities 
affecting quality. The records must include the instructions, 
procedures, and drawings required by Sec. 71.111 to prescribe quality 
assurance activities and must include closely related specifications 
such as required qualifications of personnel, procedures, and 
equipment. The records must include the instructions or procedures 
which establish a records retention program that is consistent with 
applicable regulations and designates factors such as duration, 
location, and assigned responsibility. The licensee, certificate 
holder, and applicant for a CoC shall retain these records for 3 years 
beyond the date when the licensee, certificate holder, and applicant 
for a CoC last engage in the activity for which the quality assurance 
program was developed. If any portion of the written procedures or 
instructions is superseded, the licensee, certificate holder, and 
applicant for a CoC shall retain the superseded material for 3 years 
after it is superseded.


Sec. 71.137  Audits.

    The licensee, certificate holder, and applicant for a CoC shall 
carry out a comprehensive system of planned and periodic audits to 
verify compliance with all aspects of the quality assurance program and 
to determine the effectiveness of the program. The audits must be 
performed in accordance with written procedures or checklists by 
appropriately trained personnel not having direct responsibilities in 
the areas being audited. Audited results must be documented and 
reviewed by management having responsibility in the area audited. 
Followup action, including reaudit of deficient areas, must be taken 
where indicated.

0
17. Appendix A to part 71 is revised to read as follows:

Appendix A to Part 71--Determination of A1 and A2

    I. Values of A1 and A2 for individual radionuclides, which are 
the bases for many activity limits elsewhere in these regulations, 
are given in Table A-1. The curie (Ci) values specified are obtained 
by converting from the Terabecquerel (TBq) figure. The curie values 
are expressed to three significant figures to assure that the 
difference in the TBq and Ci quantities is one tenth of one percent 
or less. Where values of A1 and A2 are unlimited, it is for 
radiation control purposes only. For nuclear criticality safety, 
some materials are subject to controls placed on fissile material.
    II. a. For individual radionuclides whose identities are known, 
but which are not listed in Table A-1, the A1 and A2 values 
contained in Table A-3 may be used. Otherwise, the licensee shall 
obtain prior Commission approval of the A1 and A2 values for 
radionuclides not listed in Table A-1, before shipping the material.
    b. For individual radionuclides whose identities are known, but 
which are not listed in Table A-2, the exempt material activity 
concentration and exempt consignment activity values contained in 
Table A-3 may be used. Otherwise, the licensee shall obtain prior 
Commission approval of the exempt material activity concentration 
and exempt consignment activity values for radionuclides not listed 
in Table A-2, before shipping the material.
    c. The licensee shall submit requests for prior approval, 
described under paragraphs II.a. and II.b. of this Appendix, to the 
Commission, in accordance with Sec. 71.1 of this part.
    III. In the calculations of A1 and A2 for 
a radionuclide not in Table A-1, a single radioactive decay chain, 
in which radionuclides are present in their naturally occurring 
proportions, and in which no daughter radionuclide has a half-life 
either longer than 10 days, or longer than that of the parent 
radionuclide, shall be considered as a single radionuclide, and the 
activity to be taken into account, and the A1 and A2 value to be 
applied, shall be those corresponding to the parent radionuclide of 
that chain. In the case of radioactive decay chains in which any 
daughter radionuclide has a half-life either longer than 10 days, or 
greater than that of the parent radionuclide, the parent and those 
daughter radionuclides shall be considered as mixtures of different 
radionuclides.
    IV. For mixtures of radionuclides whose identities and 
respective activities are known, the following conditions apply:
    a. For special form radioactive material, the maximum quantity 
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.014

where B(i) is the activity of radionuclide I, and A1(i) is the A1 
value for radionuclide I.
    b. For normal form radioactive material, the maximum quantity 
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.015

where B(i) is the activity of radionuclide I, and A2(i) is the A2(i) 
value for radionuclide I.
    c. Alternatively, the A1 value for mixtures of 
special form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.016

where f(i) is the fraction of activity for radionuclide I in the 
mixture, and A1(i) is the appropriate A1 value for radionuclide I.
    d. Alternatively, the A2 value for mixtures of normal form 
material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.017

where f(i) is the fraction of activity for radionuclide I in the 
mixture, and A2(i) is the appropriate A2 value for 
radionuclide I.
    e. The exempt activity concentration for mixtures of nuclides 
may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TR26JA04.018

where f(i) is the fraction of activity concentration of radionuclide 
I in the mixture, and [A] is the activity concentration for exempt 
material containing radionuclide I.
    f. The activity limit for an exempt consignment for mixtures of 
radionuclides may be determined as follows:

[[Page 3801]]

[GRAPHIC] [TIFF OMITTED] TR26JA04.019

where f(i) is the fraction of activity of radionuclide I in the 
mixture, and A is the activity limit for exempt consignments for 
radionuclide I.
    V. When the identity of each radionuclide is known, but the 
individual activities of some of the radionuclides are not known, 
the radionuclides may be grouped, and the lowest A1 or A2 value, as 
appropriate, for the radionuclides in each group may be used in 
applying the formulas in paragraph IV. Groups may be based on the 
total alpha activity and the total beta/gamma activity when these 
are known, using the lowest A1 or A2 values 
for the alpha emitters and beta/gamma emitters.

