[Title 10 CFR ]
[Code of Federal Regulations (annual edition) - January 1, 2011 Edition]
[From the U.S. Government Printing Office]
[[Page i]]
Title 10
Parts 51 to 199
Revised as of January 1, 2011
Energy
________________________
Containing a codification of documents of general
applicability and future effect
As of January 1, 2011
Published by the Office of the Federal Register
National Archives and Records Administration as a
Special Edition of the Federal Register
A Special Edition of the Federal Register
[[Page ii]]
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[[Page iii]]
Table of Contents
Page
Explanation................................................. v
Title 10:
Chapter I--Nuclear Regulatory Commission (Continued) 3
Finding Aids:
Table of CFR Titles and Chapters........................ 839
Alphabetical List of Agencies Appearing in the CFR...... 859
List of CFR Sections Affected........................... 869
[[Page iv]]
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Cite this Code: CFR
To cite the regulations in
this volume use title,
part and section number.
Thus, 10 CFR 51.1 refers
to title 10, part 51,
section 1.
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[[Page v]]
EXPLANATION
The Code of Federal Regulations is a codification of the general and
permanent rules published in the Federal Register by the Executive
departments and agencies of the Federal Government. The Code is divided
into 50 titles which represent broad areas subject to Federal
regulation. Each title is divided into chapters which usually bear the
name of the issuing agency. Each chapter is further subdivided into
parts covering specific regulatory areas.
Each volume of the Code is revised at least once each calendar year
and issued on a quarterly basis approximately as follows:
Title 1 through Title 16.................................as of January 1
Title 17 through Title 27..................................as of April 1
Title 28 through Title 41...................................as of July 1
Title 42 through Title 50................................as of October 1
The appropriate revision date is printed on the cover of each
volume.
LEGAL STATUS
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HOW TO USE THE CODE OF FEDERAL REGULATIONS
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OMB CONTROL NUMBERS
The Paperwork Reduction Act of 1980 (Pub. L. 96-511) requires
Federal agencies to display an OMB control number with their information
collection request.
[[Page vi]]
Many agencies have begun publishing numerous OMB control numbers as
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OBSOLETE PROVISIONS
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Sections Affected'' is published at the end of each CFR volume.
``[RESERVED]'' TERMINOLOGY
The term ``[Reserved]'' is used as a place holder within the Code of
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What is a proper incorporation by reference? The Director of the
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(a) The incorporation will substantially reduce the volume of
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(b) The matter incorporated is in fact available to the extent
necessary to afford fairness and uniformity in the administrative
process.
(c) The incorporating document is drafted and submitted for
publication in accordance with 1 CFR part 51.
What if the material incorporated by reference cannot be found? If
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CFR INDEXES AND TABULAR GUIDES
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that volume.
[[Page vii]]
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Director,
Office of the Federal Register.
January 1, 2011.
[[Page ix]]
THIS TITLE
Title 10--Energy is composed of four volumes. The parts in these
volumes are arranged in the following order: Parts 1-50, 51-199, 200-499
and part 500-end. The first and second volumes containing parts 1-199
are comprised of chapter I--Nuclear Regulatory Commission. The third and
fourth volumes containing part 200-end are comprised of chapters II, III
and X--Department of Energy, chapter XIII--Nuclear Waste Technical
Review Board, and chapter XVII--Defense Nuclear Facilities Safety Board.
The contents of these volumes represent all current regulations codified
under this title of the CFR as of January 1, 2011.
For this volume, Cheryl E. Sirofchuck was Chief Editor. The Code of
Federal Regulations publication program is under the direction of
Michael L. White, assisted by Ann Worley.
[[Page 1]]
TITLE 10--ENERGY
(This book contains parts 51 to 199)
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Part
chapter i--Nuclear Regulatory Commission (Continued)........ 51
[[Page 3]]
CHAPTER I--NUCLEAR REGULATORY COMMISSION (CONTINUED)
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Editorial Note: Nomenclature changes to chapter I appear at 70 FR
69421, Nov. 16, 2005, and at 72 FR 33386, June 18, 2007.
Part Page
51 Environmental protection regulations for
domestic licensing and related
regulatory functions.................... 5
52 Licenses, certifications, and approvals for
nuclear power plants.................... 68
53 [Reserved]
54 Requirements for renewal of operating
licenses for nuclear power plants....... 143
55 Operators' licenses......................... 150
60 Disposal of high-level radioactive wastes in
geologic repositories................... 166
61 Licensing requirements for land disposal of
radioactive waste....................... 202
62 Criteria and procedures for emergency access
to non-federal and regional low-level
waste disposal facilities............... 229
63 Disposal of high-level radioactive wastes in
a geologic repository at Yucca Mountain,
Nevada.................................. 237
70 Domestic licensing of special nuclear
material................................ 280
71 Packaging and transportation of radioactive
material................................ 329
72 Licensing requirements for the independent
storage of spent nuclear fuel, high-
level radioactive waste, and reactor-
related greater than Class C waste...... 382
73 Physical protection of plants and materials. 444
74 Material control and accounting of special
nuclear material........................ 560
75 Safeguards on nuclear material--
implementation of US/IAEA agreement..... 584
76 Certification of gaseous diffusion plants... 600
81 Standard specifications for the granting of
patent licenses......................... 626
[[Page 4]]
95 Facility security clearance and safeguarding
of national security information and
restricted data......................... 633
100 Reactor site criteria....................... 651
110 Export and import of nuclear equipment and
material................................ 666
140 Financial protection requirements and
indemnity agreements.................... 720
150 Exemptions and continued regulatory
authority in Agreement States and in
offshore waters under section 274....... 792
160 Trespassing on Commission property.......... 806
170 Fees for facilities, materials, import and
export licenses, and other regulatory
services under the Atomic Energy Act of
1954, as amended........................ 807
171 Annual fees for reactor licenses and fuel
cycle licenses and materials licenses,
including holders of certificates of
compliance, registrations, and quality
assurance program approvals and
government agencies licensed by the NRC. 822
172-199 [Reserved]
[[Page 5]]
PART 51_ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING
AND RELATED REGULATORY FUNCTIONS--Table of Contents
Sec.
51.1 Scope.
51.2 Subparts.
51.3 Resolution of conflict.
51.4 Definitions.
51.5 Interpretations.
51.6 Specific exemptions.
Subpart A_National Environmental Policy Act_Regulations Implementing
Section 102(2)
51.10 Purpose and scope of subpart; application of regulations of
Council on Environmental Quality.
51.11 Relationship to other subparts. [Reserved]
51.12 Application of subpart to ongoing environmental work.
51.13 Emergencies.
51.14 Definitions.
51.15 Time schedules.
51.16 Proprietary information.
51.17 Information collection requirements; OMB approval.
Preliminary Procedures
classification of licensing and regulatory actions
51.20 Criteria for and identification of licensing and regulatory
actions requiring environmental impact statements.
51.21 Criteria for and identification of licensing and regulatory
actions requiring environmental assessments.
51.22 Criterion for categorical exclusion; identification of licensing
and regulatory actions eligible for categorical exclusion or
otherwise not requiring environmental review.
51.23 Temporary storage of spent fuel after cessation of reactor
operation--generic determination of no significant
environmental impact.
determinations to prepare environmental impact statements, environmental
assessments or findings of no significant impact, and related procedures
51.25 Determination to prepare environmental impact statement or
environmental assessment; eligibility for categorical
exclusion.
51.26 Requirement to publish notice of intent and conduct scoping
process.
51.27 Notice of intent.
scoping
51.28 Scoping--participants.
51.29 Scoping-environmental impact statement and supplement to
environmental impact statement.
environmental assessment
51.30 Environmental assessment.
51.31 Determinations based on environmental assessment.
finding of no significant impact
51.32 Finding of no significant impact.
51.33 Draft finding of no significant impact; distribution.
51.34 Preparation of finding of no significant impact.
51.35 Requirement to publish finding of no significant impact;
limitation on Commission action.
Environmental Reports and Information--Requirements Applicable to
Applicants and Petitioners for Rulemaking
general
51.40 Consultation with NRC staff.
51.41 Requirement to submit environmental information.
environmental reports--general requirements
51.45 Environmental report.
environmental reports--production and utilization facilities
51.49 Environmental report--limited work authorization.
51.50 Environmental report--construction permit, early site permit, or
combined license stage.
51.51 Uranium fuel cycle environmental data--Table S-3.
51.52 Environmental effects of transportation of fuel and waste--Table
S-4.
51.53 Postconstruction environmental reports.
51.54 Environmental report--manufacturing license.
51.55 Environmental report--standard design certification.
51.58 Environmental report--number of copies; distribution.
environmental reports--materials licenses
51.60 Environmental report--materials licenses.
51.61 Environmental report--independent spent fuel storage installation
(ISFSI) or monitored retrievable storage installation (MRS)
license.
51.62 Environmental report--land disposal of radioactive waste licensed
under 10 CFR part 61.
51.66 Environmental report--number of copies; distribution.
[[Page 6]]
51.67 Environmental information concerning geologic repositories.
environmental reports--rulemaking
51.68 Environmental report--rulemaking.
Environmental Impact Statements
draft environmental impact statements--general requirements
51.70 Draft environmental impact statement--general.
51.71 Draft environmental impact statement--contents.
51.72 Supplement to draft environmental impact statement.
51.73 Request for comments on draft environmental impact statement.
51.74 Distribution of draft environmental impact statement and
supplement to draft environmental impact statement; news
releases.
draft environmental impact statements--production and utilization
facilities
51.75 Draft environmental impact statement--construction permit, early
site permit, or combined license.
51.76 Draft environmental impact statement--limited work authorization.
51.77 Distribution of draft environmental impact statement.
draft environmental impact statements--materials licenses
51.80 Draft environmental impact statement--materials license.
51.81 Distribution of draft environmental impact statement.
draft environmental impact statements--rulemaking
51.85 Draft environmental impact statement--rulemaking.
51.86 Distribution of draft environmental impact statement.
legislative environmental impact statements--proposals for legislation
51.88 Proposals for legislation.
final environmental impact statements--general requirements
51.90 Final environmental impact statement--general.
51.91 Final environmental impact statement--contents.
51.92 Supplement to the final environmental impact statement.
51.93 Distribution of final environmental impact statement and
supplement to final environmental impact statement; news
releases.
51.94 Requirement to consider final environmental impact statement.
final environmental impact statements--production and utilization
facilities
51.95 Postconstruction environmental impact statements.
final environmental impact statements--materials licenses
51.97 Final environmental impact statement--materials license.
final environmental impact statements--rulemaking
51.99 [Reserved]
NEPA Procedure and Administrative Action
general
51.100 Timing of Commission action.
51.101 Limitations on actions.
51.102 Requirement to provide a record of decision; preparation.
51.103 Record of decision--general.
51.104 NRC proceeding using public hearings; consideration of
environmental impact statement.
production and utilization facilities
51.105 Public hearings in proceedings for issuance of construction
permits or early site permits; limited work authorizations.
51.105a Public hearings in proceedings for issuance of manufacturing
licenses.
51.106 Public hearings in proceedings for issuance of operating
licenses.
51.107 Public hearings in proceedings for issuance of combined licenses;
limited work authorizations.
51.108 Public hearings on Commission findings that inspections, tests,
analyses, and acceptance criteria of combined licenses are
met.
materials licenses
51.109 Public hearings in proceedings for issuance of materials license
with respect to a geologic repository.
rulemaking
51.110 [Reserved]
Public Notice of and Access to Environmental Documents
51.116 Notice of intent.
51.117 Draft environmental impact statement--notice of availability.
51.118 Final environmental impact statement--notice of availability.
51.119 Publication of finding of no significant impact; distribution.
51.120 Availability of environmental documents for public inspection.
51.121 Status of NEPA actions.
[[Page 7]]
51.122 List of interested organizations and groups.
51.123 Charges for environmental documents; distribution to public;
distribution to governmental agencies.
Commenting
51.124 Commission duty to comment.
Responsible Official
51.125 Responsible official.
Appendix A to Subpart A of Part 51--Format for Presentation of Material
in Environmental Impact Statements
Appendix B to Subpart A of Part 51--Environmental Effect of Renewing the
Operating License of a Nuclear Power Plant
Subpart B [Reserved]
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 1701, 106 Stat.
2951, 2952, 2953, (42 U.S.C. 2201, 2297f); secs. 201, as amended, 202,
88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841, 5842); sec. 1704, 112
Stat. 2750 (44 U.S.C. 3504 note).
Subpart A also issued under National Environmental Policy Act of
1969, secs. 102, 104, 105, 83 Stat. 853-854, as amended (42 U.S.C. 4332,
4334, 4335); and Pub. L. 95-604, Title II, 92 Stat. 3033-3041; and sec.
193, Pub. L. 101-575, 104 Stat. 2835 (42 U.S.C. 2243).
Sections 51.20, 51.30, 51.60, 51.80. and 51.97 also issued under
secs. 135, 141, Pub. L. 97-425, 96 Stat. 2232, 2241, and sec. 148, Pub.
L. 100-203, 101 Stat. 1330-223 (42 U.S.C. 10155, 10161, 10168).
Section 51.22 also issued under sec. 274, 73 Stat. 688, as amended
by 92 Stat. 3036-3038 (42 U.S.C. 2021) and under Nuclear Waste Policy
Act of 1982, sec 121, 96 Stat. 2228 (42 U.S.C. 10141).
Sections 51.43, 51.67, and 51.109 also under Nuclear Waste Policy
Act of 1982, sec 114(f), 96 Stat. 2216, as amended (42 U.S.C. 10134(f)).
Source: 49 FR 9381, Mar. 12, 1984, unless otherwise noted.
Sec. 51.1 Scope.
This part contains environmental protection regulations applicable
to NRC's domestic licensing and related regulatory functions. These
regulations do not apply to export licensing matters within the scope of
part 110 of this chapter or to any environmental effects which NRC's
domestic licensing and related regulatory functions may have upon the
environment of foreign nations. Subject to these limitations, the
regulations in this part implement:
(a) Section 102(2) of the National Environmental Policy Act of 1969,
as amended.
Sec. 51.2 Subparts.
(a) The regulations in subpart A of this part implement section
102(2) of the National Environmental Policy Act of 1969, as amended.
Sec. 51.3 Resolution of conflict.
In any conflict between a general rule in subpart A of this part and
a special rule in another subpart of this part or another part of this
chapter applicable to a particular type of proceeding, the special rule
governs.
Sec. 51.4 Definitions.
As used in this part:
Act means the Atomic Energy Act of 1954 (Pub. L. 83-703, 68 Stat.
919) including any amendments thereto.
Commission means the Nuclear Regulatory Commission or its authorized
representatives.
Construction means the activities in paragraph (1) of this
definition, and does not mean the activities in paragraph (2) of this
definition.
(1) Activities constituting construction are the driving of piles,
subsurface preparation, placement of backfill, concrete, or permanent
retaining walls within an excavation, installation of foundations, or
in-place assembly, erection, fabrication, or testing, which are for:
(i) Safety-related structures, systems, or components (SSCs) of a
facility, as defined in 10 CFR 50.2;
(ii) SSCs relied upon to mitigate accidents or transients or used in
plant emergency operating procedures;
(iii) SSCs whose failure could prevent safety-related SSCs from
fulfilling their safety-related function;
(iv) SSCs whose failure could cause a reactor scram or actuation of
a safety-related system;
(v) SSCs necessary to comply with 10 CFR part 73;
(vi) SSCs necessary to comply with 10 CFR 50.48 and criterion 3 of
10 CFR part 50, appendix A; and
(vii) Onsite emergency facilities (i.e., technical support and
operations support centers), necessary to comply with
[[Page 8]]
10 CFR 50.47 and 10 CFR part 50, appendix E.
(2) Construction does not include:
(i) Changes for temporary use of the land for public recreational
purposes;
(ii) Site exploration, including necessary borings to determine
foundation conditions or other preconstruction monitoring to establish
background information related to the suitability of the site, the
environmental impacts of construction or operation, or the protection of
environmental values;
(iii) Preparation of a site for construction of a facility,
including clearing of the site, grading, installation of drainage,
erosion and other environmental mitigation measures, and construction of
temporary roads and borrow areas;
(iv) Erection of fences and other access control measures;
(v) Excavation;
(vi) Erection of support buildings (such as, construction equipment
storage sheds, warehouse and shop facilities, utilities, concrete mixing
plants, docking and unloading facilities, and office buildings) for use
in connection with the construction of the facility;
(vii) Building of service facilities, such as paved roads, parking
lots, railroad spurs, exterior utility and lighting systems, potable
water systems, sanitary sewerage treatment facilities, transmission
lines;
(viii) Procurement or fabrication of components or portions of the
proposed facility occurring at other than the final, in-place location
at the facility;
(ix) Manufacture of a nuclear power reactor under a manufacturing
license under subpart F of part 52 of this chapter to be installed at
the proposed site and to be part of the proposed facility; or
(x) With respect to production or utilization facilities, other than
testing facilities and nuclear power plants, required to be licensed
under Section 104.a or Section 104.c of the Act, the erection of
buildings which will be used for activities other than operation of a
facility and which may also be used to house a facility (e.g., the
construction of a college laboratory building with space for
installation of a training reactor).
NRC means the Nuclear Regulatory Commission, the agency established
by Title II of the Energy Reorganization Act of 1974, as amended.
NRC staff means any NRC officer or employee or his/her authorized
representative, except a Commissioner, a member of a Commissioner's
immediate staff, an Atomic Safety and Licensing Board, an Atomic Safety
and Licensing Appeal Board, a presiding officer, an administrative
judge, an administrative law judge, or any other officer or employee of
the Commission who performs adjudicatory functions.
NRC Staff Director means:
Executive Director for Operations;
Director, Office of Nuclear Reactor Regulation;
Director, Office of New Reactors;
Director, Office of Nuclear Material Safety and Safeguards;
Director, Office of Federal and State Materials and Environmental
Management Programs;
Director, Office of Nuclear Regulatory Research;
Director, Office of Governmental and Public Affairs; and
The designee of any NRC staff director.
[49 FR 9381, Mar. 12, 1984, as amended at 51 FR 35999, Oct. 8, 1986; 52
FR 31612, Aug. 21, 1987; 72 FR 57443, Oct. 9, 2007; 73 FR 5723, Jan. 31,
2008]
Sec. 51.5 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the regulations in this part by any officer or
employee of the Commission other than a written interpretation by the
General Counsel will be recognized to be binding upon the Commission.
Sec. 51.6 Specific exemptions.
The Commission may, upon application of any interested person or
upon its own initiative, grant such exemptions from the requirements of
the regulations in this part as it determines are authorized by law and
are otherwise in the public interest.
[[Page 9]]
Subpart A_National Environmental Policy Act_Regulations Implementing
Section 102(2)
Sec. 51.10 Purpose and scope of subpart; application of regulations of
Council on Environmental Quality.
(a) The National Environmental Policy Act of 1969, as amended (NEPA)
directs that, to the fullest extent possible: (1) The policies,
regulations, and public laws of the United States shall be interpreted
and administered in accordance with the policies set forth in NEPA, and
(2) all agencies of the Federal Government shall comply with the
procedures in section 102(2) of NEPA except where compliance would be
inconsistent with other statutory requirements. The regulations in this
subpart implement section 102(2) of NEPA in a manner which is consistent
with the NRC's domestic licensing and related regulatory authority under
the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act
of 1974, as amended, and the Uranium Mill Tailings Radiation Control Act
of 1978, and which reflects the Commission's announced policy to take
account of the regulations of the Council on Environmental Quality
published November 29, 1978 (43 FR 55978-56007) voluntarily, subject to
certain conditions. This subpart does not apply to export licensing
matters within the scope of part 110 of this chapter nor does it apply
to any environmental effects which NRC's domestic licensing and related
regulatory functions may have upon the environment of foreign nations.
(b) The Commission recognizes a continuing obligation to conduct its
domestic licensing and related regulatory functions in a manner which is
both receptive to environmental concerns and consistent with the
Commission's responsibility as an independent regulatory agency for
protecting the radiological health and safety of the public.
Accordingly, the Commission will:
(1) Examine any future interpretation or change to the Council's
NEPA regulations;
(2) Follow the provisions of 40 CFR 1501.5 and 1501.6 relating to
lead agencies and cooperating agencies, except that the Commission
reserves the right to prepare an independent environmental impact
statement whenever the NRC has regulatory jurisdiction over an acitivity
even though the NRC has not been designated as lead agency for
preparation of the statement; and
(3) Reserve the right to make a final decision on any matter within
the NRC's regulatory authority even though another agency has made a
predecisional referral of an NRC action to the Council under the
procedures of 40 CFR part 1504.
(c) The regulations in this subpart \1\ also address the limitations
imposed on NRC's authority and responsibility under the National
Environmental Policy Act of 1969, as amended, by the Federal Water
Pollution Control Act Amendments of 1972, Pub. L. 92-500, 86 Stat. 816
et seq. (33 U.S.C. 1251 et seq.) In accordance with section 511(c)(2) of
the Federal Water Pollution Control Act (86 Stat. 893, 33 U.S.C
1371(c)(2)) the NRC recognizes that responsibility for Federal
regulation of nonradiological pollutant discharges \2\ into receiving
waters rests by statute with the Environmental Protection Agency.
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\1\ See also Second Memorandum of Understanding Regarding
Implementation of Certain NRC and EPA Responsibilities and Policy
Statement on Implementation of Section 511 of the Federal Water
Pollution Control Act (FWPCA) attached as Appendix A thereto, which were
published in the Federal Register on December 31, 1975 (40 FR 60115) and
became effective January 30, 1976.
\2\ On June 1, 1976, the U.S. Supreme Court held that ``
`pollutants' subject to regulation under the FWPCA [Federal Water
Pollution Control Act] do not include source, byproduct, and special
nuclear materials, . . .'' Train v. Colorado PIRG, 426 U.S. 1 at 25.
---------------------------------------------------------------------------
(d) Commission actions initiating or relating to administrative or
judicial civil or criminal enforcement actions or proceedings are not
subject to Section 102(2) of NEPA. These actions include issuance of
notices of violation, orders, and denials of requests for action
pursuant to subpart B of part 2 of this chapter; matters covered by part
15 and part 160 of this chapter; and issuance of confirmatory action
letters, bulletins, generic letters, notices
[[Page 10]]
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of deviation, and notices of nonconformance.
