[Federal Register Volume 63, Number 203 (Wednesday, October 21, 1998)]
[Proposed Rules]
[Pages 56098-56125]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 98-28066]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50, 52 and 72

RIN 3150-AF94


Changes, Tests, and Experiments

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission is proposing to amend its 
regulations concerning the authority for licensees of production or 
utilization facilities, such as nuclear reactors, and independent spent 
fuel storage facilities, to make changes to the facility or procedures, 
or to conduct tests or experiments, without prior NRC approval. The 
proposed rule would clarify which changes, tests and experiments 
conducted at a licensed facility require evaluation, and the criteria 
that determine when NRC approval is needed before such changes to a 
licensed facility can be implemented. The proposed rule would also add 
definitions for terms that have been subject to differing 
interpretations, reorganize the rule language for clarity, and revise 
the criteria for when prior NRC approval is needed. The Commission is 
also seeking comment on several specific issues as discussed below.

DATES: Submit comments by December 21, 1998. Comments received after 
this date will be considered if it is practical to do so, but the 
Commission is able to assure consideration only for comments received 
on or before this date.

ADDRESSES: Send comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001. ATTN: Rulemakings and 
Adjudications Staff.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland, between 7:45 a.m. and 4:15 p.m. Federal workdays.

FOR FURTHER INFORMATION CONTACT: Eileen McKenna, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone (301) 415-2189. (emm@nrc.gov) or Naiem Tanious, 
Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear 
Regulatory Commission, Washington DC 20555-0001, telephone (301) 415-
6103 (nst@nrc.gov).

SUPPLEMENTARY INFORMATION:
I. Background
II. Proposed Rule Topics and Issues
    A. Organization of the rule requirements
    B. Change to the facility as described in the Safety Analysis 
Report
    C. Change to the procedures as described in the Safety Analysis 
Report
    D. Tests and experiments not described in the Safety Analysis 
Report
    E. Safety Analysis Report
    F. Probability of occurrence or consequences of an accident or 
malfunction of equipment important to safety previously evaluated in 
the safety analysis report may be increased
    G. More than a minimal increase in probability or consequences
    H. Possibility of an accident of a different type from any 
previously evaluated in the Safety Analysis Report may be created
    I. Possibility of a malfunction of a different type from any 
previously evaluated in the Safety Analysis Report may be created
    J. Margin of safety as defined in the basis for any technical 
specification is Reduced
    K. Safety Evaluation
    L. Reporting and record keeping requirements
    M. Part 72 changes
III. Section by Section Analysis
IV. Commission Voting Record on SECY-98-171
V. Rule Language Proposed by the Nuclear Energy Institute
VI. Request for Public Comments
VII. Availability of Documents and Electronic Access
VIII. Finding of No Significant Environmental Impact
IX. Paperwork Reduction Act Statement
X. Regulatory Analysis
XI. Regulatory Flexibility Certification
XII. Backfit Analysis
XIII. Criminal Penalties
XIV. Compatibility Agreement State Regulations

I. Background

    The existing requirements governing the authority of production and 
utilization facility licensees to make changes to their facilities and 
procedures, or to conduct tests or experiments, without prior NRC 
approval are contained in 10 CFR 50.59. (Comparable provisions exist in 
10 CFR 72.48 for licensees of facilities for the independent storage of 
spent nuclear fuel and high-level radioactive waste. This proposed 
rulemaking affects the requirements for 10 CFR parts 50, 52 and 72; for 
simplicity, the discussion will focus primarily on the language in 10 
CFR 50.59). These regulations provide that licensees may make changes 
to the facility or procedures as described in the safety analysis 
report, or conduct tests or experiments not described in the safety 
analysis report, without prior Commission approval, unless the proposed 
change, test or experiment involves a change to the Technical 
Specifications incorporated in the license or an unreviewed safety

[[Page 56099]]

question. Section 50.59(a)(2), as currently codified, states:

    ``A proposed change, test or experiment shall be deemed to 
involve an unreviewed safety question (i) if the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report may be increased; or (ii) if a possibility for an 
accident or malfunction of a different type than any evaluated 
previously in the safety analysis report may be created; or (iii) if 
the margin of safety as defined in the basis for any technical 
specification is reduced''.

The rule also specifies record keeping and reporting requirements 
associated with such changes, tests or experiments.
    In order to understand the reasons for the provisions of the 
current rule, and how the Commission proposes to revise it, it is 
helpful to understand how this process fits within the overall 
requirements undergirding licensing and oversight of nuclear reactors.

Overview of Licensing Process

    The application for an operating license includes the final safety 
analysis report (FSAR) which is to contain: a description of the 
facility; the design bases and limits on operation; and the safety 
analysis for the structures, systems, and components (SSC) and of the 
facility as a whole. The safety analysis emphasizes performance 
requirements, analytical bases and technical justifications, and 
evaluations that show how safety functions will be accomplished. Design 
bases include the specific functions that the SSC need to perform, the 
parameters that need to be controlled to assure the function, and the 
range of values for these parameters. As part of the FSAR, the 
applicant is required to propose, for NRC approval, Technical 
Specifications(TS) that will become part of the license.
    The NRC issues a license after finding, among other things, that 
the plant has been built according to its design and can be operated 
within its design limits. The NRC prepares a safety evaluation report 
that documents the basis for its findings, including its review of the 
design information provided in the FSAR (and supporting documents) and 
the applicable acceptance criteria (established either in regulations, 
standards or guidance documents). In some cases, the NRC staff performs 
independent analyses to confirm the adequacy of the facility design to 
meet regulatory requirements. One example of this practice is the staff 
calculation of radiological consequences (doses) for design basis 
accidents.
    The licensee is required to operate the facility in accordance with 
NRC regulations and with requirements contained in the license. The 
license describes the facility in general terms, and includes specific 
conditions imposed on the facility and the licensee, as well as 
incorporates the TS. Section 50.36 of the regulations defines for 
inclusion in the TS, those limits and parameters of most immediate 
significance for protection of public health and safety: safety limits, 
limiting safety system settings, limiting conditions for operation, 
surveillance requirements, and design features to which changes would 
have a significant effect on safety, and administrative controls. The 
TS are derived from the safety analysis, evaluations, and design bases 
described in the FSAR. Any changes to the TS must receive NRC review 
and approval before they are made.
    Engineering evaluations demonstrate that the fundamental safety 
principles of the plant design are met. Design basis events play a 
central role in plant design. These are a combination of postulated 
challenges and failure events against which plants are designed to 
ensure adequate and safe plant response. Design basis events are 
defined as conditions of normal operation, anticipated operational 
occurrences and design basis accidents, external events and natural 
phenomena for which the plant has been designed to ensure the integrity 
of the pressure boundary, the capability to shutdown safely, and the 
capability to prevent or mitigate the consequences of accidents. For 
events with high frequency, NRC requires that consequences be low (such 
as by preventing fuel damage). For more severe, but less probable 
accidents, the allowable consequences are higher, but must still meet 
the regulatory guidelines established in 10 CFR part 100. Adequacy of 
the reactor design is evaluated by consideration of postulated design 
basis events viewed as sufficiently credible that the facility should 
be designed to prevent or mitigate their effects.
    During the design process, plant response is evaluated using 
assumptions that are intended to be conservative to account for 
uncertainties in analysis or data. In the Final Safety Analysis Report 
(FSAR), analyses are done conservatively to account for uncertainties 
in the design, construction, and operation of nuclear power plants. 
These conservatisms are introduced into FSAR analyses in numerous ways. 
For example, some computer codes model systems and processes in a 
simplified but bounding fashion. Analysis input assumptions are 
typically worst case values (consistent with the design and operating 
limits) of instrument drift or error, temperature, pressure, fluid 
volume and enthalpy, flow rate, system response time, heat transfer 
rate and heat capacity, reactivity coefficients, power history and 
decay heat. An FSAR analysis also typically assumes the worst-case 
single-active failure of equipment.
    National standards and other regulatory policies, such as defense-
in-depth, constitute additional engineering considerations that 
influence plant design and operation. Commensurate with expected 
frequency and consequences of challenges to the system, defense-in-
depth could require: (1) Multiple means to accomplish safety functions 
and prevent release of radioactive material (multiple barriers); (2) 
reasonable balance among prevention of core damage, prevention of 
containment failure and consequence mitigation; (3) system redundancy; 
(4) independence; and (5) diversity.
    Various margins exist in a facility design. These margins are based 
on, for example, assumptions of initial conditions, conservatisms in 
computer modeling and codes, allowance for instrument drift and system 
response time, redundancy and independence of components in safety 
trains, and plant response during operating transient and accident 
conditions. Margin is provided by meeting codes and standards or 
alternatives approved for use by NRC, including the safety analysis 
acceptance criteria in the FSAR and in supporting analyses. Not all 
margin that exists falls within the purview of ``reduction in margin of 
safety \1\ as defined in the basis for any technical specification.''
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    \1\ Margin of safety is not defined in the regulations, although 
it is mentioned in Sec. 50.34(a) (``the margins of safety during 
normal operations and transient conditions anticipated during the 
life of the facility''); Sec. 50.92(c) (``No significant hazards 
considerations if the proposed amendment would not involve a 
significant reduction in a margin of safety'') as well as 
Sec. 50.59.
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    When a plant is licensed, the NRC states in its Safety Evaluation 
Report (SER) why it found each FSAR analysis acceptable. An FSAR 
analysis may be accepted because it was considered to be adequately 
conservative and because the NRC's acceptance criteria for that 
analysis are met. Frequently, the SER states specific conditions the 
NRC relied upon for concluding that the analysis was conservative. 
Examples of such conditions may be the use of an NRC-approved computer 
code, correlation, or setpoint methodology, specific limitations on one 
or more input assumptions, or penalties put into a calculation to 
account for uncertainties. In addition to being stated in a plant-

[[Page 56100]]

specific SER, these conditions may be found in other safety evaluations 
such as for an analysis method proposed by a topical report.
    Changes to the basis for licensing occur over the life of the plant 
through promulgation of new rules, plant-specific license amendments 
and other analyses and reviews that may be conducted, such as in 
response to NRC bulletins and generic letters. The NRC prepares a 
safety evaluation for many of these issues based upon either licensee 
requests for changes or licensee responses to NRC requests for 
information. The licensee is required to periodically update the final 
safety analysis report to reflect effects of these changes so that the 
safety analysis report (as updated) remains a complete and accurate 
description and analysis of the facility such that it can serve as the 
reference document for evaluation of changes made under 10 CFR 50.59.

10 CFR 50.59 Evaluation Process

    Section 50.59 was promulgated in 1962 to allow licensees to make 
certain changes that affect systems, structures, components, or 
procedures described in the SAR without prior approval provided certain 
conditions were met. In 1968, the rule was revised to modify some of 
the criteria for when approval was required. The intent of the 
Sec. 50.59 process is to permit licensees to make changes to the 
facility, provided the changes maintain the level of safety documented 
in the original licensing basis, such as in the safety analysis report. 
The process is thus structured around the licensing approach of design 
basis events (anticipated operational occurrences and accidents); 
safety-related mitigation systems, and consequence calculations for the 
design basis accidents. Margins and equipment functionality, 
reliability and availability also may be impacted by facility changes. 
Therefore, the criteria for requiring NRC approval were directly 
related to: (1) Preserving licensing assumptions concerning initiation 
of design basis events by not allowing a different type of initiating 
event or probability of occurrence larger than previously considered; 
(2) preserving effectiveness (reliability) of the mitigation systems by 
not allowing introduction of different equipment malfunctions and by 
limiting increases in probability of malfunction, or reductions in the 
margin of safety (which reflects the capability of the system); and (3) 
preserving acceptability of consequences by limiting increases in 
consequences of the postulated design basis events.

Implementation Guidance

    In 1989, an industry guidance document, NSAC-125, ``Guidelines for 
10 CFR 50.59 Safety Evaluations'' was published to assist licensees in 
the conduct of the evaluations required under Sec. 50.59. The NRC 
neither endorsed nor disapproved this document. While the staff 
concluded that the evaluation process established in NSAC-125 was 
generally sound, the staff was unable to endorse the document because 
of some inconsistencies between the implementation guidance and the 
language of Sec. 50.59.
    On October 31, 1997, the Nuclear Energy Institute (NEI) submitted 
for staff review a revised guidance document, NEI 96-07, ``Guidelines 
for 10 CFR 50.59 Safety Evaluations.'' This document is an updated 
version of NSAC-125 that NEI modified in response to some of the staff 
positions, and other implementation issues arising from licensee use of 
the NSAC-125 guidance. Along with the submittal of the guidance 
document, NEI included an industry-wide initiative that would require 
industry adoption and implementation of the revised guidance by June 
1998. The NRC provided comments to NEI concerning this guidance in a 
letter dated January 9, 1998. This letter noted that certain aspects of 
this guidance were unacceptable for implementation of Sec. 50.59 as 
presently written.
    Staff efforts to develop guidance on implementation of Sec. 50.59 
were prompted by a reassessment of the 10 CFR 50.59 evaluation process, 
conducted in 1995, that examined existing guidance and practice, with 
the goal of identifying how the process could be improved, or where 
additional guidance was needed. The staff provided an action plan to 
the Commission on April 15, 1996, outlining the actions the staff 
proposed to complete with respect to guidance and oversight of 
implementation of Sec. 50.59. The staff review identified a number of 
areas in which the meaning of the rule language is not clear, or where 
staff and industry interpretations (such as those in NSAC-125) are 
different. In SECY-97-035, dated February 12, 1997, the staff forwarded 
to the Commission proposed regulatory guidance on implementation of 
Sec. 50.59. In this SECY, the staff presented positions on a number of 
topic areas. These positions in some cases reaffirmed existing 
regulatory practice or clarified staff expectations, and in other 
areas, established positions where guidance did not previously exist. 
In its proposed guidance, the staff compared its proposed regulatory 
guidance to industry guidance contained in NSAC-125. In accordance with 
a Commission Staff Requirements Memorandum dated April 25, 1997, the 
staff guidance was published in the Federal Register as draft NUREG-
1606 (Proposed Regulatory Guidance Related to Implementation of 10 CFR 
50.59), for public comment on May 7, 1997 (62 FR 24947).
    In response to the Federal Register notice, many comments were 
submitted that voiced strong opposition to a number of the positions 
proposed by the staff. These comments were summarized in Attachment 1 
to SECY-97-205, Integration and Evaluation of Results from Recent 
Lessons-Learned Reviews, dated September 10, 1997. Since that time, the 
NRC has conducted a more detailed review of the comments and concludes 
that some issues can be resolved through guidance, while in other 
areas, rulemaking is necessary to clarify the implementation issues. A 
copy of this analysis of comments is available for review in the NRC 
Public Document Room. As noted, the staff concluded that rulemaking was 
necessary to resolve some of the issues associated with implementation 
of the rule.

II. Proposed Rule Topics and Issues

    The NRC is proposing rulemaking on Sec. 50.59 (and Sec. 72.48) to 
address a number of issues concerning implementation of the current 
rule, and suitability of the criteria that determine when an unreviewed 
safety question exists. The implementation issues primarily relate to 
cases involving judgment as to whether a proposed change requires NRC 
approval before it can be implemented. The differing interpretations of 
the rule as it relates to an increase in probability of an accident, or 
an increase in consequences have contributed to disputed inspection and 
enforcement findings. Too stringent an interpretation of the meaning of 
the requirements could result in diversion of licensee and staff 
resources for review of inconsequential changes. Too high a threshold 
for NRC review could lead to erosion of safety margins without NRC 
review, particularly from the cumulative effect of more than one 
change. In developing the proposed rule, the Commission has carefully 
weighed these matters in trying to establish an appropriate threshold 
for NRC review.
    Conforming changes are proposed in other portions of the rules, 
including Sec. 50.66, 50.71(e) for production and utilization 
facilities licensed under part 50. Conforming changes are also

[[Page 56101]]

required in Sec. 72.212(b)(4) and Appendices A and B to part 52 (Design 
Certification Rules for ABWR and System 80+ respectively).
    In addition, the Commission is proposing to make parallel changes 
applicable to facilities for independent spent fuel storage facilities 
licensed in accordance with part 72. These changes are included in the 
sections below (in some cases, the discussion of the issue focuses on 
Sec. 50.59 for simplicity; except where noted, the discussion is also 
applicable to the changes for Sec. 72.48). As part of the proposed 
changes to part 72, the Commission is also proposing to extend the 
change control process authority granted to ISFSI or MRS license 
holders (in Sec. 72.48) to holders of NRC Certificates of Compliance 
(CoC) for a spent fuel storage cask design.
    In addition to changes to the requirements within Secs. 50.59 and 
72.48, the Commission is also proposing to rearrange certain provisions 
of these rules to provide a more logical structure. These changes do 
not affect the substance of the requirements, but rather affect only 
where they are located and how they are stated. These organizational 
changes are discussed first, followed by discussion of each of the 
issues where revisions to requirements are proposed by this rulemaking. 
The proposed rule revisions are presented in the order that the issues 
currently arise in the regulations.

A. Organization of the Rule Requirements

    The organizational changes being proposed include the following:
(1) Applicability
    In the existing rule, language concerning applicability to 
different facilities is contained in three different paragraphs. These 
facilities are: Production and utilization facilities (including power 
and non-power reactors) that are authorized to operate, and reactors 
(both power and non-power) that have permanently ceased operations. The 
Commission proposes to place all of these provisions in one paragraph 
that is clearly labeled ``Applicability.'' 2
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    \2\ Section 50.59(a) refers to holders of a license authorizing 
operation of a production or utilization facility. Section 50.59(d) 
explicitly refers to power reactor licensees who have submitted 
certification of permanent cessation of operation required under 
Sec. 50.82(a)(1)(i). As noted in Sec. 50.82(a)(iii), for power 
reactors whose licenses were modified to allow possession but not 
operation, before the effective date of this rule (that is of 
Sec. 50.82), the certification of Sec. 50.82(a)(1)(i) shall be 
deemed to have been submitted. Section 50.59(e) refers to non-power 
reactors whose license no longer authorizes operation. The net 
effect is that Sec. 50.59 applies to both power and nonpower 
reactors, whether authorized to operate or no longer authorized to 
operate (and to other production or utilization facilities).
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(2) Form of prior Commission approval
    Existing Sec. 50.59(a) refers to the need for prior Commission 
approval of changes, tests, and experiments under certain conditions, 
but the method of receiving that approval is not discussed until 
paragraph (c), which states that the licensee shall submit an 
application for amendment under Sec. 50.90. The Commission proposes to 
combine these two paragraphs and to revise the regulation to state more 
clearly that a licensee must apply for and obtain a license amendment, 
pursuant to Sec. 50.90, before implementing such changes, tests, or 
experiments. This organizational change to the rule of combining 
(existing) paragraphs (a) and (c) will also facilitate some of the 
other proposed changes, such as the criteria for when approval is 
needed.
(3) Criteria for needing Commission approval of changes, tests and 
experiments and Unreviewed Safety Question (USQ) designation
    The Commission proposes to remove the reference in the rule to the 
term ``unreviewed safety question'' and instead to refer to the need to 
obtain a license amendment. The Commission believes that the 
terminology of ``USQ'' has sometimes led to confusion about the purpose 
of the evaluation required by Sec. 50.59. Some licensees have concluded 
that if they determined a change was safe, there could be no need for 
NRC approval.
    The Commission notes that the purpose of performing evaluations 
against the criteria specified in Sec. 50.59 is to identify possible 
changes that might affect the basis for licensing of the facility so 
that any changes that might pose a safety concern are either reviewed 
by the NRC or not implemented by the licensee. This evaluation process 
will thus distinguish those changes which by their nature do not raise 
safety concerns and therefore do not require prior NRC approval to 
confirm their safety, from those that must be reviewed by the NRC to 
independently confirm their safety before implementation. To avoid 
confusion between a determination of safety and a determination of the 
need for NRC approval, the Commission proposes to revise Sec. 50.59 to 
delete use of the term ``unreviewed safety question'' and instead to 
list the criteria (in new Sec. 50.59(c)(2)) that require prior 
Commission approval, in the form of a license amendment. It is also 
noted that many facility technical specifications refer to unreviewed 
safety question determinations and such TS should ultimately be revised 
in accordance with the final wording of Sec. 50.59. The deletion of 
reference to USQ also requires a number of conforming changes to other 
parts of the regulations, including Part 52 (Appendices A and B), in 
which the term is presently used.
    This proposed rule would revise the existing compound statements 
contained with the evaluation criteria to state each specific criterion 
individually. This will make the regulation more consistent with how it 
is generally implemented by licensees. Changes to the criteria are 
discussed in the sections below.
    Finally, the Commission would simplify existing Sec. 50.59(c) by 
removing the following statement: ``The holder of a license . . . who 
desires (1) a change to its technical specifications . . . shall submit 
an application for amendment of his license pursuant to Sec. 50.90.'' 
This statement refers to changes to the TS not associated with a 
change, test or experiment. The Commission concludes that a more 
suitable place for this provision is within Sec. 50.90, and therefore 
as part of this rulemaking, proposes to modify Sec. 50.90 to state that 
if a licensee wishes to amend its license (including the TS 
incorporated into it), the licensee must file an application as 
specified in Sec. 50.90. Revised Sec. 50.59(c)(i) would be revised to 
state that if a proposed change, test, or experiment would involve a TS 
change, the Sec. 50.90 process must be followed in order to change the 
technical specification such that the proposed change, test or 
experiment may be implemented.