                                                                         Table A-1.--A1 and A2 Values For Radionuclides
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                                                                                                  Specific activity
      Symbol of radionuclide        Element and atomic          A1 (TBq)               A1 (Ci)                A2 (TBq)               A2 (Ci)        --------------------------------------------
                                          number                                                                                                            (TBq/g)                (Ci/g)
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225 (a).......................  Actinium (89).......  8.0x10-1               2.2x101                6.0x10-3               1.6x10-1               2.1x103                5.8x104
Ac-227 (a).......................  ....................  9.0x10-1               2.4x101                9.0x10-5               2.4x10-3               2.7                    7.2x101
Ac-228...........................  ....................  6.0x10-1               1.6x101                5.0x10-1               1.4x101                8.4x104                2.2x106
Ag-105...........................  Silver (47).........  2.0                    5.4x101                2.0                    5.4x101                1.1x103                3.0x104
Ag-108m (a)......................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                9.7x10-1               2.6x101
Ag-110m (a)......................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                1.8x102                4.7x103
Ag-111...........................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                5.8x103                1.6x105
Al-26............................  Aluminum (13).......  1.0x10-1               2.7                    1.0x10-1               2.7                    7.0x10-4               1.9x10-2
Am-241...........................  Americium (95)......  1.0x101                2.7x102                1.0x10-3               2.7x10-2               1.3x10-1               3.4
Am-242m (a)......................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               3.6x10-1               1.0x101
Am-243 (a).......................  ....................  5.0                    1.4x102                1.0x10-3               2.7x10-2               7.4x10-3               2.0x10-1
Ar-37............................  Argon (18)..........  4.0x101                1.1x103                4.0x101                1.1x103                3.7x103                9.9x104
Ar-39............................  ....................  4.0x101                1.1x103                2.0x101                5.4x102                1.3                    3.4x101
Ar-41............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.5x106                4.2x107
As-72............................  Arsenic (33)........  3.0x10-1               8.1                    3.0x10-1               8.1                    6.2x104                1.7x106
As-73............................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                8.2x102                2.2x104
As-74............................  ....................  1.0                    2.7x101                9.0x10-1               2.4x101                3.7x103                9.9x104
As-76............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    5.8x104                1.6x106
As-77............................  ....................  2.0x101                5.4x102                7.0x10-1               1.9x101                3.9x104                1.0x106
At-211 (a).......................  Astatine (85).......  2.0x101                5.4x102                5.0x10-1               1.4x101                7.6x104                2.1x106
Au-193...........................  Gold (79)...........  7.0                    1.9x102                2.0                    5.4x101                3.4x104                9.2x105
Au-194...........................  ....................  1.0                    2.7x101                1.0                    2.7x101                1.5x104                4.1x105
Au-195...........................  ....................  1.0x101                2.7x102                6.0                    1.6x102                1.4x102                3.7x103
Au-198...........................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                9.0x103                2.4x105
Au-199...........................  ....................  1.0x101                2.7x102                6.0x10-1               1.6x101                7.7x103                2.1x105
Ba-131 (a).......................  Barium (56).........  2.0                    5.4x101                2.0                    5.4x101                3.1x103                8.4x104
Ba-133...........................  ....................  3.0                    8.1x101                3.0                    8.1x101                9.4                    2.6x102
Ba-133m..........................  ....................  2.0x101                5.4x102                6.0x10-1               1.6x101                2.2x104                6.1x105
Ba-140 (a).......................  ....................  5.0x10-1               1.4x101                3.0x10-1               8.1                    2.7x103                7.3x104
Be-7.............................  Beryllium (4).......  2.0x101                5.4x102                2.0x101                5.4x102                1.3x104                3.5x105
Be-10............................  ....................  4.0x101                1.1x103                6.0x10-1               1.6x101                8.3x10-4               2.2x10-2
Bi-205...........................  Bismuth (83)........  7.0x10-1               1.9x101                7.0x10-1               1.9x101                1.5x10-3               4.2x104
Bi-206...........................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    3.8x103                1.0x105
Bi-207...........................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                1.9                    5.2x101
Bi-210...........................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                4.6x103                1.2x105
Bi-210m (a)......................  ....................  6.0x10-1               1.6x101                2.0x10-2               5.4x10-1               2.1x10-5               5.7x10-4
Bi-212 (a).......................  ....................  7.0x10-1               1.9x101                6.0x10-1               1.6x101                5.4x105                1.5x107
Bk-247...........................  Berkelium (97)......  8.0                    2.2x102                8.0x10-4               2.2x10-2               3.8x10-2               1.0
Bk-249 (a).......................  ....................  4.0x101                1.1x103                3.0x10-1               8.1                    6.1x101                1.6x103
Br-76............................  Bromine (35)........  4.0x10-1               1.1x101                4.0x10-1               1.1x101                9.4x104                2.5x106
Br-77............................  ....................  3.0                    8.1x101                3.0                    8.1x101                2.6x104                7.1x105
Br-82............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                4.0x104                1.1x106
C-11.............................  Carbon (6)..........  1.0                    2.7x101                6.0x10-1               1.6x101                3.1x107                8.4x108
C-14.............................  ....................  4.0x101                1.1x103                3.0                    8.1x101                1.6x10-1               4.5
Ca-41............................  Calcium (20)........  Unlimited              Unlimited              Unlimited              Unlimited              3.1x10-3               8.5x10-2
Ca-45............................  ....................  4.0x101                1.1x103                1.0                    2.7x101                6.6x102                1.8x104
Ca-47 (a)........................  ....................  3.0                    8.1x101                3.0x10-1               8.1                    2.3x104                6.1x105
Cd-109...........................  Cadmium (48)........  3.0x101                8.1x102                2.0                    5.4x101                9.6x101                2.6x103
Cd-113m..........................  ....................  4.0x101                1.1x103                5.0x10-1               1.4x101                8.3                    2.2x102
Cd-115 (a).......................  ....................  3.0                    8.1x101                4.0x10-1               1.1x101                1.9x104                5.1x105
Cd-115m..........................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                9.4x102                2.5x104
Ce-139...........................  Cerium (58).........  7.0                    1.9x102                2.0                    5.4x101                2.5x102                6.8x103
Ce-141...........................  ....................  2.0x101                5.4x102                6.0x10-1               1.6x101                1.1x103                2.8x104
Ce-143...........................  ....................  9.0x10-1               2.4x101                6.0x10-1               1.6x101                2.5x104                6.6x105
Ce-144 (a).......................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    1.2x102                3.2x103
Cf-248...........................  Californium (98)....  4.0x101                1.1x103                6.0x10-3               1.6x10-1               5.8x101                1.6x103
Cf-249...........................  ....................  3.0                    8.1x101                8.0x10-4               2.2x10-2               1.5x10-1               4.1

[[Page 3802]]

 
Cf-250...........................  ....................  2.0x101                5.4x102                2.0x10-3               5.4x10-2               4.0                    1.1x102
Cf-251...........................  ....................  7.0                    1.9x102                7.0x10-4               1.9x10-2               5.9x10-2               1.6
Cf-252 (h).......................  ....................  5.0x10-2               1.4                    3.0x10-3               8.1x10-2               2.0x101                5.4x102
Cf-253 (a).......................  ....................  4.0x101                1.1x103                4.0x10-2               1.1                    1.1x103                2.9x104
Cf-254...........................  ....................  1.0x10-3               2.7x10-2               1.0x10-3               2.7x10-2               3.1x102                8.5x103
Cl-36............................  Chlorine (17).......  1.0x101                2.7x102                6.0x10-1               1.6x101                1.2x10-3               3.3x10-2
Cl-38............................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    4.9x106                1.3x108
Cm-240...........................  Curium (96).........  4.0x101                1.1x103                2.0x10-2               5.4x10-1               7.5x102                2.0x104
Cm-241...........................  ....................  2.0                    5.4x101                1.0                    2.7x101                6.1x102                1.7x104
Cm-242...........................  ....................  4.0x101                1.1x103                1.0x10-2               2.7x10-1               1.2x102                3.3x103
Cm-243...........................  ....................  9.0                    2.4x102                1.0x10-3               2.7x10-2               1.9x10-3               5.2x101
Cm-244...........................  ....................  2.0x101                5.4x102                2.0x10-3               5.4x10-2               3.0                    8.1x101
Cm-245...........................  ....................  9.0                    2.4x102                9.0x10-4               2.4x10-2               6.4x10-3               1.7x10-1
Cm-246...........................  ....................  9.0                    2.4x102                9.0x10-4               2.4x10-2               1.1x10-2               3.1x10-1
Cm-247 (a).......................  ....................  3.0                    8.1x101                1.0x10-3               2.7x10-2               3.4x10-6               9.3x10-5
Cm-248...........................  ....................  2.0x10-2               5.4x10-1               3.0x10-4               8.1x10-3               1.6x10-5               4.2x10-3
Co-55............................  Cobalt (27).........  5.0x10-1               1.4x101                5.0x10-1               1.4x101                1.1x105                3.1x106
Co-56............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.1x103                3.0x104
Co-57............................  ....................  1.0x101                2.7x102                1.0x101                2.7x102                3.1x102                8.4x103
Co-58............................  ....................  1.0                    2.7x101                1.0                    2.7x101                1.2x103                3.2x104
Co-58m...........................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                2.2x105                5.9x106
Co-60............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                4.2x101                1.1x103
Cr-51............................  Chromium (24).......  3.0x101                8.1x102                3.0x101                8.1x102                3.4x103                9.2x104
Cs-129...........................  Cesium (55).........  4.0                    1.1x102                4.0                    1.1x102                2.8x104                7.6x105
Cs-131...........................  ....................  3.0x101                8.1x102                3.0x101                8.1x102                3.8x103                1.0x105
Cs-132...........................  ....................  1.0                    2.7x101                1.0                    2.7x101                5.7x103                1.5x105
Cs-134...........................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                4.8x101                1.3x103
Cs-134m..........................  ....................  4.0x101                1.1x103                6.0x10-1               1.6x101                3.0x105                8.0x106
Cs-135...........................  ....................  4.0x101                1.1x103                1.0                    2.7x101                4.3x10-5               1.2x10-3
Cs-136...........................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                2.7x103                7.3x104
Cs-137 (a).......................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                3.2                    8.7x101
Cu-64............................  Copper (29).........  6.0                    1.6x102                1.0                    2.7x101                1.4x105                3.9x106
Cu-67............................  ....................  1.0x101                2.7x102                7.0x10-1               1.9x101                2.8x104                7.6x105
Dy-159...........................  Dysprosium (66).....  2.0x101                5.4x102                2.0x101                5.4x102                2.1x102                5.7x103
Dy-165...........................  ....................  9.0x10-1               2.4x101                6.0x10-1               1.6x101                3.0x105                8.2x106
Dy-166 (a).......................  ....................  9.0x10-1               2.4x101                3.0x10-1               8.1                    8.6x103                2.3x105
Er-169...........................  Erbium (68).........  4.0x101                1.1x103                1.0                    2.7x101                3.1x103                8.3x104
Er-171...........................  ....................  8.0x10-1               2.2x101                5.0x10-1               1.4x101                9.0x104                2.4x106
Eu-147...........................  Europium (63).......  2.0                    5.4x101                2.0                    5.4x101                1.4x103                3.7x104
Eu-148...........................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                6.0x102                1.6x104
Eu-149...........................  ....................  2.0x101                5.4x102                2.0x101                5.4x102                3.5x102                9.4x103
Eu-150 (short lived).............  ....................  2.0                    5.4x101                7.0x10-1               1.9x101                6.1x104                1.6x106
Eu-150 (long lived)..............  ....................  7 x 10-1               1.9x101                7.0x10-1               1.9x101                6.1x104                1.6x106
Eu-152...........................  ....................  1.0                    2.7x101                1.0                    2.7x101                6.5                    1.8x102
Eu-152m..........................  ....................  8.0x10-1               2.2x101                8.0x10-1               2.2x101                8.2x104                2.2x106
Eu-154...........................  ....................  9.0x10-1               2.4x101                6.0x10-1               1.6x101                9.8                    2.6x102
Eu-155...........................  ....................  2.0x101                5.4x102                3.0                    8.1x101                1.8x101                4.9x102
Eu-156...........................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                2.0x103                5.5x104
F-18.............................  Fluorine (9)........  1.0                    2.7x101                6.0x10-1               1.6x101                3.5x106                9.5x107
Fe-52 (a)........................  Iron (26)...........  3.0x10-1               8.1                    3.0x10-1               8.1                    2.7x105                7.3x106
Fe-55............................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                8.8x101                2.4x103
Fe-59............................  ....................  9.0x10-1               2.4x101                9.0x10-1               2.4x101                1.8x103                5.0x104
Fe-60 (a)........................  ....................  4.0x101                1.1x103                2.0x10-1               5.4                    7.4x10-4               2.0x10-2
Ga-67............................  Gallium (31)........  7.0                    1.9x102                3.0                    8.1x101                2.2x104                6.0x105
Ga-68............................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                1.5x106                4.1x107
Ga-72............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                1.1x105                3.1x106
Gd-146 (a).......................  Gadolinium (64).....  5.0x10-1               1.4x101                5.0x10-1               1.4x101                6.9x102                1.9x104
Gd-148...........................  ....................  2.0x101                5.4x102                2.0x10-3               5.4x10-2               1.2                    3.2x101
Gd-153...........................  ....................  1.0x101                2.7x102                9.0                    2.4x102                1.3x102                3.5x103
Gd-159...........................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                3.9x104                1.1x106
Ge-68 (a)........................  Germanium (32)......  5.0x10-1               1.4x101                5.0x10-1               1.4x101                2.6x102                7.1x103
Ge-71............................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                5.8x103                1.6x105
Ge-77............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.3x105                3.6x106
Hf-172 (a).......................  Hafnium (72)........  6.0x10-1               1.6x101                6.0x10-1               1.6x101                4.1x101                1.1x103
Hf-175...........................  ....................  3.0                    8.1x101                3.0                    8.1x101                3.9x102                1.1x104
Hf-181...........................  ....................  2.0                    5.4x101                5.0x10-1               1.4x101                6.3x102                1.7x104
Hf-182...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              8.1x10-6               2.2x10-4
Hg-194 (a).......................  Mercury (80)........  1.0                    2.7x101                1.0                    2.7x101                1.3x10-1               3.5