[49 FR 9381, Mar. 12, 1984, as amended at 54 FR 43578, Oct. 26, 1989; 61
FR 43408, Aug. 22, 1996]
Sec. 51.11 Relationship to other subparts. [Reserved]
Sec. 51.12 Application of subpart to ongoing environmental work.
(a) Except as otherwise provided in this section, the regulations in
this subpart shall apply to the fullest extent practicable to NRC's
ongoing environmental work.
(b) No environmental report or any supplement to an environmental
report filed with the NRC and no environmental assessment, environmental
impact statement or finding of no significant impact or any supplement
to any of the foregoing issued by the NRC before June 7, 1984, need be
redone and no notice of intent to prepare an environmental impact
statement or notice of availability of these environmental documents
need be republished solely by reason of the promulgation on March 12,
1984, of this revision of part 51.
[49 FR 9381, Mar. 12, 1984, as amended at 49 FR 24513, June 14, 1984]
Sec. 51.13 Emergencies.
Whenever emergency circumstances make it necessary and whenever, in
other situations, the health and safety of the public may be adversely
affected if mitigative or remedial actions are delayed, the Commission
may take an action with significant environmental impact without
observing the provisions of these regulations. In taking an action
covered by this section, the Commission will consult with the Council as
soon as feasible concerning appropriate alternative NEPA arrangements.
Sec. 51.14 Definitions.
(a) As used in this subpart:
Categorical Exclusion means a category of actions which do not
individually or cumulatively have a significant effect on the human
environment and which the Commission has found to have no such effect in
accordance with procedures set out in Sec. 51.22, and for which,
therefore, neither an environmental assessment nor an environmental
impact statement is required.
Cooperating Agency means any Federal agency other than the NRC which
has jurisdiction by law or special expertise with respect to any
environmental impact involved in a proposal (or a reasonable
alternative) for legislation or other major Federal action significantly
affecting the quality of the human environment. By agreement with the
Commission, a State or local agency of similar qualifications or, when
the effects are on a reservation, an Indian Tribe, may become a
cooperating agency.
Council means the Council on Environmental Quality (CEQ) established
by Title II of NEPA.
DOE means the U.S. Department of Energy or its duly authorized
representatives.
Environmental Assessment means a concise public document for which
the Commission is responsible that serves to:
(1) Briefly provide sufficient evidence and analysis for determining
whether to prepare an environmental impact statement or a finding of no
significant impact.
(2) Aid the Commission's compliance with NEPA when no environmental
impact statement is necessary.
(3) Facilitate preparation of an environmental impact statement when
one is necessary.
Environmental document includes an environmental assessment, an
environmental impact statement, a finding of no significant impact, an
environmental report and any supplements to or comments upon those
documents, and a notice of intent.
Environmental Impact Statement means a detailed written statement as
required by section 102(2)(C) of NEPA.
Environmental report means a document submitted to the Commission by
an applicant for a permit, license, or other form of permission, or an
amendment to or renewal of a permit, license or other form of
permission, or by a petitioner for rulemaking, in order to aid the
Commission in complying with section 102(2) of NEPA.
Finding of No Significant Impact means a concise public document for
[[Page 11]]
which the Commission is responsible that briefly states the reasons why
an action, not otherwise excluded, will not have a significant effect on
the human environment and for which therefore an environmental impact
statement will not be prepared.
NEPA means the National Environmental Policy Act of 1969, as amended
(Pub. L. 91-190, 83 Stat. 852, 856, as amended by Pub. L. 94-83, 89
Stat. 424, 42 U.S.C. 4321, et seq.).
Notice of Intent means a notice that an environmental impact
statement will be prepared and considered.
Uranium enrichment facility means:
(1) Any facility used for separating the isotopes for uranium or
enriching uranium in the isotope 235, except laboratory scale facilities
designed or used for experimental or analytical purposes only; or
(2) Any equipment or device, or important component part especially
designed for such equipment or device, capable of separating the
isotopes of uranium or enriching uranium in the isotope 235.
(b) The definitions in 40 CFR 1508.3, 1508.7, 1508.8, 1508.14,
1508.15, 1508.16, 1508.17, 1508.18, 1508.20, 1508.23, 1508.25, 1508.26,
and 1508.27, will also be used in implementing section 102(2) of NEPA.
[49 FR 9381, Mar. 12, 1984, as amended at 57 FR 18391, Apr. 30, 1992]
Sec. 51.15 Time schedules.
Consistent with the purposes of NEPA, the Administrative Procedure
Act, the Commission's rules of practice in part 2 of this chapter,
Sec. Sec. 51.100 and 51.101, and with other essential considerations of
national policy:
(a) The appropriate NRC staff director may, and upon the request of
an applicant for a proposed action or a petitioner for rulemaking shall,
establish a time schedule for all or any constituent part of the NRC
staff NEPA process. To the maximum extent practicable, the NRC staff
will conduct its NEPA review in accordance with any time schedule
established under this section.
(b) As specified in 10 CFR part 2, the presiding officer, the Atomic
Safety and Licensing Board or the Commissioners acting as a collegial
body may establish a time schedule for all or any part of an
adjudicatory or rulemaking proceeding to the extent that each has
jurisdiction.
[49 FR 9381, Mar. 12, 1984, as amended at 69 FR 2276, Jan. 14, 2004]
Sec. 51.16 Proprietary information.
(a) Proprietary information, such as trade secrets or privileged or
confidential commercial or financial information, will be treated in
accordance with the procedures provided in Sec. 2.390 of this chapter.
(b) Any proprietary information which a person seeks to have
withheld from public disclosure shall be submitted in accordance with
Sec. 2.390 of this chapter. When submitted, the proprietary information
should be clearly identified and accompanied by a request, containing
detailed reasons and justifications, that the proprietary information be
withheld from public disclosure. A non-proprietary summary describing
the general content of the proprietary information should also be
provided.
[69 FR 2276, Jan. 14, 2004]
Sec. 51.17 Information collection requirements; OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number. OMB
has approved the information collection requirements contained in this
part under control number 3150-0021.
(b) The approved information collection requirements in this part
appear in Sec. Sec. 51.6, 51.16, 51.41, 51.45, 51.49, 51.50, 51.51,
51.52, 51.53, 51.54, 51.55, 51.58, 51.60, 51.61, 51.62, 51.66, 51.68,
and 51.69.
[49 FR 24513, June 14, 1984, as amended at 62 FR 52188, Oct. 6, 1997; 67
FR 67100, Nov. 4, 2002; 72 FR 57443, Oct. 9, 2007]
[[Page 12]]
Preliminary Procedures
classification of licensing and regulatory actions
Sec. 51.20 Criteria for and identification of licensing and regulatory
actions requiring environmental impact statements.
(a) Licensing and regulatory actions requiring an environmental
impact statement shall meet at least one of the following criteria:
(1) The proposed action is a major Federal action significantly
affecting the quality of the human environment.
(2) The proposed action involves a matter which the Commission, in
the exercise of its discretion, has determined should be covered by an
environmental impact statement.
(b) The following types of actions require an environmental impact
statement or a supplement to an environmental impact statement:
(1) Issuance of a limited work authorization or a permit to
construct a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, or issuance of an
early site permit under part 52 of this chapter.
(2) Issuance or renewal of a full power or design capacity license
to operate a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, or a combined license
under part 52 of this chapter.
(3) Issuance of a permit to construct or a design capacity license
to operate or renewal of a design capacity license to operate an
isotopic enrichment plant pursuant to part 50 of this chapter.
(4) Conversion of a provisional operating license for a nuclear
power reactor, testing facility or fuel reprocessing plant to a full
term or design capacity license pursuant to part 50 of this chapter if a
final environmental impact statement covering full term or design
capacity operation has not been previously prepared.
(5)-(6) [Reserved]
(7) Issuance of a license to possess and use special nuclear
material for processing and fuel fabrication, scrap recovery, or
conversion of uranium hexafluoride pursuant to part 70 of this chapter.
(8) Issuance of a license to possess and use source material for
uranium milling or production of uranium hexafluoride pursuant to part
40 of this chapter.
(9) Issuance of a license pursuant to part 72 of this chapter for
the storage of spent fuel in an independent spent fuel storage
installation (ISFSI) at a site not occupied by a nuclear power reactor,
or for the storage of spent fuel or high-level radioactive waste in a
monitored retrievable storage installation (MRS).
(10) Issuance of a license for a uranium enrichment facility.
(11) Issuance of renewal of a license authorizing receipt and
disposal of radioactive waste from other persons pursuant to part 61 of
this chapter.
(12) Issuance of a license amendment pursuant to part 61 of this
chapter authorizing (i) closure of a land disposal site, (ii) transfer
of the license to the disposal site owner for the purpose of
institutional control, or (iii) termination of the license at the end of
the institutional control period.
(13) Issuance of a construction authorization and license pursuant
to part 60 or part 63 of this chapter.
(14) Any other action which the Commission determines is a major
Commission action significantly affecting the quality of the human
environment. As provided in Sec. 51.22(b), the Commission may, in
special circumstances, prepare an environmental impact statement on an
action covered by a categorical exclusion.
[49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31681, Aug. 19, 1988; 53
FR 24052, June 27, 1988; 54 FR 15398, Apr. 18, 1989; 54 FR 27870, July
3, 1989; 57 FR 18392, Apr. 30, 1992; 66 FR 55790, Nov. 2, 2001; 72 FR
49509, Aug. 28, 2007]
Sec. 51.21 Criteria for and identification of licensing and regulatory
actions requiring environmental assessments.
All licensing and regulatory actions subject to this subpart require
an environmental assessment except those identified in Sec. 51.20(b) as
requiring an environmental impact statement, those identified in Sec.
51.22(c) as categorical exclusions, and those identified in Sec.
51.22(d) as other actions not requiring environmental review. As
provided in
[[Page 13]]
Sec. 51.22(b), the Commission may, in special circumstances, prepare an
environmental assessment on an action covered by a categorical
exclusion.
[54 FR 27870, July 3, 1989]
Sec. 51.22 Criterion for categorical exclusion; identification of
licensing and regulatory actions eligible for categorical exclusion
or otherwise not
requiring environmental review.
(a) Licensing, regulatory, and administrative actions eligible for
categorical exclusion shall meet the following criterion: The action
belongs to a category of actions which the Commission, by rule or
regulation, has declared to be a categorical exclusion, after first
finding that the category of actions does not individually or
cumulatively have a significant effect on the human environment.
(b) Except in special circumstances, as determined by the Commission
upon its own initiative or upon request of any interested person, an
environmental assessment or an environmental impact statement is not
required for any action within a category of actions included in the
list of categorical exclusions set out in paragraph (c) of this section.
Special circumstances include the circumstance where the proposed action
involves unresolved conflicts concerning alternative uses of available
resources within the meaning of section 102(2)(E) of NEPA.
(c) The following categories of actions are categorical exclusions:
(1) Amendments to parts 1, 2, 4, 5, 7, 8, 9, 10, 11, 12, 13, 15, 16,
19, 21, 25, 26, 55, 75, 95, 110, 140, 150, 160, 170, or 171 of this
chapter, and actions on petitions for rulemaking relating to parts 1, 2,
4, 5, 7, 9, 10, 11, 12, 13, 14, 15, 16, 19, 21, 25, 26, 55, 75, 95, 110,
140, 150, 160, 170, or 171 of this chapter.
(2) Amendments to the regulations in this chapter which are
corrective or of a minor or nonpolicy nature and do not substantially
modify existing regulations, and actions on petitions for rulemaking
relating to these amendments.
(3) Amendments to any part in this chapter which relate to--
(i) Procedures for filing and reviewing applications for licenses or
construction permits or early site permits or other forms of permission
or for amendments to or renewals of licenses or construction permits or
early site permits or other forms of permission;
(ii) Recordkeeping requirements;
(iii) Reporting requirements;
(iv) Education, training, experience, qualification or other
employment suitability requirements or
(v) Actions on petitions for rulemaking relating to these
amendments.
(4) Entrance into or amendment, suspension, or termination of all or
part of an agreement with a State pursuant to section 274 of the Atomic
Energy Act of 1954, as amended, providing for assumption by the State
and discontinuance by the Commission of certain regulatory authority of
the Commission.
(5) Procurement of general equipment and supplies.
(6) Procurement of technical assistance, confirmatory research
provided that the confirmatory research does not involve any significant
construction impacts, and personal services relating to the safe
operation and protection of commercial reactors, other facilities, and
materials subject to NRC licensing and regulation.
(7) Personnel actions.
(8) Issuance, amendment, or renewal of operators' licenses pursuant
to part 55 of this chapter.
(9) Issuance of an amendment to a permit or license for a reactor
under part 50 or part 52 of this chapter, which changes a requirement,
or grants an exemption from any such requirement, with respect to
installation or use of a facility component located within the
restricted area, as defined in part 20 of this chapter, or which changes
an inspection or a surveillance requirement, provided that:
(i) The amendment or exemption involves no significant hazards
consideration;
(ii) There is no significant change in the types or significant
increase in the amounts of any effluents that may be released offsite;
and
(iii) There is no significant increase in individual or cumulative
occupational radiation exposure.
(10) Issuance of an amendment to a permit or license issued under
this chapter which--
[[Page 14]]
(i) Changes surety, insurance and/or indemnity requirements;
(ii) Changes recordkeeping, reporting, or administrative procedures
or requirements;
(iii) Changes the licensee's or permit holder's name, phone number,
business or e-mail address;
(iv) Changes the name, position, or title of an officer of the
licensee or permit holder, including but not limited to, the radiation
safety officer or quality assurance manager; or
(v) Changes the format of the license or permit or otherwise makes
editorial, corrective or other minor revisions, including the updating
of NRC approved references.
(11) Issuance of amendments to licenses for fuel cycle plants and
radioactive waste disposal sites and amendments to materials licenses
identified in Sec. 51.60(b)(1) which are administrative,
organizational, or procedural in nature, or which result in a change in
process operations or equipment, provided that (i) there is no
significant change in the types or significant increase in the amounts
of any effluents that may be released offsite, (ii) there is no
significant increase in individual or cumulative occupational radiation
exposure, (iii) there is no significant construction impact, and (iv)
there is no significant increase in the potential for or consequences
from radiological accidents.
(12) Issuance of an amendment to a license under parts 50, 52, 60,
61, 63, 70, 72, or 75 of this chapter relating solely to safeguards
matters (i.e., protection against sabotage or loss or diversion of
special nuclear material) or issuance of an approval of a safeguards
plan submitted under parts 50, 52, 70, 72, and 73 of this chapter,
provided that the amendment or approval does not involve any significant
construction impacts. These amendments and approvals are confined to--
(i) Organizational and procedural matters;
(ii) Modifications to systems used for security and/or materials
accountability;
(iii) Administrative changes; and
(iv) Review and approval of transportation routes pursuant to 10 CFR
73.37.
(13) Approval of package designs for packages to be used for the
transportation of licensed materials.
(14) Issuance, amendment, or renewal of materials licenses issued
pursuant to 10 CFR parts 30, 31, 32, 33, 34, 35, 36, 39, 40 or part 70
authorizing the following types of activities:
(i) Distribution of radioactive material and devices or products
containing radioactive material to general licensees and to persons
exempt from licensing.
(ii) Distribution of radiopharmaceuticals, generators, reagent kits
and/or sealed sources to persons licensed pursuant to 10 CFR 35.18.
(iii) Nuclear pharmacies.
(iv) Medical and veterinary.
(v) Use of radioactive materials for research and development and
for educational purposes.
(vi) Industrial radiography.
(vii) Irradiators.
(viii) Use of sealed sources and use of gauging devices, analytical
instruments and other devices containing sealed sources.
(ix) Use of uranium as shielding material in containers or devices.
(x) Possession of radioactive material incident to performing
services such as installation, maintenance, leak tests and calibration.
(xi) Use of sealed sources and/or radioactive tracers in well-
logging procedures.
(xii) Acceptance of packaged radioactive wastes from others for
transfer to licensed land burial facilities provided the interim storage
period for any package does not exceed 180 days and the total possession
limit for all packages held in interim storage at the same time does not
exceed 50 curies.
(xiii) Manufacturing or processing of source, byproduct, or special
nuclear materials for distribution to other licensees, except processing
of source material for extraction of rare earth and other metals.
(xiv) Nuclear laundries.
(xv) Possession, manufacturing, processing, shipment, testing, or
other use of depleted uranium military munitions.
(xvi) Any use of source, byproduct, or special nuclear material not
listed above which involves quantities and
[[Page 15]]
forms of source, byproduct, or special nuclear material similar to those
listed in paragraphs (c)(14) (i) through (xv) of this section (Category
14).
(15) Issuance, amendment or renewal of licenses for import of
nuclear facilities and materials pursuant to part 110 of this chapter,
except for import of spent power reactor fuel.
(16) Issuance or amendment of guides for the implementation of
regulations in this chapter, and issuance or amendment of other
informational and procedural documents that do not impose any legal
requirements.
(17) Issuance of an amendment to a permit or license under parts 30,
40, 50, 52, or part 70 of this chapter which deletes any limiting
condition of operation or monitoring requirement based on or applicable
to any matter subject to the provisions of the Federal Water Pollution
Control Act.
(18) Issuance of amendments or orders authorizing licensees of
production or utilization facilities to resume operation, provided the
basis for the authorization rests solely on a determination or
redetermination by the Commission that applicable emergency planning
requirements are met.
(19) Issuance, amendment, modification, or renewal of a certificate
of compliance of gaseous diffusion enrichment facilities pursuant to 10
CFR part 76.
(20) Decommissioning of sites where licensed operations have been
limited to the use of--
(i) Small quantities of short-lived radioactive materials;
(ii) Radioactive materials in sealed sources, provided there is no
evidence of leakage of radioactive material from these sealed sources;
or
(iii) Radioactive materials in such a manner that a decommissioning
plan is not required by 10 CFR 30.36(g)(1), 40.42(g)(1), or 70.38(g)(1),
and the NRC has determined that the facility meets the radiological
criteria for unrestricted use in 10 CFR 20.1402 without further
remediation or analysis.
(21) Approvals of direct or indirect transfers of any license issued
by NRC and any associated amendments of license required to reflect the
approval of a direct or indirect transfer of an NRC license.
(22) Issuance of a standard design approval under part 52 of this
chapter.
(23) The Commission finding for a combined license under Sec.
52.103(g) of this chapter.
(24) Grants to institutions of higher education in the United
States, to fund scholarships, fellowships, and stipends for the study of
science, engineering, or another field of study that the NRC determines
is in a critical skill area related to its regulatory mission, to
support faculty and curricular development in such fields, and to
support other domestic educational, technical assistance, or training
programs (including those of trade schools) in such fields, except to
the extent that such grants or programs include activities directly
affecting the environment, such as:
(i) The construction of facilities;
(ii) A major disturbance brought about by blasting, drilling,
excavating or other means;
(iii) Field work, except that which only involves noninvasive or
non-harmful techniques such as taking water or soil samples or
collecting non-protected species of flora and fauna; or
(iv) The release of radioactive material.
(25) Granting of an exemption from the requirements of any
regulation of this chapter, provided that--
(i) There is no significant hazards consideration;
(ii) There is no significant change in the types or significant
increase in the amounts of any effluents that may be released offsite;
(iii) There is no significant increase in individual or cumulative
public or occupational radiation exposure;
(iv) There is no significant construction impact;
(v) There is no significant increase in the potential for or
consequences from radiological accidents; and
(vi) The requirements from which an exemption is sought involve:
(A) Recordkeeping requirements;
(B) Reporting requirements;
(C) Inspection or surveillance requirements;
(D) Equipment servicing or maintenance scheduling requirements;
[[Page 16]]
(E) Education, training, experience, qualification, requalification
or other employment suitability requirements;
(F) Safeguard plans, and materials control and accounting inventory
scheduling requirements;
(G) Scheduling requirements;
(H) Surety, insurance or indemnity requirements; or
(I) Other requirements of an administrative, managerial, or
organizational nature.
(d) In accordance with section 121 of the Nuclear Waste Policy Act
of 1982 (42 U.S.C. 10141), the promulgation of technical requirements
and criteria that the Commission will apply in approving or disapproving
applications under part 60 or 63 of this chapter shall not require an
environmental impact statement, an environmental assessment, or any
environmental review under subparagraph (E) or (F) of section 102(2) of
NEPA.
[49 FR 9381, Mar. 12, 1984]
Editorial Note: For Federal Register citations affecting Sec.
51.22, see the List of CFR Sections Affected, which appears in the
Finding Aids section of the printed volume and at www.fdsys.gov.
Sec. 51.23 Temporary storage of spent fuel after cessation of reactor
operation--generic determination of no significant environmental impact.
(a) The Commission has made a generic determination that, if
necessary, spent fuel generated in any reactor can be stored safely and
without significant environmental impacts for at least 30 years beyond
the licensed life for operation (which may include the term of a revised
or renewed license) of that reactor at its spent fuel storage basin or
at either onsite or offsite independent spent fuel storage
installations. Further, the Commission believes there is reasonable
assurance that at least one mined geologic repository will be available
within the first quarter of the twenty-first century, and sufficient
repository capacity will be available within 30 years beyond the
licensed life for operation of any reactor to dispose of the commercial
high-level waste and spent fuel originating in such reactor and
generated up to that time.
(b) Accordingly, as provided in Sec. Sec. 51.30(b), 51.53, 51.61,
51.80(b), 51.95, and 51.97(a), and within the scope of the generic
determination in paragraph (a) of this section, no discussion of any
environmental impact of spent fuel storage in reactor facility storage
pools or independent spent fuel storage installations (ISFSI) for the
period following the term of the reactor operating license or amendment,
reactor combined license or amendment, or initial ISFSI license or
amendment for which application is made, is required in any
environmental report, environmental impact statement, environmental
assessment, or other analysis prepared in connection with the issuance
or amendment of an operating license for a nuclear power reactor under
parts 50 and 54 of this chapter, or issuance or amendment of a combined
license for a nuclear power reactor under parts 52 and 54 of this
chapter, or the issuance of an initial license for storage of spent fuel
at an ISFSI, or any amendment thereto.