B. Change to the Facility as Described in the Safety Analysis Report

    Section 50.59 states that ``changes to the facility as described in 
the safety analysis report'' must be evaluated to determine whether 
prior approval is needed before implementation. As discussed in NUREG-
1606 and in the comment discussions, a common understanding between the 
NRC and the industry on what constitutes a ``change to the facility as 
described in the safety analysis report'' is necessary for effective 
functioning of the review process. Guidance on preparation of 
Sec. 50.59 evaluations provides the means for review of the effects of 
changes, but these reviews are not conducted if the activity is not 
considered to be a ``change . . .'' The Commission concludes that 
modification of an existing provision (e.g., SSC, design requirement, 
analysis method or

[[Page 56102]]

parameter), additions, and removals (physical removals or non-reliance 
on a system to meet a requirement) are all changes to the facility as 
described in the final safety analysis. The Commission believes that 
additions to the facility which were not previously evaluated, could 
adversely impact facility performance and the bases upon which the NRC 
previously determined the acceptability of the design as described in 
the SAR. Accordingly, the Commission concludes that additions should be 
considered ``changes to the facility as described in the SAR'' in order 
to assure that such changes are subject to evaluation using the 
Sec. 50.59 criteria for determining whether prior NRC review and 
approval are necessary.
    Differences in interpretation have occurred about whether changes 
that do not actually change the physical plant (the ``hardware'') 
require a Sec. 50.59 evaluation. As an example, consider a change being 
made to the basis (documented in the SAR) for demonstrating adequacy of 
the facility without a physical change to the facility. Such changes 
might include changes to evaluative methods, acceptance standards, 
procurement specifications, or other information for SSC described in 
the FSAR. The Commission believes that Sec. 50.59 does apply to the 
requirements for design, construction and operation, and the safety 
analyses for the facility that are documented in the FSAR. Section 
50.34(b), ``Final safety analysis report,'' requires the FSAR to 
contain a presentation of the design bases and the limits on its 
operation, a description and analysis of the SSC of the facility, with 
emphasis upon performance requirements, the bases, with technical 
justifications therefore, upon which such requirements have been 
established, and the evaluations required to show that safety functions 
will be accomplished. The original licensing decision was based in part 
upon the margins provided by performance requirements, analysis methods 
and assumptions described in the SAR, and reviewed by the staff in the 
SER. Therefore, the Commission concludes that changes to such 
information (e.g., performance requirements, methods of operation, the 
bases upon which the requirements have been established, and the 
evaluations) should be considered to constitute a change to the 
``facility as described in the SAR'' in order to assure that such 
changes are subject to evaluation using the Sec. 50.59 criteria for 
determining whether prior NRC review and approval are necessary.
    If changes to methods and assumptions were not controlled, a 
licensee might revise its analyses and then subsequently conclude that 
a later facility change did not require NRC approval because the 
results of the (new) analysis with this change were bounded by the 
previous analysis. This proposed rulemaking would add definitions in 
Sec. 50.59 of ``change'' and of ``facility as described in the final 
safety analysis report(as updated)'' to more explicitly establish that 
evaluation is required for changes to the analyses and bases for the 
facility as well as for physical or hardware changes to the facility.
    Accordingly, the Commission proposes to add the following as 
definitions in section Sec. 50.59:
    Change means a modification, addition, or removal.
    Facility as described in the final safety analysis report (as 
updated) means (i) the structures, systems, and components (SSC) that 
are described in the final safety analysis report (as updated), (ii) 
design or performance requirements or methods of operation for such SSC 
required to be included or described in the final safety analysis 
report (as updated), and (iii) evaluations or methods of evaluation 
required to be included in the FSAR (as updated) for such SSC that 
demonstrate that their intended functions will be accomplished or that 
their design bases can be met.
    The Commission endorses the staff's previously stated position (in 
draft NUREG-1606) about what constitutes a single change, as compared 
to packaging of several changes with offsetting effects. Interdependent 
changes (i.e., where a second change is caused by the first, with 
respect to function or performance), can be treated as a single change, 
whereas treating as one change the combination of changes (whether to 
the facility directly or to the safety analysis) to offset one that 
would otherwise require prior approval is not an appropriate 
application of Sec. 50.59.

C. Change to the Procedures as Described in the Safety Analysis Report

    The Commission proposes to provide a definition of ``procedures as 
described in the safety analysis report'' in order to have definitions 
in the rule for all the major terms and criteria. This definition would 
include the evaluations demonstrating that requirements are met, such 
as assumed operator actions and response times.
    The Commission also notes that Sec. 50.34(b) states that the final 
SAR is to contain the managerial and administrative controls to be used 
to meet Appendix B (Quality Assurance), and plans for coping with 
emergencies, per Appendix E. Section 50.59 applies to changes to 
procedures as described in the SAR. Quality assurance and emergency 
planning program requirements are subject to the change control 
provisions of Secs. 50.54(a) and 50.54(q) respectively. Based on this 
set of rule provisions, it could be inferred that changes to quality 
assurance or emergency plans would require both a Sec. 50.59 evaluation 
and a Sec. 50.54 [either (a) or (q)] evaluation. The Sec. 50.54 
3 regulations provide criteria and reporting requirements 
specific to the plans and which were promulgated after Sec. 50.59. To 
reduce duplication of effort, the Commission proposes that changes to 
these programs be governed by Sec. 50.54 requirements, and that a 
Sec. 50.59 evaluation would not be required unless other information 
described in the FSAR is also being changed. The proposed rule would 
add language to specifically exclude from the scope of Sec. 50.59 
changes to procedures where other more specific requirements and 
criteria have been established by regulation for controlling these 
changes (e.g., for information required by Sec. 50.34(b)(6) (ii) and 
(v)), through a provision in the Sec. 50.59(c)(1) of the proposed rule.
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    \3\ Section 50.54(p) establishes change control requirements for 
safeguards contingency plans. While these plans are part of the 
application submitted pursuant to Sec. 50.34, they are not part of 
the FSAR, and thus Sec. 50.59 would not apply to these plans.
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    The proposed definition for ``procedures as described in the final 
safety analysis report (as updated)'' is as follows:

    Procedures as described in the final safety analysis report (as 
updated) means information in the final safety analysis report (as 
updated) regarding how systems, structures and components are 
operated and controlled (including assumed operator actions and 
response times), including assumed operator actions and response 
times, and information on conduct of operations.

D. Tests and Experiments Not Described in the Safety Analysis Report

    Section 50.59 also discusses the conduct of tests or experiments 
not described in the safety analysis report. ``Test'' is, of course, 
subject to many meanings including both routine verifications of 
function, and also more unusual evolutions. In the former category, 
there are many tests that are conducted that are not explicitly 
described in the SAR. For example, a licensee conducts tests of 
component and system performance that verify the

[[Page 56103]]

SSCs perform the functions as described or required. (Performance of 
tests is typically controlled by procedure.) However, there also may be 
tests of new materials or means of plant operation that may put the 
plant in a situation that has not been previously evaluated and that 
could affect the capability of SSC to perform their required functions. 
The existing rule was designed to ensure that the latter type of tests 
would be reviewed before they were conducted. Therefore, to assure that 
there is clear definition with respect to the tests that are subject to 
prior NRC review and approval before they are conducted, the Commission 
proposes that a definition of ``tests and experiments not described in 
the safety analysis report'' be provided in Sec. 50.59 as follows:

    Tests or experiments not described in the final safety analysis 
report (as updated) means any activity where the reactor or any of 
its systems, structures, or components are used or controlled in a 
manner which cannot be shown to be within (i) the controlling 
parameters of their design bases as described in the final safety 
analysis report (as updated) or (ii) consistent with the analyses in 
the final safety analysis report (as updated).

E. Safety Analysis Report

    In developing the proposed rule changes, the Commission noted the 
varying references to the safety analysis report within related 
sections of part 50. For example, in Sec. 50.59, the phrase used is 
``safety analysis report,'' in Sec. 50.66, the reference is to the 
``updated final safety analysis report;'' and Sec. 50.71(e) refers to 
the updated FSAR. (Other sections and parts generally refer to the 
final safety analysis report (e.g. part 55), but this is not 
universally true (e.g. Sec. 50.54(a)). For purposes of Sec. 50.59, 
``safety analysis report'' refers to the current revision of the FSAR, 
so that the changes are evaluated against the most complete and 
accurate description of the facility. When performing evaluations, a 
licensee needs to consider changes already made for which the FSAR 
update has not yet been submitted to the NRC. The Commission emphasizes 
the need for as current a reference base as possible for Sec. 50.59 
evaluations, in order that the evaluations appropriately consider other 
changes already made that may have impacted the facility or procedures. 
However, a licensee is not required to submit an update to its FSAR in 
the form specified by Sec. 50.71(e) except at the required frequency. 
To enhance consistency, the Commission is proposing to revise the rule 
language in these sections to add a definition of the final safety 
analysis report (as updated) and to clarify in the evaluation criteria 
that evaluations need to account for changes made through other 
processes that have not yet been included in an update to the FSAR. The 
Commission did not use ``Updated FSAR'' for this purpose in order to 
take into account two special circumstances: (1) Nonpower reactors, who 
are not required to submit updates to the FSAR, although they still 
need to consider other changes previously made when performing 
Sec. 50.59 evaluations, and (2) a plant licensed to operate, during the 
period between initial licensing and the first update. This revision is 
reflected in the definitions in the earlier sections and in the 
following sections. The definition also refers to ``Final Hazards 
Summary Report,'' which is the applicable document for some early 
plants whose application was submitted before the regulatory term 
``safety analysis report'' was adopted.
    The proposed definition is as follows:

    Final safety analysis report (as updated) means the final safety 
analysis report (or Final Hazards Summary Report) submitted in 
accordance with Sec. 50.34, as amended and supplemented, and as 
modified as a result of changes made pursuant to Sec. 50.59 and 
Sec. 50.90, and, as applicable, Sec. 50.71 (e) and (f).

F. Probability of Occurrence or Consequences of an Accident or 
Malfunction of Equipment Important to Safety Previously Evaluated in 
the Safety Analysis Report may be Increased

    The current language of the rule states that an unreviewed safety 
question exists when the probability of occurrence or consequences of 
an accident or malfunction of equipment important to safety previously 
evaluated may be increased [emphasis added]. Many of the concerns with 
current implementation relate to the appropriate interpretation of the 
words ``probability of occurrence . . . or consequences . . . may be 
increased.'' In the draft NUREG-1606, the NRC staff stated that the 
plain reading of the words would mean that uncertainty about whether 
there has been an increase must lead to the conclusion that the 
criterion is met. As a result of trying to deal with the question of 
uncertainty, licensees were placed in the position of having to prove 
there could not be an increase, even when there was no reason to 
believe that the proposed change, test or experiment would have that 
effect. A similar problem was experienced in considering whether the 
possibility of an accident or malfunction of a different type may be 
created.
    Many of the commenters on the staff's proposed positions viewed 
this as overly restrictive and stated that it would result in many 
changes requiring prior NRC approval that are below the level of 
significance warranting such review. The position espoused in the 
revised industry guidance document (NEI 96-07) is that an increase in 
probability or consequences must be discernable in order for approval 
to be needed. The Commission concludes that the plain reading of the 
existing rule language is not consistent with this interpretation.
    Although the current rule language would not permit discernable 
increases in probability or consequences, the Commission has concluded 
that at minimum, this would be a reasonable standard for requiring 
prior approval of changes, tests or experiment for increases in 
probability of occurrence of an accident or malfunction. The existing 
rule language dates from early in the development of reactor 
regulation, where with the knowledge base at the time, the then-AEC 
found it appropriate to set a very low threshold for changes. Over the 
last thirty years, the Commission has garnered experience with 
implementation of Sec. 50.59 and insights from probabilistic risk 
assessments, both of which indicate that this threshold can be adjusted 
without adversely impacting safety. Further, the analytical 
capabilities to calculate probabilities have greatly advanced, such 
that the effect of even minor changes on probabilities can be 
evaluated. Therefore, the Commission proposes to revise existing 
paragraph Sec. 50.59(a)(2)(i) of the rule by replacing ``may be 
increased'' with ``would result in more than a minimal increase,'' in 
order to provide that there must be a clearly discernable change to 
require approval, the ``minimal increase'' concept is described in the 
next section. As noted above, the (a)(2) paragraph would be broken into 
four statements and renumbered as (c)(2)(i) through (iv).

G. More than a Minimal Increase in Probability or Consequences

    The Commission notes that Sec. 50.59 permits changes that do not 
otherwise require approval (such as would be the case if the provisions 
being changed are in TS or license, quality assurance or emergency 
plans, or inservice inspection and testing programs). Because the 
information being revised is of less immediate importance to public 
health and safety, and in consideration of the conservatisms in NRC 
design and analysis requirements, acceptance criteria, and the 
precision with which safety analyses are performed, ``minimal'' 
variations in probability of occurrence or consequences of accidents 
and malfunctions should not affect the

[[Page 56104]]

basis for the licensing decision. This conclusion is based upon the 
qualitative consideration of probability during plant licensing; 
accident probabilities were assessed in relative frequencies; equipment 
failures were generally postulated to gauge the robustness of the 
design, without estimating their likelihood of occurrence. Therefore, 
minimal increases in probability could not even have been identifiable, 
and could not impact the conclusions reached about acceptability of the 
facility design. Radiological consequences for accidents are calculated 
and reported at a level of precision such that minimal increases also 
would not impact the safety determination. The Commission therefore 
concludes that the proposed criteria would provide reasonable assurance 
that those changes that would affect the NRC's basis for licensing 
would be identified as requiring NRC approval before implementation. 
The revised criteria would also provide some degree of flexibility for 
licensees to make changes with smaller impacts without the need to 
obtain a license amendment.
    On the other hand, the Commission intends to limit the amount of 
increase in probability or consequences of accidents such that it 
remains substantially less than a ``significant increase'' as referred 
to in Sec. 50.92 (in accordance with Sec. 50.92, a license amendment 
involving a significant increase in the probability or consequences of 
an accident previously evaluated involves a ``significant hazards 
considerations;'' any hearing for an amendment constituting a 
``significant hazards consideration'' must be completed prior to the 
grant of the amendment.) The standard in the proposed rule is 
qualitative (probability or consequences no more than minimally 
increased). The intent of this proposed rule is to allow changes that 
are small enough that they would not affect the facility's licensing 
basis, or adversely affect safety performance. While the proposed rule 
would allow minimal increases, licensee still must meet applicable 
regulatory limits and other acceptance criteria to which they are 
committed (such as contained in Regulatory Guides, etc.) Because the 
``more than minimal'' standard allows for there to be a discernable 
increase, NRC needs to establish a point beyond which one would 
conclude that the increase is not minimal. The following guidance is 
offered, including values as to when the Commission would conclude that 
the revised criteria are not met. Quantitative calculations are not 
required except for those instances in which a licensee offers other 
than qualitative arguments as part of its evaluation.

Probability of Occurrence of an Accident

    The current guidance in NEI 96-07 states: ``Where a change in 
probability is so small or the uncertainties in determining whether a 
change in probability has occurred are such that it cannot be 
reasonably concluded that the probability has actually changed (i.e. 
there is no clear trend towards increasing the probability), the change 
need not be considered an increase in probability.'' The Commission 
believes this satisfies the proposed NRC standard.
    In order to be considered as a minimal increase, the resulting 
probability (considering the change, test or experiment) must still 
satisfy the event frequency classification provided in the licensee's 
FSAR (as updated), e.g., for an anticipated operational occurrence 
(expected once a year) or for a design basis accident (not expected 
during life of plant, but sufficiently credible to require mitigation).

Probability of Equipment Malfunction

    The Commission believes that the probability of malfunction is more 
than minimally increased if a new failure mode as likely as existing 
modes is introduced. The determination should be made either at the 
component level, or consistent with the failure modes and effects 
analyses, taking into account single failure assumptions, and the level 
of the change being made.
    Guidance in NEI 96-07 states: ``Where a change in probability is so 
small or the uncertainties in determining whether a change in 
probability has occurred are such that it cannot be reasonably 
concluded that the probability has actually changed (i.e. there is no 
clear trend towards increasing the probability), the change need not be 
considered an increase in probability.'' The Commission believes this 
satisfies this criterion.
    The probability of malfunction of equipment important to safety 
previously evaluated in the FSAR (as updated) is no more than minimally 
increased if ``design bases'' assumptions and requirements are still 
satisfied (i.e., the seismic or wind loadings, qualification 
specifications, procurement requirements). As part of this guidance, 
note that NRC concludes that licensees can treat changes in external 
hazard design requirements as potentially affecting equipment 
malfunction probability rather than as ``accident probability.''

Consequences of Accident or Malfunction

    Guidance in NEI 96-07 states: ``Where a change in consequences is 
so small or the uncertainties in determining whether a change in 
consequences has occurred are such that it cannot be reasonably 
concluded that the consequences have actually changed (i.e. there is no 
clear trend towards increasing the consequences), the change need not 
be considered an increase in consequences.'' The NRC believes this 
satisfies the revised NRC standard.
    If a licensee has performed an analysis with certain bounding 
assumptions, and the change would increase a specific parameter from 
its present value to a different value that is still bounded by the 
value assumed in the analysis, NRC concludes that such a change 
satisfies the criteria of no more than a minimal increase in 
consequences.
    As a quantitative measure, the Commission is considering some 
options. One would be to establish that a 0.5 rem increase in 
calculated dose as a result of the change be used to assess whether a 
minimal increase has occurred. This range of change would generally be 
in the decimal place for accident analyses where doses are reported in 
rem. The facility must still satisfy applicable acceptance values 
(e.g., the SRP) or regulatory requirements (e.g., part 100) for the 
particular accident. If a licensee would need to change its design 
basis assumptions or analytical methods, or both, to demonstrate that 
the change in consequences is less than 0.5 rem, then the NRC does not 
view the change as minimal and would expect the licensee to submit a 
license amendment for such a change.
    In addition, the Commission is considering a graduated approach, 
consistent with the concept of ``minimal'' being small enough so as not 
to impact the basis for acceptability. When the facility is far from 
the limit, a larger increase can be accommodated without concern about 
impact on the basis for acceptability. The values proposed take into 
account such factors as differences between licensee calculated values 
and staff estimation of existing performance, potential for a single 
change with a large increase, or for several ``minimal'' increases to 
approach the regulatory limits. The specific proposal offered for 
comment is:

[[Page 56105]]

    Example using 300 rem thyroid dose as the limit.

----------------------------------------------------------------------------------------------------------------
      Existing calculated dose             ``Minimal'' change             Pre-change          After the  change
----------------------------------------------------------------------------------------------------------------
<50% of limit.......................  10% increase.....  140 rem...............  170 rem.
80% of limit.............  5% increase......  205 rem...............  220 rem.
more than 80%.......................  1% increase (NTE   245 rem...............  248 rem.
                                       limit).
----------------------------------------------------------------------------------------------------------------

    A third option under consideration, similar to option 2, would 
limit the fraction of remaining margin that can be consumed by a 
particular change. By defining ``minimal'' as being 10% of the 
remaining margin between current conditions and acceptance guidelines, 
the amount of change would decrease as the limit is approached, and the 
limit could not be exceeded.