[[Page 3803]]

 
Hg-195m (a)......................  ....................  3.0                    8.1x101                7.0x10-1               1.9x101                1.5x104                4.0x105
Hg-197...........................  ....................  2.0x101                5.4x102                1.0x101                2.7x102                9.2x103                2.5x105
Hg-197m..........................  ....................  1.0x101                2.7x102                4.0x10-1               1.1x101                2.5x104                6.7x105
Hg-203...........................  ....................  5.0                    1.4x102                1.0                    2.7x101                5.1x102                1.4x104
Ho-166...........................  Holmium (67)........  4.0x10-1               1.1x101                4.0x10-1               1.1x101                2.6x104                7.0x105
Ho-166m..........................  ....................  6.0x10-1               1.6x101                5.0x10-1               1.4x101                6.6x10-2               1.8
I-123............................  Iodine (53).........  6.0                    1.6x102                3.0                    8.1x101                7.1x104                1.9x106
I-124............................  ....................  1.0                    2.7x101                1.0                    2.7x101                9.3x103                2.5x105
I-125............................  ....................  2.0x101                5.4x102                3.0                    8.1x101                6.4x102                1.7x104
I-126............................  ....................  2.0                    5.4x101                1.0                    2.7x101                2.9x103                8.0x104
I-129............................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              6.5x10-6               1.8x10-4
I-131............................  ....................  3.0                    8.1x101                7.0x10-1               1.9x101                4.6x103                1.2x105
I-132............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                3.8x105                1.0x107
I-133............................  ....................  7.0x10-1               1.9x101                6.0x10-1               1.6x101                4.2x104                1.1x106
I-134............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    9.9x105                2.7x107
I-135 (a)........................  ....................  6.0x10-1               1.6x101                6.0x10-1               1.6x101                1.3x105                3.5x106
In-111...........................  Indium (49).........  3.0                    8.1x101                3.0                    8.1x101                1.5x104                4.2x105
In-113m..........................  ....................  4.0                    1.1x102                2.0                    5.4x101                6.2x105                1.7x107
In-114m (a)......................  ....................  1.0x101                2.7x102                5.0x10-1               1.4x101                8.6x102                2.3x104
In-115m..........................  ....................  7.0                    1.9x102                1.0                    2.7x101                2.2x105                6.1x106
Ir-189 (a).......................  Iridium (77)........  1.0x101                2.7x102                1.0x101                2.7x102                1.9x103                5.2x104
Ir-190...........................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                2.3x103                6.2x104
Ir-192 (c).......................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                3.4x102                9.2x103
Ir-194...........................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    3.1x104                8.4x105
K-40.............................  Potassium (19)......  9.0x10-1               2.4x101                9.0x10-1               2.4x101                2.4x10-7               6.4x10-6
K-42.............................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    2.2x105                6.0x106
K-43.............................  ....................  7.0x10-1               1.9x101                6.0x10-1               1.6x101                1.2x105                3.3x106
Kr-81............................  Krypton (36)........  4.0x101                1.1x103                4.0x101                1.1x103                7.8x10-4               2.1x10-2
Kr-85............................  ....................  1.0x101                2.7x102                1.0x101                2.7x102                1.5x101                3.9x102
Kr-85m...........................  ....................  8.0                    2.2x102                3.0                    8.1x101                3.0x105                8.2x106
Kr-87............................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    1.0x106                2.8x107
La-137...........................  Lanthanum (57)......  3.0x101                8.1x102                6.0                    1.6x102                1.6x10-3               4.4x10-2
La-140...........................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                2.1x104                5.6x105
Lu-172...........................  Lutetium (71).......  6.0x10-1               1.6x101                6.0x10-1               1.6x101                4.2x103                1.1x105
Lu-173...........................  ....................  8.0                    2.2x102                8.0                    2.2x102                5.6x101                1.5x103
Lu-174...........................  ....................  9.0                    2.4x102                9.0                    2.4x102                2.3x101                6.2x102
Lu-174m..........................  ....................  2.0x101                5.4x102                1.0x101                2.7x102                2.0x102                5.3x103
Lu-177...........................  ....................  3.0x101                8.1x102                7.0x10-1               1.9x101                4.1x103                1.1x105
Mg-28 (a)........................  Magnesium (12)......  3.0x10-1               8.1                    3.0x10-1               8.1                    2.0x105                5.4x106
Mn-52............................  Manganese (25)......  3.0x10-1               8.1                    3.0x10-1               8.1                    1.6x104                4.4x105
Mn-53............................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              6.8x10-5               1.8x10-3
Mn-54............................  ....................  1.0                    2.7x101                1.0                    2.7x101                2.9x102                7.7x103
Mn-56............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    8.0x105                2.2x107
Mo-93............................  Molybdenum (42).....  4.0x101                1.1x103                2.0x101                5.4x102                4.1x10-2               1.1
Mo-99 (a) (i)....................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                1.8x104                4.8x105
N-13.............................  Nitrogen (7)........  9.0x10-1               2.4x101                6.0x10-1               1.6x101                5.4x107                1.5x109
Na-22............................  Sodium (11).........  5.0x10-1               1.4x101                5.0x10-1               1.4x101                2.3x102                6.3x103
Na-24............................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    3.2x105                8.7x106
Nb-93m...........................  Niobium (41)........  4.0x101                1.1x103                3.0x101                8.1x102                8.8                    2.4x102
Nb-94............................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                6.9x10-3               1.9x10-1
Nb-95............................  ....................  1.0                    2.7x101                1.0                    2.7x101                1.5x103                3.9x104
Nb-97............................  ....................  9.0x10-1               2.4x101                6.0x10-1               1.6x101                9.9x105                2.7x107
Nd-147...........................  Neodymium (60)......  6.0                    1.6x102                6.0x10-1               1.6x101                3.0x103                8.1x104
Nd-149...........................  ....................  6.0x10-1               1.6x101                5.0x10-1               1.4x101                4.5x105                1.2x107
Ni-59............................  Nickel (28).........  Unlimited              Unlimited              Unlimited              Unlimited              3.0x10-3               8.0x10-2
Ni-63............................  ....................  4.0x101                1.1x103                3.0x101                8.1x102                2.1                    5.7x101
Ni-65............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                7.1x105                1.9x107
Np-235...........................  Neptunium (93)......  4.0x101                1.1x103                4.0x101                1.1x103                5.2x101                1.4x103
Np-236 (short-lived).............  ....................  2.0x101                5.4x102                2.0                    5.4x101                4.7x10-4               1.3x10-2
Np-236 (long-lived)..............  ....................  9.0x100                2.4x102                2.0x10-2               5.4x10-1               4.7x10-4               1.3x10-2
Np-237...........................  ....................  2.0x101                5.4x102                2.0x10-3               5.4x10-2               2.6x10-5               7.1x10-4
Np-239...........................  ....................  7.0                    1.9x102                4.0x10-1               1.1x101                8.6x103                2.3x105
Os-185...........................  Osmium (76).........  1.0                    2.7x101                1.0                    2.7x101                2.8x102                7.5x103
Os-191...........................  ....................  1.0x101                2.7x102                2.0                    5.4x101                1.6x103                4.4x104
Os-191m..........................  ....................  4.0x101                1.1x103                3.0x101                8.1x102                4.6x104                1.3x106
Os-193...........................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                2.0x104                5.3x105
Os-194 (a).......................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.1x101                3.1x102
P-32.............................  Phosphorus (15).....  5.0x10-1               1.4x101                5.0x10-1               1.4x101                1.1x104                2.9x105