(c) This section does not alter any requirements to consider the
environmental impacts of spent fuel storage during the term of a reactor
operating license or combined license, or a license for an ISFSI in a
licensing proceeding.
[49 FR 34694, Aug. 31, 1984, as amended at 55 FR 38474, Sept. 18, 1990;
72 FR 49509, Aug. 28, 2007]
Effective Date Note: At 75 FR 81037, December 23, 2010, Sec. 51.23
was amended by revising paragraph (a), effective Jan. 24, 2011. For the
convenience of the user, the revised text is set forth as follows:
Sec. 51.23 Temporary storage of spent fuel after cessation of reactor
operation--generic determination of no significant
environmental impact.
(a) The Commission has made a generic determination that, if
necessary, spent fuel generated in any reactor can be stored safely and
without significant environmental impacts for at least 60 years beyond
the licensed life for operation (which may include the term of a revised
or renewed license) of that reactor in a combination of storage in its
spent fuel storage basin and at either onsite or offsite independent
spent fuel storage installations. Further, the Commission believes there
is reasonable assurance that sufficient mined geologic repository
capacity will be available to dispose of the commercial high-level
radioactive waste and spent
[[Page 17]]
fuel generated in any reactor when necessary.
* * * * *
determinations to prepare environmental impact statements, environmental
assessments or findings of no significant impact, and related procedures
Sec. 51.25 Determination to prepare environmental impact statement or
environmental assessment; eligibility for categorical exclusion.
Before taking a proposed action subject to the provisions of this
subpart, the appropriate NRC staff director will determine on the basis
of the criteria and classifications of types of actions in Sec. Sec.
51.20, 51.21 and 51.22 of this subpart whether the proposed action is of
the type listed in Sec. 51.22(c) as a categorical exclusion or whether
an environmental impact statement or an environmental assessment should
be prepared. An environmental assessment is not necessary if it is
determined that an environmental impact statement will be prepared.
Sec. 51.26 Requirement to publish notice of intent and conduct
scoping process.
(a) Whenever the appropriate NRC staff director determines that an
environmental impact statement will be prepared by NRC in connection
with a proposed action, a notice of intent will be prepared as provided
in Sec. 51.27, and will be published in the Federal Register as
provided in Sec. 51.116, and an appropriate scoping process (see
Sec. Sec. 51.27, 51.28, and 51.29) will be conducted.
(b) The scoping process may include a public scoping meeting.
(c) Upon receipt of an application and accompanying environmental
impact statement under Sec. 60.22 or Sec. 63.22 of this chapter
(pertaining to geologic repositories for high-level radioactive waste),
the appropriate NRC staff director will include in the notice of
docketing required to be published by Sec. 2.101(f)(8) of this chapter
a statement of Commission intention to adopt the environmental impact
statement to the extent practicable. However, if the appropriate NRC
staff director determines, at the time of such publication or at any
time thereafter, that NRC should prepare a supplemental environmental
impact statement in connection with the Commission's action on the
license application, the NRC shall follow the procedures set out in
paragraph (a) of this section.
(d) Whenever the appropriate NRC staff director determines that a
supplement to an environmental impact statement will be prepared by the
NRC, a notice of intent will be prepared as provided in Sec. 51.27, and
will be published in the Federal Register as provided in Sec. 51.116.
The NRC staff need not conduct a scoping process (see Sec. Sec. 51.27,
51.28, and 51.29), provided, however, that if scoping is conducted, then
the scoping must be directed at matters to be addressed in the
supplement. If scoping is conducted in a proceeding for a combined
license referencing an early site permit under part 52, then the scoping
must be directed at matters to be addressed in the supplement as
described in Sec. 51.92(e).
[49 FR 9381, Mar. 12, 1984, as amended at 54 FR 27870, July 3, 1989; 66
FR 55791, Nov. 2, 2001; 72 FR 49510, Aug. 28, 2007]
Sec. 51.27 Notice of intent.
(a) The notice of intent required by Sec. 51.26(a) shall:
(1) State that an environmental impact statement will be prepared;
(2) Describe the proposed action and, to the extent sufficient
information is available, possible alternatives;
(3) State whether the applicant or petitioner for rulemaking has
filed an environmental report, and, if so, where copies are available
for public inspection;
(4) Describe the proposed scoping process, including the role of
participants, whether written comments will be accepted, the last date
for submitting comments and where comments should be sent, whether a
public scoping meeting will be held, the time and place of any scoping
meeting or when the time and place of the meeting will be announced; and
(5) State the name, address and telephone number of an individual in
NRC who can provide information about the proposed action, the scoping
process, and the environmental impact statement.
[[Page 18]]
(b) The notice of intent required by Sec. 51.26(d) shall:
(1) State that a supplement to a final environmental impact
statement will be prepared in accordance with Sec. 51.72 or Sec.
51.92. For a combined license application that references an early site
permit, the supplement to the early site permit environmental impact
statement will be prepared in accordance with Sec. 51.92(e);
(2) Describe the proposed action and, to the extent required,
possible alternatives. For the case of a combined license referencing an
early site permit, identify the proposed action as the issuance of a
combined license for the construction and operation of a nuclear power
plant as described in the combined license application at the site
described in the early site permit referenced in the combined license
application;
(3) Identify the environmental report prepared by the applicant and
information on where copies are available for public inspection;
(4) Describe the matters to be addressed in the supplement to the
final environmental impact statement;
(5) Describe any proposed scoping process that the NRC staff may
conduct, including the role of participants, whether written comments
will be accepted, the last date for submitting comments and where
comments should be sent, whether a public scoping meeting will be held,
the time and place of any scoping meeting or when the time and place of
the meeting will be announced; and
(6) State the name, address, and telephone number of an individual
in NRC who can provide information about the proposed action, the
scoping process, and the supplement to the environmental impact
statement.
[49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49510, Aug. 28, 2007]
scoping
Sec. 51.28 Scoping--participants.
(a) The appropriate NRC staff director shall invite the following
persons to participate in the scoping process:
(1) The applicant or the petitioner for rulemaking;
(2) Any person who has petitioned for leave to intervene in the
proceeding or who has been admitted as a party to the proceeding;
(3) Any other Federal agency which has jurisdiction by law or
special expertise with respect to any environmental impact involved or
which is authorized to develop and enforce relevant environmental
standards;
(4) Affected State and local agencies, including those authorized to
develop and enforce relevant environmental standards;
(5) Any affected Indian tribe; and
(6) Any person who has requested an opportunity to participate in
the scoping process.
(b) The appropriate NRC staff director may also invite any other
appropriate person to participate in the scoping process.
(c) Participation in the scoping process for an environmental impact
statement does not entitle the participant to become a party to the
proceeding to which the environmental impact statement relates.
Participation in an adjudicatory proceeding is governed by the
procedures in Sec. Sec. 2.309 and 2.315 of this chapter. Participation
in a rulemaking proceeding in which the Commission has decided to have a
hearing is governed by the provisions in the notice of hearing.
[49 FR 9381, Mar. 12, 1984, as amended at 74 FR 62682, Dec. 1, 2009]
Sec. 51.29 Scoping-environmental impact statement and supplement to
environmental impact statement.
(a) The scoping process for an environmental impact statement shall
begin as soon as practicable after publication of the notice of intent
as provided in Sec. 51.116, and shall be used to:
(1) Define the proposed action which is to be the subject of the
statement or supplement. For environmental impact statements other than
a supplement to an early site permit final environmental impact
statement prepared for a combined license application, the provisions of
40 CFR 1502.4 will be used for this purpose. For a supplement to an
early site permit final environmental impact statement prepared for a
combined license application, the
[[Page 19]]
proposed action shall be as set forth in the relevant provisions of
Sec. 51.92(e).
(2) Determine the scope of the statement and identify the
significant issues to be analyzed in depth.
(3) Identify and eliminate from detailed study issues which are
peripheral or are not significant or which have been covered by prior
environmental review. Discussion of these issues in the statement will
be limited to a brief presentation of why they are peripheral or will
not have a significant effect on the quality of the human environment or
a reference to their coverage elsewhere.
(4) Identify any environmental assessments and other environmental
impact statements which are being or will be prepared that are related
to but are not part of the scope of the statement under consideration.
(5) Identify other environmental review and consultation
requirements related to the proposed action so that other required
analyses and studies may be prepared concurrently and integrated with
the environmental impact statement.
(6) Indicate the relationship between the timing of the preparation
of environmental analyses and the Commission's tentative planning and
decision-making schedule.
(7) Identify any cooperating agencies, and as appropriate, allocate
assignments for preparation and schedules for completion of the
statement to the NRC and any cooperating agencies.
(8) Describe the means by which the environmental impact statement
will be prepared, including any contractor assistance to be used.
(b) At the conclusion of the scoping process, the appropriate NRC
staff director will prepare a concise summary of the determinations and
conclusions reached, including the significant issues identified, and
will send a copy of the summary to each participant in the scoping
process.
(c) At any time prior to issuance of the draft environmental impact
statement, the appropriate NRC staff director may revise the
determinations made under paragraph (b) of this section, as appropriate,
if substantial changes are made in the proposed action, or if
significant new circumstances or information arise which bear on the
proposed action or its impacts.
[49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49510, Aug. 28, 2007]
environmental assessment
Sec. 51.30 Environmental assessment.
(a) An environmental assessment for proposed actions, other than
those for a standard design certification under 10 CFR part 52 or a
manufacturing license under part 52, shall identify the proposed action
and include:
(1) A brief discussion of:
(i) The need for the proposed action;
(ii) Alternatives as required by section 102(2)(E) of NEPA;
(iii) The environmental impacts of the proposed action and
alternatives as appropriate; and
(2) A list of agencies and persons consulted, and identification of
sources used.
(b) Unless otherwise determined by the Commission, an environmental
assessment will not include discussion of any aspect of the storage of
spent fuel within the scope of the generic determination in Sec.
51.23(a) and in accordance with the provisions of Sec. 51.23(b).
(c) An environmental assessment for a proposed action regarding a
monitored retrievable storage installation (MRS) will not address the
need for the MRS or any alternative to the design criteria for an MRS
set forth in section 141(b)(1) of the Nuclear Waste Policy Act of 1982
(96 Stat. 2242, 42 U.S.C. 10161(b)(1)).
(d) An environmental assessment for a standard design certification
under subpart B of part 52 of this chapter must identify the proposed
action, and will be limited to the consideration of the costs and
benefits of severe accident mitigation design alternatives and the bases
for not incorporating severe accident mitigation design alternatives in
the design certification. An environmental assessment for an amendment
to a design certification will be limited to the consideration of
whether the design change which is the subject of the proposed amendment
renders a severe accident mitigation design alternative previously
rejected
[[Page 20]]
in the earlier environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives, in which case the costs and benefits of new severe
accident mitigation design alternatives and the bases for not
incorporating new severe accident mitigation design alternatives in the
design certification must be addressed.
(e) An environmental assessment for a manufacturing license under
subpart F of part 52 of this chapter must identify the proposed action,
and will be limited to the consideration of the costs and benefits of
severe accident mitigation design alternatives and the bases for not
incorporating severe accident mitigation design alternatives in the
manufacturing license. An environmental assessment for an amendment to a
manufacturing license will be limited to consideration of whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives, in which case the costs and benefits of new severe
accident mitigation design alternatives and the bases for not
incorporating new severe accident mitigation design alternatives in the
manufacturing license must be addressed. In either case, the
environmental assessment will not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
[49 FR 9381, Mar. 12, 1984, as amended at 49 FR 34694, Aug. 31, 1984; 53
FR 31681, Aug. 19, 1988; 72 FR 49510, Aug. 28, 2007]
Sec. 51.31 Determinations based on environmental assessment.
(a) General. Upon completion of an environmental assessment for
proposed actions other than those involving a standard design
certification or a manufacturing license under part 52 of this chapter,
the appropriate NRC staff director will determine whether to prepare an
environmental impact statement or a finding of no significant impact on
the proposed action. As provided in Sec. 51.33, a determination to
prepare a draft finding of no significant impact may be made.
(b) Standard design certification. (1) For actions involving the
issuance or amendment of a standard design certification, the Commission
shall prepare a draft environmental assessment for public comment as
part of the proposed rule. The proposed rule must state that:
(i) The Commission has determined in Sec. 51.32 that there is no
significant environmental impact associated with the issuance of the
standard design certification or its amendment, as applicable; and
(ii) Comments on the environmental assessment will be limited to the
consideration of SAMDAs as required by Sec. 51.30(d).
(2) The Commission will prepare a final environmental assessment
following the close of the public comment period for the proposed
standard design certification.
(c) Manufacturing license. (1) Upon completion of the environmental
assessment for actions involving issuance or amendment of a
manufacturing license (manufacturing license environmental assessment),
the appropriate NRC staff director will determine the costs and benefits
of severe accident mitigation design alternatives and the bases for not
incorporating severe accident mitigation design alternatives in the
design of the reactor to be manufactured under the manufacturing
license. The NRC staff director may determine to prepare a draft
environmental assessment.
(2) The manufacturing license environmental assessment must state
that:
(i) The Commission has determined in Sec. 51.32 that there is no
significant environmental impact associated with the issuance of a
manufacturing license or an amendment to a manufacturing license, as
applicable;
(ii) The environmental assessment will not address the environmental
impacts associated with manufacturing the reactor under the
manufacturing license; and
(iii) Comments on the environmental assessment will be limited to
the consideration of severe accident mitigation design alternatives as
required by Sec. 51.30(e).
[[Page 21]]
(3) If the NRC staff director makes a determination to prepare and
issue a draft environmental assessment for public review and comment
before making a final determination on the manufacturing license
application, the assessment will be marked, ``Draft.'' The NRC notice of
availability on the draft environmental assessment will include a
request for comments which specifies where comments should be submitted
and when the comment period expires. The notice will state that copies
of the environmental assessment and any related environmental documents
are available for public inspection and where inspections can be made. A
copy of the final environmental assessment will be sent to the U.S.
Environmental Protection Agency, the applicant, any party to a
proceeding, each commenter, and any other Federal, State, and local
agencies, and Indian tribes, State, regional, and metropolitan
clearinghouses expressing an interest in the action. Additional copies
will be made available in accordance with Sec. 51.123.
(4) When a hearing is held under the regulations in part 2 of this
chapter on the proposed issuance of the manufacturing license or
amendment, the NRC staff director will prepare a final environmental
assessment which may be subject to modification as a result of review
and decision as appropriate to the nature and scope of the proceeding.
(5) Only a party admitted into the proceeding with respect to a
contention on the environmental assessment, or an entity participating
in the proceeding pursuant to Sec. 2.315(c) of this chapter, may take a
position and offer evidence on the matters within the scope of the
environmental assessment.
[72 FR 49510, Aug. 28, 2007]
finding of no significant impact
Sec. 51.32 Finding of no significant impact.
(a) A finding of no significant impact will:
(1) Identify the proposed action;
(2) State that the Commission has determined not to prepare an
environmental impact statement for the proposed action;
(3) Briefly present the reasons why the proposed action will not
have a significant effect on the quality of the human environment;
(4) Include the environmental assessment or a summary of the
environmental assessment. If the assessment is included, the finding
need not repeat any of the discussion in the assessment but may
incorporate it by reference;
(5) Note any other related environmental documents; and
(6) State that the finding and any related environmental documents
are available for public inspection and where the documents may be
inspected.
(b) The Commission finds that there is no significant environmental
impact associated with the issuance of:
(1) A standard design certification under subpart B of part 52 of
this chapter;
(2) An amendment to a design certification;
(3) A manufacturing license under subpart F of part 52 of this
chapter; or
(4) An amendment to a manufacturing license.
[49 FR 9381, Mar. 12, 1984, as amended at 72 FR 49511, Aug. 28, 2007]
Sec. 51.33 Draft finding of no significant impact; distribution.
(a) As provided in paragraph (b) of this section, the appropriate
NRC staff director may make a determination to prepare and issue a draft
finding of no significant impact for public review and comment before
making a final determination whether to prepare an environmental impact
statement or a final finding of no significant impact on the proposed
action.
(b) Circumstances in which a draft finding of no significant impact
may be prepared will ordinarily include the following:
(1) A finding of no significant impact appears warranted for the
proposed action but the proposed action is (i) closely similar to one
which normally requires the preparation of an environmental impact
statement, or (ii) without precedent; and
(2) The appropriate NRC staff director determines that preparation
of a draft finding of no significant impact will further the purposes of
NEPA.
[[Page 22]]
(c) A draft finding of no significant impact will (1) be marked
``Draft'', (2) contain the information specified in Sec. 51.32, (3) be
accompanied by or include a request for comments on the proposed action
and on the draft finding within thirty (30) days, or such longer period
as may be specified in the notice of the draft finding, and (4) be
published in the Federal Register as required by Sec. Sec. 51.35 and
51.119.
(d) A draft finding will be distributed as provided in Sec.
51.74(a). Additional copies will be made available in accordance with
Sec. 51.123.
(e) When a draft finding of no significant impact is issued for a
proposed action, a final determination to prepare an environmental
impact statement or a final finding of no significant impact for that
action shall not be made until the last day of the public comment period
has expired.
Sec. 51.34 Preparation of finding of no significant impact.
(a) Except as provided in paragraph (b) of this section, the finding
of no significant impact will be prepared by the NRC staff director
authorized to take the action.
(b) When a hearing is held on the proposed action under the
regulations in subpart G of part 2 of this chapter or when the action
can only be taken by the Commissioners acting as a collegial body, the
appropriate NRC staff director will prepare a proposed finding of no
significant impact which may be subject to modification as a result of
review and decision as appropriate to the nature and scope of the
proceeding. In such cases, the presiding officer, the Atomic Safety and
Licensing Appeal Board, or the Commission acting as a collegial body, as
appropriate, will issue the final finding of no significant impact.
Sec. 51.35 Requirement to publish finding of no significant impact;
limitation on Commission action.
(a) Whenever the Commission makes a draft or final finding of no
significant impact on a proposed action, the finding will be published
in the Federal Register as provided in Sec. 51.119.
(b) Except as provided in Sec. 51.13, the Commission shall not take
the proposed action until after the final finding has been published in
the Federal Register.
Environmental Reports and Information--Requirements Applicable to
Applicants and Petitioners for Rulemaking
general
Sec. 51.40 Consultation with NRC staff.
(a) A prospective applicant or petitioner for rulemaking is
encouraged to confer with NRC staff as early as possible in its planning
process before submitting environmental information or filing an
environmental report.
(b) Requests for guidance or information on environmental matters
may include inquiries relating to:
(1) Applicable NRC rules and regulations;
(2) Format, content and procedures for filing environmental reports
and other environmental information, including the type and quantity of
environmental information likely to be needed to address issues and
concerns identified in the scoping process described in Sec. 51.29 in a
manner appropriate to their relative significance;
(3) Availability of relevant environmental studies and environmental
information;
(4) Need for, appropriate level and scope of any environmental
studies or information which the Commission may require to be submitted
in connection with an application or petition for rulemaking;
(5) Public meetings with NRC staff.
(c) Questions concerning environmental matters should be addressed
to the following NRC staff offices as appropriate:
(1) Utilization facilities: ATTN: Document Control Desk, Director,
Office of Nuclear Reactor Regulation or Director, Office of New
Reactors, as appropriate, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone (301) 415-1270, e-mail
[email protected].
(2) Production facilities: ATTN: Document Control Desk, Director,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001,
[[Page 23]]
telephone (301) 415-7800, e-mail [email protected].
(3) Materials licenses: ATTN: Document Control Desk, Director,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
7800, e-mail [email protected].
(4) Rulemaking: ATTN: Chief, Rules and Directives Branch, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (800) 368-5642, e-mail [email protected].
(5) General Environmental Matters: Executive Director for
Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555,
Telephone: (301) 415-1700.
[49 FR 9381, Mar. 12, 1984, as amended at 53 FR 13399, Apr. 25, 1988; 60
FR 24552, May 9, 1995; 68 FR 58810, Oct. 10, 2003; 73 FR 5724, Jan. 31,
2008]
Sec. 51.41 Requirement to submit environmental information.
The Commission may require an applicant for a permit, license, or
other form of permission, or amendment to or renewal of a permit,
license or other form of permission, or a petitioner for rulemaking to
submit such information to the Commission as may be useful in aiding the
Commission in complying with section 102(2) of NEPA. The Commission will
independently evaluate and be responsible for the reliability of any
information which it uses.
environmental reports--general requirements
Sec. 51.45 Environmental report.
(a) General. As required by Sec. Sec. 51.50, 51.53, 51.54, 51.55,
51.60, 51.61, 51.62, or 51.68, as appropriate, each applicant or
petitioner for rulemaking shall submit with its application or petition
for rulemaking one signed original of a separate document entitled
``Applicant's'' or ``Petitioner's Environmental Report,'' as
appropriate. An applicant or petitioner for rulemaking may submit a
supplement to an environmental report at any time.
(b) Environmental considerations. The environmental report shall
contain a description of the proposed action, a statement of its
purposes, a description of the environment affected, and discuss the
following considerations:
(1) The impact of the proposed action on the environment. Impacts
shall be discussed in proportion to their significance;
(2) Any adverse environmental effects which cannot be avoided should
the proposal be implemented;
(3) Alternatives to the proposed action. The discussion of
alternatives shall be sufficiently complete to aid the Commission in
developing and exploring, pursuant to section 102(2)(E) of NEPA,
``appropriate alternatives to recommended courses of action in any
proposal which involves unresolved conflicts concerning alternative uses
of available resources.'' To the extent practicable, the environmental
impacts of the proposal and the alternatives should be presented in
comparative form;
(4) The relationship between local short-term uses of man's
environment and the maintenance and enhancement of long-term
productivity; and
(5) Any irreversible and irretrievable commitments of resources
which would be involved in the proposed action should it be implemented.