Cumulative Effect

    The Commission is concerned about the cumulative effect of minimal 
increases. Since some increases are allowed, the Commission believes 
that the proposed process would place greater importance on: (1) 
Complete and accurate SAR updating; (2) the licensee's evaluation 
process taking into account other changes made since last update; (3) 
the licensee's screening process examining plant changes to determine 
whether they are indeed changes requiring evaluation; and (4) reporting 
requirements so that staff can assess the ongoing nature of cumulative 
impact.
    The issue then becomes how the NRC can best oversee the process 
such that several ``minimal'' changes do not result in unacceptable 
results. The Commission has decided to require licensees to report 
effects of changes in a different manner to facilitate evaluation of 
cumulative effect, as discussed in a later section on reporting 
requirements, in which the Commission proposes to require that the SAR 
update in accordance with Sec. 50.71(e) discuss the effects of the 
changes upon calculated doses and other information.

H. Possibility of an Accident of a Different Type from any Previously 
Evaluated in the Safety Analysis Report may be Created

    As noted in Section F above, the uncertainty connected with 
demonstrating that no accident or malfunction may have been created is 
a major source of confusion and difficulty in implementing the existing 
rule; and is unnecessary for purposes of identifying when NRC review of 
a change is needed. Accordingly, the Commission proposes that the 
language in existing Sec. 50.59(a)(2)(ii) be revised as discussed below 
in this section and the following one. As noted earlier, the Commission 
is proposing to separate the requirements into distinct criteria for 
clarity. This criterion would now read ``if a possibility for an 
accident of a different type from any previously evaluated in the final 
safety analysis report (as updated) is created.'' Under the proposed 
rule, a license amendment would be needed only if the licensee 
reasonably concluded that the possibility of an accident of a different 
type is created. This contrasts with the current rule, which would 
require a license amendment if the licensee is uncertain or unable to 
reasonably conclude that a new accident of a different type is not 
created. The Commission concludes that this proposed rule change will 
still identify those proposed changes, tests, or experiments that the 
NRC should review, without also including other changes of lesser 
significance that may be viewed as meeting the existing criteria.

Need for Definition of Accident

    In determining whether a proposed change requires prior NRC 
approval under Sec. 50.59, the rule refers to whether ``accidents'' 
previously evaluated in the SAR are impacted, or whether an accident of 
a different type may be created (see also Sec. 50.92 criteria for ``no 
significant hazards consideration)''. Those accidents evaluated in the 
SAR, that is, those events that a plant must show that it can 
withstand, are derived from a number of regulatory requirements, and 
the safety analyses are included in the FSAR.
    The regulations and NRC guidance documents, refer to ``a design 
basis accident'' (Sec. 50.36), to design basis events (Sec. 50.49), to 
loss-of-coolant accidents (Appendix A), to anticipated operational 
occurrences (Appendix A) and to accidents that could result in release 
of significant quantities of radioactive fission products (part 100). 
The PSAR, and by extension the FSAR, pursuant to Sec. 50.34, is to 
contain ``analysis and evaluation of the design and performance of SSC 
of the facility with the objective of assessing the risk to public 
health and safety resulting from operation of the facility and 
including determination of (i) the margins of safety during normal 
operations and transient conditions anticipated during the life of the 
facility and (ii) the adequacy of SSC provided for the prevention of 
accidents and the mitigation of the consequences of accidents.'' RG 
1.70 states that the FSAR is to include postulated anticipated 
operational occurrences; postulated off-design transients that induce 
fuel failures above those expected for normal operational experience, 
and design basis accidents. The Standard Review Plan for Chapter 15, 
refers to anticipated operational occurrences and to postulated 
accidents, and also to ``transients and accidents'' (the SRP notes that 
other events, such as response to external phenomena, are covered in 
other chapters).
    Design basis accident(s) has been used in regulatory practice both 
singularly and generally. The regulations also include the concept of a 
design basis accident (DBA), for purposes of evaluating siting, which 
is an assumed fission product release, based upon a major accident that 
would result in potential hazards not exceeded by those from any 
accident considered credible. Such accidents have generally been 
assumed to result in substantial meltdown of the core with subsequent 
release of appreciable quantities of fission products. The set of 
``accidents'' that a plant must postulate for purposes of FSAR design 
and safety analyses, including LOCA, other pipe ruptures, rod ejection, 
etc., are often referred to as ``design basis accidents''.
    The terms of accidents and transients are often used in regulatory 
documents (as for example in Chapter 15 of the Standard Review Plan), 
where transients are viewed as the more likely, low consequence events 
and accidents as more serious. In the context of probabilistic risk 
assessment, transients are typically viewed as initiating events, and 
accidents as the sequences that result from various combinations of 
plant and safety system response.
    However, the meaning of the term ``accident'' as it is used more 
generally in Part 50, is somewhat obscured by the

[[Page 56106]]

use of the term ``design basis event.'' In Sec. 50.49, design basis 
event is defined as:

normal operations including anticipated operational occurrences, 
design basis accidents, external events, natural phenomena 
(earthquakes, tornados, hurricanes, floods, tsunami and seiches), 
for which the plant must be designed to ensure safety-related 
functions.

    In view of the range of language presently used to describe the 
types of events evaluated as part of the licensing basis, the 
Commission is contemplating the need to clarify its intent as to the 
extent of events that are within the purview of the criteria in 
Sec. 50.59 and in Sec. 72.48). For purposes of stimulating discussion, 
the Commission offers two proposals. One would be to set forth a 
definition for the term ``accident'' as follows:

an initiating event or combination of events and/or conditions that 
could occur from equipment failure, human error, natural or manmade 
hazards which challenges the integrity of one or more fission 
product barriers (fuel, reactor coolant system, release of 
radionuclides (confinement/containment)), required to be analyzed 
and/or accounted for by the Commission and addressed in the 
licensee's safety analysis report.

    Such a definition would make it clear that the Commission's intent 
in referring to ``accidents'' in Sec. 50.59 (and in Sec. 72.48) is to 
refer to the design basis accidents that are addressed in the SAR. The 
second approach is to add the phrase ``design basis accident'' into the 
existing criteria. This could be done for each of the three criteria 
that refer to ``accident'' or just for the one on accident of a 
different type. Since the criteria on probability and consequences also 
contain language about ``previously evaluated in the SAR,'' there may 
be less need for a reference to ``design basis accident'' in these 
criteria. The proposed rule language includes use of the phrase 
``design basis accident'' in the one criterion, for purposes of 
obtaining public comment.

I. Possibility of a Malfunction of a Different Type from any Previously 
Evaluated in the Safety Analysis Report may be Created

    In a similar fashion, the Commission proposes to modify the 
remaining part of existing Sec. 50.59(a)(2)(ii), concerning 
malfunctions of a different type by creating a new criterion that would 
read ``if a possibility for a malfunction of equipment important to 
safety with a different result than any evaluated previously in the 
final safety analysis report (as updated) is created.'' This criterion 
involves three revisions to the existing rule. The first change is the 
use of the phrase ``is created'' which would require a determination 
that the possibility has been created, rather than uncertainty as to 
exclusion.
    The second change is to insert the words ``of equipment important 
to safety.'' The existing rule does not provide this characterization 
within paragraph (ii), but it is included in paragraph (i). It has 
generally been inferred that the statement in paragraph (ii) is an 
abbreviated version of that in paragraph (i). A review of the history 
of the 1968 rulemaking adopting revisions to Sec. 50.59 did not 
disclose any discussion suggesting that the Commission intended to 
distinguish between the (a)(2)(i) and the (a)(2)(ii) criteria with 
respect to the scope of equipment covered. Therefore, the Commission 
concludes that the rule was intended to apply to the same scope of 
equipment in each cases, and therefore, proposes to include the words 
in this criterion to eliminate any doubt.
    The final change is being proposed in response to the comments on 
the staff-proposed guidance (NUREG-1606) on the interpretation of 
malfunction (of equipment important to safety) of a different type. The 
commenters believe that the cause of the malfunction should be a 
consideration in determining whether the probability of the malfunction 
may have increased, and that a malfunction of a different type would 
only be created if the effects of the malfunction are not already 
bounded by the FSAR analysis. The recent industry guidance states that 
if a component were subject to failure from a new failure mode but the 
failure of the component is already considered in the safety analysis, 
then there would not be a failure of a different type. The Commission 
does not agree that the industry interpretation is consistent with the 
rule as written, which refers to creation or possibility of a 
malfunction of a different type, not of a different result. However, 
the Commission recognizes that in its reviews, equipment malfunctions 
are generally postulated as potential single failures to evaluate plant 
performance; thus, the focus of the NRC review was on the result, 
rather than the cause/type of malfunction. Unless the equipment would 
fail in a way not already evaluated in the safety analysis, there is no 
need for NRC review of the change that led to the new type of 
malfunction. Therefore, as the third change in Sec. 50.59(a)(2)(ii), 
the Commission is proposing to change the phrase ``of a different 
type'' to ``with a different result''. Therefore, this criterion would 
read: ``if a possibility for a malfunction of equipment important to 
safety with a different result . . . is created.''
    In implementing this position, attention must be given to whether 
the malfunction is evaluated at the component level or the overall 
system level. While the evaluation should take into account the level 
that was previously evaluated in terms of malfunctions and resulting 
event initiators or mitigation impacts, it also needs to consider the 
nature of the change. Thus for instance, if failures were previously 
postulated on a train level because the trains were independent, a 
change that introduces a cross-tie might need to be evaluated to see 
whether new outcomes have been introduced. The staff has provided 
guidance on this issue in Generic Letter (GL) 95-02, concerning 
replacement of analog systems with digital instrumentation. The GL 
states that in considering whether new types of failures are created, 
this must be done at the level of equipment being replaced--not at the 
overall system level. Further, it is not sufficient for a licensee to 
state that since failure of a system or train was postulated in the 
SAR, any other equipment failure is bounded by this assumption, unless 
there is some assurance that the mode of failure can be detected and 
that there are no consequential effects (electrical interference, 
materials interactions, etc), such that it can be reasonably concluded 
that the SAR analysis was truly bounding and applicable. Otherwise, the 
Commission would conclude that there was increase in probability of 
malfunction or that a malfunction with a different result has been 
created.

J. Margin of Safety as Defined in the Basis for any Technical 
Specification is Reduced

    Two criteria in the current regulations (Sec. 50.59) specifically 
focus upon accidents and equipment malfunction (creation, consequences 
and likelihood) as the measures for determining when a change requires 
prior NRC approval. However, the phrases ``margin of safety'' and ``as 
defined in the basis for any technical specification'' in the third 
criterion have been the subject of differing interpretations because 
the rule does not define what constitutes a margin of safety or a basis 
for any technical specification in the context of Secs. 50.59 and 
72.48. In addition, some have questioned the need for the third 
criterion on ``margin of safety.''
    The Commission has under consideration a number of proposals on 
margin. In the proposed rule text specifically being offered for 
comment, one option has been inserted so that commenters can examine 
the

[[Page 56107]]

relationship of this aspect of the proposed rule to other changes being 
offered. This should not be viewed as meaning that this option is 
preferred by the Commission. The range of options under consideration 
is discussed in more detail below.
    Questions of margin are commonly judged in terms of the degree of 
confidence that the response of the facility, or of particular SSC, to 
postulated challenges is acceptable. Various margins exist in a 
facility design. These margins are based on, for example, assumptions 
of initial conditions, conservatisms in computer modeling and codes, 
allowance for instrument drift and system response time, redundancy and 
independence of components in safety trains, and plant response during 
operating transient and accident conditions. Margin to conditions that 
might be detrimental to safety is also determined by establishing 
acceptance criteria to be met for response to various accidents and 
transients. Acceptance criteria are established at a value that 
accounts for uncertainty about physical properties and other 
variability and thus provides margin to unacceptable plant conditions. 
Margins are built into the facility to account for routine plant 
fluctuations and transients. Margins are also built into the plant to 
establish the regulatory envelope within which a plant has demonstrated 
its ability to respond to a spectrum of design basis accidents. It is 
in this category termed the ``regulatory envelope,'' that the NRC 
believes that regulatory oversight of changes in margin may be needed 
from the standpoint of Sec. 50.59. Thus the Commission notes that not 
all margins fall within the purview in which changes to the margin 
require prior NRC approval. As part of this rulemaking, the Commission 
wants to clarify which margins fall within the regulatory envelope and 
how possible reductions in margin resulting from facility or procedure 
changes, or from conduct of tests and experiments should be evaluated.
    In defining in the rule a standard for NRC review and approval of 
changes to margins in the regulatory envelope, the Commission may want 
to preserve the NRC's ability to review changes when there is a 
potentially significant reduction in a margin of safety,\4\ but clearly 
would not want to unduly affect licensee operations. Therefore, for 
this proposed rulemaking, the Commission is offering the public the 
opportunity to comment on a range of options for treating margin. 
Commenters are requested to present opinions about the merits, or 
concerns about the specific proposals, or both, and also to offer any 
other suggestions for wording.
---------------------------------------------------------------------------

    \4\ In accordance with 10 CFR 50.92(c)(3), license amendments 
involving a significant reduction in a margin of safety do not meet 
the criteria for a ``no significant hazards consideration'' 
determination; thus, changes involving a significant reduction in a 
margin of safety are not to be performed under 10 CFR 50.59.
---------------------------------------------------------------------------

Option 1: Control Inputs to Analyses and Methods that Establish TS

    The Commission believes it is reasonable to interpret the specific 
reference to ``basis for any technical specification'' in the 1968 
rulemaking that added the ``margin of safety'' criterion as preserving 
the margins in the analyses that established the TS requirements. For 
instance, the minimum plant performance conditions and configurations 
stated in the TS are the limiting conditions for operation, limiting 
safety system settings, and safety limits. Margins of safety exist 
within the safety analyses as a result of the specific input 
assumptions, methods, or other limits that were used. These parameters 
and methods were proposed by the licensee and reviewed by NRC to 
account for uncertainties, instrumentation response, and ranges of 
possible operating conditions. Because Sec. 50.59 requires prior NRC 
approval for a change to the TS, a change that could invalidate the 
basis upon which the TS values were established should also receive 
prior approval. In accordance with this interpretation, changes that 
invalidate these specific conditions described in the FSAR for analyses 
that established the TS requirement (such as a limiting condition of 
operation, or a limiting safety system setting) would reduce the margin 
of safety associated with the TS.
    Under this option, the Commission would conclude that the analyses 
and information in the FSAR establish the basis for the margins of 
safety for the TS. Thus, the Commission would propose to add a 
definition for ``reduction in margin of safety associated with any 
technical specification'' and to conform the criterion for needing a 
license amendment in new Sec. 50.59(c)(2). The existing terminology of 
``basis for any TS'' would be replaced by ``associated with any TS.''
    The following definition would be added:

    Reduction in margin of safety associated with any technical 
specification means that the input assumptions, analytical methods, 
acceptance conditions, criteria and limits of the safety analyses, 
presented in the final safety analysis report (as updated), that 
established any technical specification requirement, are altered in 
a nonconservative manner.

    Although this option would maintain the safety analyses that 
underlie the TS, this approach would also have the effect of giving 
input values and assumptions the weight of TS, which is inconsistent 
with the philosophy in Sec. 50.36 of establishing TS only on those 
values of most immediate safety importance. In many instances, changes 
to inputs can be accommodated by other available margins so that the 
licensing envelope is preserved.

Option 2: Delete ``margin of safety'' as a Criterion.

    Under this option, the Commission would delete any criterion 
focusing upon margins. Instead, the Commission would rely upon the 
other criteria in Sec. 50.59, as well as the regulatory requirement 
that all changes to TS be reviewed and approved by the NRC, to assure 
that there are no significant adverse changes to margins in design and 
operation. The Commission would argue that there is no need for prior 
review of changes that do not satisfy any of the other evaluation 
criteria in view of ``risk-informed'' insights and greater 
understanding of the margins that exist through meeting the body of 
regulatory requirements. The Commission seeks comment on whether any of 
the other evaluation criteria should be revised were this approach to 
be adopted.

Option 3: Control margins associated with results of analyses

    Instead of focusing on the inputs to safety analyses, another 
interpretation would be to examine the results of the safety analyses, 
and to determine whether changes to operational characteristics or 
other information described in the FSAR (as updated) would reduce the 
level of protection afforded by the TS (i.e., by the limiting safety 
system settings and limiting conditions of operation), as reflected in 
the results of safety analyses.
    As part of the licensing review for a facility, the NRC established 
a level of required performance (which will be referred to in this 
discussion as acceptance criteria) for certain physical parameters, 
such as those that define the integrity of the fission product barriers 
(fuel cladding, reactor coolant system boundary and containment). 
Satisfying these acceptance criteria (or regulatory limits) produces a 
margin of safety to loss of barrier integrity. The safety analyses 
presented in the FSAR (as updated) demonstrate that the response of the 
barriers to the postulated accidents, transients, and malfunctions 
meets the acceptance criteria. For

[[Page 56108]]

certain of these parameters, TS safety limits have been established; 
these safety limits are limits upon important process variables that 
are found necessary to reasonably protect the integrity of physical 
barriers that guard against the uncontrolled release of radioactivity.
    However, for other parameters, a licensee must determine the 
licensing basis of the parameter in question by reviewing the plant-
specific safety analyses. The acceptance criterion is that value 
approved by the NRC for a particular parameter or process variable 
(e.g., ASME Code stress limits, a departure from nucleate boiling ratio 
limit or maximum critical power ratio limit or containment design 
pressure). These acceptance criteria may be stated in the FSAR, may be 
in NRC regulations, or may be presented in the NRC Standard Review 
Plan. (Note: This approach may require some licensees to revise their 
FSAR to accurately describe the regulatory values for the set of 
critical parameters. For example, licensees would need to identify the 
expected operating or design values and then specify the minimum 
performance capabilities for the related parameters, which cannot be 
modified with NRC review).
    In constructing the requirements for controlling margin through 
consideration of results of analyses, there are three aspects to take 
into account: (a) Which results/parameters are to be controlled through 
the Sec. 50.59 process, (b) the degree of change to be allowed without 
review, and (c) how the changes should be evaluated in demonstrating 
that the criterion is satisfied.
    In the sections below, these three aspects are separately discussed 
in order to amplify upon the issues under consideration. However, any 
rule language option would need to include some provision for each of 
the three aspects.
    (a) Which parameters should be controlled?
    The margins of safety that would be controlled by the 10 CFR 50.59 
process can be characterized in different ways.

OPTION 3(A)(1)--Safety and Regulatory Limits

    The margin between regulatory limits and the failure of physical 
barriers is protected in the regulations (and also in the portion of 
the Technical Specifications (TSs) called ``safety limits''). The 
margin, as reflected in approved safety and accident analyses, between 
the protection afforded by the TSs (e.g., the limiting safety system 
settings and limiting conditions of operations) and the associated 
regulatory limits is a possible interpretation as to ``the margin of 
safety as defined in the basis for any TS'', which would be subject to 
the 10 CFR 50.59 evaluation process. Thus, one proposal under 
consideration would be to define ``margin of safety'' as follows:

    The ``margin of safety as defined in any technical 
specification'' (margin of safety) is the amount (quantitative or 
qualitative) of margin between the operation of the facility as 
described in the technical specifications and the exceedance of 
safety limits listed in the technical specifications or other 
regulatory limits. In relation to accident analysis, the margin of 
safety is typically the difference between calculated parameters 
(e.g., peak fuel clad temperature, maximum RCS pressure, etc.) and 
the associated regulatory or safety limit. The margin of safety is a 
product of specific values and limits contained in the technical 
specifications (which cannot be changed without NRC approval) and 
other values, such as assumed accident or transient initial 
conditions or assumed safety system response times, which are not 
specifically contained in the technical specifications. Any change 
to the values not specifically contained in the technical 
specifications must be evaluated for impact on the margin between 
the calculated result of an accident or transient and the safety or 
regulatory limit.