[[Page 3804]]

 
P-33.............................  ....................  4.0x101                1.1x103                1.0                    2.7x101                5.8x103                1.6x105
Pa-230 (a).......................  Protactinium (91)...  2.0                    5.4x101                7.0x10-2               1.9                    1.2x103                3.3x104
Pa-231...........................  ....................  4.0                    1.1x102                4.0x10-4               1.1x10-2               1.7x10-3               4.7x10-2
Pa-233...........................  ....................  5.0                    1.4x102                7.0x10-1               1.9x101                7.7x102                2.1x104
Pb-201...........................  Lead (82)...........  1.0                    2.7x101                1.0                    2.7x101                6.2x104                1.7x106
Pb-202...........................  ....................  4.0x101                1.1x103                2.0x101                5.4x102                1.2x10-4               3.4x10-3
Pb-203...........................  ....................  4.0                    1.1x102                3.0                    8.1x101                1.1x104                3.0x105
Pb-205...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              4.5x10-6               1.2x10-4
Pb-210 (a).......................  ....................  1.0                    2.7x101                5.0x10-2               1.4                    2.8                    7.6x101
Pb-212 (a).......................  ....................  7.0x10-1               1.9x101                2.0x10-1               5.4                    5.1x104                1.4x106
Pd-103 (a).......................  Palladium (46)......  4.0x101                1.1x103                4.0x101                1.1x103                2.8x103                7.5x104
Pd-107...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              1.9x10-5               5.1x10-4
Pd-109...........................  ....................  2.0                    5.4x101                5.0x10-1               1.4x101                7.9x104                2.1x106
Pm-143...........................  Promethium (61).....  3.0                    8.1x101                3.0                    8.1x101                1.3x102                3.4x103
Pm-144...........................  ....................  7.0x10-1               1.9x101                7.0x10-1               1.9x101                9.2x101                2.5x103
Pm-145...........................  ....................  3.0x101                8.1x102                1.0x101                2.7x102                5.2                    1.4x102
Pm-147...........................  ....................  4.0x101                1.1x103                2.0                    5.4x101                3.4x101                9.3x102
Pm-148m (a)......................  ....................  8.0x10-1               2.2x101                7.0x10-1               1.9x101                7.9x102                2.1x104
Pm-149...........................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                1.5x104                4.0x105
Pm-151...........................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                2.7x104                7.3x105
Po-210...........................  Polonium (84).......  4.0x101                1.1x103                2.0x10-2               5.4x10-1               1.7x102                4.5x103
Pr-142...........................  Praseodymium (59)...  4.0x10-1               1.1x101                4.0x10-1               1.1x101                4.3x104                1.2x106
Pr-143...........................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                2.5x103                6.7x104
Pt-188 (a).......................  Platinum (78).......  1.0                    2.7x101                8.0x10-1               2.2x101                2.5x103                6.8x104
Pt-191...........................  ....................  4.0                    1.1x102                3.0                    8.1x101                8.7x103                2.4x105
Pt-193...........................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                1.4                    3.7x101
Pt-193m..........................  ....................  4.0x101                1.1x103                5.0x10-1               1.4x101                5.8x103                1.6x105
Pt-195m..........................  ....................  1.0x101                2.7x102                5.0x10-1               1.4x101                6.2x103                1.7x105
Pt-197...........................  ....................  2.0x101                5.4x102                6.0x10-1               1.6x101                3.2x104                8.7x105
Pt-197m..........................  ....................  1.0x101                2.7x102                6.0x10-1               1.6x101                3.7x105                1.0x107
Pu-236...........................  Plutonium (94)......  3.0x101                8.1x102                3.0x10-3               8.1x10-2               2.0x101                5.3x102
Pu-237...........................  ....................  2.0x101                5.4x102                2.0x101                5.4x102                4.5x102                1.2x104
Pu-238...........................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               6.3x10-1               1.7x101
Pu-239...........................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               2.3x10-3               6.2x10-2
Pu-240...........................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               8.4x10-3               2.3x10-1
Pu-241 (a).......................  ....................  4.0x101                1.1x103                6.0x10-2               1.6                    3.8                    1.0x102
Pu-242...........................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               1.5x10-4               3.9x10-3
Pu-244 (a).......................  ....................  4.0x10-1               1.1x101                1.0x10-3               2.7x10-2               6.7x10-7               1.8x10-5
Ra-223 (a).......................  Radium (88).........  4.0x10-1               1.1x101                7.0x10-3               1.9x10-1               1.9x103                5.1x104
Ra-224 (a).......................  ....................  4.0x10-1               1.1x101                2.0x10-2               5.4x10-1               5.9x103                1.6x105
Ra-225 (a).......................  ....................  2.0x10-1               5.4                    4.0x10-3               1.1x10-1               1.5x103                3.9x104
Ra-226 (a).......................  ....................  2.0x10-1               5.4                    3.0x10-3               8.1x10-2               3.7x10-2               1.0
Ra-228 (a).......................  ....................  6.0x10-1               1.6x101                2.0x10-2               5.4x10-1               1.0x101                2.7x102
Rb-81............................  Rubidium (37).......  2.0                    5.4x101                8.0x10-1               2.2x101                3.1x105                8.4x106
Rb-83 (a)........................  ....................  2.0                    5.4x101                2.0                    5.4x101                6.8x102                1.8x104
Rb-84............................  ....................  1.0                    2.7x101                1.0                    2.7x101                1.8x103                4.7x104
Rb-86............................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                3.0x103                8.1x104
Rb-87............................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              3.2x10-9               8.6x10-8
Rb(nat)..........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              6.7x106                1.8x108
Re-184...........................  Rhenium (75)........  1.0                    2.7x101                1.0                    2.7x101                6.9x102                1.9x104
Re-184m..........................  ....................  3.0                    8.1x101                1.0                    2.7x101                1.6x102                4.3x103
Re-186...........................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                6.9x103                1.9x105
Re-187...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              1.4x10-9               3.8x10-8
Re-188...........................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                3.6x104                9.8x105
Re-189 (a).......................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                2.5x104                6.8x105
Re(nat)..........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              0.0                    2.4x10-8
Rh-99............................  Rhodium (45)........  2.0                    5.4x101                2.0                    5.4x101                3.0x103                8.2x104
Rh-101...........................  ....................  4.0                    1.1x102                3.0                    8.1x101                4.1x101                1.1x103
Rh-102...........................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                4.5x101                1.2x103
Rh-102m..........................  ....................  2.0                    5.4x101                2.0                    5.4x101                2.3x102                6.2x103
Rh-103m..........................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                1.2x106                3.3x107
Rh-105...........................  ....................  1.0x101                2.7x102                8.0x10-1               2.2x101                3.1x104                8.4x105
Rn-222 (a).......................  Radon (86)..........  3.0x10-1               8.1                    4.0x10-3               1.1x10-1               5.7x103                1.5x105
Ru-97............................  Ruthenium (44)......  5.0                    1.4x102                5.0                    1.4x102                1.7x104                4.6x105
Ru-103 (a).......................  ....................  2.0                    5.4x101                2.0                    5.4x101                1.2x103                3.2x104
Ru-105...........................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                2.5x105                6.7x106
Ru-106 (a).......................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    1.2x102                3.3x103
S-35.............................  Sulphur (16)........  4.0x101                1.1x103                3.0                    8.1x101                1.6x103                4.3x104
Sb-122...........................  Antimony (51).......  4.0x10-1               1.1x101                4.0x10-1               1.1x101                1.5x104                4.0x105
Sb-124...........................  ....................  6.0x10-1               1.6x101                6.0x10-1               1.6x101                6.5x102                1.7x104