(c) Analysis. The environmental report must include an analysis that
considers and balances the environmental effects of the proposed action,
the environmental impacts of alternatives to the proposed action, and
alternatives available for reducing or avoiding adverse environmental
effects. An environmental report prepared at the early site permit stage
under Sec. 51.50(b), limited work authorization stage under Sec.
51.49, construction permit stage under Sec. 51.50(a), or combined
license stage under Sec. 51.50(c) must include a description of impacts
of the preconstruction activities performed by the applicant at the
proposed site (i.e., those activities listed in paragraphs (2)(i)
through (2)(x) in the definition of ``construction'' contained in Sec.
51.4), necessary to support the construction and operation of the
facility which is the subject of the early site permit, limited work
authorization, construction permit, or combined license application. The
environmental report must also contain an
[[Page 24]]
analysis of the cumulative impacts of the activities to be authorized by
the limited work authorization, construction permit, or combined license
in light of the preconstruction impacts described in the environmental
report. Except for an environmental report prepared at the early site
permit stage, or an environmental report prepared at the license renewal
stage under Sec. 51.53(c), the analysis in the environmental report
should also include consideration of the economic, technical, and other
benefits and costs of the proposed action and its alternatives.
Environmental reports prepared at the license renewal stage under Sec.
51.53(c) need not discuss the economic or technical benefits and costs
of either the proposed action or alternatives except if these benefits
and costs are either essential for a determination regarding the
inclusion of an alternative in the range of alternatives considered or
relevant to mitigation. In addition, environmental reports prepared
under Sec. 51.53(c) need not discuss issues not related to the
environmental effects of the proposed action and its alternatives. The
analyses for environmental reports shall, to the fullest extent
practicable, quantify the various factors considered. To the extent that
there are important qualitative considerations or factors that cannot be
quantified, those considerations or factors shall be discussed in
qualitative terms. The environmental report should contain sufficient
data to aid the Commission in its development of an independent
analysis.
(d) Status of compliance. The environmental report shall list all
Federal permits, licenses, approvals and other entitlements which must
be obtained in connection with the proposed action and shall describe
the status of compliance with these requirements. The environmental
report shall also include a discussion of the status of compliance with
applicable environmental quality standards and requirements including,
but not limited to, applicable zoning and land-use regulations, and
thermal and other water pollution limitations or requirements which have
been imposed by Federal, State, regional, and local agencies having
responsibility for environmental protection. The discussion of
alternatives in the report shall include a discussion of whether the
alternatives will comply with such applicable environmental quality
standards and requirements.
(e) Adverse information. The information submitted pursuant to
paragraphs (b) through (d) of this section should not be confined to
information supporting the proposed action but should also include
adverse information.
[49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28486, June 5, 1996; 61
FR 66542, Dec. 18, 1996; 68 FR 58810, Oct. 10, 2003; 72 FR 49511, Aug.
28, 2007; 72 FR 57443, Oct. 9, 2007; 73 FR 22787, Apr. 28, 2008]
environmental reports--production and utilization facilities
Sec. 51.49 Environmental report--limited work authorization.
(a) Limited work authorization submitted as part of complete
construction permit or combined license application. Each applicant for
a construction permit or combined license applying for a limited work
authorization under Sec. 50.10(d) of this chapter in a complete
application under 10 CFR 2.101(a)(1) through (a)(4), shall submit with
its application a separate document, entitled, ``Applicant's
Environmental Report--Limited Work Authorization Stage,'' which is in
addition to the environmental report required by Sec. 51.50 of this
part. Each environmental report must also contain the following
information:
(1) A description of the activities proposed to be conducted under
the limited work authorization;
(2) A statement of the need for the activities; and
(3) A description of the environmental impacts that may reasonably
be expected to result from the activities, the mitigation measures that
the applicant proposes to implement to achieve the level of
environmental impacts described, and a discussion of the reasons for
rejecting mitigation measures that could be employed by the applicant to
further reduce environmental impacts.
(b) Phased application for limited work authorization and
construction permit or combined license. If the construction permit or
combined license application
[[Page 25]]
is filed in accordance with Sec. 2.101(a)(9) of this chapter, then the
environmental report for part one of the application may be limited to a
discussion of the activities proposed to be conducted under the limited
work authorization. If the scope of the environmental report for part
one is so limited, then part two of the application must include the
information required by Sec. 51.50, as applicable.
(c) Limited work authorization submitted as part of an early site
permit application. Each applicant for an early site permit under
subpart A of part 52 of this chapter requesting a limited work
authorization shall submit with its application the environmental report
required by Sec. 51.50(b). Each environmental report must contain the
following information:
(1) A description of the activities proposed to be conducted under
the limited work authorization;
(2) A statement of the need for the activities; and
(3) A description of the environmental impacts that may reasonably
be expected to result from the activities, the mitigation measures that
the applicant proposes to implement to achieve the level of
environmental impacts described, and a discussion of the reasons for
rejecting mitigation measures that could be employed by the applicant to
further reduce environmental impacts.
(d) Limited work authorization request submitted by early site
permit holder. Each holder of an early site permit requesting a limited
work authorization shall submit with its application a document
entitled, ``Applicant's Environmental Report--Limited Work Authorization
under Early Site Permit,'' containing the following information:
(1) A description of the activities proposed to be conducted under
the limited work authorization;
(2) A statement of the need for the activities;
(3) A description of the environmental impacts that may reasonably
be expected to result from the activities, the mitigation measures that
the applicant proposes to implement to achieve the level of
environmental impacts described, and a discussion of the reasons for
rejecting mitigation measures that could be employed by the applicant to
further reduce environmental impacts; and
(4) Any new and significant information for issues related to the
impacts of construction of the facility that were resolved in the early
site permit proceeding with respect to the environmental impacts of the
activities to be conducted under the limited work authorization.
(5) A description of the process used to identify new and
significant information regarding NRC's conclusions in the early site
permit environmental impact statement. The process must be a reasonable
methodology for identifying this new and significant information.
(e) Limited work authorization for a site where an environmental
impact statement was prepared, but the facility construction was not
completed. If the limited work authorization is for activities to be
conducted at a site for which the Commission has previously prepared an
environmental impact statement for the construction and operation of a
nuclear power plant, and a construction permit was issued but
construction of the plant was never completed, then the applicant's
environmental report may incorporate by reference the earlier
environmental impact statement. In the event of such referencing, the
environmental report must identify:
(1) Any new and significant information material to issues related
to the impacts of construction of the facility that were resolved in the
construction permit proceeding for the matters required to be addressed
in paragraph (a) of this section; and
(2) A description of the process used to identify new and
significant information regarding the NRC's conclusions in the
construction permit environmental impact statement. The process must use
a reasonable methodology for identifying this new and significant
information.
(f) Environmental report. An environmental report submitted in
accordance with this section must separately evaluate the environmental
impacts and proposed alternatives attributable to the activities
proposed to be conducted under the limited work authorization. At the
option of the applicant,
[[Page 26]]
the ``Applicant's Environmental Report--Limited Work Authorization
Stage,'' may contain the information required to be submitted in the
environmental report required under Sec. 51.50, which addresses the
impacts of construction and operation for the proposed facility
(including the environmental impacts attributable to the limited work
authorization), and discusses the overall costs and benefits balancing
for the proposed action.
[72 FR 57444, Oct. 9, 2007]
Sec. 51.50 Environmental report--construction permit, early site
permit, or combined license stage.
(a) Construction permit stage. Each applicant for a permit to
construct a production or utilization facility covered by Sec. 51.20
shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Construction Permit Stage,'' which
shall contain the information specified in Sec. Sec. 51.45, 51.51, and
51.52. Each environmental report shall identify procedures for reporting
and keeping records of environmental data, and any conditions and
monitoring requirements for protecting the non-aquatic environment,
proposed for possible inclusion in the license as environmental
conditions in accordance with Sec. 50.36b of this chapter.
(b) Early site permit stage. Each applicant for an early site permit
shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Early Site Permit Stage,'' which
shall contain the information specified in Sec. Sec. 51.45, 51.51, and
51.52, as modified in this paragraph.
(1) The environmental report must include an evaluation of
alternative sites to determine whether there is any obviously superior
alternative to the site proposed.
(2) The environmental report may address one or more of the
environmental effects of construction and operation of a reactor, or
reactors, which have design characteristics that fall within the site
characteristics and design parameters for the early site permit
application, provided however, that the environmental report must
address all environmental effects of construction and operation
necessary to determine whether there is any obviously superior
alternative to the site proposed. The environmental report need not
include an assessment of the economic, technical, or other benefits (for
example, need for power) and costs of the proposed action or an
evaluation of alternative energy sources.
(3) For other than light-water-cooled nuclear power reactors, the
environmental report must contain the basis for evaluating the
contribution of the environmental effects of fuel cycle activities for
the nuclear power reactor.
(4) Each environmental report must identify the procedures for
reporting and keeping records of environmental data, and any conditions
and monitoring requirements for protecting the non-aquatic environment,
proposed for possible inclusion in the license as environmental
conditions in accordance with Sec. 50.36b of this chapter.
(c) Combined license stage. Each applicant for a combined license
shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Combined License Stage.'' Each
environmental report shall contain the information specified in
Sec. Sec. 51.45, 51.51, and 51.52, as modified in this paragraph. For
other than light-water-cooled nuclear power reactors, the environmental
report shall contain the basis for evaluating the contribution of the
environmental effects of fuel cycle activities for the nuclear power
reactor. Each environmental report shall identify procedures for
reporting and keeping records of environmental data, and any conditions
and monitoring requirements for protecting the non-aquatic environment,
proposed for possible inclusion in the license as environmental
conditions in accordance with Sec. 50.36b of this chapter. The combined
license environmental report may reference information contained in a
final environmental document previously prepared by the NRC staff.
(1) Application referencing an early site permit. If the combined
license application references an early site permit, then the
``Applicant's Environmental Report--Combined License Stage'' need not
contain information or analyses submitted to the Commission in
``Applicant's Environmental Report--Early
[[Page 27]]
Site Permit Stage,'' or resolved in the Commission's early site permit
environmental impact statement, but must contain, in addition to the
environmental information and analyses otherwise required:
(i) Information to demonstrate that the design of the facility falls
within the site characteristics and design parameters specified in the
early site permit;
(ii) Information to resolve any significant environmental issue that
was not resolved in the early site permit proceeding;
(iii) Any new and significant information for issues related to the
impacts of construction and operation of the facility that were resolved
in the early site permit proceeding;
(iv) A description of the process used to identify new and
significant information regarding the NRC's conclusions in the early
site permit environmental impact statement. The process must use a
reasonable methodology for identifying such new and significant
information; and
(v) A demonstration that all environmental terms and conditions that
have been included in the early site permit will be satisfied by the
date of issuance of the combined license. Any terms or conditions of the
early site permit that could not be met by the time of issuance of the
combined license, must be set forth as terms or conditions of the
combined license.
(2) Application referencing standard design certification. If the
combined license references a standard design certification, then the
combined license environmental report may incorporate by reference the
environmental assessment previously prepared by the NRC for the
referenced design certification. If the design certification
environmental assessment is referenced, then the combined license
environmental report must contain information to demonstrate that the
site characteristics for the combined license site fall within the site
parameters in the design certification environmental assessment.
(3) Application referencing a manufactured reactor. If the combined
license application proposes to use a manufactured reactor, then the
combined license environmental report may incorporate by reference the
environmental assessment previously prepared by the NRC for the
underlying manufacturing license. If the manufacturing license
environmental assessment is referenced, then the combined license
environmental report must contain information to demonstrate that the
site characteristics for the combined license site fall within the site
parameters in the manufacturing license environmental assessment. The
environmental report need not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
[72 FR 49511, Aug. 28, 2007]
Sec. 51.51 Uranium fuel cycle environmental data--Table S-3.
(a) Under Sec. 51.50, every environmental report prepared for the
construction permit stage or early site permit stage or combined license
stage of a light-water-cooled nuclear power reactor, and submitted on or
after September 4, 1979, shall take Table S-3, Table of Uranium Fuel
Cycle Environmental Data, as the basis for evaluating the contribution
of the environmental effects of uranium mining and milling, the
production of uranium hexafluoride, isotopic enrichment, fuel
fabrication, reprocessing of irradiated fuel, transportation of
radioactive materials and management of low-level wastes and high-level
wastes related to uranium fuel cycle activities to the environmental
costs of licensing the nuclear power reactor. Table S-3 shall be
included in the environmental report and may be supplemented by a
discussion of the environmental significance of the data set forth in
the table as weighed in the analysis for the proposed facility.
(b) Table S-3.
[[Page 28]]
Table S-3--Table of Uranium Fuel Cycle Environmental Data \1\
[Normalized to model LWR annual fuel requirement [WASH-1248] or
reference reactor year [NUREG-0116]]
[See footnotes at end of this table]
------------------------------------------------------------------------
Maximum effect per
annual fuel requirement
Environmental considerations Total or reference reactor
year of model 1,000 MWe
LWR
------------------------------------------------------------------------
Natural Resource Use
Land (acres):
Temporarily committed \2\......... 100
Undisturbed area................ 79
Disturbed area.................. 22 Equivalent to a 110 MWe
coal-fired power
plant.
Permanently committed............. 13
Overburden moved (millions of MT). 2.8 Equivalent to 95 MWe
coal-fired power
plant.
-----------
Water (millions of gallons):
Discharged to air................. 160 =2 percent of model
1,000 MWe LWR with
cooling tower.
Discharged to water bodies........ 11,090
Discharged to ground.............. 127
-----------
Total......................... 11,377 <4 percent of model
1,000 MWe LWR with
once-through cooling.
-----------
Fossil fuel:
Electrical energy (thousands of MW- 323 <5 percent of model
hour). 1,000 MWe LWR output.
Equivalent coal (thousands of MT). 118 Equivalent to the
consumption of a 45
MWe coal-fired power
plant.
Natural gas (millions of scf)..... 135 <0.4 percent of model
1,000 MWe energy
output.
Effluents--Chemical (MT)
Gases (including entrainment): \3\
SOX............................... 4,400
NOX\4\............................ 1,190 Equivalent to emissions
from 45 MWe coal-fired
plant for a year.
Hydrocarbons...................... 14
CO................................ 29.6
Particulates...................... 1,154
Other gases:
F................................. .67 Principally from UF6
production,
enrichment, and
reprocessing.
Concentration within
range of state
standards--below level
that has effects on
human health.
HCl............................... .014
Liquids:
SO-4................................ 9.9 From enrichment, fuel
NO-3................................ 25.8 fabrication, and
Fluoride............................ 12.9 reprocessing steps.
Ca++................................ 5.4 Components that
C1-................................. 8.5 constitute a potential
Na+................................. 12.1 for adverse
NH3................................. 10.0 environmental effect
Fe.................................. .4 are present in dilute
concentrations and
receive additional
dilution by receiving
bodies of water to
levels below
permissible standards.
The constituents that
require dilution and
the flow of dilution
water are: NH3--600
cfs., NO3--20 cfs.,
Fluoride--70 cfs.
Tailings solutions (thousands of MT) 240 From mills only--no
significant effluents
to environment.
Solids.............................. 91,000 Principally from mills--
no significant
effluents to
environment.
Effluents--Radiological (curies)
Gases (including entrainment):
Rn-222............................ ......... Presently under
reconsideration by the
Commission.
Ra-226............................ .02
Th-230............................ .02
Uranium........................... .034
Tritium (thousands)............... 18.1
C-14.............................. 24
Kr-85 (thousands)................. 400
Ru-106............................ .14 Principally from fuel
reprocessing plants.
I-129............................. 1.3
I-131............................. .83
Tc-99............................. ......... Presently under
consideration by the
Commission.
Fission products and transuranics. .203
Liquids:
Uranium and daughters............. 2.1 Principally from
milling--included
tailings liquor and
returned to ground--no
effluents; therefore,
no effect on
environment.
Ra-226............................ .0034 From UF6 production.
Th-230............................ .0015
[[Page 29]]
Th-234............................ .01 From fuel fabrication
plants--concentration
10 percent of 10 CFR
20 for total
processing 26 annual
fuel requirements for
model LWR.
Fission and activation products... 5.9x10-6
Solids (buried on site):
Other than high level (shallow)... 11,300 9,100 Ci comes from low
level reactor wastes
and 1,500 Ci comes
from reactor
decontamination and
decommissioning--burie
d at land burial
facilities. 600 Ci
comes from mills--
included in tailings
returned to ground.
Approximately 60 Ci
comes from conversion
and spent fuel
storage. No
significant effluent
to the environment.
TRU and HLW (deep)................ 1.1x10\7\ Buried at Federal
Repository.
Effluents--thermal (billions of 4,063 <5 percent of model
British thermal units). 1,000 MWe LWR.
Transportation (person-rem):
Exposure of workers and general 2.5
public.
Occupational exposure (person-rem) 22.6 From reprocessing and
waste management.
------------------------------------------------------------------------
\1\ In some cases where no entry appears it is clear from the background
documents that the matter was addressed and that, in effect, the Table
should be read as if a specific zero entry had been made. However,
there are other areas that are not addressed at all in the Table.
Table S-3 does not include health effects from the effluents described
in the Table, or estimates of releases of Radon-222 from the uranium
fuel cycle or estimates of Technetium-99 released from waste
management or reprocessing activities. These issues may be the subject
of litigation in the individual licensing proceedings.
Data supporting this table are given in the ``Environmental Survey of
the Uranium Fuel Cycle,'' WASH-1248, April 1974; the ``Environmental
Survey of the Reprocessing and Waste Management Portion of the LWR
Fuel Cycle,'' NUREG-0116 (Supp.1 to WASH-1248); the ``Public Comments
and Task Force Responses Regarding the Environmental Survey of the
Reprocessing and Waste Management Portions of the LWR Fuel Cycle,''
NUREG-0216 (Supp. 2 to WASH-1248); and in the record of the final
rulemaking pertaining to Uranium Fuel Cycle Impacts from Spent Fuel
Reprocessing and Radioactive Waste Management, Docket RM-50-3. The
contributions from reprocessing, waste management and transportation
of wastes are maximized for either of the two fuel cycles (uranium
only and no recycle). The contribution from transportation excludes
transportation of cold fuel to a reactor and of irradiated fuel and
radioactive wastes from a reactor which are considered in Table S-4 of
Sec. 51.20(g). The contributions from the other steps of the fuel
cycle are given in columns A-E of Table S-3A of WASH-1248.
\2\ The contributions to temporarily committed land from reprocessing
are not prorated over 30 years, since the complete temporary impact
accrues regardless of whether the plant services one reactor for one
year or 57 reactors for 30 years.
\3\ Estimated effluents based upon combustion of equivalent coal for
power generation.
\4\ 1.2 percent from natural gas use and process.
[49 FR 9381, Mar. 12, 1984; 49 FR 10922, Mar. 23, 1984, as amended at 67
FR 77652, Dec. 19, 2002; 72 FR 49512, Aug. 28, 2007]
Sec. 51.52 Environmental effects of transportation of fuel and waste--
Table S-4.
Under Sec. 51.50, every environmental report prepared for the
construction permit stage or early site permit stage or combined license
stage of a light-water-cooled nuclear power reactor, and submitted after
February 4, 1975, shall contain a statement concerning transportation of
fuel and radioactive wastes to and from the reactor. That statement
shall indicate that the reactor and this transportation either meet all
of the conditions in paragraph (a) of this section or all of the
conditions of paragraph (b) of this section.
(a)(1) The reactor has a core thermal power level not exceeding
3,800 megawatts;
(2) The reactor fuel is in the form of sintered uranium dioxide
pellets having a uranium-235 enrichment not exceeding 4% by weight, and
the pellets are encapsulated in zircaloy rods;
(3) The average level of irradiation of the irradiated fuel from the
reactor does not exceed 33,000 megawatt-days per metric ton, and no
irradiated fuel assembly is shipped until at least 90 days after it is
discharged from the reactor;
(4) With the exception of irradiated fuel, all radioactive waste
shipped from the reactor is packaged and in a solid form;
(5) Unirradiated fuel is shipped to the reactor by truck; irradiated
fuel is shipped from the reactor by truck, rail, or barge; and
radioactive waste other than irradiated fuel is shipped from the reactor
by truck or rail; and
(6) The environmental impacts of transportation of fuel and waste to
and
[[Page 30]]
from the reactor, with respect to normal conditions of transport and
possible accidents in transport, are as set forth in Summary Table S-4
in paragraph (c) of this section; and the values in the table represent
the contribution of the transportation to the environmental costs of
licensing the reactor.
(b) For reactors not meeting the conditions of paragraph (a) of this
section, the statement shall contain a full description and detailed
analysis of the environmental effects of transportation of fuel and
wastes to and from the reactor, including values for the environmental
impact under normal conditions of transport and for the environmental
risk from accidents in transport. The statement shall indicate that the
values determined by the analysis represent the contribution of such
effects to the environmental costs of licensing the reactor.
(c)
Summary Table S-4--Environmental Impact of Transportation of Fuel and
Waste to and From One Light-Water-Cooled Nuclear Power Reactor \1\
Normal Conditions of Transport
------------------------------------------------------------------------
Environmental impact
------------------------------------------------------------------------
Heat (per irradiated fuel cask in 250,000 Btu/hr.
transit).
Weight (governed by Federal or State 73,000 lbs. per truck; 100 tons
restrictions). per cask per rail car.
Traffic density:
Truck................................ Less than 1 per day.
Rail................................. Less than 3 per month
------------------------------------------------------------------------
----------------------------------------------------------------------------------------------------------------
Estimated
number of Range of doses to exposed Cumulative dose to exposed
Exposed population persons individuals \2\ (per population (per reactor
exposed reactor year) year) \3\
----------------------------------------------------------------------------------------------------------------
Transportation workers................... 200 0.01 to 300 millirem....... 4 man-rem.
General public:
Onlookers.............................. 1,100 0.003 to 1.3 millirem...... 3 man-rem.
Along Route............................ 600,000 0.0001 to 0.06 millirem.... ............................
----------------------------------------------------------------------------------------------------------------
Accidents in Transport
------------------------------------------------------------------------
Environmental risk
------------------------------------------------------------------------
Radiological effects................... Small \4\
Common (nonradiological) causes........ 1 fatal injury in 100 reactor
years; 1 nonfatal injury in 10
reactor years; $475 property
damage per reactor year.