    With this option, before changing operational characteristics 
described in the UFSAR (not directly controlled by TS), a safety 
evaluation must be performed to determine, among other things, if the 
change results in a reduction in the level of protection afforded by 
the TS (margin of safety as defined in any TS). Such a reduction would 
typically occur only if the operational characteristic had been used as 
a bounding condition in the analysis upon which the selection of TS was 
based, or in analysis where the acceptability of selected TS values was 
demonstrated. Licensees could make desired changes to operational 
characteristics without prior NRC approval, provided that the change 
does not result in accident analysis results that are nearer the 
regulatory, or safety, limits than the corresponding results that the 
NRC used in evaluating the acceptability of the TS during licensing of 
the facility.

OPTION 3(A)(2)--Fission product barriers--definition

    The NRC notes that Sec. 50.36 (requirements for Technical 
Specifications) has criteria for when TS are to be provided that 
specifically are tied to design basis accident or transient analysis 
that either assumes the failure of or presents a challenge to the 
integrity of a fission product barrier. Thus, the margin as defined in 
the basis for any TS can be reasonably viewed as that margin associated 
with preserving integrity of these barriers. Therefore, the NRC is also 
considering a more explicit linkage to the response of the three 
fission product barriers generally relied upon to provide protection 
from uncontrolled release of radioactive materials from a reactor 
facility. Under such a proposal, the text of the rule would explicitly 
state that it is the response of fission product barriers (fuel, 
reactor coolant system, and containment) to accidents, transients, and 
malfunctions that is being controlled.
    The following could be given as a definition of margin of safety 
and of fission product barrier response. Regulatory guidance would 
explicitly list the parameters (for PWRs and BWRs) that are to be 
controlled.

    The margin of safety for any fission product barrier response is 
the difference between the calculated value and its associated 
acceptance criteria. Fission product barrier response means those 
parameters that must be satisfied in the event of postulated design 
basis events to demonstrate integrity of the fuel, reactor coolant 
system and containment system barriers.

    The following parameters would be included: Fuel and cladding 
performance (peak cladding temperature, or energy deposition, DNBR or 
MCPR, oxidation), RCS performance (pressure, flows, stress), and 
containment performance (peak pressure, containment leakage).

OPTION 3(A)(3)--Specified Parameters

    A variant on the previous option would be to actually list the 
parameters of interest directly in the criterion for prior review, as 
for instance, the criterion could read:

    (vii) Result in a change to the FSAR (as updated) calculated 
value of RCS peak pressure, containment peak pressure, or fuel 
performance (DNBR/MCPR, others), etc.

    This variant has the advantage of being more precise, but the rule 
language would need to be crafted to account for various reactor types.

OPTION 3(A)(4)--Include Mitigation Capability

    The Commission is interested in preserving the integrity of both 
prevention and mitigation capabilities available in the plant, and is 
therefore considering an option that would include both features within 
the ``margin'' criterion if the margin criterion is maintained. If this 
approach were adopted, the definition or the list of parameters would 
be supplemented with the performance parameters for the

[[Page 56109]]

accident mitigation capability of the plant, as for instance, ECCS 
performance (pressures, flows, actuation values), engineered safety 
feature performance (flows, pressures, spray effectiveness, system 
efficiencies).
    Finally, in conjunction with any of these approaches, the 
Commission is also considering whether there are other parameters 
important to preservation of barriers that should be explicitly 
defined. For instance, for fuel stored in spent fuel pools, or for the 
reactor during periods of shutdown or refueling, there may be other 
analysis results (water level, pool temperature) in lieu of reactor 
coolant system pressure. Therefore, the Commission seeks input as to 
whether there are other parameters of interest beyond those previously 
offered that should be included within the ``margin of safety'' 
criterion if that criterion is maintained, and how should the rule 
language be revised to specify what those parameters might be.
    (b) Determination of reduction in margin requiring review
    Once the parameters of interest are determined, it is also 
necessary to define when a reduction in margin warranting NRC review 
and approval has occurred. The Commission is evaluating options ranging 
from any ``nonconservative change in calculated values,'' to a 
``minimal change'' standard, and ultimately an option that would allow 
increases up to ``specified limits (acceptance criteria)'' for those 
parameters that may be established in the regulations or NRC guidance 
(such approaches to the limits might be controlled in a graduated 
fashion as was discussed in the section of this notice relating to 
``minimal increases''). An option for the degree of reduction would be 
paired with an option (such as one of those listed in (a) above) to 
provide the text of the rule.

OPTION 3(B)(1)--No Reduction

    One approach would be require that the safety analysis, considering 
the effect of the change, must show that the accident analysis results 
are not nearer to any safety or regulatory limit, thus, a ``no 
reduction in margin'' standard. Possible rule text:

    Changes, or the net effect of multiple changes, which result in 
a reduction in the margin of safety require prior NRC approval. 
Changes, or the net effect of multiple changes, which do not cause a 
reduction in the margin of safety do not require prior NRC approval.

OPTION 3(B)(2)--Minimal Amount--Definition of Margin Reduction

    As discussed in other sections of this notice, the Commission 
concludes that the revised rule should allow licensees some flexibility 
in making changes, through development of a ``minimal increase'' 
standard. In considering margins, the Commission is thus weighing how 
such a concept could be applied. One option would be that NRC approval 
would be required for a change, test, or experiment if the output 
values (calculated in the SAR) are altered by more than a minimal 
amount. The ``margin'' criterion would be modified to state that a 
change in calculated result of ``more than a minimal amount'' would 
require prior review and approval. Either in the rule itself, or in 
guidance, the Commission would define ``minimal amount'', modeled upon 
the options offered for minimal increases in consequences (see section 
II.G. of this notice). For example, there could be a fixed amount 
(percent change) in margin, as long as regulatory limits are still met. 
If guidance itemizes the parameters, such guidance could also customize 
how ``minimal'' should be judged for each particular parameter 
(allowing greater amounts for certain parameters depending on precision 
of calculations, sensitivity of results and other considerations).
    For instance, the definition of ``margin of safety reduction * * 
*'' might be stated as follows:

    Reduction in margin of safety means that as a result of a 
change, the [MARGIN] is altered in a nonconservative manner by more 
than a minimal amount.

OPTION 3(B)(3)--Minimal Determined With Respect to Acceptance Criteria 
(Available Margin)

    It is also possible to achieve this result by removing the language 
referring to margin of safety (and to TS), and defining ``minimal'' in 
the rule itself in terms of the results or analyses for barrier 
response, with respect to meeting the acceptance criteria for those 
barriers. For example, rule language could read as follows:
    License amendment needed if as a result of a change, test or 
experiment:

    (vii) there is more than a 10% reduction in the difference 
between the calculated value and the acceptance criteria for fission 
product barrier response to accidents evaluated in the SAR.

    If such an approach is followed, the Commission would propose to 
include a definition of acceptance criteria, such as follows:

    Acceptance criteria are those values, established by NRC 
regulation or review guidance, to which the licensee is committed 
through its FSAR (as updated), as the basis for acceptability of 
response to the postulated accident, transient or malfunction.

    (c) Evaluation of effect of the change upon analysis results.
    The Commission also notes that the results of safety analyses are 
subject to variance depending upon the assumptions, analysis methods or 
analytical techniques used. In many instances, these factors were 
reviewed by the NRC during its licensing deliberations, and their use 
may have formed part of the basis for the conclusion that acceptable 
safety margins were demonstrated. Therefore, the Commission wishes to 
ensure that proposed changes by a licensee would not invalidate these 
conclusions by requiring a demonstration that the evaluation techniques 
and analyses are suitable.
    To accomplish this, the Commission is considering having as part of 
whichever definition of ``margin of safety reduction'' is selected the 
following statement [Option 3(c)]:

    All analyses and evaluations for assessing the impacts of 
proposed changes must be performed using methodology and analytical 
techniques which are either reviewed and approved by the NRC or 
which are shown to meet applicable review guidance and standards for 
such analyses.

    The alternative to this proposed language would be to rely upon a 
licensee's design control processes under their quality assurance 
requirements and program, to provide the assurance that any evaluative 
work has been conducted with methods and techniques commensurate with 
the safety significance of the analyses being performed.

Impacts for Part 72 Changes

    Certain of the options discussed above may need to be modified for 
application to independent spent fuel storage facilities or spent fuel 
storage cask designs in Part 72. While the overall philosophy would be 
the same, the particular outputs or barriers that would be specified 
for reductions in margin would have to be defined in terms of the 
barriers against release of radioactivity afforded by fuel storage 
facilities. For instance, these might include calculated fuel 
temperature or cladding oxidation, and stresses (or pressures) on the 
cask structure. Comment is also requested on the appropriate parameters 
for facilities licensed under Part 72.

K. Safety Evaluation

    Section 50.59(b)(1) requires licensees to maintain records that 
must include a written safety evaluation that provides

[[Page 56110]]

the bases for the determination that the change, test, or experiment 
does not involve an unreviewed safety question. Section 50.59(b)(2) 
requires submittal of a report containing a brief description of any 
changes, tests, or experiment, including a summary of the safety 
evaluation of each. In the interest of emphasizing the regulatory 
purpose of the evaluation required under Sec. 50.59, which led the 
Commission to propose deletion of the term ``unreviewed safety 
question,'' the Commission proposes to delete the word ``safety'' in 
referring to the required evaluation for determining whether the 
change, test, or experiment requires a license amendment. For purposes 
of the summary report of tests and experiments submitted to NRC, the 
staff would propose that the rule specify that a summary of the 
evaluation be provided (rather than a summary of the safety 
evaluation).
    A similar change is proposed for Sec. 50.71(e), which presently 
refers to safety evaluations either in support of license amendments or 
of conclusions that changes did not involve USQs. The Commission 
proposes to change ``safety evaluation in support of license 
amendments'' to ``safety analysis in support of license amendments,'' 
to reduce confusion between the information prepared by the licensee 
for the amendment (safety analysis) and the NRC review (safety 
evaluation). The second part of this phrase would be revised to refer 
to the ``evaluation that changes did not require a license amendment in 
accordance with Sec. 50.59(c)(2) of this part.'' (In this case, it is a 
licensee evaluation against the regulatory criteria in Sec. 50.59 that 
is being referred to). In addition, other minor wording changes are 
proposed such as with respect to terminology on ``final safety analysis 
report'' and ``effects of'' (see reporting requirements discussion 
below). Conforming changes in the appendices to part 52 and in part 72 
to revise language to refer to ``evaluation'' are also proposed.

L. Reporting and Recordkeeping Requirements

    In view of the ``minimal increase'' criteria in Sec. 50.59, the 
Commission concludes that the reporting requirements for the SAR update 
should be enhanced to enable the NRC to better understand the potential 
cumulative impact of changes that might have been made since the last 
update. Therefore, the Commission proposes to supplement the reporting 
requirements on ``effects'' of changes to require that in the FSAR 
update submittal (with the replacement pages), the licensee shall 
include a description of each change affecting that part of the SAR 
that provides sufficient information to document the effect of the 
change upon the probability or consequences of accidents or 
malfunctions, or reductions in margin associated with that part of the 
SAR. Accordingly, the Commission proposes to revise Sec. 50.71(e) to 
read as follows:

    ``(e) Each person licensed to operate a nuclear power reactor 
pursuant to the provisions of Sec. 50.21 or Sec. 50.22 of this part 
shall update periodically, as provided in paragraphs (e)(3) and (4) 
of this section, the final safety analysis report (FSAR) originally 
submitted as part of the application for the operating license, to 
assure that the information included in the FSAR (as updated) 
contains the latest information developed. The submittal must 
describe the effects \1\ of: (1) All changes made in the facility or 
procedures as described in the FSAR; (2) all safety analyses and 
evaluations performed by the licensee either in support of requested 
license amendments, or in support of conclusions that changes did 
not require a license amendment in accordance with Sec. 50.59(c)(2) 
of this part; (3) all analyses of new safety issues performed by or 
on behalf of the licensee at Commission request; and (4) the net 
effect of all changes made since the last update on the safety 
analyses, including probabilities, consequences, calculated values, 
system or component performance, that are in the FSAR (as updated). 
The updated information shall be appropriately located within the 
update to the FSAR.

    \1\ Effects of changes includes appropriate revisions of 
descriptions in the FSAR such that the FSAR (as updated) is complete 
and accurate.
---------------------------------------------------------------------------

    Finally, the Commission is proposing a change to the record 
retention requirements in existing Sec. 50.59 (b)(3) (renumbered by 
this rulemaking to (c)(3)). The change would add to the requirement 
that the records of changes to the facility be maintained until the 
termination of the license, the statement ``or until the termination of 
a license issued pursuant to 10 CFR part 54, whichever is later.'' This 
change would make more clear the requirement that records must be 
maintained through the life of the facility so that they will remain 
available until such time as they are no longer needed (that is, when 
the license is terminated, not just at the end of the initial licensing 
term).

M. Part 72 Changes

    In part 72 the Commission is proposing to make conforming changes 
to Sec. 72.48 with those made to Sec. 50.59 and to expand the scope of 
Sec. 72.48 so that holders of a Certificate of Compliance (CoC) are 
also subject to it. In addition to the proposed changes to Sec. 72.48, 
the Commission proposes to make changes in other sections of part 72. 
When subpart L--Approval of Spent Fuel Storage Casks, was originally 
added to part 72, no provisions were included to address potential 
amendments of CoCs. However, regulations in this area are necessary to 
provide requirements for certificate holders in instances where a 
proposed change does not meet the tests of Sec. 72.48, and an amendment 
to the CoC is necessary. Therefore Secs. 72.244 and 72.246 would be 
added to subpart L, to provide regulations on applying for, and 
approving, amendments to CoCs. Section 72.248 would also be added to 
provide regulations for the certificate holder submitting an updated 
final safety analysis report, which would document the changes it made 
to procedures or structures, systems, and components under the 
provisions of Sec. 72.48. The Commission notes that a general licensee 
is not precluded from loading spent fuel into an approved spent fuel 
storage cask during the 90-day period allowed for the certificate 
holder to submit a final safety analysis report. This approach is the 
same as that required for part 72 license holders to update their final 
safety analysis report under Sec. 72.70. The Commission also notes, 
that for dual-purpose spent fuel casks (i.e., casks which have been 
issued CoCs for transportation and storage under parts 71 and 72, 
respectively), no regulation equivalent to Sec. 72.48 exists in part 
71. Consequently, a certificate holder could make changes to the design 
of a spent fuel storage cask under the authority of Sec. 72.48 (i.e., 
without prior NRC approval); however, if the change also affected the 
transportation aspects of the cask's design and involved a modification 
to the part 71 certificate, then NRC approval and amendment of the 
transportation CoC would be required before the cask could be used to 
transport spent fuel to another site. Additionally, a transportation 
cask CoC has a term of 5 years, compared to the 20-year term for a 
storage CoC. Consequently, the Commission envisions that most of this 
type of change would be captured during the periodic renewal of a 
transportation CoC and this delay would not have a significant adverse 
impact on a licensee's ability to transport spent fuel in a dual 
purpose cask.
    In Sec. 72.3 the definition for independent spent fuel storage 
installation (ISFSI) would be revised to remove the tests for 
evaluation of the acceptability of sharing common utilities and 
services between the ISFSI and other facilities. The existing 
requirement in Sec. 72.24(a)--Contents of application: Technical 
Information,

[[Page 56111]]

would be revised to reference shared common utilities and services in 
the applicant's assessment of potential interactions between the ISFSI 
and another facility. The Commission would remove the existing 
requirement in Sec. 72.3 for the applicant to evaluate the impact of 
sharing common utilities and services on the ``other facility.'' The 
Commission believes that evaluation of the impact on the ``other 
facility'' should not be part of the licensing process for an ISFSI. 
Rather, such evaluation should be part of the license amendment process 
for that ``other facility'' and should be performed under the 
regulations used to license that ``other facility.''
    Changes to Sec. 72.56 would be conforming changes to those made to 
Sec. 50.90. Changes to Sec. 72.70 are also conforming changes to those 
made to Sec. 50.71(e); additionally, requirements would be added to 
Sec. 72.70 on standards for submitting revised Final Safety Analysis 
Report (FSAR) pages. The Commission notes that the proposed Sec. 72.70 
would retain the requirement that the site-specific licensee submit a 
final safety analysis report at least 90 days prior to the planned 
receipt of spent fuel or high-level waste. The Commission has not 
received any requests for exemption from this regulation and believes 
that this regulation does not impose an undue burden or schedule impact 
on licensees. The proposed rule also modifies the requirements for 
filing of updates (through reference to Sec. 72.4) to be consistent 
with other changes being made to part 72. Changes to Sec. 72.216 for a 
general licensee are similar to the changes made to Sec. 72.70 for a 
site-specific licensee and are also conforming changes to those made to 
Sec. 50.71(e). The Commission also envisions that a general licensee 
who wishes to adopt a change to the design of a spent fuel storage cask 
it possesses--which was previously made to the generic design by the 
certificate holder under the provisions of Sec. 72.48--would be 
required to perform a separate evaluation under the provisions of 
Sec. 72.48 to determine the suitability of the change for itself. The 
changes to Secs. 72.9 and 72.86 are conforming changes due to the 
addition of new Secs. 72.244, 72.246, and 72.248.
    Changes to part 72 Record keeping requirements would include the 
clarification that records required by Sec. 72.48 shall also include 
determinations that significant increases in occupational exposure or 
unreviewed environmental impacts did not exist, such that a license 
amendment would have been required. (The existing language linked the 
written evaluation only to the ``unreviewed safety question'' 
determination, and thus did not explicitly require Record keeping for 
the determinations of whether the change would cause a significant 
increase in occupational exposure or a significant unreviewed 
environmental impact). Certificate holders would also be required to 
keep records of such changes as would be allowed under Sec. 72.48.
    Requirements in Sec. 72.70 would be established for reporting 
changes to procedures. The Commission notes that Sec. 72.70 presently 
requires that the update include 5 a description and 
analysis of changes in the structures, systems, and components with 
emphasis upon performance requirements; the bases, with technical 
justification therefor, upon which such requirements are based; and 
evaluations showing that safety functions will be accomplished. It also 
requires an analysis of the significance of any changes to codes, 
standards, regulations, or regulatory guides which the licensee has 
committed to meeting the requirements of which are applicable to the 
design, construction, or operation of the facility. New reporting 
requirements for certificate holders would be added in Secs. 72.244 and 
72.248, similar to existing requirements imposed on licensees in 
Secs. 72.56 and 72.70, respectively. New reporting requirements for 
general licensees would be added as Sec. 72.216(d), similar to existing 
reporting requirements for site-specific licensees in Sec. 72.70 and 
proposed requirements for certificate holders in Sec. 72.248. In both 
of these sections, the Commission is adding a requirement that the 
entity making a change to the cask, either the general licensee or the 
certificate holder, provide a copy of the submittal to the other party 
for their information.
---------------------------------------------------------------------------

    \5\ The similarity in the language between Secs. 72.24 and 
50.34(a) and between Secs. 72.70 and 50.34(b)(2) is noteworthy.
---------------------------------------------------------------------------

III. Section By Section Analysis

10 CFR Part 50

10 CFR 50.59

    As discussed in more detail above, Sec. 50.59 would be restructured 
and revised to have the following components.
    Paragraph (a)--This is a new paragraph that provides definitions of 
terms such as ``change'', ``facility as described * * *,'' in order to 
specify more clearly which changes, tests and experiments require 
further evaluation and how reductions in margin of safety are to be 
determined. The references to ``safety analysis report'' are being 
revised to ``final safety analysis report (as updated)'' to state that 
the evaluations are to be performed that take into account other 
changes made that have affected the final safety analysis report since 
its original submittal.
    Paragraph (b)--Relocation of existing applicability provisions.
    Paragraph (c)(1)--Relocation of existing provisions establishing 
which changes, tests, or experiments require evaluation, using the 
defined terms. The terminology of ``unreviewed safety question'' has 
been replaced by referring to the need to obtain a license amendment. 
This paragraph also clarifies that the licensee must submit its request 
for license amendment, and obtain the amendment prior to implementing 
those changes, tests or experiments that involve TS or otherwise meet 
the criteria for prior NRC approval as specified in (new) paragraph 
(c)(2).
    Paragraph (c)(2)--Reformatting of the evaluation requirements into 
seven distinct statements of the criteria and revision of the criteria 
for when prior NRC approval of a change, test or experiment is 
required. Specifically, language of ``more than a minimal increase'' 
was inserted in the criteria concerning increases in probability and 
consequences, and revisions to the rule requirements were made 
concerning creation of accidents of a different type and malfunctions 
of equipment with a different result. Clarification is also being 
provided that the margins of safety are those associated with TS 
requirements established by the FSAR analyses, and are not confined to 
the BASES section of the TS. These revisions clarify the criteria for 
when prior approval is needed and allow some flexibility for licensees 
to make changes that would not affect the NRC basis for licensing of 
the facility.
    Paragraph (d)(1)--Renumbered paragraph with record keeping 
requirements. Also includes change from ``safety evaluation'' to 
``evaluation.''
    Paragraph (d)(2)--Renumbered paragraph with reporting requirements.
    Paragraph (d)(3)--Renumbered and revised paragraph on retention of 
records, to cover the term of any renewed license.