[[Page 3805]]

 
Sb-125...........................  ....................  2.0                    5.4x101                1.0                    2.7x101                3.9x101                1.0x103
Sb-126...........................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                3.1x103                8.4x104
Sc-44............................  Scandium (21).......  5.0x10-1               1.4x101                5.0x10-1               1.4x101                6.7x105                1.8x107
Sc-46............................  ....................  5.0x10-1               1.4x101                5.0x10-1               1.4x101                1.3x103                3.4x104
Sc-47............................  ....................  1.0x101                2.7x102                7.0x10-1               1.9x101                3.1x104                8.3x105
Sc-48............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    5.5x104                1.5x106
Se-75............................  Selenium (34).......  3.0                    8.1x101                3.0                    8.1x101                5.4x102                1.5x104
Se-79............................  ....................  4.0x101                1.1x103                2.0                    5.4x101                2.6x10-3               7.0x10-2
Si-31............................  Silicon (14)........  6.0x10-1               1.6x101                6.0x10-1               1.6x101                1.4x106                3.9x107
Si-32............................  ....................  4.0x101                1.1x103                5.0x10-1               1.4x101                3.9                    1.1x102
Sm-145...........................  Samarium (62).......  1.0x101                2.7x102                1.0x101                2.7x102                9.8x101                2.6x103
Sm-147...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              8.5x10-1               2.3x10-8
Sm-151...........................  ....................  4.0x101                1.1x103                1.0x101                2.7x102                9.7x10-1               2.6x101
Sm-153...........................  ....................  9.0                    2.4x102                6.0x10-1               1.6x101                1.6x104                4.4x105
Sn-113 (a).......................  Tin (50)............  4.0                    1.1x102                2.0                    5.4x101                3.7x102                1.0x104
Sn-117m..........................  ....................  7.0                    1.9x102                4.0x10-1               1.1x101                3.0x103                8.2x104
Sn-119m..........................  ....................  4.0x101                1.1x103                3.0x101                8.1x102                1.4x102                3.7x103
Sn-121m (a)......................  ....................  4.0x101                1.1x103                9.0x10-1               2.4x101                2.0                    5.4x101
Sn-123...........................  ....................  8.0x10-1               2.2x101                6.0x10-1               1.6x101                3.0x102                8.2x103
Sn-125...........................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                4.0x103                1.1x105
Sn-126 (a).......................  ....................  6.0x10-1               1.6x101                4.0x10-1               1.1x101                1.0x10-3               2.8x10-2
Sr-82 (a)........................  Strontium (38)......  2.0x10-1               5.4                    2.0x10-1               5.4                    2.3x103                6.2x104
Sr-85............................  ....................  2.0                    5.4x101                2.0                    5.4x101                8.8x102                2.4x104
Sr-85m...........................  ....................  5.0                    1.4x102                5.0                    1.4x102                1.2x106                3.3x107
Sr-87m...........................  ....................  3.0                    8.1x101                3.0                    8.1x101                4.8x105                1.3x107
Sr-89............................  ....................  6.0x10-1               1.6x101                6.0x10-1               1.6x101                1.1x103                2.9x104
Sr-90 (a)........................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    5.1                    1.4x102
Sr-91 (a)........................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.3x105                3.6x106
Sr-92 (a)........................  ....................  1.0                    2.7x101                3.0x10-1               8.1                    4.7x105                1.3x107
T(H-3)...........................  Tritium (1).........  4.0x101                1.1x103                4.0x101                1.1x103                3.6x102                9.7x103
Ta-178 (long-lived)..............  Tantalum (73).......  1.0                    2.7x101                8.0x10-1               2.2x101                4.2x106                1.1x108
Ta-179...........................  ....................  3.0x101                8.1x102                3.0x101                8.1x102                4.1x101                1.1x103
Ta-182...........................  ....................  9.0x10-1               2.4x101                5.0x10-1               1.4x101                2.3x102                6.2x103
Tb-157...........................  Terbium (65)........  4.0x101                1.1x103                4.0x101                1.1x103                5.6x10-1               1.5x101
Tb-158...........................  ....................  1.0                    2.7x101                1.0                    2.7x101                5.6x10-1               1.5x101
Tb-160...........................  ....................  1.0                    2.7x101                6.0x10-1               1.6x101                4.2x102                1.1x104
Tc-95m (a).......................  Technetium (43).....  2.0                    5.4x101                2.0                    5.4x101                8.3x102                2.2x104
Tc-96............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                1.2x104                3.2x105
Tc-96m (a).......................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                1.4x106                3.8x107
Tc-97............................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              5.2x10-5               1.4x10-3
Tc-97m...........................  ....................  4.0x101                1.1x103                1.0                    2.7x101                5.6x102                1.5x104
Tc-98............................  ....................  8.0x10-1               2.2x101                7.0x10-1               1.9x101                3.2x10-5               8.7x10-4
Tc-99............................  ....................  4.0x101                1.1x103                9.0x10-1               2.4x101                6.3x10-4               1.7x10-2
Tc-99m...........................  ....................  1.0x101                2.7x102                4.0                    1.1x102                1.9x105                5.3x106
Te-121...........................  Tellurium (52)......  2.0                    5.4x101                2.0                    5.4x101                2.4x103                6.4x104
Te-121m..........................  ....................  5.0                    1.4x102                3.0                    8.1x101                2.6x102                7.0x103
Te-123m..........................  ....................  8.0                    2.2x102                1.0                    2.7x101                3.3x102                8.9x103
Te-125m..........................  ....................  2.0x101                5.4x102                9.0x10-1               2.4x101                6.7x102                1.8x104
Te-127...........................  ....................  2.0x101                5.4x102                7.0x10-1               1.9x101                9.8x104                2.6x106
Te-127m (a)......................  ....................  2.0x101                5.4x102                5.0x10-1               1.4x101                3.5x102                9.4x103
Te-129...........................  ....................  7.0x10-1               1.9x101                6.0x10-1               1.6x101                7.7x105                2.1x107
Te-129m (a)......................  ....................  8.0x10-1               2.2x101                4.0x10-1               1.1x101                1.1x103                3.0x104
Te-131m (a)......................  ....................  7.0x10-1               1.9x101                5.0x10-1               1.4x101                3.0x104                8.0x105
Te-132 (a).......................  ....................  5.0x10-1               1.4x101                4.0x10-1               1.1x101                1.1x104                8.0x105
Th-227...........................  Thorium (90)........  1.0x101                2.7x102                5.0x10-3               1.4x10-1               1.1x103                3.1x104
Th-228 (a).......................  ....................  5.0x10-1               1.4x101                1.0x10-3               2.7x10-2               3.0x101                8.2x102
Th-229...........................  ....................  5.0                    1.4x102                5.0x10-4               1.4x10-2               7.9x10-3               2.1x10-1
Th-230...........................  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               7.6x10-4               2.1x10-2
Th-231...........................  ....................  4.0x101                1.1x103                2.0x10-2               5.4x10-1               2.0x104                5.3x105
Th-232...........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              4.0x10-9               1.1x10-7
Th-234 (a).......................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    8.6x102                2.3x104
Th(nat)..........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              8.1x10-9               2.2x10-7
Ti-44 (a)........................  Titanium (22).......  5.0x10-1               1.4x101                4.0x10-1               1.1x101                6.4                    1.7x102
Tl-200...........................  Thallium (81).......  9.0x10-1               2.4x101                9.0x10-1               2.4x101                2.2x104                6.0x105
Tl-201...........................  ....................  1.0x101                2.7x102                4.0                    1.1x102                7.9x103                2.1x105
Tl-202...........................  ....................  2.0                    5.4x101                2.0                    5.4x101                2.0x103                5.3x104
Tl-204...........................  ....................  1.0x101                2.7x102                7.0x10-1               1.9x101                1.7x101                4.6x102
Tm-167...........................  Thulium (69)........  7.0                    1.9x102                8.0x10-1               2.2x101                3.1x103                8.5x104
Tm-170...........................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                2.2x102                6.0x103