------------------------------------------------------------------------
\1\ Data supporting this table are given in the Commission's
``Environmental Survey of Transportation of Radioactive Materials to
and from Nuclear Power Plants,'' WASH-1238, December 1972, and Supp. 1
NUREG-75/038 April 1975. Both documents are available for inspection
and copying at the Commission's Public Document Room, 2120 L Street
NW., Washington, DC and may be obtained from National Technical
Information Service, Springfield, VA 22161. WASH-1238 is available
from NTIS at a cost of $5.45 (microfiche, $2.25) and NUREG-75/038 is
available at a cost of $3.25 (microfiche, $2.25).
\2\ The Federal Radiation Council has recommended that the radiation
doses from all sources of radiation other than natural background and
medical exposures should be limited to 5,000 millirem per year for
individuals as a result of occupational exposure and should be limited
to 500 millirem per year for individuals in the general population.
The dose to individuals due to average natural background radiation is
about 130 millirem per year.
\3\ Man-rem is an expression for the summation of whole body doses to
individuals in a group. Thus, if each member of a population group of
1,000 people were to receive a dose of 0.001 rem (1 millirem), or if 2
people were to receive a dose of 0.5 rem (500 millirem) each, the
total man-rem dose in each case would be 1 man-rem.
\4\ Athough the environmental risk of radiological effects stemming from
transportation accidents is currently incapable of being numerically
quantified, the risk remains small regardless of whether it is being
appiled to a single reactor or a multireactor site.
[49 FR 9381, Mar. 12, 1984; 49 FR 10922, Mar. 23, 1984, as amended at 53
FR 43420, Oct. 27, 1988; 72 FR 49512, Aug. 28, 2007]
Sec. 51.53 Postconstruction environmental reports.
(a) General. Any environmental report prepared under the provisions
of this section may incorporate by reference any information contained
in a prior environmental report or supplement thereto that relates to
the production or utilization facility or site, or any information
contained in a final environmental document previously prepared by the
NRC staff that relates to the production or utilization facility or
site. Documents that may be referenced include, but are not limited to,
[[Page 31]]
the final environmental impact statement; supplements to the final
environmental impact statement, including supplements prepared at the
license renewal stage; NRC staff-prepared final generic environmental
impact statements; and environmental assessments and records of
decisions prepared in connection with the construction permit, operating
license, early site permit, combined license and any license amendment
for that facility.
(b) Operating license stage. Each applicant for a license to operate
a production or utilization facility covered by Sec. 51.20 shall submit
with its application a separate document entitled ``Supplement to
Applicant's Environmental Report--Operating License Stage,'' which will
update ``Applicant's Environmental Report--Construction Permit Stage.''
Unless otherwise required by the Commission, the applicant for an
operating license for a nuclear power reactor shall submit this report
only in connection with the first licensing action authorizing full-
power operation. In this report, the applicant shall discuss the same
matters described in Sec. Sec. 51.45, 51.51, and 51.52, but only to the
extent that they differ from those discussed or reflect new information
in addition to that discussed in the final environmental impact
statement prepared by the Commission in connection with the construction
permit. No discussion of need for power, or of alternative energy
sources, or of alternative sites for the facility, or of any aspect of
the storage of spent fuel for the facility within the scope of the
generic determination in Sec. 51.23(a) and in accordance with Sec.
51.23(b) is required in this report.
(c) Operating license renewal stage. (1) Each applicant for renewal
of a license to operate a nuclear power plant under part 54 of this
chapter shall submit with its application a separate document entitled
``Applicant's Environmental Report--Operating License Renewal Stage.''
(2) The report must contain a description of the proposed action,
including the applicant's plans to modify the facility or its
administrative control procedures as described in accordance with Sec.
54.21 of this chapter. This report must describe in detail the
modifications directly affecting the environment or affecting plant
effluents that affect the environment. In addition, the applicant shall
discuss in this report the environmental impacts of alternatives and any
other matters described in Sec. 51.45. The report is not required to
include discussion of need for power or the economic costs and economic
benefits of the proposed action or of alternatives to the proposed
action except insofar as such costs and benefits are either essential
for a determination regarding the inclusion of an alternative in the
range of alternatives considered or relevant to mitigation. The
environmental report need not discuss other issues not related to the
environmental effects of the proposed action and the alternatives. In
addition, the environmental report need not discuss any aspect of the
storage of spent fuel for the facility within the scope of the generic
determination in Sec. 51.23(a) and in accordance with Sec. 51.23(b).
(3) For those applicants seeking an initial renewed license and
holding an operating license, construction permit, or combined license
as of June 30, 1995, the environmental report shall include the
information required in paragraph (c)(2) of this section subject to the
following conditions and considerations:
(i) The environmental report for the operating license renewal stage
is not required to contain analyses of the environmental impacts of the
license renewal issues identified as Category 1 issues in appendix B to
subpart A of this part.
(ii) The environmental report must contain analyses of the
environmental impacts of the proposed action, including the impacts of
refurbishment activities, if any, associated with license renewal and
the impacts of operation during the renewal term, for those issues
identified as Category 2 issues in appendix B to subpart A of this part.
The required analyses are as follows:
(A) If the applicant's plant utilizes cooling towers or cooling
ponds and withdraws make-up water from a river whose annual flow rate is
less than 3.15x10\12\ ft\3\/year (9x10\10\ m\3\/year), an assessment of
the impact of the proposed action on the flow of the river and related
impacts on instream and riparian
[[Page 32]]
ecological communities must be provided. The applicant shall also
provide an assessment of the impacts of the withdrawal of water from the
river on alluvial aquifers during low flow.
(B) If the applicant's plant utilizes once-through cooling or
cooling pond heat dissipation systems, the applicant shall provide a
copy of current Clean Water Act 316(b) determinations and, if necessary,
a 316(a) variance in accordance with 40 CFR part 125, or equivalent
State permits and supporting documentation. If the applicant can not
provide these documents, it shall assess the impact of the proposed
action on fish and shellfish resources resulting from heat shock and
impingement and entrainment.
(C) If the applicant's plant uses Ranney wells or pumps more than
100 gallons (total onsite) of ground water per minute, an assessment of
the impact of the proposed action on ground-water use must be provided.
(D) If the applicant's plant is located at an inland site and
utilizes cooling ponds, an assessment of the impact of the proposed
action on groundwater quality must be provided.
(E) All license renewal applicants shall assess the impact of
refurbishment and other license-renewal-related construction activities
on important plant and animal habitats. Additionally, the applicant
shall assess the impact of the proposed action on threatened or
endangered species in accordance with the Endangered Species Act.
(F) If the applicant's plant is located in or near a nonattainment
or maintenance area, an assessment of vehicle exhaust emissions
anticipated at the time of peak refurbishment workforce must be provided
in accordance with the Clean Air Act as amended.
(G) If the applicant's plant uses a cooling pond, lake, or canal or
discharges into a river having an annual average flow rate of less than
3.15x10\12\ ft\3\/year (9x10\10\ m\3\/year), an assessment of the impact
of the proposed action on public health from thermophilic organisms in
the affected water must be provided.
(H) If the applicant's transmission lines that were constructed for
the specific purpose of connecting the plant to the transmission system
do not meet the recommendations of the National Electric Safety Code for
preventing electric shock from induced currents, an assessment of the
impact of the proposed action on the potential shock hazard from the
transmission lines must be provided.
(I) An assessment of the impact of the proposed action on housing
availability, land-use, and public schools (impacts from refurbishment
activities only) within the vicinity of the plant must be provided.
Additionally, the applicant shall provide an assessment of the impact of
population increases attributable to the proposed project on the public
water supply.
(J) All applicants shall assess the impact of highway traffic
generated by the proposed project on the level of service of local
highways during periods of license renewal refurbishment activities and
during the term of the renewed license.
(K) All applicants shall assess whether any historic or
archaeological properties will be affected by the proposed project.
(L) If the staff has not previously considered severe accident
mitigation alternatives for the applicant's plant in an environmental
impact statement or related supplement or in an environmental
assessment, a consideration of alternatives to mitigate severe accidents
must be provided.
(M) [Reserved]
(iii) The report must contain a consideration of alternatives for
reducing adverse impacts, as required by Sec. 51.45(c), for all
Category 2 license renewal issues in appendix B to subpart A of this
part. No such consideration is required for Category 1 issues in
appendix B to subpart A of this part.
(iv) The environmental report must contain any new and significant
information regarding the environmental impacts of license renewal of
which the applicant is aware.
(d) Postoperating license stage. Each applicant for a license
amendment authorizing decommissioning activities for a production or
utilization facility either for unrestricted use or based on continuing
use restrictions applicable to the site; and each applicant for a
license amendment approving a license termination plan or
decommissioning
[[Page 33]]
plan under Sec. 50.82 of this chapter either for unrestricted use or
based on continuing use restrictions applicable to the site; and each
applicant for a license or license amendment to store spent fuel at a
nuclear power reactor after expiration of the operating license for the
nuclear power reactor shall submit with its application a separate
document, entitled ``Supplement to Applicant's Environmental Report--
Post Operating License Stage,'' which will update ``Applicant's
Environmental Report--Operating License Stage,'' as appropriate, to
reflect any new information or significant environmental change
associated with the applicant's proposed decommissioning activities or
with the applicant's proposed activities with respect to the planned
storage of spent fuel. Unless otherwise required by the Commission, in
accordance with the generic determination in Sec. 51.23(a) and the
provisions in Sec. 51.23(b), the applicant shall only address the
environmental impact of spent fuel storage for the term of the license
applied for. The ``Supplement to Applicant's Environmental Report--Post
Operating License Stage'' may incorporate by reference any information
contained in ``Applicants Environmental Report--Construction Permit
Stage.
[61 FR 66543, Dec. 18, 1996, as amended at 64 FR 48506, Sept. 3, 1999;
68 FR 58810, Oct. 10, 2003; 72 FR 49513, Aug. 28, 2007]
Sec. 51.54 Environmental report--manufacturing license.
(a) Each applicant for a manufacturing license under subpart F of
part 52 of this chapter shall submit with its application a separate
document entitled, ``Applicant's Environmental Report--Manufacturing
License.'' The environmental report must address the costs and benefits
of severe accident mitigation design alternatives, and the bases for not
incorporating severe accident mitigation design alternatives into the
design of the reactor to be manufactured. The environmental report need
not address the environmental impacts associated with manufacturing the
reactor under the manufacturing license, the benefits and impacts of
utilizing the reactor in a nuclear power plant, or an evaluation of
alternative energy sources.
(b) Each applicant for an amendment to a manufacturing license shall
submit with its application a separate document entitled, ``Applicant's
Supplemental Environmental Report--Amendment to Manufacturing License.''
The environmental report must address whether the design change which is
the subject of the proposed amendment either renders a severe accident
mitigation design alternative previously rejected in an environmental
assessment to become cost beneficial, or results in the identification
of new severe accident mitigation design alternatives that may be
reasonably incorporated into the design of the manufactured reactor. The
environmental report need not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
[72 FR 49513, Aug. 28, 2007]
Sec. 51.55 Environmental report--standard design certification.
(a) Each applicant for a standard design certification under subpart
B of part 52 of this chapter shall submit with its application a
separate document entitled, ``Applicant's Environmental Report--Standard
Design Certification.'' The environmental report must address the costs
and benefits of severe accident mitigation design alternatives, and the
bases for not incorporating severe accident mitigation design
alternatives in the design to be certified.
(b) Each applicant for an amendment to a design certification shall
submit with its application a separate document entitled, ``Applicant's
Supplemental Environmental Report--Amendment to Standard Design
Certification.'' The environmental report must address whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation
[[Page 34]]
design alternatives that may be reasonably incorporated into the design
certification.
[72 FR 49513, Aug. 28, 2007]
Sec. 51.58 Environmental report--number of copies; distribution.
(a) Each applicant for a license or permit to site, construct,
manufacture, or operate a production or utilization facility covered by
Sec. Sec. 51.20(b)(1), (b)(2), (b)(3), or (b)(4), each applicant for
renewal of an operating or combined license for a nuclear power plant,
each applicant for a license amendment authorizing the decommissioning
of a production or utilization facility covered by Sec. 51.20, and each
applicant for a license or license amendment to store spent fuel at a
nuclear power plant after expiration of the operating license or
combined license for the nuclear power plant shall submit a copy to the
Director of the Office of Nuclear Reactor Regulation, the Director of
the Office of New Reactors, the Director of the Office of Nuclear
Material Safety and Safeguards, as appropriate, of an environmental
report or any supplement to an environmental report. These reports must
be sent either by mail addressed: ATTN: Document Control Desk; U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand
delivery to the NRC's offices at 11555 Rockville Pike, Rockville,
Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; or,
where practicable, by electronic submission, for example, via Electronic
Information Exchange, or CD-ROM. Electronic submissions must be made in
a manner that enables the NRC to receive, read, authenticate,
distribute, and archive the submission, and process and retrieve it a
single page at a time. Detailed guidance on making electronic
submissions can be obtained by visiting the NRC's Web site at http://
www.nrc.gov/site-help/e-submittals.html; by e-mail to
[email protected]; or by writing the Office of Information Services,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The
guidance discusses, among other topics, the formats the NRC can accept,
the use of electronic signatures, and the treatment of nonpublic
information. If the communication is on paper, the signed original must
be sent. If a submission due date falls on a Saturday, Sunday, or
Federal holiday, the next Federal working day becomes the official due
date. The applicant shall maintain the capability to generate additional
copies of the environmental report or any supplement to the
environmental report for subsequent distribution to parties and Boards
in the NRC proceedings; Federal, State, and local officials; and any
affected Indian tribes, in accordance with written instructions issued
by the Director of the Office of New Reactors, the Director of the
Office of Nuclear Reactor Regulation, or the Director of the Office of
Nuclear Material Safety and Safeguards, as appropriate.
(b) Each applicant for a license to manufacture a nuclear power
reactor, or for an amendment to a license to manufacture, seeking
approval of the final design of the nuclear power reactor under subpart
F of part 52 of this chapter, shall submit to the Commission an
environmental report or any supplement to an environmental report in the
manner specified in Sec. 50.3 of this chapter. The applicant shall
maintain the capability to generate additional copies of the
environmental report or any supplement to the environmental report for
subsequent distribution to parties and Boards in the NRC proceeding;
Federal, State, and local officials; and any affected Indian tribes, in
accordance with written instructions issued by the Director of the
Office of New Reactors or the Director of the Office of Nuclear Reactor
Regulation.
[72 FR 49513, Aug. 28, 2007, as amended at 74 FR 62682, Dec. 1, 2009]
environmental reports--materials licenses
Sec. 51.60 Environmental report--materials licenses.
(a) Each applicant for a license or other form of permission, or an
amendment to or renewal of a license or other form of permission issued
pursuant to parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70 and/or 72 of
this chapter, and covered by paragraphs (b)(1) through (b)(5) of this
section, shall submit with its application to: ATTN: Document Control
[[Page 35]]
Desk, Director, Nuclear Material Safety and Safeguards, a separate
document, entitled ``Applicant's Environmental Report'' or ``Supplement
to Applicant's Environmental Report,'' as appropriate. The ``Applicant's
Environmental Report'' shall contain the information specified in Sec.
51.45. If the application is for an amendment to or a renewal of a
license or other form of permission for which the applicant has
previously submitted an environmental report, the supplement to
applicant's environmental report may be limited to incorporating by
reference, updating or supplementing the information previously
submitted to reflect any significant environmental change, including any
significant environmental change resulting from operational experience
or a change in operations or proposed decommissioning activities. If the
applicant is the U.S. Department of Energy, the environmental report may
be in the form of either an environmental impact statement or an
environmental assessment, as appropriate.
(b) As required by paragraph (a) of this section, each applicant
shall prepare an environmental report for the following types of
actions:
(1) Issuance or renewal of a license or other form of permission
for:
(i) Possession and use of special nuclear material for processing
and fuel fabrication, scrap recovery, or conversion of uranium
hexafluoride pursuant to part 70 of this chapter.
(ii) Possession and use of source material for uranium milling or
production of uranium hexafluoride pursuant to part 40 of this chapter.
(iii) Storage of spent fuel in an independent spent fuel storage
installation (ISFSI) or the storage of spent fuel or high-level radio-
active waste in a monitored retrievable storage installation (MRS)
pursuant to part 72 of this chapter.
(iv) Receipt and disposal of radioactive waste from other persons
pursuant to part 61 of this chapter.
(v) Processing of source material for extraction of rare earth and
other metals.
(vi) Use of radioactive tracers in field flood studies involving
secondary and tertiary oil and gas recovery.
(vii) Construction and operation of a uranium enrichment facility.
(2) Issuance of an amendment that would authorize or result in (i) a
significant expansion of a site, (ii) a significant change in the types
of effluents, (iii) a significant increase in the amounts of effluents,
(iv) a significant increase in individual or cumulative occupational
radiation exposure, (v) a significant increase in the potential for or
consequences from radiological accidents, or (vi) a significant increase
in spent fuel storage capacity, in a license or other form of permission
to conduct an activity listed in paragraph (b)(1) of this section.
(3) Amendment of a license to authorize the decommissioning of an
independent spent fuel storage installation (ISFSI) or a monitored
retrievable storage installation (MRS) pursuant to part 72 of this
chapter.
(4) Issuance of a license amendment pursuant to part 61 of this
chapter authorizing (i) closure of a land disposal site, (ii) transfer
of the license to the disposal site owner for the purpose of
institutional control, or (iii) termination of the license at the end of
the institutional control period.
(5) Any other licensing action for which the Commission determines
an Environmental Report is necessary.
[49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31681, Aug. 19, 1988; 57
FR 18392, Apr. 30, 1992; 58 FR 7737, Feb. 9, 1993; 62 FR 26732, May 14,
1997; 68 FR 58811, Oct. 10, 2003]
Sec. 51.61 Environmental report--independent spent fuel storage
installation (ISFSI) or monitored retrievable storage installation
(MRS) license.
Each applicant for issuance of a license for storage of spent fuel
in an independent spent fuel storage installation (ISFSI) or for the
storage of spent fuel and high-level radioactive waste in a monitored
retrievable storage installation (MRS) pursuant to part 72 of this
chapter shall submit with its application to: ATTN: Document Control
Desk, Director, Office of Nuclear Material Safety and Safeguards, a
separate document entitled ``Applicant's Environmental Report--ISFSI
License'' or ``Applicant's Environmental Report--MRS License,'' as
[[Page 36]]
appropriate. If the applicant is the U.S. Department of Energy, the
environmental report may be in the form of either an environmental
impact statement or an environmental assessment, as appropriate. The
environmental report shall contain the information specified in Sec.
51.45 and shall address the siting evaluation factors contained in
subpart E of part 72 of this chapter. Unless otherwise required by the
Commission, in accordance with the generic determination in Sec.
51.23(a) and the provisions in Sec. 51.23(b), no discussion of the
environmental impact of the storage of spent fuel at an ISFSI beyond the
term of the license or amendment applied for is required in an
environmental report submitted by an applicant for an initial license
for storage of spent fuel in an ISFSI, or any amendment thereto.
[53 FR 31681, Aug. 19, 1988, as amended at 68 FR 58811, Oct. 10, 2003]
Sec. 51.62 Environmental report--land disposal of radioactive waste
licensed under 10 CFR part 61.
(a) Each applicant for issuance of a license for land disposal of
radioactive waste pursuant to part 61 of this chapter shall submit with
its application to: ATTN: Document Control Desk, Director of Nuclear
Material Safety and Safeguards, a separate document, entitled
``Applicant's Environmental Report--License for Land Disposal of
Radioactive Waste.'' The environmental report and any supplement to the
environmental report may incorporate by reference information contained
in the application or in any previous application, statement or report
filed with the Commission provided that such references are clear and
specific and that copies of the information so incorporated are
available at the NRC Web site, http://www.nrc.gov, and/or at the NRC
Public Document Room.
(b) The environmental report shall contain the information specified
in Sec. 51.45, shall address the applicant's environmental monitoring
program required by Sec. Sec. 61.12(l), 61.53 and 61.59(b) of this
chapter, and shall be as complete as possible in the light of
information that is available at the time the environmental report is
submitted.
(c) The applicant shall supplement the environmental report in a
timely manner as necessary to permit the Commission to review, prior to
issuance, amendment or renewal of a license, new information regarding
the environmental impact of previously proposed activities, information
regarding the environmental impact of any changes in previously proposed
activities, or any significant new information regarding the
environmental impact of closure activities and long-term performance of
the disposal site.
[49 FR 9381, Mar. 12, 1984, as amended at 53 FR 43420, Oct. 27, 1988; 64
FR 48952, Sept. 9, 1999; 68 FR 58811, Oct. 10, 2003]
Sec. 51.66 Environmental report--number of copies; distribution.
Each applicant for a license or other form of permission, or an
amendment to or renewal of a license or other form of permission issued
under parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70, and/or 72 of this
chapter, and covered by Sec. Sec. 51.60(b)(1) through (6); or by
Sec. Sec. 51.61 or 51.62 shall submit to the Director of Nuclear
Material Safety and Safeguards an environmental report or any supplement
to an environmental report in the manner specified in Sec. 51.58(a).
The applicant shall maintain the capability to generate additional
copies of the environmental report or any supplement to the
environmental report for subsequent distribution to Federal, State, and
local officials, and any affected Indian tribes in accordance with
written instructions issued by the Director of Nuclear Material Safety
and Safeguards.
[72 FR 49514, Aug. 28, 2007]
Sec. 51.67 Environmental information concerning geologic repositories.
(a) In lieu of an environmental report, the Department of Energy, as
an applicant for a license or license amendment pursuant to part 60 or
63 of this chapter, shall submit to the Commission any final
environmental impact statement which the Department prepares in
connection with any geologic repository developed under Subtitle A of
Title I, or under Title IV, of the Nuclear Waste Policy Act of 1982, as
amended. (See Sec. 60.22 or Sec. 63.22 of this chapter as to the
required time and
[[Page 37]]
manner of submission.) The statement shall include, among the
alternatives under consideration, denial of a license or construction
authorization by the Commission.