10 CFR 50.66

    The proposed changes for Sec. 50.66 are to conform existing 
language referring to unreviewed safety questions, and references to 
updated final safety analysis report, to the language

[[Page 56112]]

proposed in revised Sec. 50.59 for consistency.

10 CFR 50.71(e)

    The proposed changes to this section are to conform language with 
respect to unreviewed safety question, safety evaluation, and reference 
to final safety analysis report (as updated), with the proposed 
language in Sec. 50.59, and to clarify reporting requirements relating 
to ``effects of'' changes such that cumulative effects of minimal 
increases in probability and consequences are included in the update to 
the FSAR.

10 CFR 50.90

    A portion of existing Sec. 50.59(c) would be relocated into this 
section. This change would place the requirements for changes to 
technical specifications in the rule section on amendments to licenses.

10 CFR Part 52

Appendix A and Appendix B to 10 CFR Part 52

    The proposed changes to these sections are to conform references to 
unreviewed safety question, safety evaluation and the evaluation 
criteria concerning when prior NRC approval is needed, to the language 
in the proposed revision to Sec. 50.59.

10 CFR Part 72

10 CFR 72.3

    The definition for independent spent fuel storage installation 
would be revised to remove the tests for evaluation of the 
acceptability of sharing common utilities and services between the 
ISFSI and other facilities. (Section 72.24 is also proposed to be 
revised to include this evaluation).

10 CFR 72.9

    Paragraph (b) would be revised as a conforming change to include in 
the list of information collection requirements the new reporting 
requirements in Secs. 72.244 and 72.248 for reports of changes made by 
CoC holders and for updates to the safety analysis reports by CoC 
holders.

10 CFR 72.24

    This section would be revised to reference shared common utilities 
and services in the applicant's assessment of potential interactions 
between the ISFSI and another facility (previously covered by 
Sec. 72.3).

10 CFR 72.48

    New definitions have been added for terms such as ``change'' and 
``facility as described in the Final Safety Analysis Report (as 
updated).'' The specific criteria in existing paragraph (a)(2) have 
been revised to separate out the various statements, to insert the 
language of ``more than a minimal increase,'' and to modify the 
criterion from ``malfunction of a different type'' to ``malfunction of 
a different result.'' The text for Record keeping requirements was 
revised to refer to the need for license or certificate of compliance 
(CoC) amendments, rather than involving an unreviewed safety question. 
As part of this revision, the Commission is also clarifying that the 
records shall also provide a basis for why a proposed change, test, or 
experiment did not require a license or CoC amendment with respect to 
significant increases in occupational exposure or significant 
unreviewed environmental impacts. Additionally, the term ``Final Safety 
Analysis Report (FSAR) (as updated)'' has been used to provide greater 
clarity and consistency with Sec. 50.59 and other sections of Part 72. 
The filing requirements for the summary reports are modified to be 
consistent with Sec. 72.4 (Communications).

10 CFR 72.56

    Existing Sec. 72.48 (c)(2) is being relocated into this section. 
This is a parallel change to that proposed for Sec. 50.59 and 
Sec. 50.90, wherein the Commission would place the requirements for 
changes to license conditions in the rule section on amendments to 
licenses.

10 CFR 72.70

    Paragraphs (a) and (b) would be revised to use the terms ``Final 
Safety Analysis Report,'' ``FSAR,'' and ``as updated.'' Paragraph 
(b)(2) would be revised to add changes to procedures to the annual 
updates of the FSAR. New paragraph (c) would be added to provide 
requirements on submitting revisions to the FSAR.

10 CFR 72.86

    Paragraph (b) currently includes those sections under which 
criminal sanctions are not issued. This paragraph would be revised by 
adding Secs. 72.244 and 72.246 as a conforming change to reflect that 
certificate holders who fail to comply with these new sections would 
not be subject to the criminal penalty provisions of section 223 of the 
Atomic Energy Act (AEA). New Sec. 72.248 has not been included in 
paragraph (b) to reflect that certificate holders who fail to comply 
with this new section would be subject to the criminal penalty 
provisions of section 223 of the AEA.

10 CFR 72.212(b)(4)

    The change to this section is to conform the reference to 10 CFR 
50.59 provisions, specifically to change from the terminology of 
unreviewed safety question to referring to need for license amendment 
for the facility (that is, the reactor facility at whose site the 
independent spent fuel storage installation is located).

10 CFR 72.216

    New paragraph (d) provides requirements for a general licensee to 
submit annual updates to a final safety analysis report (FSAR) for the 
cask or casks approved for spent fuel storage cask that are used by the 
general licensee. The general licensee is also required to provide a 
copy of its submittal to the certificate holder. This section is 
similar to the requirements in Secs. 72.70 and 72.248 for submission of 
annual updates to the FSAR associated with a site-specific Part 72 
licensee or a certificate holder, respectively.

10 CFR 72.244

    This new section provides requirements for a certificate holder to 
submit an application to amend the certificate of compliance (CoC). 
This section is similar to the requirements in Sec. 72.56 for licensees 
to apply for an amendment to their license.

10 CFR 72.246

    This new section provides requirements for approval of an amendment 
to a CoC. This section is similar to the requirements in Sec. 72.58 for 
approval of an amendment to a license.

10 CFR 72.248

    This new section provides requirements for submittal of annual 
updates to a FSAR associated with the design of a spent fuel storage 
cask which has been issued a CoC. This new section also provides that 
the changes to procedures and structures, systems, and components 
associated with the spent fuel storage cask and which are made pursuant 
to Sec. 72.48 would be included in the annual update. The proposed 
revisions would also require that the certificate holder provide a copy 
of the FSAR submittal to each general licensee using that cask. This 
section is similar to the requirements in Sec. 72.70 for submission of 
annual updates to the FSAR associated with a site-specific part 72 
license and new section 72.216 for general licensees to provide updates 
to the FSAR.

[[Page 56113]]

IV. Commission Voting Record on SECY-98-171

    The staff forwarded to the Commission a proposed rulemaking package 
on Sec. 50.59 and related regulations in SECY-98-171, dated July 10, 
1998. This document was placed in the Public Document Room on July 29, 
1998. Subsequently, the Commission voted to approve issuance of a 
proposed rule for public comments with several additions and changes 
that are reflected in this notice. The Commission also directed that 
the record of their decision on SECY-98-171 be included as part of this 
notice to clearly inform stakeholders on preliminary positions taken by 
the Commission. The text of the resultant staff requirements memorandum 
and of the individual Commissioner vote sheets, is presented below.

Commission SRM on SECY-98-171, Dated September 25, 1998

    The Commission has approved publication, for a 60 day public 
comment period, the proposed rulemaking that would revise 10 CFR 50.59 
and related provisions in parts 50, 52 and 72 concerning the processes 
controlling licensee changes, tests and experiments for production and 
utilization facilities and for facilities for independent storage of 
spent nuclear fuel and high-level radioactive waste. The Voting Record, 
which includes the Commissioner votes and this Staff Requirements 
Memorandum, should be published in the Federal Register notice to 
clearly inform stakeholders on preliminary positions taken by the 
Commission (Enclosed).
    The Commission also approves the staff's recommendations for 
handling violations of 10 CFR 50.59 and 72.48, including staff plans 
for exercise of enforcement discretion, while rulemaking is underway.
    The Commission requested that the staff specifically solicit public 
comment in the Federal Register notice on:
    1. A wide array of options for the margin of safety criterion 
(50.59(c)(2)(vii) in the proposed rule) and its definition including: 
(a) Deleting the criterion and definition, (b) a new definition as 
described in Chairman Jackson's vote, and (c) an option which would 
decouple the last criterion from technical specifications and focus 
instead on a new criterion relating to performance of fission product 
barriers (e.g., reactor coolant system pressure, containment pressure, 
etc), with minimal changes being allowed up to specified limits, 
perhaps utilizing a graduated approach similar to the approaches 
proposed for other criteria.
    2. Options for defining ``minimal'' as it pertains to ``probability 
of occurrence of an accident'' or ``probability of equipment 
malfunction.''
    3. The definitions of ``facility,'' ``procedures,'' and ``tests or 
experiments,'' including elimination of the definitions.
    4. A clear definition of ``accident.''
    (This action scheduled for completion October 9, 1998).
    The Commission requests the staff to complete the revised 50.59 
rule on an expedited schedule.
    (This action scheduled for completion February 19, 1999).
    All Commissioners approved in part and disapproved in part the 
proposed rulemaking on 10 CFR parts 50, 52 and 72 requirements 
concerning changes, tests and experiments and staff recommendations on 
changes to other regulations and enforcement policy, and provided 
additional comments. In their vote sheets, all Commissioners approved 
the staff's recommendations to approve publication of the proposed rule 
for public comment, and use of the enforcement discretion guidance in 
its assessment of severity levels for violations while the rulemaking 
is underway, and provided some additional comments. In particular, all 
Commissioners disapproved the staff's proposed margin of safety 
criterion (Sec. 50.59(c)(2)(vii) in the proposed rule) and its 
definition and each Commissioner provided an option for evaluation 
during the comment period. The Commissioners also specifically 
requested comments on a number of other issues. Because of the need to 
finalize this rule as expeditiously as possible and because SECY-98-171 
has already been publicly available since July 29, 1998, the Commission 
agreed to a 60 day comment period, and that the staff complete the 
revised Sec. 50.59 rule by February 19, 1999. Subsequently, the 
comments of the Commission were incorporated into the guidance to staff 
as reflected in the SRM issued on September 25, 1998.

Chairman Jackson's Comments on SECY-98-171

    I approve, in part, and disapprove, in part, the staffs proposal 
for rulemaking. I approve the staff's proceeding with issuance of the 
proposed rule language for public comment in order to support the 
expedited finalization of a revision to these processes. I disapprove 
of the specific language proposed by the staff for 
Sec. 50.59(c)(2)(vii), ``reductions in the margin of safety.''
    I agree with the recent letter from ACRS on this rulemaking, in 
that: (1) 10 CFR 50.59 can accommodate risk-informed decisionmaking. 
(2) the positions, as presented, on margin of safety may add regulatory 
burden without a commensurate safety benefit.
    I disagree with ACRS in that I believe:
    (1) The rulemaking should go out for public comment to foster 
comment on this high priority issue, and
    (2) The regulatory guidance can be worked in parallel with the 
rulemaking.
    I note that a further reason for issuing this package for public 
comment at this time is that the paper calls for the proper use of 
enforcement discretion as this rulemaking progresses, thereby providing 
further stability in the implementation of this rule in the industry.
    Further, I propose that the SRM on this SECY, and the voting 
record, be placed in the FR notice to clearly inform stakeholders on 
preliminary positions taken by the Commission.

Giving Definition to Minimal

    Attached to the recent ACRS letter was ``A Proposal for the 
Development of a Risk-Informed Framework for 10 CFR 50.59 and Related 
Matters.'' The proposal forwarded by the ACRS parallels an existing 
risk-informed approach described in Regulatory Guide 1.174. Regulatory 
Guide 1.174 describes a method for determining the level of review, 
based on severe accident implications, for proposed licensing actions. 
The proposal forwarded by the ACRS describes methodology for creating 
frequency-consequence curves for Class 1-8 accidents. The proposal 
states that existing processes could be extended to provide appropriate 
context for whether the results of a change are ``minimal.'' The 
proposal also notes that aspects of this type of approach are in use in 
the international regulatory community. The approach utilized in the 
proposal forwarded by the ACRS is consistent with the Commission 
guidance in the Staff Requirements Memorandum of March 24, 1998 on 
SECY-97-205.
    Without commenting on the specifics of the proposal forwarded by 
the ACRS, I am convinced that changes to nuclear plants can be 
evaluated in a risk-informed context. Any such approach would benefit 
from paralleling existing methodology. Careful consideration would be 
required to ensure that the ``consequence'' and ``frequency'' standards 
are appropriate for a Sec. 50.59 type application. For instance, 
``consequences'' could be evaluated at one of the following levels: 
Fractional releases, off-site or on-site doses, or

[[Page 56114]]

challenges to fission product release barriers. ``Frequency'' could be 
evaluated for Class 1-8 accidents or for design basis accidents using 
existing guidelines for risk-informed regulation. The level at which 
consequences and frequency of events were tracked would also impact the 
type of parallel, deterministic (e.g., protection of redundancy, 
defense in depth, etc.), considerations against which changes would 
have to be evaluated. For instance, evaluating consequences at the 
level of the loss of a single barrier, or occurrences of accident 
sequence initiators, might allow elimination of parallel, 
deterministic, considerations such as ``margin.''
    It is of some concern to me that the whole staff has pursued risk-
informed approaches to issues like the review of TSs, the use of Graded 
Quality Assurance, and programs like Inservice Inspection and Inservice 
Testing, the staff appears to be more reluctant to allow risk-informed 
approaches if the result is the relinquishment of review and approval 
authority. Because prior NRC review and approval impacts the cost and 
schedule of licensed activities, we must ensure that we require such 
prior review and approval only when justified or required by mandate. 
We should not limit the application of risk-informed regulation as a 
means to ensure continued NRC reviews and approvals of licensed 
activities. This message is complimentary to my oft repeated message to 
industry that the use of risk information is ``double-edged,'' that is 
that relief and additional regulatory scrutiny may both result from its 
use.

Margin of safety

    The staff proposes to provide a specific definition of ``Reduction 
in margin of safety associated with any technical specification,'' and 
to revise the current provisions of 10 CFR 50.59(a)(2)(iii) to 
explicitly refer to this definition. While I commend the staff on its 
efforts to provide clear, definitive, requirements in this proposed 
rulemaking, I am concerned that the proposed rule is not consistent 
with policy direction established by the Commission in the SRM dated 
March 24, 1998. I concur that it is important that the staff has the 
independence to (and, I believe, has the responsibility to) inform the 
Commission when there are concerns with Commission guidance (as it did 
in COMSECY 98-013). However, I believe that when the staff proposes to 
take action that is inconsistent with Commission direction, it is 
obliged to provide a clear and complete rationale for the proposed 
departure. I do not feel that the staff has met that obligation for the 
``margin of safety'' aspect of this proposed rule. However, this said, 
I do not disagree with the staff's conclusion that we should be careful 
to understand, and maintain, a consistent regulatory basis on ``margin 
of safety.'' We must proceed in a manner that does not call into 
question the existing deterministic basis for ``reasonable assurance'' 
of public safety embodied in plants Technical Specifications (TSs).
    My previous discussions with the staff have indicated that it is 
extremely difficult (and probably not legally defensible) to allow 
decreases in the ``margin of safety'' when the upper and lower limits 
between which ``margin'' may exist are not defined in relation to the 
regulatory requirements for safe operation. Based upon these 
discussions, I can only assume that the staff is hesitant to allow 
direct reductions in margin within the ``basis'' for TSs because some 
such changes could create a de-facto change in the TSs themselves. The 
staff may also be concerned by the lack of consistency in the ``margin 
of safety in the basis for TSs'' associated with the different 
generations of existing licenses (e.g., older customized TSs compared 
to improved standardized TSs), and associated with the different 
methods utilized in the technical review and approval of the TS (e.g., 
some TSs might be based on maintaining margin between accident analysis 
results and acceptance limits, while other TSs might be based on margin 
which was built into analytical techniques and methodologies used in 
the accident and safety analysis, with no ``margin'' between the 
results and the acceptance limits, etc.).
    The staff's proposed method of requiring prior agency approval to 
changes of input assumptions, analytical methods, etc., for those 
parameters which affected the selection of TSs, results in the newly 
controlled parameters being treated essentially the same way as values 
in the TSs. It also appears that implementation of the staffs proposed 
control over a broad range of parameters used in the safety analysis 
would effectively prevent any change to the facility that would result 
in a ``minimal change in consequence,'' a condition allowed elsewhere 
in the proposed rule. In other words, it is not clear what type of 
changes would successfully pass the 10 CFR 50.59 test for allowed 
``minimal increases in consequences,'' without failing the test for 
``no reductions in the margin of safety.'' I do not believe that the 
potential safety significance of all the parameters to be covered under 
the proposed definition of a reduction in the margin of safety always 
justify the requirement of prior NRC approval.
    The staff should continue to work to establish a technically sound 
method for allowing licensees to make plant changes where there is only 
``minimal'' impact on safety. If fundamental conflicts exist with 
allowing reductions in some ``margins of safety,'' especially those on 
which the validity of TSs are based, then staff should provide a clear 
explanation of this, and should address how other changes to the 
structure of the regulation, which do not create fundamental conflicts, 
can be made in a manner which achieves the Commission's objective of 
removing unnecessary burdens from licensees.
    Attachment ``A'' to this vote describes one alternate method for 
addressing the issue of ``margin of safety.'' This alternative would 
maintain existing margins of safety (associated with TSs), while 
providing greater flexibility to licensees in implementing changes to 
their facilities. This alternative is based on methodology similar to 
that described in NEI 96-07. This methodology requires evaluating the 
effect of proposed tests and changes on the accident analysis results 
(rather than inputs, as proposed by the staff), in cases where TSs are 
based on accident analysis considerations. Prior NRC approval of 
changes, tests, and experiments would be limited to those cases where 
there was a net effect on the accident analysis results. The 
alternative also recognizes the significance of the analytical 
techniques used in the safety or accident analysis, and would require 
some form of prior approval for analytical methods used to support 
changes when the change did not have prior NRC approval. This approach 
could provide staff reasonable assurance that the assumptions made by 
the license reviews are not invalidated. The staff should evaluate this 
option, along with other comments in this area, during the comment 
period.
    In considering the technical and regulatory underpinning of this 
clause of Sec. 50.59, I have become concerned that we are evaluating 
incremental changes to a provision which is not well suited to such 
changes. I am concerned that the result may be the addition of yet 
another layer of regulatory process rather than the elimination of any 
unnecessary layers. For this reason, the staff should be receptive to 
internal or public comments on feasible alternatives which eliminate 
the discussion of ``the margin of safety in the basis of TSs,'' while 
maintaining the integrity of the plant's licensing basis. I envision 
that it may be possible to eliminate the rule

[[Page 56115]]

language criteria on ``margin of safety'' if evaluations of 
``frequency'' and ``consequences'' are performed at a level of 
significance which bounds allowable ``minimal'' reductions in margin.

Accident of a Different Type

    In determining the effect of any proposed change to Sec. 50.59, it 
will be necessary to more clearly understand what an ``accident of a 
different type'' is. The staff should provide a more definitive 
definition of an accident than was included in COMSECY-98-013. The 
information provided by the staff should address, as a minimum, the 
following:
    (1) What is an ``accident'' under this section, and is it 
consistent with other existing regulations (e.g., Sec. 50.92, 
Sec. 50.34, Appendix A of part 50, etc.)?
    (2) Is an ``accident of a different type'' better described as an 
``initiating event (e.g., loss of feedwater, loss of offsite power, new 
common mode failure mechanism, etc.) of a different Type?''
    (3) What are the bounds which limit those ``accidents'' which are 
the subject of this Section (e.g., only those initiating events which, 
when evaluated using approved analytical techniques, result in 
transients with the potential to challenge fission product barriers, 
etc.)?