[[Page 3806]]

 
Tm-171...........................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                4.0x101                1.1x103
U-230 (fast lung absorption)       Uranium (92)........  4.0x101                1.1x103                1.0x10-1               2.7                    1.0x103                2.7x104
 (a)(d).
U-230 (medium lung absorption)     ....................  4.0x101                1.1x103                4.0x10-3               1.1x10-1               1.0x103                2.7x104
 (a)(e).
U-230 (slow lung absorption)       ....................  3.0x101                8.1x102                3.0x10-3               8.1x10-2               1.0x103                2.7x104
 (a)(f).
U-232 (fast lung absorption) (d).  ....................  4.0x101                1.1x103                1.0x10-2               2.7x10-1               8.3x10-1               2.2x101
U-232 (medium lung absorption)     ....................  4.0x101                1.1x103                7.0x10-3               1.9x10-1               8.3x10-1               2.2x101
 (e).
U-232 (slow lung absorption) (f).  ....................  1.0x101                2.7x102                1.0x10-3               2.7x10-2               8.3x10-1               2.2x101
U-233 (fast lung absorption) (d).  ....................  4.0x101                1.1x103                9.0x10-2               2.4                    3.6x10-4               9.7x10-3
U-233 (medium lung absorption)     ....................  4.0x101                1.1x103                2.0x10-2               5.4x10-1               3.6x10-4               9.7x10-3
 (e).
U-233 (slow lung absorption) (f).  ....................  4.0x101                1.1x103                6.0x10-3               1.6x10-1               3.6x10-4               9.7x10-3
U-234 (fast lung absorption) (d).  ....................  4.0x101                1.1x103                9.0x10-2               2.4                    2.3x10-4               6.2x10-3
U-234 (medium lung absorption)     ....................  4.0x101                1.1x103                2.0x10-2               5.4x10-1               2.3x10-4               6.2x10-3
 (e).
U-234 (slow lung absorption) (f).  ....................  4.0x101                1.1x103                6.0x10-3               1.6x10-1               2.3x10-4               6.2x10-3
U-235 (all lung absorption types)  ....................  Unlimited              Unlimited              Unlimited              Unlimited              8.0x10-8               2.2x10-6
 (a),(d),(e),(f).
U-236 (fast lung absorption) (d).  ....................  Unlimited              Unlimited              Unlimited              Unlimited              2.4x10-6               6.5x10-5
U-236 (medium lung absorption)     ....................  4.0x101                1.1x103                2.0x10-2               5.4x10-1               2.4x10-6               6.5x10-5
 (e).
U-236 (slow lung absorption) (f).  ....................  4.0x101                1.1x103                6.0x10-3               1.6x10-1               2.4x10-6               6.5x10-5
U-238 (all lung absorption types)  ....................  Unlimited              Unlimited              Unlimited              Unlimited              1.2x10-8               3.4x10-7
 (d),(e),(f).
U (nat)..........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              2.6x10-8               7.1x10-7
U (enriched to 20% or less)(g)...  ....................  Unlimited              Unlimited              Unlimited              Unlimited              See Table A-4          See Table A-4
U (dep)..........................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              See Table A-4          See Table A-4
V-48.............................  Vanadium (23).......  4.0x10-1               1.1x101                4.0x10-1               1.1x101                6.3x103                1.7x105
V-49.............................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                3.0x102                8.1x103
W-178 (a)........................  Tungsten (74).......  9.0                    2.4x102                5.0                    1.4x102                1.3x103                3.4x104
W-181............................  ....................  3.0x101                8.1x102                3.0x101                8.1x102                2.2x102                6.0x103
W-185............................  ....................  4.0x101                1.1x103                8.0x10-1               2.2x101                3.5x102                9.4x103
W-187............................  ....................  2.0                    5.4x101                6.0x10-1               1.6x101                2.6x104                7.0x105
W-188 (a)........................  ....................  4.0x10-1               1.1x101                3.0x10-1               8.1                    3.7x102                1.0x104
Xe-122 (a).......................  Xenon (54)..........  4.0x10-1               1.1x101                4.0x10-1               1.1x101                4.8x104                1.3x106
Xe-123...........................  ....................  2.0                    5.4x101                7.0x10-1               1.9x101                4.4x105                1.2x107
Xe-127...........................  ....................  4.0                    1.1x102                2.0                    5.4x101                1.0x103                2.8x104
Xe-131m..........................  ....................  4.0x101                1.1x103                4.0x101                1.1x103                3.1x103                8.4x104
Xe-133...........................  ....................  2.0x101                5.4x102                1.0x101                2.7x102                6.9x103                1.9x105
Xe-135...........................  ....................  3.0                    8.1x101                2.0                    5.4x101                9.5x104                2.6x106
Y-87 (a).........................  Yttrium (39)........  1.0                    2.7x101                1.0                    2.7x101                1.7x104                4.5x105

[[Page 3807]]

 
Y-88.............................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                5.2x102                1.4x104
Y-90.............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    2.0x104                5.4x105
Y-91.............................  ....................  6.0x10-1               1.6x101                6.0x10-1               1.6x101                9.1x102                2.5x104
Y-91m............................  ....................  2.0                    5.4x101                2.0                    5.4x101                1.5x106                4.2x107
Y-92.............................  ....................  2.0x10-1               5.4                    2.0x10-1               5.4                    3.6x105                9.6x106
Y-93.............................  ....................  3.0x10-1               8.1                    3.0x10-1               8.1                    1.2x105                3.3x106
Yb-169...........................  Ytterbium (70)......  4.0                    1.1x102                1.0                    2.7x101                8.9x102                2.4x104
Yb-175...........................  ....................  3.0x101                8.1x102                9.0x10-1               2.4x101                6.6x103                1.8x105
Zn-65............................  Zinc (30)...........  2.0                    5.4x101                2.0                    5.4x101                3.0x102                8.2x103
Zn-69............................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                1.8x106                4.9x107
Zn-69m (a).......................  ....................  3.0                    8.1x101                6.0x10-1               1.6x101                1.2x105                3.3x106
Zr-88............................  Zirconium (40)......  3.0                    8.1x101                3.0                    8.1x101                6.6x102                1.8x104
Zr-93............................  ....................  Unlimited              Unlimited              Unlimited              Unlimited              9.3x10-5               2.5x10-3
Zr-95 (a)........................  ....................  2.0                    5.4x101                8.0x10-1               2.2x101                7.9x102                2.1x104
Zr-97 (a)........................  ....................  4.0x10-1               1.1x101                4.0x10-1               1.1x101                7.1x104                1.9x106
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
a A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days.
b [Reserved]
c The quantity may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.
g These values apply to unirradiated uranium only.
h A1 = 0.1 TBq (2.7 Ci) and A2 = 0.001 TBq (0.027 Ci) for Cf-252 for domestic use.
i A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.


                      Table A-2.--Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                           Activity concentration  Activity concentration    Activity limit for      Activity limit for
      Symbol of radionuclide          Element and atomic     for exempt material     for exempt material     exempt consignment      exempt consignment
                                            number                 (Bq/g)                  (Ci/g)                   (Bq)                    (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225............................  Actinium (89)........  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Ac-227............................  .....................  1.0x10-1                2.7x10-12               1.0x103                 2.7x10-8
Ac-228............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ag-105............................  Silver (47)..........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ag-108m (b).......................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ag-110m...........................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ag-111............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Al-26.............................  Aluminum (13)........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Am-241............................  Americium (95).......  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Am-242m (b).......................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Am-243 (b)........................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Ar-37.............................  Argon (18)...........  1.0x106                 2.7x10-5                1.0x108                 2.7x10-3
Ar-39.............................  .....................  1.0x107                 2.7x10-4                1.0x104                 2.7x10-7
Ar-41.............................  .....................  1.0x102                 2.7x10-9                1.0x109                 2.7x10-2
As-72.............................  Arsenic (33).........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
As-73.............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
As-74.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
As-76.............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
As-77.............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
At-211............................  Astatine (85)........  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Au-193............................  Gold (79)............  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Au-194............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Au-195............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Au-198............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Au-199............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ba-131............................  Barium (56)..........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ba-133............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ba-133m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ba-140 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Be-7..............................  Beryllium (4)........  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Be-10.............................  .....................  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Bi-205............................  Bismuth (83).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Bi-206............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Bi-207............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5

[[Page 3808]]