(b) Under applicable provisions of law, the Department of Energy may
be required to supplement its final environmental impact statement if it
makes a substantial change in its proposed action that is relevant to
environmental concerns or determines that there are significant new
circumstances or information relevant to environmental concerns and
bearing on the proposed action or its impacts. The Department shall
submit any supplement to its final environmental impact statement to the
Commission. (See Sec. 60.22 or Sec. 63.22 of this chapter as to the
required time and manner of submission.)
(c) Whenever the Department of Energy submits a final environmental
impact statement, or a final supplement to an environmental impact
statement, to the Commission pursuant to this section, it shall also
inform the Commission of the status of any civil action for judicial
review initiated pursuant to section 119 of the Nuclear Waste Policy Act
of 1982. This status report, which the Department shall update from time
to time to reflect changes in status, shall:
(1) State whether the environmental impact statement has been found
by the courts of the United States to be adequate or inadequate; and
(2) Identify any issues relating to the adequacy of the
environmental impact statement that may remain subject to judicial
review.
[54 FR 27870, July 3, 1989, as amended at 66 FR 55791, Nov. 2, 2001]
environmental reports--rulemaking
Sec. 51.68 Environmental report--rulemaking.
Petitioners for rulemaking requesting amendments of parts 30, 31,
32, 33, 34, 35, 36, 39, 40 or part 70 of this chapter concerning the
exemption from licensing and regulatory requirements of or authorizing
general licenses for any equipment, device, commodity or other product
containing byproduct material, source material or special nuclear
material shall submit with the petition a separate document entitled
``Petitioner's Environmental Report,'' which shall contain the
information specified in Sec. 51.45.
[68 FR 58811, Oct. 10, 2003]
Environmental Impact Statements
draft environmental impact statements--general requirements
Sec. 51.70 Draft environmental impact statement--general.
(a) The NRC staff will prepare a draft environmental impact
statement as soon as practicable after publication of the notice of
intent to prepare an environmental impact statement and completion of
the scoping process. To the fullest extent practicable, environmental
impact statements will be prepared concurrently or integrated with
environmental impact analyses and related surveys and studies required
by other Federal law.
(b) The draft environmental impact statement will be concise, clear
and analytic, will be written in plain language with appropriate
graphics, will state how alternatives considered in it and decisions
based on it will or will not achieve the requirements of sections 101
and 102(1) of NEPA and of any other relevant and applicable
environmental laws and policies, will identify any methodologies used
and sources relied upon, and will be supported by evidence that the
necessary environmental analyses have been made. The format provided in
section 1(a) of appendix A of this subpart should be used. The NRC staff
will independently evaluate and be responsible for the reliability of
all information used in the draft environmental impact statement.
(c) The Commission will cooperate with State and local agencies to
the fullest extent possible to reduce duplication between NEPA and State
and local requirements, in accordance with 40 CFR 1506.2 (b) and (c).
Sec. 51.71 Draft environmental impact statement--contents.
(a) Scope. The draft environmental impact statement will be prepared
in accordance with the scope decided upon in the scoping process
required by
[[Page 38]]
Sec. Sec. 51.26 and 51.29. As appropriate and to the extent required by
the scope, the draft statement will address the topics in paragraphs
(b), (c), (d) and (e) of this section and the matters specified in
Sec. Sec. 51.45, 51.50, 51.51, 51.52, 51.53, 51.54, 51.61 and 51.62.
(b) Analysis of major points of view. To the extent sufficient
information is available, the draft environmental impact statement will
include consideration of major points of view concerning the
environmental impacts of the proposed action and the alternatives, and
contain an analysis of significant problems and objections raised by
other Federal, State, and local agencies, by any affected Indian tribes,
and by other interested persons.
(c) Status of compliance. The draft environmental impact statement
will list all Federal permits, licenses, approvals, and other
entitlements which must be obtained in implementing the proposed action
and will describe the status of compliance with those requirements. If
it is uncertain whether a Federal permit, license, approval, or other
entitlement is necessary, the draft environmental impact statement will
so indicate.
(d) Analysis. Unless excepted in this paragraph or Sec. 51.75, the
draft environmental impact statement will include a preliminary analysis
that considers and weighs the environmental effects of the proposed
action; the environmental impacts of alternatives to the proposed
action; and alternatives available for reducing or avoiding adverse
environmental effects and consideration of the economic, technical, and
other benefits and costs of the proposed action and alternatives and
indicate what other interests and considerations of Federal policy,
including factors not related to environmental quality if applicable,
are relevant to the consideration of environmental effects of the
proposed action identified under paragraph (a) of this section. The
draft supplemental environmental impact statement prepared at the
license renewal stage under Sec. 51.95(c) need not discuss the economic
or technical benefits and costs of either the proposed action or
alternatives except if benefits and costs are either essential for a
determination regarding the inclusion of an alternative in the range of
alternatives considered or relevant to mitigation. In addition, the
supplemental environmental impact statement prepared at the license
renewal stage need not discuss other issues not related to the
environmental effects of the proposed action and associated
alternatives. The draft supplemental environmental impact statement for
license renewal prepared under Sec. 51.95(c) will rely on conclusions
as amplified by the supporting information in the GEIS for issues
designated as Category 1 in appendix B to subpart A of this part. The
draft supplemental environmental impact statement must contain an
analysis of those issues identified as Category 2 in appendix B to
subpart A of this part that are open for the proposed action. The
analysis for all draft environmental impact statements will, to the
fullest extent practicable, quantify the various factors considered. To
the extent that there are important qualitative considerations or
factors that cannot be quantified, these considerations or factors will
be discussed in qualitative terms. Consideration will be given to
compliance with environmental quality standards and requirements that
have been imposed by Federal, State, regional, and local agencies having
responsibility for environmental protection, including applicable zoning
and land-use regulations and water pollution limitations or requirements
issued or imposed under the Federal Water Pollution Control Act. The
environmental impact of the proposed action will be considered in the
analysis with respect to matters covered by environmental quality
standards and requirements irrespective of whether a certification or
license from the appropriate authority has been obtained. \3\ While
satisfaction
[[Page 39]]
of Commission standards and criteria pertaining to radiological effects
will be necessary to meet the licensing requirements of the Atomic
Energy Act, the analysis will, for the purposes of NEPA, consider the
radiological effects of the proposed action and alternatives.
---------------------------------------------------------------------------
\3\ Compliance with the environmental quality standards and
requirements of the Federal Water Pollution Control Act (imposed by EPA
or designated permitting states) is not a substitute for, and does not
negate the requirement for NRC to weigh all environmental effects of the
proposed action, including the degradation, if any, of water quality,
and to consider alternatives to the proposed action that are available
for reducing adverse effects. Where an environmental assessment of
aquatic impact from plant discharges is available from the permitting
authority, the NRC will consider the assessment in its determination of
the magnitude of environmental impacts for striking an overall cost-
benefit balance at the construction permit and operating license and
early site permit and combined license stages, and in its determination
of whether the adverse environmental impacts of license renewal are so
great that preserving the option of license renewal for energy planning
decision-makers would be unreasonable at the license renewal stage. When
no such assessment of aquatic impacts is available from the permitting
authority, NRC will establish on its own, or in conjunction with the
permitting authority and other agencies having relevant expertise, the
magnitude of potential impacts for striking an overall cost-benefit
balance for the facility at the construction permit and operating
license and early site permit and combined license stages, and in its
determination of whether the adverse environmental impacts of license
renewal are so great that preserving the option of license renewal for
energy planning decision-makers would be unreasonable at the license
renewal stage.
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(e) Effect of limited work authorization. If a limited work
authorization was issued either in connection with or subsequent to an
early site permit, or in connection with a construction permit or
combined license application, then the environmental impact statement
for the construction permit or combined license application will not
address or consider the sunk costs associated with the limited work
authorization.
(f) Preliminary recommendation. The draft environmental impact
statement normally will include a preliminary recommendation by the NRC
staff respecting the proposed action. This preliminary recommendation
will be based on the information and analysis described in paragraphs
(a) through (d) of this section and Sec. Sec. 51.75, 51.76, 51.80,
51.85, and 51.95, as appropriate, and will be reached after considering
the environmental effects of the proposed action and reasonable
alternatives, \4\ and, except for supplemental environmental impact
statements for the operating license renewal stage prepared pursuant to
Sec. 51.95(c), after weighing the costs and benefits of the proposed
action. In lieu of a recommendation, the NRC staff may indicate in the
draft statement that two or more alternatives remain under
consideration.
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\4\ The consideration of reasonable alternatives to a proposed
action involving nuclear power reactors (e.g., alternative energy
sources) is intended to assist the NRC in meeting its NEPA obligations
and does not preclude any State authority from making separate
determinations with respect to these alternatives and in no way
preempts, displaces, or affects the authority of States or other Federal
agencies to address these issues.
[49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28488, June 5, 1996; 61
FR 66544, Dec. 18, 1996; 72 FR 49514, Aug. 28, 2007; 72 FR 57445, Oct.
9, 2007]
Sec. 51.72 Supplement to draft environmental impact statement.
(a) The NRC staff will prepare a supplement to a draft environmental
impact statement for which a notice of availability has been published
in the Federal Register as provided in Sec. 51.117, if:
(1) There are substantial changes in the proposed action that are
relevant to environmental concerns; or
(2) There are significant new circumstances or information relevant
to environmental concerns and bearing on the proposed action or its
impacts.
(b) The NRC staff may prepare a supplement to a draft environmental
impact statement when, in its opinion, preparation of a supplement will
further the purposes of NEPA.
(c) The supplement to a draft environmental impact statement will be
prepared and noticed in the same manner as the draft environmental
impact statement except that a scoping process need not be used.
Sec. 51.73 Request for comments on draft environmental impact statement.
Each draft environmental impact statement and each supplement to a
[[Page 40]]
draft environmental impact statement distributed in accordance with
Sec. 51.74, and each news release provided pursuant to Sec. 51.74(d)
will be accompanied by or include a request for comments on the proposed
action and on the draft environmental impact statement or any supplement
to the draft environmental impact statement and will state where
comments should be submitted and the date on which the comment period
closes. A minimum comment period of 45 days will be provided. The
comment period will be calculated from the date on which the
Environmental Protection Agency notice stating that the draft statement
or the supplement to the draft statement has been filed with EPA is
published in the Federal Register. If no comments are provided within
the time specified, it will be presumed, unless the agency or person
requests an extension of time, that the agency or person has no comment
to make. To the extent practicable, NRC staff will grant reasonable
requests for extensions of time of up to fifteen (15) days.
Sec. 51.74 Distribution of draft environmental impact statement and
supplement to draft environmental impact statement; news releases.
(a) A copy of the draft environmental impact statement will be
distributed to:
(1) The Environmental Protection Agency.
(2) Any other Federal agency which has special expertise or
jurisdiction by law with respect to any environmental impact involved or
which is authorized to develop and enforce relevant environmental
standards.
(3) The applicant or petitioner for rulemaking and any other party
to the proceeding.
(4) Appropriate State and local agencies authorized to develop and
enforce relevant environmental standards.
(5) Appropriate State, regional and metropolitan clearinghouses.
(6) Appropriate Indian tribes when the proposed action may have an
environmental impact on a reservation.
(7) Upon written request, any organization or group included in the
master list of interested organizations and groups maintained under
Sec. 51.122.
(8) Upon written request, any other person to the extent available.
(b) Additional copies will be made available in accordance with
Sec. 51.123.
(c) A supplement to a draft environmental impact statement will be
distributed in the same manner as the draft environmental impact
statement to which it relates.
(d) News releases stating the availability for comment and place for
obtaining or inspecting a draft environmental statement or supplement
will be provided to local newspapers and other appropriate media.
(e) A notice of availability will be published in the Federal
Register in accordance with Sec. 51.117.
draft environmental impact statements--production and utilization
facilities
Sec. 51.75 Draft environmental impact statement--construction permit,
early site permit, or combined licesne.
(a) Construction permit stage. A draft environmental impact
statement relating to issuance of a construction permit for a production
or utilization facility will be prepared in accordance with the
procedures and measures described in Sec. Sec. 51.70, 51.71, 51.72, and
51.73. The contribution of the environmental effects of the uranium fuel
cycle activities specified in Sec. 51.51 shall be evaluated on the
basis of impact values set forth in Table S-3, Table of Uranium Fuel
Cycle Environmental Data, which shall be set out in the draft
environmental impact statement. With the exception of radon-222 and
technetium-99 releases, no further discussion of fuel cycle release
values and other numerical data that appear explicitly in the table
shall be required. \5\ The impact statement shall take account of dose
commitments and health effects from fuel cycle effluents set forth in
Table S-3 and shall in addition take account of economic, socioeconomic,
and possible cumulative impacts and other fuel cycle impacts as may
reasonably appear significant.
(b) Early site permit stage. A draft environmental impact statement
relating to issuance of an early site permit for a production or
utilization facility will
[[Page 41]]
be prepared in accordance with the procedures and measures described in
Sec. Sec. 51.70, 51.71, 51.72, 51.73, and this section. The
contribution of the environmental effects of the uranium fuel cycle
activities specified in Sec. 51.51 shall be evaluated on the basis of
impact values set forth in Table S-3, Table of Uranium Fuel Cycle
Environmental Data, which shall be set out in the draft environmental
impact statement. With the exception of radon-222 and technetium-99
releases, no further discussion of fuel cycle release values and other
numerical data that appear explicitly in the table shall be required.
\5\ The impact statement shall take account of dose commitments and
health effects from fuel cycle effluents set forth in Table S-3 and
shall in addition take account of economic, socioeconomic, and possible
cumulative impacts and other fuel cycle impacts as may reasonably appear
significant. The draft environmental impact statement must include an
evaluation of alternative sites to determine whether there is any
obviously superior alternative to the site proposed. The draft
environmental impact statement must also include an evaluation of the
environmental effects of construction and operation of a reactor, or
reactors, which have design characteristics that fall within the site
characteristics and design parameters for the early site permit
application, but only to the extent addressed in the early site permit
environmental report or otherwise necessary to determine whether there
is any obviously superior alternative to the site proposed. The draft
environmental impact statement must not include an assessment of the
economic, technical, or other benefits (for example, need for power) and
costs of the proposed action or an evaluation of alternative energy
sources, unless these matters are addressed in the early site permit
environmental report.
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\5\ Values for releases of Rn-222 and Tc-99 are not given in the
table. The amount and significance of Rn-222 releases from the fuel
cycle and Tc-99 releases from waste management or reprocessing
activities shall be considered in the draft environmental impact
statement and may be the subject of litigation in individual licensing
proceedings.
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(c) Combined license stage. A draft environmental impact statement
relating to issuance of a combined license that does not reference an
early site permit will be prepared in accordance with the procedures and
measures described in Sec. Sec. 51.70, 51.71, 51.72, and 51.73. The
contribution of the environmental effects of the uranium fuel cycle
activities specified in Sec. 51.51 shall be evaluated on the basis of
impact values set forth in Table S-3, Table of Uranium Fuel Cycle
Environmental Data, which shall be set out in the draft environmental
impact statement. With the exception of radon-222 and technetium-99
releases, no further discussion of fuel cycle release values and other
numerical data that appear explicitly in the table shall be required.
\5\ The impact statement shall take account of dose commitments and
health effects from fuel cycle effluents set forth in Table S-3 and
shall in addition take account of economic, socioeconomic, and possible
cumulative impacts and other fuel cycle impacts as may reasonably appear
significant. The impact statement will include a discussion of the
storage of spent fuel for the nuclear power plant within the scope of
the generic determination in Sec. 51.23(a) and in accordance with Sec.
51.23(b).
(1) Combined license application referencing an early site permit.
If the combined license application references an early site permit,
then the NRC staff shall prepare a draft supplement to the early site
permit environmental impact statement. The supplement must be prepared
in accordance with Sec. 51.92(e).
(2) Combined license application referencing a standard design
certification. If the combined license application references a standard
design certification and the site characteristics of the combined
license's site fall within the site parameters specified in the design
certification environmental assessment, then the draft combined license
environmental impact statement shall incorporate by reference the design
certification environmental assessment, and summarize the findings and
conclusions of the environmental assessment with respect to severe
accident mitigation design alternatives.
[[Page 42]]
(3) Combined license application referencing a manufactured reactor.
If the combined license application proposes to use a manufactured
reactor and the site characteristics of the combined license's site fall
within the site parameters specified in the manufacturing license
environmental assessment, then the draft combined license environmental
impact statement shall incorporate by reference the manufacturing
license environmental assessment, and summarize the findings and
conclusions of the environmental assessment with respect to severe
accident mitigation design alternatives. The combined license
environmental impact statement report will not address the environmental
impacts associated with manufacturing the reactor under the
manufacturing license.
[72 FR 49514, Aug. 28, 2007]
Sec. 51.76 Draft environmental impact statement--limited work
authorization.
The NRC will prepare a draft environmental impact statement relating
to issuance of a limited work authorization in accordance with the
procedures and measures described in Sec. Sec. 51.70, 51.71, and 51.73,
as further supplemented or modified in the following paragraphs.
(a) Limited work authorization submitted as part of complete
construction permit or combined license application. If the application
for a limited work authorization is submitted as part of a complete
construction permit or combined license application, then the NRC may
prepare a partial draft environmental impact statement. The analysis
called for by Sec. 51.71(d) must be limited to the activities proposed
to be conducted under the limited work authorization. Alternatively, the
NRC may prepare a complete draft environmental impact statement prepared
in accordance with Sec. 51.75(a) or (c), as applicable.
(b) Phased application for limited work authorization under Sec.
2.101(a)(9) of this chapter. If the application for a limited work
authorization is submitted in accordance with Sec. 2.101(a)(9) of this
chapter, then the draft environmental impact statement for part one of
the application may be limited to consideration of the activities
proposed to be conducted under the limited work authorization, and the
proposed redress plan. However, if the environmental report contains the
full set of information required to be submitted under Sec. 51.50(a) or
(c), then a draft environmental impact statement must be prepared in
accordance with Sec. 51.75(a) or (c), as applicable. Siting issues,
including whether there is an obviously superior alternative site, or
issues related to operation of the proposed nuclear power plant at the
site, including need for power, may not be considered. After part two of
the application is docketed, the NRC will prepare a draft supplement to
the final environmental impact statement for part two of the application
under Sec. 51.72. No updating of the information contained in the final
environmental impact statement prepared for part one is necessary in
preparation of the supplemental environmental impact statement. The
draft supplement must consider all environmental impacts associated with
the prior issuance of the limited work authorization, but may not
address or consider the sunk costs associated with the limited work
authorization.
(c) Limited work authorization submitted as part of an early site
permit application. If the application for a limited work authorization
is submitted as part of an application for an early site permit, then
the NRC will prepare an environmental impact statement in accordance
with Sec. 51.75(b). However, the analysis called for by Sec. 51.71(d)
must also address the activities proposed to be conducted under the
limited work authorization.
(d) Limited work authorization request submitted by an early site
permit holder. If the application for a limited work authorization is
submitted by a holder of an early site permit, then the NRC will prepare
a draft supplement to the environmental impact statement for the early
site permit. The supplement is limited to consideration of the
activities proposed to be conducted under the limited work
authorization, the adequacy of the proposed redress plan, and whether
there is new and significant information identified with respect to
issues related to the impacts of construction of the facility that
[[Page 43]]
were resolved in the early site permit proceeding with respect to the
environmental impacts of the activities to be conducted under the
limited work authorization. No other updating of the information
contained in the final environmental impact statement prepared for the
early site permit is required.
(e) Limited work authorization for a site where an environmental
impact statement was prepared, but the facility construction was not
completed. If the limited work authorization is for activities to be
conducted at a site for which the Commission has previously prepared an
environmental impact statement for the construction and operation of a
nuclear power plant, and a construction permit was issued but
construction of the plant was not completed, then the draft
environmental impact statement shall incorporate by reference the
earlier environmental impact statement. The draft environmental impact
statement must be limited to a consideration of whether there is
significant new information with respect to the environmental impacts of
construction, relevant to the activities to be conducted under the
limited work authority, so that the conclusion of the referenced
environmental impact statement on the impacts of construction would,
when analyzed in accordance with Sec. 51.71, lead to the conclusion
that the limited work authorization should not be issued or should be
issued with appropriate conditions.
(f) Draft environmental impact statement. A draft environmental
impact statement prepared under this section must separately evaluate
the environmental impacts and proposed alternatives attributable to the
activities proposed to be conducted under the limited work
authorization. However, if the ``Applicant's Environmental Report--
Limited Work Authorization Stage,'' also contains the information
required to be submitted in the environmental report required under
Sec. 51.50, then the environmental impact statement must address the
impacts of construction and operation for the proposed facility
(including the environmental impacts attributable to the limited work
authorization), and discuss the overall costs and benefits balancing for
the underlying proposed action, in accordance with Sec. 51.71, and
Sec. 51.75(a) or (c), as applicable.
[72 FR 57445, Oct. 9, 2007]
Sec. 51.77 Distribution of draft environmental impact statement.
(a) In addition to the distribution authorized by Sec. 51.74, a
copy of a draft environmental statement for a licensing action for a
production or utilization facility, except an action authorizing
issuance, amendment or renewal of a license to manufacture a nuclear
power reactor pursuant to 10 CFR part 52, appendix M will also be
distributed to:
(1) The chief executive of the municipality or county identified in
the draft environmental impact statement as the preferred site for the
proposed facility or activity.
(2) Upon request, the chief executive of each municipality or county
identified in the draft environmental impact statement as an alternative
site.
(b) Additional copies will be made available in accordance with
Sec. 51.123.
[49 FR 9381, Mar. 12, 1984, as amended at 54 FR 15398, Apr. 18, 1989]
draft environmental impact statements--materials licenses
Sec. 51.80 Draft environmental impact statement--materials license.
(a) The NRC staff will either prepare a draft environmental impact
statement or as provided in Sec. 51.92, a supplement to a final
environmental impact statement for each type of action identified in
Sec. 51.20(b) (7) through (12). Except as the context may otherwise
require, procedures and measures similar to those described in
Sec. Sec. 51.70, 51.71, 51.72 and 51.73 will be followed.