Procedures

    I commend staff on inserting a definition for the term ``Procedures 
as described in the final safety analysis report (as updated).'' 
However, I am concerned that the definition provided may cloud the 
distinction between: (1) Those procedures which must be screened, or 
evaluated, under Sec. 50.59, and (2) the criteria which necessitates a 
full safety evaluation. I believe that staff seeks to indicate that all 
procedures which are described as being required in the FSAR are 
subject to a Sec. 50.59 screening. The screening would identify the 
need for a full safety evaluation only if a proposed procedure change 
created a change to the ``information in the FSAR regarding how 
structures, systems, and components are operated and controlled. . . 
.'' Staff should solicit comment on this definition and clarify the 
proposed definition, as required, in the final rule.

Making the Rule Risk Informed

    I note with interest that members of the ACRS believe that there 
are substantial barriers in the existing deterministic framework of 10 
CFR part 50 to the concept of allowing ``minimal'' changes in accident 
probabilities or consequences. In my previous vote on SECY-97-205, 
``Integration and Evaluation of Results from Recent Lessons-Learned 
Reviews,'' I approved the staff's proposal to develop the framework for 
risk-informed regulatory processes. In particular, I called for the 
staff to develop a series of milestones by which the Commission could 
``chart its course in its move to more risk-informed regulatory 
processes.'' Additionally, I promoted the idea of promulgating a new 
regulation in 10 CFR part 50, that would make clear how the Commission 
uses risk information in its decision-making. In proceeding with the 
``short-term'' changes to 10 CFR 50.59 (and related regulations; 
``short-term'' actions from SECY-97-205), and in responding to the 
ACRS, the staff should re-evaluate whether the Agency should initiate 
action to provide for a risk-informed framework that would allow for 
the efficiencies to be gained through use of risk-informed, 
performance-based revisions to our regulatory processes.

Attachment ``A'' to Chairman Jackson's vote sheet on SECY-98-171

``Straw Man'' on Margin of Safety

    Regarding margin:
     The margin between regulatory limits and the failure of 
physical barriers is protected in the regulations (and also in the 
portion of the Technical Specifications (TSs) called ``safety 
limits'').
     The margin, as reflected in approved safety and 
accident analyses, between the protection afforded by the TSs (e.g., 
the limiting safety system settings and limiting conditions of 
operations) and the associated regulatory limits is ``the margin of 
safety as defined in the basis for any TS.''
     The margin between normal plant or system operation and 
the ``bounding'' assumptions used in accident analysis is below the 
threshold of safety significance that requires NRC prior approval 
for changes.
     The results of safety and accident analyses are subject 
to significant variance, depending on the analytical techniques and 
methods used in the analysis. Where a licensee wishes to make a 
change in their facility without prior NRC approval, the effects of 
the change must be evaluated using analytical techniques and methods 
which are NRC approved for the application, or which are reviewed 
and vetted (but not subject to specific NRC approval) in a NRC 
approved manner.

    Direct changes to technical specifications require prior NRC 
approval. Before changing other operational characteristics described 
in the UFSAR, a safety evaluation must be performed to determine, among 
other things, if the change results in a reduction in the level of 
protection afforded by the TS (margin of safety as defined in any TS). 
Such a reduction would typically occur only if the operational 
characteristic had been used as a bounding condition in the analysis 
upon which the selection of TS was based, or in analysis where the 
acceptability of selected TS values was demonstrated. Licensees can 
make desired changes to operational characteristics without prior NRC 
approval, provided that the change does not result in accident analysis 
results that are nearer the regulatory, or safety, limits than the 
corresponding results that the NRC used in evaluating the acceptability 
of the TS during licensing of the facility.
    This regulatory position could be codified by adding the following 
footnote to Section 50.59(a)(2)(iii):

    The ``margin of safety as defined in any technical 
specification'' (margin of safety) is the amount (quantitative or 
qualitative) of margin between the operation of the facility as 
described in the technical specifications and the exceedance of 
safety limits listed in the technical specifications or other 
regulatory limits. In relation to accident analysis, the margin of 
safety is typically the difference between calculated parameters 
(e.g., peak fuel clad temperature, maximum RCS pressure, etc.) and 
the associated regulatory or safety limit. The margin of safety is a 
product of specific values and limits contained in the technical 
specifications (which cannot be changed without NRC approval) and 
other values, such as assumed accident or transient initial 
conditions or assumed safety system response times, which are not 
specifically contained in the technical specifications. Any change 
to the values not specifically contained in technical specifications 
must be evaluated for impact on the margin between the calculated 
result of an accident or transient and the safety or regulatory 
limit. Changes, or the net effect of multiple changes, which result 
in a reduction in the margin of safety require prior NRC approval. 
Changes, or the net effect of multiple changes, which do not cause a 
reduction in margin of safety do not require prior NRC approval. All 
evaluatory work in assessing the impact of proposed changes must be 
performed using methodology and analytical techniques which are 
either reviewed and approved by the NRC or which are reviewed and 
vetted in a manner approved by the NRC.

Commissioner Diaz's Comments on SECY-98-171

    I consider this rulemaking effort to be our short term fix for the 
50.59 rule, not the longer term risk-informed rule enhancement 
discussed in SECY-97-205.
    I approve the publication of this rulemaking package for a 90-day 
public comment period, contingent upon the additions described in the 
last paragraph of my comments. I propose that the package also include 
the Commissioners' votes for public consideration. The purpose of 
issuing the rulemaking package is to expedite rulemaking by opening the 
process for

[[Page 56116]]

public comments during the Commission's continuing deliberation on this 
matter. It should be made very clear to all stakeholders that 
publication of the package is an invitation to participate in improving 
the rulemaking. In fact, I do not agree with several of the proposed 
positions in this paper, as delineated in my specific comments below.
    I agree with the staff's recommendation to remove the reference to 
``unreviewed safety question'' from Sec. 50.59 and to make conforming 
changes in parts 50, 52, and 72. I also agree with staff's proposal to 
allow a minimal increase in the probability of occurrence or 
consequence of an accident or malfunction previously evaluated, and to 
not allow the creation of an accident of a different type or 
malfunction of equipment important to safety with a different result 
than any previously evaluated.
    I agree with the ACRS comments in their June 16, 1998, letter 
regarding the definition of ``reduction in margin of safety.'' 
Notwithstanding the staff's suggestion of a possible Commission 
interpretation, the language ``altered in a nonconservative manner'' 
can still be interpreted as a de facto ``zero increase'' standard for 
the 50.59 criterion on margin of safety. I believe the risk-informed 
Sec. 50.59 approach suggested in the ACRS letter deserves serious 
consideration as part of longer term improvements and should be 
considered in the staff's response, due in February 1999, to the SRM 
for SECY-97-205.
    The current language in Sec. 50.59(a)(2)(iii) (``margin of safety 
as defined in the basis for any technical specification'') is, in fact, 
defined and bounded by the technical specifications. Therefore, as long 
as the licensee proposed change, test, or experiment under Sec. 50.59 
is not in violation of the technical specification requirements, the 
requisite margin of safety is maintained, and it is possible to 
eliminate ``reduction of margin of safety'' from the rule as a 
condition requiring prior staff approval. This change will eliminate 
the existing ambiguity in the use of Sec. 50.59 for changes with 
minimal safety significance. This alternative should also be published 
for public comment; it is consistent with the safety envelope provided 
by the technical specifications and is a straightforward improvement 
that will match with the eventual conversion to a risk-informed rule.
    I support the staff's recommended changes in the reporting and 
record keeping requirements relating to Sec. 50.59. The enforcement 
policy and its corresponding implementation guidance should be changed 
in accordance with the revised Sec. 50.59 rule. I recommend that, 
during the rulemaking period, the enforcement policy be revised to 
grant discretion (i.e., suspend issuance of Level IV violations) under 
Section VII.B.6 for those Sec. 50.59 violations of little or no safety 
significance.
    I do not agree with the recommended definitions of ``facility'', 
``procedures'', ``reduction in margin of safety'', and ``tests or 
experiments.'' These definitions appear to increase prescriptiveness at 
the input of the licensees' change process instead of the output, and 
therefore, are more broad-based than the definitions to date. I believe 
that these definitions will create more burden for the NRC and 
licensees, are not consistent with the original intent of the 
Sec. 50.59 rule, i.e., to evaluate whether the licensee proposed 
changes will result in inadequate protection of public health and 
safety, and therefore, are not necessary.
    On the other hand, the ``accident'' in the proposed revisions to 
Sec. 50.59 should be defined. The ``accident of a different type than 
any previously evaluated'' as described in the proposed 
Sec. 50.59(c)(2)(v) should be of the same safety significance as the 
``accident'' in the proposed Sec. 50.59(c)(2)(I) and (c)(2)(iii). The 
staff should determine if the anticipated operational transients and 
the postulated design basis accidents described in the FSAR form a 
sufficient basis for the Sec. 50.59 evaluation.
    The staff should continue its interactions with NEI in resolving 
the differences between the NRC's position on Sec. 50.59 implementation 
guidance and that contained in NEI 96-07. The regulatory guide for 
Sec. 50.59 that endorses a revised NEI 96-07, with exceptions and 
clarifications, as appropriate, should be developed concurrently with 
the rulemaking process.
    In summary, the staff should proceed with publishing the existing 
rulemaking package, and concurrently solicit public comment on the 
following alternatives: (1) eliminate ``reduction of margin of safety'' 
as a condition requiring prior staff approval, (2) eliminate the 
broadened definitions of ``facility'', ``procedures'', ``reduction in 
margin of safety'', and ``tests or experiments,'' and (3) clearly 
define ``accident'' in the proposed revisions to Sec. 50.59. I urge the 
staff to complete the revised Sec. 50.59 rule and the associated 
regulatory guide by the end of March, 1999.

Commissioner McGaffigan's Comments on SECY-98-171

    I approve publishing this rulemaking package for a ninety-day 
public comment period. However, like my colleagues, I do not agree with 
the staff proposal regarding ``reduction in the margin of safety 
associated with any technical specification.''
    As the Chairman points out, the definition of ``reduction in margin 
of safety * * *'' would extend the requirements for prior agency 
approval to underlying aspects (e.g., input assumptions) of parameters 
that affected the selection of technical specifications, and result in 
the newly controlled parameters being treated essentially the same way 
as values in the technical specifications. This is the wrong way to go.
    It is clear from my colleagues' and my vote that the margin of 
safety criterion (Sec. 50.59(c)(2)(vii) in the proposed rule) and the 
definition will need to be fixed in the final rule. My concern at this 
point is that the staff discuss a wide enough array of options in the 
Federal Register notice to ensure that the proposed rule will not have 
to be renoticed before being finalized. Commissioner Diaz has proposed 
to simply delete the criterion and definition as not needed. The 
Chairman has proposed essentially a new definition. Another option 
would decouple the last criterion from technical specifications and 
focus instead on a new criterion relating to performance of fission 
product barriers (e.g., RCS pressure, containment pressure. etc), with 
minimal changes being allowed up to specified limits, perhaps utilizing 
a graduated approach similar to the approaches proposed for other 
criteria. Comment should be solicited on this option as well.
    I believe that the staff has done a good job in proposing options 
for defining ``minimal'' for consequences of an accident or 
malfunction. On probability, however, the staff has essentially only 
said that NEI 96-07 satisfies the proposed NRC standard for a 
``minimal'' increase. That is a good step forward, and will bring 
regulatory stability. I believe that in choosing the word ``minimal'' 
the Commission intended to grant greater flexibility than the NEI 96-07 
``so small'' or negligible standard. The staff should continue to try 
to give better definition to ``minimal'' as it pertains to 
``probability of occurrence of an accident'' or ``probability of 
equipment malfunction'' and solicit comment on this.
    Finally, I endorse the use of enforcement discretion under Section

[[Page 56117]]

VII of the Enforcement Policy as the rulemaking proceeds for those 
Sec. 50.59 violations of little or no safety/risk significance. The 
staff should treat (vice ``consider treating'' as proposed by staff) as 
minor violations cases where the violation of existing rule 
requirements would not constitute a violation under the rule were it 
revised as proposed. I do not object to documenting such minor 
violations in inspection reports because the rule is still in a 
proposed revision stage.

V. Rule Language Proposed by The Nuclear Energy Institute

    In a letter dated November 14, 1997, the Nuclear Energy Institute 
provided to the NRC suggested language for revising 10 CFR 50.59 that 
they believed would enable the NRC to endorse NEI 96-07. This language 
is included here in this Statement of Considerations so that interested 
parties can offer comment on whether this language should be adopted by 
the NRC. The supporting information for NEI's proposal is contained in 
the referenced letter which is available for review in the Public 
Document Room.
    Specifically, NEI proposed that [existing] section 50.59(a)(2) be 
revised to read:

    (a)(2) A proposed change, test, or experiment shall be deemed to 
involve an unreviewed safety question: (i) If there is more than a 
negligible increase in the probability of occurrence of an accident 
or malfunction of equipment important to safety previously evaluated 
in the safety analysis report; or (ii) if the consequences of an 
accident or malfunction important to safety previously evaluated in 
the safety analysis report exceeds the established acceptance limit; 
or (iii) if a possibility for an accident of a different type or 
malfunction with a different result from any evaluated previously in 
the safety analysis report may be created; or (iv) if the margin of 
safety provided by any technical specification is reduced.

    In this rulemaking, the Commission is proposing to adopt certain 
aspects of the changes offered by NEI (e.g., on malfunction with a 
different result). The Commission is seeking comment as to whether 
other aspects of this proposal should be adopted. The Commission also 
offers the following observations about this proposal for consideration 
as part of the comment process:

A. Negligible Increase in Probability of Occurrence

    NEI proposes that the rule be revised to state that a change would 
be an USQ ``if there is more than a negligible increase in the 
probability of occurrence of an accident or malfunction of equipment 
important to safety previously evaluated in the safety analysis 
report.'' As discussed above, the Commission is proposing a ``more than 
minimally increased'' criterion, which is considered comparable in 
overall intent to what was proposed by NEI.

B. Increase in Consequences of an Accident or Malfunction

    NEI proposes that the rule be revised such that a change would be a 
USQ if the consequences of an accident or malfunction previously 
evaluated exceed the established acceptance limit. As NEI discusses 
further in its letter, the established acceptance limit would be the 
value that was previously reviewed and approved by the NRC generally as 
documented in the staff's safety evaluation report (SER).6
---------------------------------------------------------------------------

    \6\ Attempting to use values from the staff's SER as acceptance 
limits would be difficult since SERs were not written for the 
purpose of establishing such limits. In a literal sense, neither the 
SAR nor the SER set an ``acceptance limit.'' Rather, the SAR 
documents an applicant's/licensee's analytically derived conclusion 
that a given event has a certain consequence which is within the 
regulatory bounds set by NRC regulations. The SER is intended only 
to confirm or modify that conclusion. The SAR value as modified 
through the staff's review and approval then becomes the baseline 
for future analyses.
---------------------------------------------------------------------------

    The current industry guidance, NEI 96-07, would permit, in some 
instances, increases in consequences up to the regulatory thresholds 
(such as Part 100), without review. As discussed in (draft) NUREG-1606, 
the staff typically performs independent evaluations of radiological 
consequences of accidents, rather than an in-depth review of the 
licensee's calculations, during licensing of the plant. As a result, 
the degree of conservatism in the licensee calculations differs from 
that used in the staff's assessments. As noted above, the Commission is 
proposing to revise the rule to allow ``minimal'' increases in 
consequences without prior approval, provided that the regulatory 
limits are still met. The Commission has some concerns about allowing 
licensee changes without review, which when evaluated with licensee 
assumptions and methods, result in doses at or very close to the 
regulatory guidelines (e.g., part 100). This is because such changes, 
if reviewed with staff assumptions (or starting from the staff's 
previous estimation of the accident dose), might result in the 
regulatory guidelines not being met. Rather than allowing one change to 
result in an increase in consequences up to the guidelines, the 
Commission concludes that minimal increases, along with NRC oversight 
of cumulative effects, is the appropriate standard for review.

C. Malfunction with a Different Result

    As discussed above, the Commission is proposing to adopt this 
particular proposed change to the rule.

D. Margin of Safety Provided by Any Technical Specification

    NEI proposes to replace the existing language of ``as defined in 
the basis for any technical specifications,'' with ``as provided by any 
technical specification'' with respect to reductions in the margin of 
safety. The proposed change is intended to clarify that the margin of 
safety is not necessarily limited to information in the BASES section 
of the technical specification. NEI 96-07 guidance notes that the SAR, 
staff SERs and other licensing basis documents should be reviewed to 
determine if a proposed change would result in a reduction in margin of 
safety. NEI intended to use this rule language in conjunction with 
guidance that the margin of safety is the range of values between the 
acceptance limit reviewed by the NRC (e.g., ASME code stress limits, 
containment design pressure, etc.) and the failure point. The 
Commission is seeking comment on a range of options relating to margin 
of safety, including the option proposed by NEI.

VI. Request for Comment

    The Commission requests comments on the proposed rule, as discussed 
in Section II above. In addition, the Commission is seeking comment on 
a number of specific issues related to this rulemaking. All commenters 
are encouraged to provide specific comments on the following issue 
areas:
    1. The Commission is seeking input on a number of options relating 
to the criterion of margin of safety reduction, and its definition. 
Some possible alternatives are presented in Section II.J as being 
representative of the range of approaches under consideration, but the 
Commission is open to other proposals that commenters may wish to put 
forth as representing the best means to provide a clear understanding 
of which margins should fall within the regulatory envelope of 
requiring approval if they would be reduced as a result of a change, 
test or experiment, if the margin of safety criterion were to be 
retained.
    2. The Commission is interested in options for defining what 
constitutes a ``minimal'' increase in the probability of occurrence of 
an accident previously evaluated in the FSAR or in the probability of 
equipment malfunction (refer to Section II.G). This might include 
suggested examples of changes

[[Page 56118]]

that commenters believe represent only a ``minimal increase'' in 
probability.
    3. The Commission is interested in comments upon the proposed 
definitions for such terms as ``facility as described in the FSAR,'' 
``procedures as described in the FSAR,'' and ``tests or experiments'' 
(refer to Sections II.B, C, and D). The Commission is soliciting views 
on whether (1) definitions are necessary, (2) the proposed definitions 
are desirable, even if not necessary, and (3) whether the suggested 
definitions are clear and focused upon the appropriate changes that 
should be evaluated. In this light, the Commission is also interested 
in comments on a broader view of the scope of changes that should be 
evaluated; for instance, should the scope be linked to the SAR, or 
should the focus of changes to the facility be linked to another set of 
regulatory information?
    4. As part of the present rulemaking, the Commission is seeking 
comment on the need for a clear definition of accident as it is used in 
Sec. 50.59 to reflect the Commission's intent that the ``accidents'' 
referred to are those dealt with in the safety analysis report (see 
Section II.H of this notice for discussion of issues related to 
definition of accident).
    5. In addition to the NRC proposals in Sections II and III, the 
Commission is also interested in receiving comments on the proposals 
and language suggested by NEI (Section V).

VII. Availability of Documents and Electronic Access

    Certain documents related to this rulemaking, including comments 
received and the regulatory analysis, may be examined at the NRC Public 
Document Room, 2120 L Street NW. (Lower Level), Washington, DC NRC 
documents also may be viewed and downloaded electronically via the 
interactive rulemaking website established by NRC for this rulemaking.
    You may also provide comments via the NRC's interactive rulemaking 
web site through the NRC home page (http://www.nrc.gov). This site 
provides the availability to upload comments as files (any format), if 
your web browser supports that function. For information about the 
interactive rulemaking site, contact Ms. Carol Gallagher, (301) 415-
5905; e-mail CAG@nrc.gov.

VIII. Finding of No Significant Environmental Impact

    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, if adopted, will not have 
a significant impact on the environment. The proposed rule changes are 
of two types: those that relate to the processes for evaluating and 
approving changes to licensed facilities and those that involve the 
degree of potential change in safety for which changes can proceed 
without NRC review. The process changes being proposed will make it 
more likely that planned changes are properly reviewed and approved by 
NRC when necessary. With respect to the criteria changes, only minimal 
increases in probability or consequences of accidents (still satisfying 
regulatory limits) would be allowed without prior NRC review. All 
changes to the Technical Specifications, which are the operating limits 
and other parameters of most immediate concern for public health and 
safety, will continue to require prior NRC review and approval. Changes 
to the facility that would involve an accident of a different type from 
any already analyzed, or reductions in defined margins of safety 
require prior approval. Further, changes which result in more than 
minimal increases in radiological consequences will continue to require 
prior NRC approval, including NRC consideration of potential impact on 
the environment. Therefore, the Commission concludes that there will be 
no significant impact on the environment from this proposed rule. This 
discussion constitutes the environmental assessment and finding of no 
significant impact for this proposed rule.