 
Bi-210............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Bi-210m...........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Bi-212 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Bk-247............................  Berkelium (97).......  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Bk-249............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Br-76.............................  Bromine (35).........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Br-77.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Br-82.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
C-11..............................  Carbon (6)...........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
C-14..............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Ca-41.............................  Calcium (20).........  1.0x105                 2.7x10-6                1.0x107                 2.7x10-4
Ca-45.............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Ca-47.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Cd-109............................  Cadmium (48).........  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Cd-113m...........................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Cd-115............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Cd-115m...........................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Ce-139............................  Cerium (58)..........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ce-141............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Ce-143............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ce-144 (b)........................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Cf-248............................  Californium (98).....  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cf-249............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Cf-250............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cf-251............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Cf-252............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cf-253............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Cf-254............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Cl-36.............................  Chlorine (17)........  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Cl-38.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Cm-240............................  Curium (96)..........  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Cm-241............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Cm-242............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Cm-243............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Cm-244............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cm-245............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Cm-246............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Cm-247............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Cm-248............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Co-55.............................  Cobalt (27)..........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Co-56.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Co-57.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Co-58.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Co-58m............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Co-60.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Cr-51.............................  Chromium (24)........  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Cs-129............................  Cesium (55)..........  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Cs-131............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Cs-132............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Cs-134............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cs-134m...........................  .....................  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
Cs-135............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Cs-136............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Cs-137 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Cu-64.............................  Copper (29)..........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Cu-67.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Dy-159............................  Dysprosium (66)......  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Dy-165............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Dy-166............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Er-169............................  Erbium (68)..........  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Er-171............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Eu-147............................  Europium (63)........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Eu-148............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Eu-149............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Eu-150 (short lived)..............  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Eu-150 (long lived)...............  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Eu-152............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5

[[Page 3809]]

 
Eu-152m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Eu-154............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Eu-155............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Eu-156............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
F-18..............................  Fluorine (9).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Fe-52.............................  Iron (26)............  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Fe-55.............................  .....................  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Fe-59.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Fe-60.............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Ga-67.............................  Gallium (31).........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ga-68.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Ga-72.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Gd-146............................  Gadolinium (64)......  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Gd-148............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Gd-153............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Gd-159............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Ge-68.............................  Germanium (32).......  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Ge-71.............................  .....................  1.0x104                 2.7x10-7                1.0x108                 2.7x10-3
Ge-77.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Hf-172............................  Hafnium (72).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Hf-175............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Hf-181............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Hf-182............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Hg-194............................  Mercury (80).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Hg-195m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Hg-197............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Hg-197m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Hg-203............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Ho-166............................  Holmium (67).........  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
Ho-166m...........................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
I-123.............................  Iodine (53)..........  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
I-124.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
I-125.............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
I-126.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
I-129.............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
I-131.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
I-132.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
I-133.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
I-134.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
I-135.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
In-111............................  Indium (49)..........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
In-113m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
In-114m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
In-115m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ir-189............................  Iridium (77).........  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Ir-190............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ir-192............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Ir-194............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
K-40..............................  Potassium (19).......  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
K-42..............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
K-43..............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Kr-81.............................  Krypton (36).........  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Kr-85.............................  .....................  1.0x105                 2.7x10-6                1.0x104                 2.7x10-7
Kr-85m............................  .....................  1.0x103                 2.7x10-8                1.0x1010                2.7x10-1
Kr-87.............................  .....................  1.0x102                 2.7x10-9                1.0x109                 2.7x10-2
La-137............................  Lanthanum (57).......  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
La-140............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Lu-172............................  Lutetium (71)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Lu-173............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Lu-174............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Lu-174m...........................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Lu-177............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Mg-28.............................  Magnesium (12).......  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Mn-52.............................  Manganese (25).......  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Mn-53.............................  .....................  1.0x104                 2.7x10-7                1.0x109                 2.7x10-2
Mn-54.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Mn-56.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6

[[Page 3810]]

 
Mo-93.............................  Molybdenum (42)......  1.0x103                 2.7x10-8                1.0x108                 2.7x10-3
Mo-99.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
N-13..............................  Nitrogen (7).........  1.0x102                 2.7x10-9                1.0x109                 2.7x10-2
Na-22.............................  Sodium (11)..........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Na-24.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Nb-93m............................  Niobium (41).........  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Nb-94.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Nb-95.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Nb-97.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Nd-147............................  Neodymium (60).......  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Nd-149............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ni-59.............................  Nickel (28)..........  1.0x104                 2.7x10-7                1.0x108                 2.7x10-3
Ni-63.............................  .....................  1.0x105                 2.7x10-6                1.0x108                 2.7x10-3
Ni-65.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Np-235............................  Neptunium (93).......  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Np-236 (short-lived)..............  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Np-236 (long-lived)...............  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Np-237 (b)........................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Np-239............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Os-185............................  Osmium (76)..........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Os-191............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Os-191m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Os-193............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Os-194............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
P-32..............................  Phosphorus (15)......  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
P-33..............................  .....................  1.0x105                 2.7x10-6                1.0x108                 2.7x10-3
Pa-230............................  Protactinium (91)....  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Pa-231............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Pa-233............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Pb-201............................  Lead (82)............  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Pb-202............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Pb-203............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Pb-205............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Pb-210 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Pb-212 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Pd-103............................  Palladium (46).......  1.0x103                 2.7x10-8                1.0x108                 2.7x10-3
Pd-107............................  .....................  1.0x105                 2.7x10-6                1.0x108                 2.7x10-3
Pd-109............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Pm-143............................  Promethium (61)......  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Pm-144............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Pm-145............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Pm-147............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Pm-148m...........................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Pm-149............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Pm-151............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Po-210............................  Polonium (84)........  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Pr-142............................  Praseodymium (59)....  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Pr-143............................  .....................  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Pt-188............................  Platinum (78)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Pt-191............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Pt-193............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Pt-193m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Pt-195m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Pt-197............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Pt-197m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Pu-236............................  Plutonium (94).......  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Pu-237............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Pu-238............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Pu-239............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Pu-240............................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Pu-241............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Pu-242............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Pu-244............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Ra-223 (b)........................  Radium (88)..........  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Ra-224 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Ra-225............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Ra-226 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7

[[Page 3811]]

 
Ra-228 (b)........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Rb-81.............................  Rubidium (37)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Rb-83.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Rb-84.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Rb-86.............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Rb-87.............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Rb(nat)...........................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Re-184............................  Rhenium (75).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Re-184m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Re-186............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Re-187............................  .....................  1.0x106                 2.7x10-5                1.0x109                 2.7x10-2
Re-188............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Re-189............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Re(nat)...........................  .....................  1.0x106                 2.7x10-5                1.0x109                 2.7x10-2
Rh-99.............................  Rhodium (45).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Rh-101............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Rh-102............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Rh-102m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Rh-103m...........................  .....................  1.0x104                 2.7x10-7                1.0x108                 2.7x10-3
Rh-105............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Rn-222 (b)........................  Radon (86)...........  1.0x101                 2.7x10-10               1.0x108                 2.7x10-3
Ru-97.............................  Ruthenium (44).......  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Ru-103............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Ru-105............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ru-106 (b)........................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
S-35..............................  Sulphur (16).........  1.0x105                 2.7x10-6                1.0x108                 2.7x10-3
Sb-122............................  Antimony (51)........  1.0x102                 2.7x10-9                1.0x104                 2.7x10-7
Sb-124............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Sb-125............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sb-126............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Sc-44.............................  Scandium (21)........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Sc-46.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Sc-47.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sc-48.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Se-75.............................  Selenium (34)........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Se-79.............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Si-31.............................  Silicon (14).........  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Si-32.............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Sm-145............................  Samarium (62)........  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Sm-147............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Sm-151............................  .....................  1.0x104                 2.7x10-7                1.0x108                 2.7x10-3
Sm-153............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sn-113............................  Tin (50).............  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Sn-117m...........................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sn-119m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Sn-121m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Sn-123............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Sn-125............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Sn-126............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Sr-82.............................  Strontium (38).......  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Sr-85.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sr-85m............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Sr-87m............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Sr-89.............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Sr-90 (b).........................  .....................  1.0x102                 2.7x10-9                1.0x104                 2.7x10-7
Sr-91.............................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Sr-92.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
T(H-3)............................  Tritium (1)..........  1.0x106                 2.7x10-5                1.0x109                 2.7x10-2
Ta-178 (long-lived)...............  Tantalum (73)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Ta-179............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Ta-182............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Tb-157............................  Terbium (65).........  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Tb-158............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tb-160............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tc-95m............................  Technetium (43)......  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tc-96.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tc-96m............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4