(b)(1) Independent spent fuel storage installation (ISFSI). Unless
otherwise determined by the Commission and in accordance with the
generic determination in Sec. 51.23(a) and the provisions of Sec.
51.23(b), a draft environmental impact statement on the issuance of an
initial license for storage of spent fuel at an independent spent fuel
storage installation (ISFSI) or any amendment thereto, will address
environmental impacts of spent fuel only for the term of the license or
amendment applied for.
[[Page 44]]
(2) Monitored retrievable storage installation (MRS). As provided in
sections 141 (c), (d), and (e) and 148 (a) and (c) of the Nuclear Waste
Policy Act of 1982, as amended (NWPA) (96 Stat. 2242, 2243, 42 U.S.C.
10161 (c), (d), (e); 101 Stat. 1330-235, 1330-236, 42 U.S.C. 10168 (a)
and (c)), a draft environmental impact statement for the construction of
a monitored retrievable storage installation (MRS) will not address the
need for the MRS or any alternative to the design criteria for an MRS
set forth in section 141(b)(1) of the NWPA (96 Stat. 2242, 42 U.S.C.
10161(b)(1)) but may consider alternative facility designs which are
consistent with these design criteria.
[49 FR 34695, Aug. 31, 1984, as amended at 53 FR 31682, Aug. 19, 1988]
Sec. 51.81 Distribution of draft environmental impact statement.
Copies of the draft environmental impact statement and any
supplement to the draft environmental impact statement will be
distributed in accordance with the provisions of Sec. 51.74.
draft environmental impact statements--rulemaking
Sec. 51.85 Draft environmental impact statement--rulemaking.
Except as the context may otherwise require, procedures and measures
similar to those described in Sec. Sec. 51.70, 51.71, 51.72 and 51.73
will be followed in proceedings for rulemaking for which the Commission
has determined to prepare an environmental impact statement.
Sec. 51.86 Distribution of draft environmental impact statement.
Copies of the draft environmental impact statement and any
supplement to the draft environmental impact statement will be
distributed in accordance with the provisions of Sec. 51.74.
legislative environmental impact statements--proposals for legislation
Sec. 51.88 Proposals for legislation.
The Commission will, as a matter of policy, follow the provisions of
40 CFR 1506.8 regarding the NEPA process for proposals for legislation.
final environmental impact statements--general requirements
Sec. 51.90 Final environmental impact statement--general.
After receipt and consideration of comments requested pursuant to
Sec. Sec. 51.73 and 51.117, the NRC staff will prepare a final
environmental impact statement in accordance with the requirements in
Sec. Sec. 51.70(b) and 51.71 for a draft environmental impact
statement. The format provided in section 1(a) of appendix A of this
subpart should be used.
Sec. 51.91 Final environmental impact statement--contents.
(a)(1) The final environmental impact statement will include
responses to any comments on the draft environmental impact statement or
on any supplement to the draft environmental impact statement. Responses
to comments may include:
(i) Modification of alternatives, including the proposed action;
(ii) Development and evaluation of alternatives not previously given
serious consideration;
(iii) Supplementation or modification of analyses;
(iv) Factual corrections;
(v) Explanation of why comments do not warrant further response,
citing sources, authorities or reasons which support this conclusion.
(2) All substantive comments received on the draft environmental
impact statement or any supplement to the draft environmental impact
statement (or summaries thereof where the response has been
exceptionally voluminous) will be attached to the final statement,
whether or not each comment is discussed individually in the text of the
statement.
(3) If changes in the draft environmental impact statement in
response to comments are minor and are confined either to factual
corrections or to explanations of why the comments do not warrant
further response, the changes may be made by attaching errata sheets to
the draft statement. The entire document with a new cover may then be
issued as the final environmental impact statement.
[[Page 45]]
(b) The final environmental impact statement will discuss any
relevant responsible opposing view not adequately discussed in the draft
environmental impact statement or in any supplement to the draft
environmental impact statement, and respond to the issues raised.
(c) The final environmental impact statement will state how the
alternatives considered in it and decisions based on it will or will not
achieve the requirements of sections 101 and 102(1) of NEPA and of any
other relevant and applicable environmental laws and policies.
(d) The final environmental impact statement will include a final
analysis and a final recommendation on the action to be taken.
Sec. 51.92 Supplement to the final environmental impact statement.
(a) If the proposed action has not been taken, the NRC staff will
prepare a supplement to a final environmental impact statement for which
a notice of availability has been published in the Federal Register as
provided in Sec. 51.118, if:
(1) There are substantial changes in the proposed action that are
relevant to environmental concerns; or
(2) There are new and significant circumstances or information
relevant to environmental concerns and bearing on the proposed action or
its impacts.
(b) In a proceeding for a combined license application under 10 CFR
part 52 referencing an early site permit under part 52, the NRC staff
shall prepare a supplement to the final environmental impact statement
for the referenced early site permit in accordance with paragraph (e) of
this section.
(c) The NRC staff may prepare a supplement to a final environmental
impact statement when, in its opinion, preparation of a supplement will
further the purposes of NEPA.
(d) The supplement to a final environmental impact statement will be
prepared in the same manner as the final environmental impact statement
except that a scoping process need not be used.
(e) The supplement to an early site permit final environmental
impact statement which is prepared for a combined license application in
accordance with Sec. 51.75(c)(1) and paragraph (b) of this section
must:
(1) Identify the proposed action as the issuance of a combined
license for the construction and operation of a nuclear power plant as
described in the combined license application at the site described in
the early site permit referenced in the combined license application;
(2) Incorporate by reference the final environmental impact
statement prepared for the early site permit;
(3) Contain no separate discussion of alternative sites;
(4) Include an analysis of the economic, technical, and other
benefits and costs of the proposed action, to the extent that the final
environmental impact statement prepared for the early site permit did
not include an assessment of these benefits and costs;
(5) Include an analysis of other energy alternatives, to the extent
that the final environmental impact statement prepared for the early
site permit did not include an assessment of energy alternatives;
(6) Include an analysis of any environmental issue related to the
impacts of construction or operation of the facility that was not
resolved in the proceeding on the early site permit; and
(7) Include an analysis of the issues related to the impacts of
construction and operation of the facility that were resolved in the
early site permit proceeding for which new and significant information
has been identified, including, but not limited to, new and significant
information demonstrating that the design of the facility falls outside
the site characteristics and design parameters specified in the early
site permit.
(f)(1) A supplement to a final environmental impact statement will
be accompanied by or will include a request for comments as provided in
Sec. 51.73 and a notice of availability will be published in the
Federal Register as provided in Sec. 51.117 if paragraphs (a) or (b) of
this section applies.
(2) If comments are not requested, a notice of availability of a
supplement
[[Page 46]]
to a final environmental impact statement will be published in the
Federal Register as provided in Sec. 51.118.
[72 FR 49515, Aug. 28, 2007]
Sec. 51.93 Distribution of final environmental impact statement and
supplement to final environmental impact statement; news releases.
(a) A copy of the final environmental impact statement will be
distributed to:
(1) The Environmental Protection Agency.
(2) The applicant or petitioner for rulemaking and any other party
to the proceeding.
(3) Appropriate State, regional and metropolitan clearinghouses.
(4) Each commenter.
(b) Additional copies will be made available in accordance with
Sec. 51.123.
(c) If the final environmental impact statement is unusually long or
there are so many comments on a draft environmental impact statement or
any supplement to a draft environmental impact statement that
distribution of the entire final statement to all commenters is
impracticable, a summary of the final statement and the substantive
comments will be distributed. When the final environmental impact
statement has been prepared by adding errata sheets to the draft
environmental impact statement as provided in Sec. 51.91(a)(3), only
the comments, the responses to the comments and the changes to the
environmental impact statement will be distributed.
(d) A supplement to a final environmental impact statement will be
distributed in the same manner as the final environmental impact
statement to which it relates.
(e) News releases stating the availability and place for obtaining
or inspecting a final environmental impact statement or supplement will
be provided to local newspapers and other appropriate media.
(f) A notice of availability will be published in the Federal
Register in accordance with Sec. 51.118.
Sec. 51.94 Requirement to consider final environmental impact statement.
The final environmental impact statement, together with any comments
and any supplement, will accompany the application or petition for
rulemaking through, and be considered in, the Commission's
decisionmaking process. The final environmental impact statement,
together with any comments and any supplement, will be made a part of
the record of the appropriate adjudicatory or rulemaking proceeding.
final environmental impact statements--production and utilization
facilities
Sec. 51.95 Postconstruction environmental impact statements.
(a) General. Any supplement to a final environmental impact
statement or any environmental assessment prepared under the provisions
of this section may incorporate by reference any information contained
in a final environmental document previously prepared by the NRC staff
that relates to the same production or utilization facility. Documents
that may be referenced include, but are not limited to, the final
environmental impact statement; supplements to the final environmental
impact statement, including supplements prepared at the operating
license stage; NRC staff-prepared final generic environmental impact
statements; environmental assessments and records of decisions prepared
in connection with the construction permit, the operating license, the
early site permit, or the combined license and any license amendment for
that facility. A supplement to a final environmental impact statement
will include a request for comments as provided in Sec. 51.73.
(b) Initial operating license stage. In connection with the issuance
of an operating license for a production or utilization facility, the
NRC staff will prepare a supplement to the final environmental impact
statement on the construction permit for that facility, which will
update the prior environmental review. The supplement will only cover
matters that differ from the final environmental impact statement or
that reflect significant new information concerning matters discussed in
the final environmental impact statement. Unless otherwise determined by
[[Page 47]]
the Commission, a supplement on the operation of a nuclear power plant
will not include a discussion of need for power, or of alternative
energy sources, or of alternative sites, or of any aspect of the storage
of spent fuel for the nuclear power plant within the scope of the
generic determination in Sec. 51.23(a) and in accordance with Sec.
51.23(b), and will only be prepared in connection with the first
licensing action authorizing full-power operation.
(c) Operating license renewal stage. In connection with the renewal
of an operating license or combined license for a nuclear power plant
under parts 52 or 54 of this chapter, the Commission shall prepare an
environmental impact statement, which is a supplement to the
Commission's NUREG-1437, ``Generic Environmental Impact Statement for
License Renewal of Nuclear Plants'' (May 1996), which is available in
the NRC Public Document Room, 11555 Rockville Pike, Rockville, Maryland.
(1) The supplemental environmental impact statement for the
operating license renewal stage shall address those issues as required
by Sec. 51.71. In addition, the NRC staff must comply with 40 CFR
1506.6(b)(3) in conducting the additional scoping process as required by
Sec. 51.71(a).
(2) The supplemental environmental impact statement for license
renewal is not required to include discussion of need for power or the
economic costs and economic benefits of the proposed action or of
alternatives to the proposed action except insofar as such benefits and
costs are either essential for a determination regarding the inclusion
of an alternative in the range of alternatives considered or relevant to
mitigation. In addition, the supplemental environmental impact statement
prepared at the license renewal stage need not discuss other issues not
related to the environmental effects of the proposed action and the
alternatives, or any aspect of the storage of spent fuel for the
facility within the scope of the generic determination in Sec. 51.23(a)
and in accordance with Sec. 51.23(b). The analysis of alternatives in
the supplemental environmental impact statement should be limited to the
environmental impacts of such alternatives and should otherwise be
prepared in accordance with Sec. 51.71 and appendix A to subpart A of
this part.
(3) The supplemental environmental impact statement shall be issued
as a final impact statement in accordance with Sec. Sec. 51.91 and
51.93 after considering any significant new information relevant to the
proposed action contained in the supplement or incorporated by
reference.
(4) The supplemental environmental impact statement must contain the
NRC staff's recommendation regarding the environmental acceptability of
the license renewal action. In order to make its recommendation and
final conclusion on the proposed action, the NRC staff, adjudicatory
officers, and Commission shall integrate the conclusions, as amplified
by the supporting information in the generic environmental impact
statement for issues designated Category 1 (with the exception of
offsite radiological impacts for collective effects and the disposal of
spent fuel and high level waste) or resolved Category 2,information
developed for those open Category 2 issues applicable to the plant in
accordance with Sec. 51.53(c)(3)(ii), and any significant new
information. Given this information, the NRC staff, adjudicatory
officers, and Commission shall determine whether or not the adverse
environmental impacts of license renewal are so great that preserving
the option of license renewal for energy planning decisionmakers would
be unreasonable.
(d) Postoperating license stage. In connection with the amendment of
an operating or combined license authorizing decommissioning activities
at a production or utilization facility covered by Sec. 51.20, either
for unrestricted use or based on continuing use restrictions applicable
to the site, or with the issuance, amendment or renewal of a license to
store spent fuel at a nuclear power reactor after expiration of the
operating or combined license for the nuclear power reactor, the NRC
staff will prepare a supplemental environmental impact statement for the
post operating or post combined license stage or an environmental
assessment, as appropriate, which will update the prior environmental
documentation prepared by the NRC for compliance
[[Page 48]]
with NEPA under the provisions of this part. The supplement or
assessment may incorporate by reference any information contained in the
final environmental impact statement--for the operating or combined
license stage, as appropriate, or in the records of decision prepared in
connection with the early site permit, construction permit, operating
license, or combined license for that facility. The supplement will
include a request for comments as provided in Sec. 51.73. Unless
otherwise required by the Commission in accordance with the generic
determination in Sec. 51.23(a) and the provisions of Sec. 51.23(b), a
supplemental environmental impact statement for the postoperating or
post combined license stage or an environmental assessment, as
appropriate, will address the environmental impacts of spent fuel
storage only for the term of the license, license amendment or license
renewal applied for.
[61 FR 66545, Dec. 18, 1996, as amended at 72 FR 49516, Aug. 28, 2007]
final environmental impact statements--materials licenses
Sec. 51.97 Final environmental impact statement--materials license.
(a) Independent spent fuel storage installation (ISFSI). Unless
otherwise determined by the Commission, and in accordance with the
generic determination in Sec. 51.23(a) and the provisions of Sec.
51.23(b), a final environmental impact statement on the issuance of an
initial license for the storage of spent fuel at an independent spent
fuel storage installation (ISFSI) or any amendment thereto, will address
environmental impacts of spent fuel storage only for the term of the
license or amendment applied for.
(b) Monitored retrievable storage facility (MRS). As provided in
sections 141 (c), (d), and (e) and 148 (a) and (c) of the Nuclear Waste
Policy Act of 1982, as amended (NWPA) (96 Stat. 2242, 2243, 42 U.S.C.
10161 (c), (d), (e); 101 Stat. 1330-235, 1330-236, 42 U.S.C. 10168 (a),
(c)) a final environmental impact statement for the construction of a
monitored retrievable storage installation (MRS) will not address the
need for the MRS or any alternative to the design criteria for an MRS
set forth in section 141(b)(1) of the NWPA (96 Stat. 2242, 42 U.S.C.
10161(b)(1)) but may consider alternative facility designs which are
consistent with these design criteria.
(c) Uranium enrichment facility. As provided in section 5(e) of the
Solar, Wind, Waste, and Geothermal Power Production Incentives Act of
1990 (104 Stat. 2834 at 2835, 42 U.S.C. 2243), a final environmental
impact statement must be prepared before the hearing on the issuance of
a license for a uranium enrichment facility is completed.
[49 FR 34695, Aug. 31, 1984, as amended at 53 FR 31682, Aug. 19, 1988;
57 FR 18392, Apr. 30, 1992]
final environmental impact statements--rulemaking
Sec. 51.99 [Reserved]
NEPA Procedure and Administrative Action
general
Sec. 51.100 Timing of Commission action.
(a)(1) Except as provided in Sec. 51.13 and paragraph (b) of this
section, no decision on a proposed action, including the issuance of a
permit, license, or other form of permission, or amendment to or renewal
of a permit, license, or other form of permission, or the issuance of an
effective regulation, for which an environmental impact statement is
required, will be made and no record of decision will be issued until
the later of the following dates:
(i) Ninety (90) days after publication by the Environmental
Protection Agency of a Federal Register notice stating that the draft
environmental impact statement has been filed with EPA.
(ii) Thirty (30) days after publication by the Environmental
Protection Agency of a Federal Register notice stating that the final
environmental impact statement has been filed with EPA.
(2) If a notice of filing of a final environmental impact statement
is published by the Environmental Protection Agency within ninety (90)
days after a notice of filing of a draft environmental impact statement
has been published by EPA, the minimum thirty
[[Page 49]]
(30) day period and the minimum ninety (90) day period may run
concurrently to the extent they overlap.
(b) In any rulemaking proceeding for the purpose of protecting the
public health or safety or the common defense and security, the
Commission may make and publish the decision on the final rule at the
same time that the Environmental Protection Agency publishes the Federal
Register notice of filing of the final environmental impact statement.
Sec. 51.101 Limitations on actions.
(a) Until a record of decision is issued in connection with a
proposed licensing or regulatory action for which an environmental
impact statement is required under Sec. 51.20, or until a final finding
of no significant impact is issued in connection with a proposed
licensing or regulatory action for which an environmental assessment is
required under Sec. 51.21:
(1) No action concerning the proposal may be taken by the Commission
which would (i) have an adverse environmental impact, or (ii) limit the
choice of reasonable alternatives.
(2) Any action concerning the proposal taken by an applicant which
would (i) have an adverse environmental impact, or (ii) limit the choice
of reasonable alternatives may be grounds for denial of the license. In
the case of an application covered by Sec. Sec. 30.32(f), 40.31(f),
50.10(c), 70.21(f), or Sec. Sec. 72.16 and 72.34 of this chapter, the
provisions of this paragraph will be applied in accordance with
Sec. Sec. 30.33(a)(5), 40.32(e), 50.10 (c) and (e), 70.23(a)(7) or
Sec. 72.40(b) of this chapter, as appropriate.
(b) While work on a required program environmental impact statement
is in progress, the Commission will not undertake in the interim any
major Federal action covered by the program which may significantly
affect the quality of the human environment unless such action:
(1) Is justified independently of the program;
(2) Is itself accompanied by an adequate environmental impact
statement; and
(3) Will not prejudice the ultimate decision on the program. Absent
any satisfactory explanation to the contrary, interim action which tends
to determine subsequent development or limit reasonable alternatives,
will be considered prejudicial.
(c) This section does not preclude any applicant for an NRC permit,
license, or other form of permission, or amendment to or renewal of an
NRC permit, license, or other form of permission, (1) from developing
any plans or designs necessary to support an application; or (2) after
prior notice and consultation with NRC staff, (i) from performing any
physical work necessary to support an application, or (ii) from
performing any other physical work relating to the proposed action if
the adverse environmental impact of that work is de minimis.
[49 FR 9381, Mar. 12, 1984, as amended at 53 FR 31682, Aug. 19, 1988]
Sec. 51.102 Requirement to provide a record of decision; preparation.
(a) A Commission decision on any action for which a final
environmental impact statement has been prepared shall be accompanied by
or include a concise public record of decision.
(b) Except as provided in paragraph (c) of this section, the record
of decision will be prepared by the NRC staff director authorized to
take the action.
(c) When a hearing is held on the proposed action under the
regulations in subpart G of part 2 of this chapter or when the action
can only be taken by the Commissioners acting as a collegial body, the
initial decision of the presiding officer or the final decision of the
Atomic Safety and Licensing Appeal Board or the final decision of the
Commissioners acting as a collegial body will constitute the record of
decision. An initial or final decision constituting the record of
decision will be distributed as provided in Sec. 51.93.
Sec. 51.103 Record of decision--general.
(a) The record of decision required by Sec. 51.102 shall be clearly
identified and shall:
(1) State the decision.
(2) Identify all alternatives considered by the Commission in
reaching the decision, state that these alternatives were included in
the range of
[[Page 50]]
alternatives discussed in the environmental impact statement, and
specify the alternative or alternatives which were considered to be
environmentally preferable.
(3) Discuss preferences among alternatives based on relevant
factors, including economic and technical considerations where
appropriate, the NRC's statutory mission, and any essential
considerations of national policy, which were balanced by the Commission
in making the decision and state how these considerations entered into
the decision.
(4) State whether the Commission has taken all practicable measures
within its jurisdiction to avoid or minimize environmental harm from the
alternative selected, and if not, to explain why those measures were not
adopted. Summarize any license conditions and monitoring programs
adopted in connection with mitigation measures.
(5) In making a final decision on a license renewal action pursuant
to part 54 of this chapter, the Commission shall determine whether or
not the adverse environmental impacts of license renewal are so great
that preserving the option of license renewal for energy planning
decisionmakers would be unreasonable.
(6) In a construction permit or a combined license proceeding where
a limited work authorization under 10 CFR 50.10 was issued, the
Commission's decision on the construction permit or combined license
application will not address or consider the sunk costs associated with
the limited work authorization in determining the proposed action.
(b) The record of decision may be integrated into any other record
prepared by the Commission in connection with the action.
(c) The record of decision may incorporate by reference material
contained in a final environmental impact statement.
[49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28490, June 5, 1996; 61
FR 66546, Dec. 18, 1996; 61 FR 68543, Dec. 30, 1996; 72 FR 57445, Oct.
9, 2007]
Sec. 51.104 NRC proceeding using public hearings; consideration of
environmental impact statement.
(a)(1) In any proceeding in which (i) a hearing is held on the
proposed action, (ii) a final environmental impact statement has been
prepared in connection with the proposed action, and (iii) matters
within the scope of NEPA and this subpart are in issue, the NRC staff
may not offer the final environmental impact statement in evidence or
present the position of the NRC staff on matters within the scope of
NEPA and this subpart until the final environmental impact statement is
filed with the Environmental Protection Agency, furnished to commenting
agencies and made available to the public.
(2) Any party to the proceeding may take a position and offer
evidence on the aspects of the proposed action within the scope of NEPA
and this subpart in accordance with the provisions of part 2 of this
chapter applicable to that proceeding or in accordance with the terms of
the notice of hearing.
(3) In the proceeding the presiding officer will decide those
matters in controversy among the parties within the scope of NEPA and
this subpart.