IX. Paperwork Reduction Act Statement

    This proposed rule amends information collection requirements that 
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). This rule has been submitted to the Office of Management and 
Budget for review and approval of the information collection 
requirements. Existing requirements were approved by the Office of 
Management and Budget approval numbers 3150-0011 and 3150-0132.
    The proposed rule changes would affect information collection 
requirements through the existing reporting requirements in Sec. 50.59 
for a summary report of changes, tests and experiments, performed under 
the authority of Sec. 50.59 and in Sec. 50.71(e) for submittal of 
updates to the FSAR, as well as record keeping requirements. To the 
extent that the definitions provided in the proposed revisions would 
require evaluations that are not presently being performed, there may 
be an increase in record keeping and reporting. The Commission 
estimates that this is a small increment over the existing burden. On 
the other hand, some changes might be screened out as not needing 
evaluation on the basis of these definitions, and thus there would 
overall be at most a small increase in the record keeping required.
    In addition, the requirements under Sec. 72.48 are also being 
revised to explicitly require records of determinations concerning 
occupational dose and environmental impact (the existing rules required 
the evaluations but did not explicitly specify record retention 
requirements for these evaluations). The Commission does not believe 
this that this change will significantly impact record keeping burden 
because records of evaluations of changes are already required (as to 
whether they involve a USQ), and the evaluation itself is already 
required by the rule. The part 72 burden associated with the 
definitions of when evaluations are required should be significantly 
less than for Sec. 50.59 since the number of licensees is smaller and 
the expected number of changes is also smaller. Further, there is a 
recordkeeping requirement established for CoC holders who make changes 
to an approved storage cask design in accordance with Sec. 72.48.
    With respect to reporting requirements, the Commission is proposing 
to modify the FSAR update requirement to state that the updates must 
include specific information on the effects of changes made. This was 
not explicitly stated in the current rule, although it could be 
inferred that this was what the update rule intended, as follows. In 
the Statement of Considerations for Sec. 50.71(e),(45 FR 30615), the 
NRC commented on the relationship between changes made under Sec. 50.59 
and FSAR updating, stating: ``The Sec. 50.59(b) reporting may not be 
detailed sufficiently to be considered adequate to fulfill the FSAR 
updating requirement. The degree of detail required for updating the 
FSAR will be generally greater than a `brief description' and a 
`summary of the safety evaluation'.'' Thus, the Commission clearly 
expected the update submittal to include sufficient information to 
appropriately reflect the changes that were made. The burden associated 
with explicitly documenting in the update the effects of the changes on 
event probabilities and consequences is therefore small.
    The public reporting burden for this information collection request 
is estimated to average 3100 hours per response, including the time for 
reviewing instructions, searching

[[Page 56119]]

existing data sources, gathering and maintaining the data needed, and 
completing and reviewing the information collection. The Commission 
estimates that there is only a slight increase in burden associated 
with these proposed changes over the existing burden. The U.S. Nuclear 
Regulatory Commission is seeking public comment on the potential impact 
of the collection of information contained in the proposed rule and on 
the following issues:
    1. Is the proposed collection of information necessary for the 
proper performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of the burden correct?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the collection of information be 
minimized, including the use of automated collection techniques?
    Send comments on any aspect of this proposed collection of 
information, including suggestions for reducing the burden, to the 
Information and Records Management Branch (T-6 F33), U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, or by Internet 
electronic mail at BJS1@NRC.GOV; and to the Desk Officer, Office of 
Information and Regulatory Affairs, NEOB-10202, (3150-0017, -0020, -
0011, -0009, and -01320), Office of Management and Budget, Washington, 
DC 20503.
    Comments to OMB on the collections of information or on the above 
issues should be submitted by November 20, 1998. Comments received 
after this date will be considered if it is practical to do so, but 
assurance of consideration cannot be given to comments received after 
this date.

Public Protection Notification

    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.

X. Regulatory Analysis

    The Commission has prepared a draft regulatory analysis on this 
proposed regulation. The analysis examines the values and impacts of 
the alternatives considered by the Commission and includes the backfit 
analysis required by Sec. 50.109 (and Sec. 72.62). The alternatives 
considered in this analysis include no action, issuance of guidance 
only, or rulemaking. The draft analysis is available for inspection in 
the NRC Public Document Room, 2120 L Street NW. (Lower Level), 
Washington, DC and is available through the NRC interactive rulemaking 
website. Single copies of the analysis may be obtained from Eileen 
McKenna, EMM@NRC.GOV (301) 415-2189, Mail stop O-11-F-1, U.S. Nuclear 
Regulatory Commission, Washington DC 20555.
    The Commission requests public comment on the draft analysis. 
Comments on the draft analysis may be submitted to the NRC as indicated 
under the ADDRESSES heading.

XI. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act of 1980, (5 
U.S.C. 605(b)), the Commission certifies that this rule will not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This proposed rule affects only the licensing and 
operation and decommissioning of nuclear power plants, nonpower 
reactors, and independent spent fuel storage facilities. The companies 
that own these facilities do not fall within the scope of the 
definition of ``small entities'' set forth in the Regulatory 
Flexibility Act or the Small Business Size Standards set out in 
regulations issued by the Small Business Administration at 13 CFR part 
121.

XII. Backfit Analysis

    As required by Sec. 50.109 and Sec. 72.62, the Commission has 
completed a backfit analysis for the proposed rule, which is included 
within the regulatory analysis. The Commission has determined, based on 
this analysis, that in most respects, the proposed rule does not impose 
new requirements, but provides more flexibility or clarification of 
existing requirements. In other respects, such as the definitions of 
change to the facility and ``reduction of margin of safety* * *'', some 
licensees may view the revised rule as imposing new requirements. 
Therefore, the Commission has prepared an analysis considering the 
factors in Sec. 50.109(c), which is included in the Regulatory 
Analysis.

XIII. Criminal Penalties

    For the purposes of Section 223 of the Atomic Energy Act (AEA), the 
Commission is issuing the proposed rule to amend 10 CFR part 50 : 
50.59,: 50.66, and : 50.71; and 10 CFR part 72: 72.48,: 72.70,: 72.212, 
and : 72.248, under one or more of sections 161b, 161i, or 161o of the 
AEA. Willful violations of the rule would be subject to criminal 
enforcement.

XIV. Compatibility of Agreement State Regulations

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the Commission on June 30, 1997, 
and published in the Federal Register (62 FR 46517, September 3, 1997), 
this rule is classified as compatibility Category ``NRC.'' 
Compatibility is not required for Category ``NRC'' regulations. The NRC 
program elements in this category are those that relate directly to 
areas of regulation reserved to the NRC by the AEA or the provisions of 
Title 10 of the Code of Federal Regulations, and although an Agreement 
State may not adopt program elements reserved to NRC, it may wish to 
inform its licensees of certain requirements via a mechanism that is 
consistent with the particular State's administrative procedure laws, 
but does not confer regulatory authority on the State.

List of Subjects

10 CFR Part 50

    Antitrust, Classified Information, Criminal penalties, Fire 
protection, Intergovernmental relations, Nuclear power plants and 
reactors, Radiation protection, Reactor siting criteria, Reporting and 
record keeping requirements.

10 CFR Part 52

    Administrative practice and procedure, Antitrust, Backfitting, 
Combined license, Early site permit, Emergency planning, Fees, 
Inspection, Limited work authorization, Nuclear power plants and 
reactors, Probabilistic risk assessment, Prototype, Reactor siting 
criteria, Redress of site, Reporting and record keeping requirements, 
Standard design, Standard design certification.

10 CFR Part 72

    Manpower training programs, Nuclear materials, Occupational safety 
and health, Reporting and record keeping requirements, Security 
measures, Spent fuel
    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. 553, the NRC is proposing to 
adopt the following amendments to 10 CFR parts 50, 52 and 72.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    1. The authority citation for part 50 continues to read as follows:


[[Page 56120]]


    Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
    Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 
185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. 
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, and 
50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as 
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 
50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also 
issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Section 50.37 
also issued under E.O. 12829, 3 CFR 1993 Comp., P. 570; E.O. 12958, 
Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-415, 
96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 
122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80--50.81 also 
issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). 
Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C 2237).

    2. Section 50.59 is revised to read as follows:


Sec. 50.59  Changes, tests and experiments.

    (a) Definitions for the purposes of this section:
    (1) Change means a modification, addition, or removal.
    (2) Facility as described in the final safety analysis report (as 
updated) means:
    (i) The systems, structures, and components that are described in 
the final safety analysis report(as updated),
    (ii) The design, performance requirements and methods of operation 
for such systems, structures and components required to be included or 
described in the final safety analysis report (as updated), and
    (iii) The evaluations or methods of evaluation required to be 
included in the FSAR (as updated) for such SSC and which demonstrate 
that their intended function(s) will be accomplished.
    (3) Final safety analysis report (as updated) means the Final 
Safety Analysis Report (or Final Hazards Summary Report) submitted in 
accordance with Sec. 50.34, as amended and supplemented, and as 
modified as a result of changes made pursuant to Sec. 50.59 and 
Sec. 50.90, and, as applicable, Sec. 50.71 (e) and (f).
    (4) Procedures as described in the final safety analysis report (as 
updated) means information in the final safety analysis report (as 
updated) regarding how structures, systems, and components are operated 
and controlled (including assumed operator actions and response times) 
and information describing the conduct of operations.
    (5) Reduction in margin of safety associated with any technical 
specification means that the input assumptions, analytical methods, 
acceptance conditions, criteria and limits of the safety analyses, 
presented in the final safety analysis report (as updated), that 
established any technical specification requirement, are altered in a 
nonconservative manner.
    (6) Tests or experiments not described in the final safety analysis 
report (as updated) means any condition where the reactor or any of its 
systems, structures or components are utilized or controlled in a 
manner which is either:
    (i) Outside the controlling parameters of the design bases as 
described in the final safety analysis report (as updated) or
    (ii) Inconsistent with the analyses in the final safety analysis 
report (as updated).
    (b) Applicability. The provisions of this section apply to each 
holder of a license authorizing operation of a production or 
utilization facility, including the holder of a license authorizing 
operation of a nuclear power reactor that has submitted the 
certification of permanent cessation of operations required under 
Sec. 50.82(a)(1) or a reactor licensee whose license has been 
permanently modified to allow possession but not operation of the 
facility.
    (c)(1) A licensee may make changes in the facility as described in 
the final safety analysis report (as updated), make changes in the 
procedures as described in the final safety analysis report (as 
updated), and conduct tests or experiments not described in the final 
safety analysis report (as updated) without obtaining a license 
amendment pursuant to Sec. 50.90 only if:
    (i) A change to the technical specifications incorporated in the 
license is not required, and
    (ii) The change, test or experiment does not meet any of the 
criteria in paragraph (c)(2) of this section. The provisions in this 
section do not apply to changes in procedures when the applicable 
regulations establish more specific criteria for accomplishing such 
changes.
    (2) A licensee shall obtain an amendment to the license pursuant to 
Sec. 50.90 prior to implementing a change, test or experiment if it 
would:
    (i) Result in more than a minimal increase in the probability of 
occurrence of an accident previously evaluated in either the final 
safety analysis report (as updated), or in evaluations performed 
pursuant to this section and safety analyses performed pursuant to 
Sec. 50.90 after the last final safety analysis report was updated 
pursuant to Sec. 50.71 of this part;
    (ii) Result in more than a minimal increase in the probability of 
occurrence of a malfunction of equipment important to safety previously 
evaluated in either the final safety analysis report (as updated), or 
in evaluations performed pursuant to this section and safety analyses 
performed pursuant to Sec. 50.90 after the last final safety analysis 
report was updated pursuant to Sec. 50.71 of this part;
    (iii) Result in more than a minimal increase in the consequences of 
an accident previously evaluated in either the final safety analysis 
report (as updated), or in evaluations performed pursuant to this 
section and safety analyses performed pursuant to Sec. 50.90 after the 
last final safety analysis report was updated pursuant to Sec. 50.71 of 
this part;
    (iv) Result in more than a minimal increase in the consequences of 
a malfunction of equipment important to safety previously evaluated in 
either the final safety analysis report (as updated), or in evaluations 
performed pursuant to this section and safety analyses performed 
pursuant to Sec. 50.90 after the last final safety analysis report was 
updated pursuant to Sec. 50.71 of this part;
    (v) Create a possibility for a design basis accident of a different 
type than any previously evaluated in either the final safety analysis 
report (as updated), or in evaluations performed pursuant to this 
section and safety analyses performed pursuant to Sec. 50.90 with 
respect to design basis accidents after the last final safety analysis 
report was updated pursuant to Sec. 50.71 of this part;
    (vi) Create a possibility for a malfunction of equipment important 
to safety with a different result than any previously evaluated in 
either the final safety analysis report (as updated), or in evaluations 
performed pursuant to this section and safety analyses performed 
pursuant to Sec. 50.90 after the last final safety analysis report was 
updated pursuant to Sec. 50.71 of this part;
    (vii) Result in a reduction in the margin of safety associated with 
any Technical Specification.
    (d)(1) The licensee shall maintain records of changes in the 
facility and of changes in procedures made pursuant to this section, to 
the extent that these changes constitute changes in the facility as 
described in the final safety analysis report (as updated) or to the 
extent that they constitute changes in procedures as described in the 
final

[[Page 56121]]

safety analysis report (as updated). The licensee shall also maintain 
records of tests and experiments carried out pursuant to paragraph (c) 
of this section. These records must include a written evaluation which 
provides the bases for the determination that the change, test or 
experiment does not require a license amendment pursuant to paragraph 
(c)(2) of this section.
    (2) The licensee shall submit, as specified in Sec. 50.4, a report 
containing a brief description of any changes, tests, and experiments, 
including a summary of the evaluation of each. The report may be 
submitted annually or along with the FSAR updates as specified by 
Sec. 50.71(e), or at such shorter intervals as may be specified in the 
license.
    (3) The records of changes in the facility must be maintained until 
the termination of a license issued pursuant to this part or the 
termination of a license issued pursuant to 10 CFR part 54, whichever 
is later. Records of changes in procedures and records of tests and 
experiments must be maintained for a period of five years.
    3. In Sec. 50.66, paragraph (b), introductory text, paragraphs 
(b)(4), (c)(2), and (c)(3)(iii) are revised to read as follows:


Sec. 50.66  Requirements for thermal annealing of the reactor pressure 
vessel.

* * * * *
    (b) Thermal Annealing Report. The Thermal Annealing Report must 
include: a Thermal Annealing Operating Plan; a Requalification 
Inspection and Test Program; a Fracture Toughness Recovery and 
Reembrittlement Trend Assurance Program; and Identification of Changes 
Requiring a License Amendment.
    (1) * * *
    (4) Identification of changes requiring a license amendment. Any 
changes to the facility as described in the final safety analysis 
report (as updated) which requires a license amendment pursuant to 
Sec. 50.59(c)(2) of this part, and any changes to the technical 
specifications, which are necessary to either conduct the thermal 
annealing or to operate the nuclear power reactor following the 
annealing must be identified. The section shall demonstrate that the 
Commission's requirements continue to be complied with, and that there 
is reasonable assurance of adequate protection to the public health and 
safety following the changes.
    (c) * * *
    (2) If the thermal annealing was completed but the annealing was 
not performed in accordance with the Thermal Annealing Operating Plan 
and the Requalification Inspection and Test Program, the licensee shall 
submit a summary of lack of compliance with the Thermal Annealing 
Operating Plan and the Requalification Inspection and Test Program and 
a justification for subsequent operation to the Director, Office of 
Nuclear Reactor Regulation. Any changes to the facility as described in 
the final safety analysis report (as updated) which are attributable to 
the noncompliances and which require a license amendment pursuant to 
Sec. 50.59(c)(2) and any changes to the technical specifications, shall 
also be identified.
    (i) If no changes requiring a license amendment pursuant to 
Sec. 50.59(c)(2) or changes to Technical Specifications are identified, 
the licensee may restart its reactor after the requirements of 
paragraph (f)(2) of this section have been met.
    (ii) If any changes requiring a license amendment pursuant to 
Sec. 50.59(c)(2) or changes to the Technical Specifications are 
identified, the licensee may not restart its reactor until approval is 
obtained from the Director, Office of Nuclear Reactor Regulation and 
the requirements of paragraph (f)(2) of this section have been met.
    (3) * * *
    (iii) If the partial annealing was not performed in accordance with 
the Thermal Annealing Operating Plan and the Requalification Inspection 
and Test Program, the licensee shall submit a summary of lack of 
compliance with the Thermal Annealing Operating Plan and the 
Requalification Inspection and Test Program and a justification for 
subsequent operation to the Director, Office of Nuclear Reactor 
Regulation. Any changes to the facility as described in the final 
safety analysis report (as updated) which are attributable to the 
noncompliances and which require a license amendment pursuant to 
Sec. 50.59(c)(2) and any changes to the technical specifications which 
are required as a result of the noncompliances, shall also be 
identified.
    (A) If no changes requiring a license amendment pursuant to 
Sec. 50.59(c)(2) or changes to technical specifications are identified, 
the licensee may restart its reactor after the requirements of 
paragraph (f)(2) of this section have been met.
    (B) If any changes requiring a license amendment pursuant to 
Sec. 50.59(c)(2) or changes to technical specifications are identified, 
the licensee may not restart its reactor until approval is obtained 
from the Director, Office of Nuclear Reactor Regulation and the 
requirements of paragraph (f)(2) of this section have been met.
* * * * *
    4. In Sec. 50.71 paragraph (e) is revised to read as follows:


Sec. 50.71  Maintenance of records, making of reports.

* * * * *
    (e) Each person licensed to operate a nuclear power reactor 
pursuant to the provisions of Sec. 50.21 or Sec. 50.22 of this part 
shall update periodically, as provided in paragraphs (e)(3) and (4) of 
this section, the final safety analysis report (FSAR) originally 
submitted as part of the application for the operating license, to 
assure that the information included in the report contains the latest 
information developed. This submittal must contain all the changes 
necessary to reflect information and analyses submitted to the 
Commission by the licensee or prepared by the licensee pursuant to 
Commission requirement since the submission of the original FSAR, or as 
appropriate the last update to the FSAR under this section. The 
submittal must include the effects \1\ of:
---------------------------------------------------------------------------

    \1\ Effects of changes includes appropriate revisions of 
descriptions in the FSAR such that the FSAR (as updated) is complete 
and accurate.''
---------------------------------------------------------------------------

    (1) All changes made in the facility or procedures as described in 
the FSAR;
    (2) All safety analyses and evaluations performed by the licensee 
either in support of requested license amendments, or in support of 
conclusions that changes did not require a license amendment in 
accordance with Sec. 50.59(c)(2) of this part;
    (3) All analyses of new safety issues performed by or on behalf of 
the licensee at Commission request; and
    (4) The net effect of all changes made since the last update on the 
safety analyses, including probabilities, consequences, calculated 
values, system or component performance, that are in the FSAR (as 
updated). The updated information shall be appropriately located within 
the update to the FSAR.
* * * * *
    5. Section 50.90 is revised to read as follows:


Sec. 50.90  Application for Amendment of license or construction 
permit.

    Whenever a holder of a license or construction permit desires to 
amend the license (including the Technical Specifications incorporated 
into the license) or permit, application for an amendment must be filed 
with the Commission, as specified in Sec. 50.4, fully describing the 
changes desired, and following as far as applicable, the form 
prescribed for original applications.

[[Page 56122]]

PART 52--EARLY SITE PERMITS, STANDARD DESIGN CERTIFICATIONS; AND 
COMBINED LICENSES FOR NUCLEAR POWER PLANTS

    6. The authority citation for part 52 continues to read as follows:

    Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, 
as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282); 
secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 
U.S.C. 5841, 5842, 5546).