[[Page 3812]]

 
Tc-97.............................  .....................  1.0x103                 2.7x10-8                1.0x108                 2.7x10-3
Tc-97m............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Tc-98.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tc-99.............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
Tc-99m............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Te-121............................  Tellurium (52).......  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Te-121m...........................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Te-123m...........................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Te-125m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Te-127............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Te-127m...........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Te-129............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Te-129m...........................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Te-131m...........................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Te-132............................  .....................  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Th-227............................  Thorium (90).........  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Th-228 (b)........................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Th-229 (b)........................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Th-230............................  .....................  1.0                     2.7x10-11               1.0x104                 2.7x10-7
Th-231............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Th-232............................  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
Th-234 (b)........................  .....................  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
Th (nat) (b)......................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
Ti-44.............................  Titanium (22)........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
Tl-200............................  Thallium (81)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Tl-201............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Tl-202............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Tl-204............................  .....................  1.0x104                 2.7x10-7                1.0x104                 2.7x10-7
Tm-167............................  Thulium (69).........  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Tm-170............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Tm-171............................  .....................  1.0x104                 2.7x10-7                1.0x108                 2.7x10-3
U-230 (fast lung absorption)        Uranium (92).........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
 (b),(d).
U-230 (medium lung absorption) (e)  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-230 (slow lung absorption) (f)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-232 (fast lung absorption)        .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
 (b),(d).
U-232 (medium lung absorption) (e)  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-232 (slow lung absorption) (f)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-233 (fast lung absorption) (d)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-233 (medium lung absorption) (e)  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
U-233 (slow lung absorption) (f)..  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
U-234 (fast lung absorption) (d)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-234 (medium lung absorption) (e)  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
U-234 (slow lung absorption) (f)..  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
U-235 (all lung absorption types)   .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
 (b),(d),(e),(f).
U-236 (fast lung absorption) (d)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-236 (medium lung absorption) (e)  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
U-236 (slow lung absorption) (f)..  .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
U-238 (all lung absorption types)   .....................  1.0x101                 2.7x10-10               1.0x104                 2.7x10-7
 (b),(d),(e),(f).
U (nat) (b).......................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8

[[Page 3813]]

 
U (enriched to 20% or less)(g)....  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
U (dep)...........................  .....................  1.0                     2.7x10-11               1.0x103                 2.7x10-8
V-48..............................  Vanadium (23)........  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
V-49..............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
W-178.............................  Tungsten (74)........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
W-181.............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
W-185.............................  .....................  1.0x104                 2.7x10-7                1.0x107                 2.7x10-4
W-187.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
W-188.............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Xe-122............................  Xenon (54)...........  1.0x102                 2.7x10-9                1.0x109                 2.7x10-2
Xe-123............................  .....................  1.0x102                 2.7x10-9                1.0x109                 2.7x10-2
Xe-127............................  .....................  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
Xe-131m...........................  .....................  1.0x104                 2.7x10-7                1.0x104                 2.7x10-7
Xe-133............................  .....................  1.0x103                 2.7x10-8                1.0x104                 2.7x10-7
Xe-135............................  .....................  1.0x103                 2.7x10-8                1.0x1010                2.7x10-1
Y-87..............................  Yttrium (39).........  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Y-88..............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Y-90..............................  .....................  1.0x103                 2.7x10-8                1.0x105                 2.7x10-6
Y-91..............................  .....................  1.0x103                 2.7x10-8                1.0x106                 2.7x10-5
Y-91m.............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Y-92..............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Y-93..............................  .....................  1.0x102                 2.7x10-9                1.0x105                 2.7x10-6
Yb-169............................  Ytterbium (70).......  1.0x102                 2.7x10-9                1.0x107                 2.7x10-4
Yb-175............................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Zn-65.............................  Zinc (30)............  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Zn-69.............................  .....................  1.0x104                 2.7x10-7                1.0x106                 2.7x10-5
Zn-69m............................  .....................  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Zr-88.............................  Zirconium (40).......  1.0x102                 2.7x10-9                1.0x106                 2.7x10-5
Zr-93 (b).........................  .....................  1.0x103                 2.7x10-8                1.0x107                 2.7x10-4
Zr-95.............................  .....................  1.0x101                 2.7x10-10               1.0x106                 2.7x10-5
Zr-97 (b).........................  .....................  1.0x101                 2.7x10-10               1.0x105                 2.7x10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
a [Reserved]
b Parent nuclides and their progeny included in secular equilibrium are listed in the following:
 Sr-90 Y-90
 Zr-93 Nb-93m
 Zr-97 Nb-97
 Ru-106 Rh-106
 Cs-137 Ba-137m
 Ce-134 La-134
 Ce-144 Pr-144
 Ba-140 La-140
 Bi-212 Tl-208 (0.36), Po-212 (0.64)
 Pb-210 Bi-210, Po-210
 Pb-212 Bi-212, Tl-208 (0.36), Po-212 (0.64)
 Rn-220 Po-216
 Rn-222 Po-218, Pb-214, Bi-214, Po-214
 Ra-223 Rn-219, Po-215, Pb-211, Bi-211, Tl-207
 Ra-224 Rn-220, Po-216, Pb-212, Bi-212, Tl-208(0.36), Po-212 (0.64)
 Ra-226 Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
 Ra-228 Ac-228
 Th-226 Ra-222, Rn-218, Po-214
 Th-228 Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
 Th-229 Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
 Th-nat Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
 Th-234 Pa-234m
 U-230 Th-226, Ra-222, Rn-218, Po-214
 U-232 Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
 U-235 Th-231
 U-238 Th-234, Pa-234m
 U-nat Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
 U-240 Np-240m
 Np-237 Pa-233
 Am-242m Am-242
 Am-243 Np-239
c [Reserved]
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of
  transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident
  conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.

[[Page 3814]]

 
g These values apply to unirradiated uranium only.


                                                        Table A-3.--General Values for A1 and A2
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                             A1                             A2                  Activity       Activity        Activity       Activity
                              -------------------------------------------------------------- concentration   concentration    limits for     limits for
           Contents                                                                            for exempt     for exempt        exempt         exempt
                                    (TBq)           (Ci)           (TBq)           (Ci)      material  (Bq/ material  (Ci/   consignments   consignments
                                                                                                   g)             g)             (Bq)           (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Only beta or gamma emitting    1 x 10 -1       2.7 x 100       2 x 10 -2      5.4 x 10 -1    1 x 10 -1      2.7 x 10 -10    1 x 10 -4      2.7 x 10 -7
 radionuclides are known to
 be present.
Only alpha emitting            2 x 10 -1       5.4 x 10 0      9 x 10 -5      2.4 x 10 -3    1 x 10 -1      2.7 x 10 -12    1 x 10 3       2.7 x 10 -8
 radionuclides are known to
 be present.
No relevant data are           1 x 10 -3       2.7 x 10 -2     9 x 10 -5      2.4 x 10 -3    1 x 10 -1      2.7 x 10 -12    1 x 10 3       2.7 x 10 -8
 available.
--------------------------------------------------------------------------------------------------------------------------------------------------------


           Table A-4.--Activity-mass Relationships for Uranium
------------------------------------------------------------------------
                                             Specific Activity
Uranium Enrichment \1\ wt % U-235 --------------------------------------
             present                      TBq/g               Ci/g
------------------------------------------------------------------------
0.45.............................  1.8 x 10 -8         5.0 x 10 -7
0.72.............................  2.6 x 10 -8         7.1 x 10 -7
1................................  2.8 x 10 -8         7.6 x 10 -7
1.5..............................  3.7 x 10 -8         1.0 x 10 -6
5................................  1.0 x 10 -7         2.7 x 10 -6
10...............................  1.8 x 10 -7         4.8 x 10 -6
20...............................  3.7 x 10 -7         1.0 x 10 -5
35...............................  7.4 x 10 -7         2.0 x 10 -5
50...............................  9.3 x 10 -7         2.5 x 10 -5
90...............................  2.2 x 10 -6         2.8 x 10 -5
93...............................  2.6 x 10 -6         7.0 x 10 -5
95...............................  3.4 x 10 -6         9.1 x 10 -5
------------------------------------------------------------------------
\1\ The figures for uranium include representative values for the
  activity of the uranium-234 that is concentrated during the enrichment
  process.


    Dated in Rockville, Maryland, this 29th day of December, 2003.

For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 04-35 Filed 1-23-04; 8:45 am]
BILLING CODE 7590-01-P