    7. Appendix A to Part 52 is amended by revising Section VIII.B, 
paragraphs 5.a,b,d, and Section X.A.3 as follows:

Appendix A--Design Certification Rule for the U.S. Advanced Boiling 
Water Reactor

VIII. Processes for Changes and Departures

* * * * *

B. Tier 2 information

    5. * * *
    a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the technical specifications, 
or otherwise requires a license amendment as defined in paragraphs 
B.5.b and B.5.c of this section. When evaluating the proposed 
departure, an applicant or licensee shall consider all matters 
described in the plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
    (1) Result in more than a minimal increase in the probability of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the probability of 
occurrence of a malfunction of equipment important to safety 
previously evaluated in the plant-specific DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of equipment important to safety previously 
evaluated in the plant-specific DCD;
    (5) Create a possibility for a design basis accident of a 
different type than any evaluated previously in the plant-specific 
DCD;
    (6) Create a possibility for a malfunction of equipment 
important to safety with a different result than any evaluated 
previously in the plant-specific DCD; or
    (7) Result in a reduction in the margin of safety associated 
with any Technical Specification for an application or license 
referencing this design certification.
* * * * *
    d. If a departure requires a license amendment pursuant to 
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR 
50.90.
* * * * *

X. Records and Reporting

A. Records.

* * * * *
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).

    8. Appendix B to part 52 is amended by revising Section VIII.B, 
paragraphs 5.a,b,d, and Section X.A.3 to read as follows:

Appendix B--Design Certification Rule for the System 80+ Design

VIII. Processes for Changes and Departures

* * * * *

B. Tier 2 information.

* * * * *
    a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the technical specifications, 
or otherwise requires a license amendment as defined in paragraphs 
B.5.b and B.5.c of this section. When evaluating the proposed 
departure, an applicant or licensee shall consider all matters 
described in the plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
    (1) Result in more than a minimal increase in the probability of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the probability of 
occurrence of a malfunction of equipment important to safety 
previously evaluated in the plant-specific DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of equipment important to safety previously 
evaluated in the plant-specific DCD;
    (5) Create a possibility for a design basis accident of a 
different type than any evaluated previously in the plant-specific 
DCD;
    (6) Create a possibility for a malfunction of equipment 
important to safety with a different result than any evaluated 
previously in the plant-specific DCD; or
    (7) Result in a reduction in the margin of safety associated 
with any Technical Specification for an application or license 
referencing this design certification.
* * * * *
    d. If a departure requires a license amendment pursuant to 
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR 
50.90.
* * * * *

X. Records and Reporting

A. Records.
* * * * *
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE

    9. The authority citation for part 72 continues to read as follows:

    Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953, 
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233, 
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat. 
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
Pub. L. 95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 102, 
Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332); Secs. 131, 132, 133, 
135, 137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 
148, Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 
10153, 10155, 10157, 10161, 10168).
    Section 72.44(g) also issued under secs. 142(b) and 148(c), (d), 
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b), 
10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat. 
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub. 
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also 
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2224 (42 U.S.C. 10101, 
10137(a), 10161(h)). Subparts K and L are also issued under sec. 
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252 
(42 U.S.C. 10198).

10. Section 72.3 is amended by revising the definition for independent 
spent fuel storage installation or ISFSI to read as follows:


Sec. 72.3  Definitions.

* * * * *
    Independent spent fuel storage installation or ISFSI means a 
complex designed and constructed for the

[[Page 56123]]

interim storage of spent nuclear fuel and other radioactive materials 
associated with spent fuel storage. An ISFSI which is located on the 
site of another facility licensed under this part or a facility 
licensed under part 50 of this chapter and which shares common 
utilities and services with such a facility or is physically connected 
with such other facility may still be considered independent.
* * * * *
    11. In Sec. 72.9, paragraph (b) is revised to read as follows:


Sec. 72.9  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Secs. 72.7, 72.11, 72.16, 72.19, 72.22 through 
72.34, 72.42, 72.44, 72.48 through 72.56, 72.62, 72.70 through 72.82, 
72.90, 72.92, 72.94, 72.98, 72.100, 72.102, 72.104, 72.108, 72.120, 
72.126, 72.140 through 72.176, 72.180 through 72.186, 72.192, 72.206, 
72.212, 72.216, 72.218, 72.230, 72.232, 72.234, 72.236, 72.240, 72.244, 
and 72.248.
    12. In Sec. 72.24, paragraph (a) is revised as follows:


Sec. 72.24  Contents of application: Technical information.

* * * * *
    (a) A description and safety assessment of the site on which the 
ISFSI or MRS is to be located, with appropriate attention to the design 
bases for external events. Such assessment must contain an analysis and 
evaluation of the major structures, systems and components of the ISFSI 
or MRS that bear on the suitability of the site when the ISFSI or MRS 
is operated at its design capacity. If the proposed ISFSI or MRS is to 
be located on the site of a nuclear power plant or other licensed 
facility, the potential interactions between the ISFSI or MRS and such 
other facility--including shared common utilities and services--must be 
evaluated.
* * * * *
    13. Section 72.48 is revised to read as follows:


Sec. 72.48  Changes, tests and experiments.

    (a) Definitions--As used in this section:
    (1) Change means a modification, addition or removal.
    (2) Final Safety Analysis Report (as updated) means:
    (i) For site-specific licensees, the Safety Analysis Report for a 
ISFSI, MRS or spent fuel storage cask, submitted in accordance with 
Sec. 72.24, as modified as a result of changes made pursuant to 
Sec. 72.48, and as updated in accordance with Sec. 72.70;
    (ii) For general licensees, the Safety Analysis Report for a ISFSI, 
MRS or spent fuel storage cask, as modified as a result of changes made 
pursuant to Sec. 72.48, and as updated in accordance with Sec. 72.216; 
and
    (iii) For certificate holders, the Safety Analysis Report for an 
approved cask, modified by as a result of changes made pursuant to 
Sec. 72.48 and as updated in accordance with Sec. 72.248.
    (3) The ISFSI, MRS, or spent fuel storage cask as described in the 
Final Safety Analysis Report (as updated) means:
    (i) The systems, structures, and components that are described in 
the Final Safety Analysis Report as updated in accordance with 
Secs. 72.70, 72.216 or Sec. 72.248,
    (ii) The design, performance requirements and methods of operation 
for such systems, structures, and components required to be included or 
described in the Final Safety Analysis Report (as updated), and
    (iii) The evaluations for such systems, structures, and components 
required to be included in the Final Safety Analysis Report (as 
updated) and which demonstrate that their intended function(s) will be 
accomplished.
    (4) Procedures as described in the Final Safety Analysis Report (as 
updated) means information in the Final Safety Analysis Report (as 
updated) regarding how structures, systems, and components are operated 
or controlled and information describing conduct of operations.
    (5) Reduction in margin of safety associated with any technical 
specification means that the input assumptions, analytical methods, 
acceptance conditions, criteria and limits of the safety analyses, 
presented in the Final Safety Analysis Report (as updated), that 
established any technical specification requirement, are altered in a 
nonconservative manner.
    (6) Tests or experiments not described in the Final Safety Analysis 
Report (as updated) means any condition where the ISFSI, MRS or spent 
fuel storage cask or any of its systems, structures, or components are 
utilized or controlled in a manner which is either:
    (i) Outside the controlling parameters of the design bases as 
described in the Final Safety Analysis Report (as updated) or
    (ii) Inconsistent with the analyses in the Final Safety Analysis 
Report (as updated).
    (b)(1) A licensee or certificate holder may make changes in the 
ISFSI, MRS, or spent fuel storage cask as described in the Final Safety 
Analysis Report (as updated), make changes in the procedures as 
described in the Final Safety Analysis Report (as updated), and conduct 
tests or experiments not described in the Final Safety Analysis Report 
(as updated), without obtaining either a license amendment pursuant to 
Sec. 72.56 (for licensees), if a change in the conditions incorporated 
in the license is not required, and the change, test, or experiment 
does not meet any of the criteria in paragraph (b)(2) of this section 
or a Certificate of Compliance (CoC) amendment pursuant to Sec. 72.244 
(for certificate holders), if a change in the terms, conditions or 
specifications incorporated in the CoC is not required; and the change, 
test, or experiment does not meet any of the criteria in paragraph 
(b)(2) of this section. The provisions in this section do not apply to 
changes in procedures when the applicable regulations establish more 
specific criteria for accomplishing such changes.
    (2) A licensee shall obtain a license amendment pursuant to 
Sec. 72.56 and a certificate holder shall obtain a CoC amendment 
pursuant to Sec. 72.244, prior to implementing a change, test, or 
experiment if it would:
    (i) Result in more than a minimal increase in the probability of 
occurrence of an accident previously evaluated in either the Final 
Safety Analysis Report (as updated), or in evaluations performed 
pursuant to this section and safety analyses performed pursuant to 
Secs. 72.56 or 72.244 after the last Final Safety Analysis Report was 
updated pursuant to Secs. 72.70, 72.216 or Sec. 72.248, of this part, 
as applicable;
    (ii) Result in more than a minimal increase in the probability of 
occurrence of a malfunction of structures, systems, and components 
important to safety which were previously evaluated in either the Final 
Safety Analysis Report (as updated), or in evaluations performed 
pursuant to this section and safety analyses performed pursuant to 
Secs. 72.56 or 72.244 after the last final safety analysis report was 
updated pursuant to Secs. 72.70, 72.216 or Sec. 72.248, of this part, 
as applicable;
    (iii) Result in more than a minimal increase in the consequences of 
an accident previously evaluated in either the Final Safety Analysis 
Report (as updated), or in evaluations performed pursuant to this 
section and safety analyses performed pursuant to Secs. 72.56 or 72.244 
after the last final safety analysis report was updated pursuant to 
section 72.70, 72.216 or Sec. 72.248, of this part, as applicable;
    (iv) Result in more than a minimal increase in the consequences of 
a

[[Page 56124]]

malfunction of structures, systems, and components important to safety 
which were previously evaluated in either the Final Safety Analysis 
Report (as updated), or in evaluations performed pursuant to this 
section and safety analyses performed pursuant to Sec. 72.56 or 
Sec. 72.244 after the last final safety analysis report was updated 
pursuant to Sec. 72.70, Sec. 72.216 or Sec. 72.248, of this part, as 
applicable;
    (v) Create the possibility for a design basis accident of a 
different type than any evaluated previously in either the Final Safety 
Analysis Report (as updated), or in evaluations performed pursuant to 
this section and safety analyses performed pursuant to Secs. 72.56 or 
Sec. 72.244 with respect to design basis accidents after the last final 
safety analysis report was updated pursuant to Sec. 72.70, Sec. 72.216 
or Sec. 72.248, of this part, as applicable;
    (vi) Create the possibility for a malfunction of structures, 
systems, and components important to safety with a different result 
than any evaluated previously in either the Final Safety Analysis 
Report (as updated), or in evaluations performed pursuant to this 
section and safety analyses performed pursuant to Secs. 72.56 or 
Sec. 72.244 after the last final safety analysis report was updated 
pursuant to Sec. 72.70, Sec. 72.216 or Sec. 72.248, of this part, as 
applicable;
    (vii) Result in a reduction in the margin of safety associated with 
any technical specification; (viii) Result in a significant increase in 
occupational exposure;
    (ix) Result in a significant unreviewed environmental impact.
    (c)(1) Each licensee or certificate holder shall maintain records 
of changes in the ISFSI, MRS, or spent fuel storage cask and of changes 
in procedures it has made pursuant to this section if these changes 
constitute changes in the ISFSI, MRS, or spent fuel storage cask or 
procedures described in the Final Safety Analysis Report (as updated). 
The licensee or certificate holder shall also maintain records of test 
and experiments carried out pursuant to paragraph (b) of this section. 
These records shall include a written evaluation that provides the 
bases for the determination that the change, test, or experiment does 
not require a license or CoC amendment pursuant to paragraph (b)(2) of 
this section. The records of changes in the ISFSI, MRS, or spent fuel 
storage cask and of changes in procedures and records of tests and 
experiments shall be maintained until spent nuclear fuel is no longer 
stored in the ISFSI, MRS or spent fuel storage cask, and the Commission 
terminates the license or CoC. For a holder of cask Certificate of 
Compliance who permanently ceases operation, any such records shall be 
provided to the new holder of cask Certificate of Compliance or to the 
Commission, as appropriate, in accordance with Sec. 72.234(d)(3).
    (2) Annually, or at such shorter interval as may be specified in 
the license or CoC, each holder of a license or cask Certificate of 
Compliance shall submit a report containing a brief description of 
changes, tests and experiments made by the license or certificate 
holder under paragraph (b) of this section, including a summary of the 
evaluation of each. Licensee and certificate holders shall submit their 
reports in accordance with Sec. 72.4. Any report submitted by a 
licensee or certificate holder pursuant to this paragraph will be made 
a part of the public record pertaining to the license or CoC.
    14. Section 72.56 is revised to read as follows:


Sec. 72.56  Application for amendment of license.

    Whenever a holder of a license desires to amend the license 
(including a change to the license conditions), an application for an 
amendment shall be filed with the Commission fully describing the 
changes desired and the reasons for such changes, and following as far 
as applicable the form prescribed for original applications.
    15. In Sec. 72.70, paragraphs (a), (b), introductory text, and 
(b)(2) are revised to read and a new paragraph (c) is added to read as 
follows:


Sec. 72.70  Safety analysis report updating.

    (a) The design, description of planned operations, and other 
information submitted in the Safety Analysis Report for an ISFSI or MRS 
shall be updated by the licensee and submitted to the Commission at 
least once every six months after issuance of the license during final 
design and construction, until preoperational testing is completed, 
with a Final Safety Analysis Report (FSAR) completed and submitted to 
the Commission at least 90 days prior to the planned receipt of spent 
fuel or high-level radioactive waste. The FSAR shall include a final 
analysis and evaluation of the design and performance of structures, 
systems, and components that are important to safety taking into 
account any pertinent information developed since the submittal of the 
license application.
    (b) After the first receipt of spent fuel or high-level radioactive 
waste for storage, the FSAR shall be updated annually and submitted to 
the Commission by the licensee. This submittal shall include the 
following:
* * * * *
    (2) A description and analysis of changes in procedures or in 
structures, systems, and components of the ISFSI or MRS, as described 
in the FSAR (as updated), with emphasis upon:
* * * * *
    (c) The licensee shall submit revisions of the FSAR to the 
Commission in accordance with Sec. 72.4, on a replacement-page basis 
that is accompanied by a list which identifies the current pages of the 
FSAR following page replacement. Each replacement page shall include 
both a change indicator for the area changed (e.g., a bold line 
vertically drawn in the margin adjacent to the portion actually 
changed) and a page change identification (date of change or change 
number or both).
    16. In Sec. 72.86, paragraph (b) is revised to read as follows:


Sec. 72.86  Criminal penalties.

* * * * *
    (b) The regulations in this part 72 that are not issued under 
sections 161b, 161i, or 161o for the purposes of section 223 are as 
follows: Secs. 72.1, 72.2, 72.3, 72.4, 72.5, 72.7, 72.8, 72.9, 72.16, 
72.18, 72.20, 72.22, 72.24, 72.26, 72.28, 72.32, 72.34, 72.40, 72.46, 
72.56, 72.58, 72.60, 72.62, 72.84, 72.86, 72.90, 72.96, 72.108, 72.120, 
72.122, 72.124, 72.126, 72.128, 72.130, 72.182, 72.194, 72.200, 72.202, 
72.204, 72.206, 72.210, 72.214, 72.220, 72.230, 72.238, 72.240, 72.244, 
and 72.246.
    17. In Sec. 72.212, paragraph (b)(4) is revised to read as follows:


Sec. 72.212  Conditions of general license issued under Sec. 72.210.

* * * * *
    (b) * * *
    (4) Prior to use of this general license, determine whether 
activities related to storage of spent fuel under this general license 
involve a change in the facility Technical Specifications or require a 
license amendment for the facility pursuant to Sec. 50.59(c)(2) of this 
chapter. Results of this determination must be documented in the 
evaluation made in paragraph (b)(2) of this section.
    18. In Sec. 72.216, new paragraph (d) is added to read as follows:


Sec. 72.216  Reports.

* * * * *
    (d) The final safety analysis report (FSAR) for each approved cask 
used by the general licensee shall be updated annually and submitted to 
the Commission by the general licensee.

[[Page 56125]]

The submittal shall include the following:
    (1) A description and analysis of changes in procedures or in 
structures, systems, and components of the spent fuel storage cask, as 
described in the FSAR (as updated), with emphasis upon:
    (i) Performance requirements,
    (ii) The bases, with technical justification therefor upon which 
such requirements have been established, and
    (iii) Evaluations showing that safety functions will be 
accomplished.
    (2) An analysis of the significance of any changes to codes, 
standards, regulations, or regulatory guides which the general licensee 
has committed to meeting the requirements of which are applicable to 
the design, construction, or fabrication of the spent fuel storage 
cask.
    (3) The general licensee shall submit revisions containing updated 
information to the Commission, in accordance with Sec. 72.4, on a 
replacement-page basis that is accompanied by a list which identifies 
the current pages of the FSAR following page replacement. The general 
licensee shall also provide a copy of the submittal to the holder of 
the certificate for the cask. Each replacement page shall include both 
a change indicator for the area changed (e.g., a bold line vertically 
drawn in the margin adjacent to the portion actually changed) and a 
page change identification (date of change or change number or both). 
Each replacement page shall also indicate the cask FSAR, including the 
certificate holder's revision number, upon which the general licensee's 
update is based.
    19. Section 72.244 is added to read as follows:


Sec. 72.244  Application for amendment of a certificate of compliance.

    Whenever a certificate holder desires to amend the CoC (including a 
change to the terms, conditions or specifications of the CoC), an 
application for an amendment shall be filed with the Commission fully 
describing the changes desired and the reasons for such changes, and 
following as far as applicable the form prescribed for original 
applications.
    20. Section 72.246 is added to read as follows:


Sec. 72.246  Issuance of amendment to a certificate of compliance.

    In determining whether an amendment to a CoC will be issued to the 
applicant, the Commission will be guided by the considerations that 
govern the issuance of an initial CoC.
    21. Section 72.248 is added to read as follows:


Sec. 72.248  Safety analysis report updating.

    (a) The design, description of planned operations, and other 
information submitted in the Safety Analysis Report for a spent fuel 
storage cask shall be updated by the certificate holder and submitted 
to the Commission after the design of the spent fuel storage cask has 
been approved pursuant to Sec. 72.238. This Final Safety Analysis 
Report (FSAR) shall be completed and submitted to the Commission within 
90 days after approval of the cask design. The FSAR shall incorporate 
all changes and requirements contained in the CoC and the staff's 
safety evaluation report (SER) associated with approval of the cask's 
design.
    (b) The FSAR shall be updated annually and submitted to the 
Commission by the certificate holder. This submittal shall include the 
following:
    (1) A description and analysis of changes in procedures or in 
structures, systems, and components of the spent fuel storage cask, as 
described in the FSAR (as updated), with emphasis upon:
    (i) Performance requirements,
    (ii) The bases, with technical justification therefor upon which 
such requirements have been established, and
    (iii) Evaluations showing that safety functions will be 
accomplished.
    (2) An analysis of the significance of any changes to codes, 
standards, regulations, or regulatory guides which the certificate 
holder has committed to meeting the requirements of which are 
applicable to the design, construction, or fabrication of the spent 
fuel storage cask.
    (c) The certificate holder shall submit revisions containing 
updated information to the Commission, in accordance with Sec. 72.4, on 
a replacement-page basis that is accompanied by a list which identifies 
the current pages of the FSAR following page replacement. The 
certificate holder shall also provide a copy of the submittal to each 
general licensee using the spent fuel storage cask. Each replacement 
page shall include both a change indicator for the area changed (e.g., 
a bold line vertically drawn in the margin adjacent to the portion 
actually changed) and a page change identification (date of change or 
change number or both).

    Dated at Rockville, Maryland, this 14th day of October, 1998.

    For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 98-28066 Filed 10-20-98; 8:45 am]
BILLING CODE 7590-01-P