(b) In any proceeding in which a hearing is held where the NRC staff
has determined that no environmental impact statement need be prepared
for the proposed action, unless the Commission orders otherwise, any
party to the proceeding may take a position and offer evidence on the
aspects of the proposed action within the scope of NEPA and this subpart
in accordance with the provisions of part 2 of this chapter applicable
to that proceeding or in accordance with the terms of the notice of
hearing. In the proceeding, the presiding officer will decide any such
matters in controversy among the parties.
(c) In any proceeding in which a limited work authorization is
requested, unless the Commission orders otherwise, a party to the
proceeding may take a position and offer evidence only on the aspects of
the proposed action within the scope of NEPA and this subpart which are
within the scope of that party's admitted contention, in accordance with
the provisions of part 2 of this chapter applicable to the limited
[[Page 51]]
work authorization or in accordance with the terms of any notice of
hearing applicable to the limited work authorization. In the proceeding,
the presiding officer will decide all matters in controversy among the
parties.
[49 FR 9381, Mar. 12, 1984, as amended at 72 FR 57445, Oct. 9, 2007]
production and utilization facilities
Sec. 51.105 Public hearings in proceedings for issuance of construction
permits or early site permits; limited work authorizations.
(a) In addition to complying with applicable requirements of Sec.
51.104, in a proceeding for the issuance of a construction permit or
early site permit for a nuclear power reactor, testing facility, fuel
reprocessing plant or isotopic enrichment plant, the presiding officer
will:
(1) Determine whether the requirements of Sections 102(2) (A), (C),
and (E) of NEPA and the regulations in this subpart have been met;
(2) Independently consider the final balance among conflicting
factors contained in the record of the proceeding with a view to
determining the appropriate action to be taken;
(3) Determine, after weighing the environmental, economic,
technical, and other benefits against environmental and other costs, and
considering reasonable alternatives, whether the construction permit or
early site permit should be issued, denied, or appropriately conditioned
to protect environmental values;
(4) Determine, in an uncontested proceeding, whether the NEPA review
conducted by the NRC staff has been adequate; and
(5) Determine, in a contested proceeding, whether in accordance with
the regulations in this subpart, the construction permit or early site
permit should be issued as proposed by the NRC's Director, Office of New
Reactors or Director, Office of Nuclear Reactor Regulation, as
appropriate.
(b) The presiding officer in an early site permit hearing shall not
admit contentions proffered by any party concerning the benefits
assessment (e.g., need for power) or alternative energy sources if those
issues were not addressed by the applicant in the early site permit
application.
(c)(1) In addition to complying with the applicable provisions of
Sec. 51.104, in any proceeding for the issuance of a construction
permit for a nuclear power plant or an early site permit under part 52
of this chapter, where the applicant requests a limited work
authorization under Sec. 50.10(d) of this chapter, the presiding
officer shall--
(i) Determine whether the requirements of Section 102(2)(A), (C),
and (E) of NEPA and the regulations in the subpart have been met, with
respect to the activities to be conducted under the limited work
authorization;
(ii) Independently consider the balance among conflicting factors
with respect to the limited work authorization which is contained in the
record of the proceeding, with a view to determining the appropriate
action to be taken;
(iii) Determine whether the redress plan will adequately redress the
activities performed under the limited work authorization, should
limited work activities be terminated by the holder or the limited work
authorization be revoked by the NRC, or upon effectiveness of the
Commission's final decision denying the associated construction permit
or early site permit, as applicable;
(iv) In an uncontested proceeding, determine whether the NEPA review
conducted by the NRC staff for the limited work authorization has been
adequate; and
(v) In a contested proceeding, determine whether, in accordance with
the regulations in this subpart, the limited work authorization should
be issued as proposed.
(2) If the limited work authorization is for activities to be
conducted at a site for which the Commission has previously prepared an
environmental impact statement for the construction and operation of a
nuclear power plant, and a construction permit was issued but
construction of the plant was never completed, then in making the
determinations in paragraph (c)(1) of this section, the presiding
officer shall be limited to a consideration whether there is, with
respect to construction
[[Page 52]]
activities encompassed by the environmental impact statement which are
analogous to the activities to be conducted under the limited work
authorization, new and significant information on the environmental
impacts of those activities, such that the limited work authorization
should not be issued as proposed.
(3) The presiding officer's determination in this paragraph shall be
made in a partial initial decision to be issued separately from, and in
advance of, the presiding officer's decision in paragraph (a) of this
section.
[72 FR 49516, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007; 73
FR 5724, Jan. 31, 2008]
Sec. 51.105a Public hearings in proceedings for issuance of
manufacturing licenses.
In addition to complying with applicable requirements of Sec.
51.31(c), in a proceeding for the issuance of a manufacturing license,
the presiding officer will determine whether, in accordance with the
regulations in this subpart, the manufacturing license should be issued
as proposed by the NRC's Director, Office of New Reactors or Director,
Office of Nuclear Reactor Regulation, as appropriate.
[73 FR 5724, Jan. 31, 2008]
Sec. 51.106 Public hearings in proceedings for issuance of operating
licenses.
(a) Consistent with the requirements of this section and as
appropriate, the presiding officer in an operating license hearing shall
comply with any applicable requirements of Sec. Sec. 51.104 and 51.105.
(b) During the course of a hearing on an application for issuance of
an operating license for a nuclear power reactor, or a testing facility,
the presiding officer may authorize, pursuant to Sec. 50.57(c) of this
chapter, the loading of nuclear fuel in the reactor core and limited
operation within the scope of Sec. 50.57(c) of this chapter, upon
compliance with the procedures described therein. In any such hearing,
where any party opposes such authorization on the basis of matters
covered by subpart A of this part, the provisions of Sec. Sec. 51.104
and 51.105 will apply, as appropriate.
(c) The presiding officer in an operating license hearing shall not
admit contentions proffered by any party concerning need for power or
alternative energy sources or alternative sites for the facility for
which an operating license is requested.
(d) The presiding officer in an operating license hearing shall not
raise issues concerning alternative sites for the facility for which an
operating license is requested sua sponte.
Sec. 51.107 Public hearings in proceedings for issuance of combined
licenses; limited work authorizations.
(a) In addition to complying with the applicable requirements of
Sec. 51.104, in a proceeding for the issuance of a combined license for
a nuclear power reactor under part 52 of this chapter, the presiding
officer will:
(1) Determine whether the requirements of Sections 102(2) (A), (C),
and (E) of NEPA and the regulations in this subpart have been met;
(2) Independently consider the final balance among conflicting
factors contained in the record of the proceeding with a view to
determining the appropriate action to be taken;
(3) Determine, after weighing the environmental, economic,
technical, and other benefits against environmental and other costs, and
considering reasonable alternatives, whether the combined license should
be issued, denied, or appropriately conditioned to protect environmental
values;
(4) Determine, in an uncontested proceeding, whether the NEPA review
conducted by the NRC staff has been adequate; and
(5) Determine, in a contested proceeding, whether in accordance with
the regulations in this subpart, the combined license should be issued
as proposed by the NRC's Director, Office of New Reactors or Director,
Office of Nuclear Reactor Regulation, as appropriate.
(b) If a combined license application references an early site
permit, then the presiding officer in the combined license hearing shall
not admit any contention proffered by any party on
[[Page 53]]
environmental issues which have been accorded finality under Sec. 52.39
of this chapter, unless the contention:
(1) Demonstrates that the nuclear power reactor proposed to be built
does not fit within one or more of the site characteristics or design
parameters included in the early site permit;
(2) Raises any significant environmental issue that was not resolved
in the early site permit proceeding; or
(3) Raises any issue involving the impacts of construction and
operation of the facility that was resolved in the early site permit
proceeding for which new and significant information has been
identified.
(c) If the combined license application references a standard design
certification, or proposes to use a manufactured reactor, then the
presiding officer in a combined license hearing shall not admit
contentions proffered by any party concerning severe accident mitigation
design alternatives unless the contention demonstrates that the site
characteristics fall outside of the site parameters in the standard
design certification or underlying manufacturing license for the
manufactured reactor.
(d)(1) In any proceeding for the issuance of a combined license
where the applicant requests a limited work authorization under Sec.
50.10(d) of this chapter, the presiding officer, in addition to
complying with any applicable provision of Sec. 51.104, shall:
(i) Determine whether the requirements of Section 102(2)(A), (C),
and (E) of NEPA and the regulations in this subpart have been met, with
respect to the activities to be conducted under the limited work
authorization;
(ii) Independently consider the balance among conflicting factors
with respect to the limited work authorization which is contained in the
record of the proceeding, with a view to determining the appropriate
action to be taken;
(iii) Determine whether the redress plan will adequately redress the
activities performed under the limited work authorization, should
limited work activities be terminated by the holder or the limited work
authorization be revoked by the NRC, or upon effectiveness of the
Commission's final decision denying the combined license application;
(iv) In an uncontested proceeding, determine whether the NEPA review
conducted by the NRC staff for the limited work authorization has been
adequate; and
(v) In a contested proceeding, determine whether, in accordance with
the regulations in this subpart, the limited work authorization should
be issued as proposed by the Director of New Reactors or the Director of
Nuclear Reactor Regulation, as applicable.
(2) If the limited work authorization is for activities to be
conducted at a site for which the Commission has previously prepared an
environmental impact statement for the construction and operation of a
nuclear power plant, and a construction permit was issued but
construction of the plant was never completed, then in making the
determinations in paragraph (c)(1) of this section, the presiding
officer shall be limited to a consideration whether there is, with
respect to construction activities encompassed by the environmental
impact statement which are analogous to the activities to be conducted
under the limited work authorization, new and significant information on
the environmental impacts of those activities, so that the limited work
authorization should not be issued as proposed by the Director of New
Reactors or the Director of Nuclear Reactor Regulation, as applicable.
(3) In making the determination required by this section, the
presiding officer may not address or consider the sunk costs associated
with the limited work authorization.
(4) The presiding officer's determination in this paragraph shall be
made in a partial initial decision to be issued separately from, and in
advance of, the presiding officer's decision in paragraph (a) of this
section on the combined license.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007; 73
FR 5724, Jan. 31, 2008]
[[Page 54]]
Sec. 51.108 Public hearings on Commission findings that inspections,
tests, analyses, and acceptance criteria of combined licenses are met.
In any public hearing requested under 10 CFR 52.103(b), the
Commission will not admit any contentions on environmental issues, the
adequacy of the environmental impact statement for the combined license
issued under subpart C of part 52, or the adequacy of any other
environmental impact statement or environmental assessment referenced in
the combined license application. The Commission will not make any
environmental findings in connection with the finding under 10 CFR
52.103(g).
[72 FR 49517, Aug. 28, 2007]
materials licenses
Sec. 51.109 Public hearings in proceedings for issuance of materials
license with respect to a geologic repository.
(a)(1) In a proceeding for issuance of a construction authorization
for a high-level radioactive waste repository at a geologic repository
operations area under parts 60 and 63 of this chapter, and in a
proceeding for issuance of a license to receive and possess source,
special nuclear, and byproduct material at a geologic repository
operations area under parts 60 and 63 of this chapter, the NRC staff
shall, upon the publication of the notice of hearing in the Federal
Register, present its position on whether it is practicable to adopt,
without further supplementation, the environmental impact statement
(including any supplement thereto) prepared by the Secretary of Energy.
If the position of the staff is that supplementation of the
environmental impact statement by NRC is required, it shall file its
final supplemental environmental impact statement with the Environmental
Protection Agency, furnish that statement to commenting agencies, and
make it available to the public, before presenting its position, or as
soon thereafter as may be practicable. In discharging its
responsibilities under this paragraph, the staff shall be guided by the
principles set forth in paragraphs (c) and (d) of this section.
(2) Any other party to the proceeding who contends that it is not
practicable to adopt the DOE environmental impact statement, as it may
have been supplemented, shall file a contention to that effect within
thirty (30) days after the publication of the notice of hearing in the
Federal Register. Such contention must be accompanied by one or more
affidavits which set forth factual and/or technical bases for the claim
that, under the principles set forth in paragraphs (c) and (d) of this
section, it is not practicable to adopt the DOE environmental impact
statement, as it may have been supplemented. The presiding officer shall
resolve disputes concerning adoption of the DOE environmental impact
statement by using, to the extent possible, the criteria and procedures
that are followed in ruling on motions to reopen under Sec. 2.326 of
this chapter.
(b) In any such proceeding, the presiding officer will determine
those matters in controversy among the parties within the scope of NEPA
and this subpart, specifically including whether, and to what extent, it
is practicable to adopt the environmental impact statement prepared by
the Secretary of Energy in connection with the issuance of a
construction authorization and license for such repository.
(c) The presiding officer will find that it is practicable to adopt
any environmental impact statement prepared by the Secretary of Energy
in connection with a geologic repository proposed to be constructed
under Title I of the Nuclear Waste Policy Act of 1982, as amended,
unless:
(1)(i) The action proposed to be taken by the Commission differs
from the action proposed in the license application submitted by the
Secretary of Energy; and
(ii) The difference may significantly affect the quality of the
human environment; or
(2) Significant and substantial new information or new
considerations render such environmental impact statement inadequate.
(d) To the extent that the presiding officer determines it to be
practicable, in accordance with paragraph (c) of this section, to adopt
the environmental impact statement prepared by
[[Page 55]]
the Secretary of Energy, such adoption shall be deemed to satisfy all
responsibilities of the Commission under NEPA and no further
consideration under NEPA or this subpart shall be required.
(e) To the extent that it is not practicable, in accordance with
paragraph (c) of this section, to adopt the environmental impact
statement prepared by the Secretary of Energy, the presiding officer
will:
(1) Determine whether the requirements of section 102(2) (A), (C),
and (E) of NEPA and the regulations in this subpart have been met;
(2) Independently consider the final balance among conflicting
factors contained in the record of the proceeding with a view to
determining the appropriate action to be taken;
(3) Determine, after weighing the environmental, economic, technical
and other benefits against environmental and other costs, whether the
construction authorization or license should be issued, denied, or
appropriately conditioned to protect environmental values;
(4) Determine, in an uncontested proceeding, whether the NEPA review
conducted by the NRC staff has been adequate; and
(5) Determine, in a contested proceeding, whether in accordance with
the regulations in this subpart, the construction authorization or
license should be issued as proposed.
(f) In making the determinations described in paragraph (e), the
environmental impact statement will be deemed modified to the extent
that findings and conclusions differ from those in the final statement
prepared by the Secretary of Energy, as it may have been supplemented.
The initial decision will be distributed to any persons not otherwise
entitled to receive it who responded to the request in the notice of
docketing, as described in Sec. 51.26(c). If the Commission or the
Atomic Safety and Licensing Appeal Board reaches conclusions different
from those of the presiding officer with respect to such matters, the
final environmental impact statement will be deemed modified to that
extent and the decision will be similarly distributed.
(g) The provisions of this section shall be followed, in place of
those set out in Sec. 51.104, in any proceedings for the issuance of a
license to receive and possess source, special nuclear, and byproduct
material at a geologic repository operations area.
[54 FR 27870, July 3, 1989, as amended at 69 FR 2276, Jan. 14, 2004]
rulemaking
Sec. 51.110 [Reserved]
Public Notice of and Access to Environmental Documents
Sec. 51.116 Notice of intent.
(a) In accordance with Sec. 51.26, the appropriate NRC staff
director will publish in the Federal Register a notice of intent stating
that an environmental impact statement will be prepared. The notice will
contain the information specified in Sec. 51.27.
(b) Copies of the notice will be sent to appropriate Federal, State,
and local agencies, and Indian tribes, appropriate State, regional, and
metropolitan clearinghouses and to interested persons upon request. A
public announcement of the notice of intent will also be made.
Sec. 51.117 Draft environmental impact statement--notice of availability.
(a) Upon completion of a draft environmental impact statement or any
supplement to a draft environmental impact statement, the appropriate
NRC staff director will publish a notice of availability of the
statement in the Federal Register.
(b) The notice will request comments on the proposed action and on
the draft statement or any supplement to the draft statement and will
specify where comments should be submitted and when the comment period
expires.
(c) The notice will (1) state that copies of the draft statement or
any supplement to the draft statement are available for public
inspection; (2) state where inspection may be made, and (3) state that
any comments of Federal, State, and local agencies, Indian tribes or
other interested persons will be made available for public inspection
when received.
[[Page 56]]
(d) Copies of the notice will be sent to appropriate Federal, State,
and local agencies, and Indian tribes, appropriate State, regional, and
metropolitan clearinghouses, and to interested persons upon request.
Sec. 51.118 Final environmental impact statement--notice of availability.
(a) Upon completion of a final environmental impact statement or any
supplement to a final environmental impact statement, the appropriate
NRC staff director will publish a notice of availability of the
statement in the Federal Register. The notice will state that copies of
the final statement or any supplement to the final statement are
available for public inspection and where inspection may be made. Copies
of the notice will be sent to appropriate Federal, State, and local
agencies, and Indian tribes, appropriate State, regional, and
metropolitan clearinghouses and to interested persons upon request.
(b) Upon adoption of a final environmental impact statement or any
supplement to a final environmental impact statement prepared by the
Department of Energy with respect to a geologic repository that is
subject to the Nuclear Waste Policy Act of 1982, the appropriate NRC
staff director shall follow the procedures set out in paragraph (a) of
this section.
[49 FR 9381, Mar. 12, 1984, as amended at 54 FR 27871, July 3, 1989]
Sec. 51.119 Publication of finding of no significant impact; distribution.
(a) As required by Sec. 51.35, the appropriate NRC staff director
will publish the finding of no significant impact in the Federal
Register. The finding of no significant impact will be identified as a
draft or final finding, and will contain the information specified in
Sec. Sec. 51.32 or 51.33, as appropriate. A draft finding of no
significant impact will include a request for comments which specifies
where comments should be submitted and when the comment period expires.
(b) The finding will state that copies of the finding, the
environmental assessment setting forth the basis for the finding and any
related environmental documents are available for public inspection and
where inspection may be made.
(c) A copy of a final finding will be sent to appropriate Federal,
State, and local agencies, and Indian tribes, appropriate State,
regional, and metropolitan clearinghouses, the applicant or petitioner
for rulemaking and any other party to the proceeding, and if a draft
finding was issued, to each commenter. Additional copies will be made
available in accordance with Sec. 51.123.
Sec. 51.120 Availability of environmental documents for public inspection.
Copies of environmental reports, draft and final environmental
impact statements, environmental assessments, and findings of no
significant impact, together with any related comments and environmental
documents, will be made available at the NRC Web site, http://
www.nrc.gov, and/or at the NRC Public Document Room.
[64 FR 48952, Sept. 9, 1999]
Sec. 51.121 Status of NEPA actions.
Individuals or organizations desiring information on the NRC's NEPA
process or on the status of specific NEPA actions should address
inquiries to:
(a) Utilization facilities: ATTN: Document Control Desk, Director,
Office of Nuclear Reactor Regulation or Director, Office of New
Reactors, as appropriate, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, telephone (301) 415-1270, e-mail
[email protected].
(b) Production facilities: ATTN: Document Control Desk, Director,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
7800, e-mail [email protected].
(c) Materials licenses: ATTN: Document Control Desk, Director,
Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-
7800, e-mail [email protected].
(d) Rulemaking: ATTN: Chief, Rules and Directives Branch, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone (800) 368-5642, e-mail [email protected].
[[Page 57]]
(e) General environmental matters: Executive Director for
Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555,
Telephone: (301) 415-1700.
[53 FR 13399, Apr. 25, 1988, as amended at 60 FR 24552, May 9, 1995; 68
FR 58811, Oct. 10, 2003; 73 FR 5724, Jan. 31, 2008]
Sec. 51.122 List of interested organizations and groups.
The NRC Office of Information Resources Management will maintain a
master list of organizations and groups, including relevant conservation
commissions, known to be interested in the Commission's licensing and
regulatory activities. The NRC Office of Information Resources
Management with the assistance of the appropriate NRC staff director
will select from this master list those organizations and groups that
may have an interest in a specific NRC NEPA action and will promptly
notify such organizations and groups of the availability of a draft
environmental impact statement or a draft finding of no significant
impact.
[49 FR 9381, Mar. 12, 1984, as amended at 52 FR 31612, Aug. 12, 1987; 54
FR 53316, Dec. 28, 1989]
Sec. 51.123 Charges for environmental documents; distribution to
public; distribution to governmental agencies.
(a) Distribution to public. Upon written request to the Office of
Information Services, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, e-mail [email protected], and to the extent available,
single copies of draft environmental impact statements and draft
findings of no significant impact will be made available to interested
persons without charge. Single copies of final environmental impact
statements and final findings of no significant impact will also be
provided without charge to the persons listed in Sec. Sec. 51.93(a) and
51.119(c), respectively. When more than one copy of an environmental
impact statement or a finding of no significant impact is requested or
when available NRC copies have been exhausted, the requestor will be
advised that the NRC will provide copies at the charges specified in
Sec. 9.35 of this chapter.
(b) Distribution to governmental agencies. Upon written request to
the Office of Information Services, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, e-mail [email protected], and to the
extent available, copies of draft and final environmental impact
statements and draft final findings of no significant impact will be
made available in the number requested to Federal, State and local
agencies, Indian Tribes, and State, regional, and metropolitan
clearinghouses. When available NRC copies have been exhausted, the
requester will be advised that the NRC will provide copies at the
charges specified in Sec. 9.35 of this chapter.
(c) Charges. Charges for the reproduction of environmental documents
by the NRC at locations other than the NRC Public Document Room located
in Washington, DC vary according to location.
[50 FR 21037, May 22, 1985, as amended at 52 FR 31612, Aug. 21, 1987; 53
FR 43421, Oct. 27, 1988; 61 FR 9902, Mar. 12, 1996; 64 FR 48952, Sept.
9, 1999; 68 FR 58812, Oct. 10, 2003]
Commenting
Sec. 51.124 Commission duty to comment.
It is the policy of the Commission to comment on draft environmental
impact statements prepared by other Federal agencies, consistent with
the provisions of 40 CFR 1503.2 and 1503.3.
Responsible Official
Sec. 51.125 Responsible official.
The Executive Director for Operations shall be responsible for
overall review of NRC NEPA compliance, except for matters under the
jurisdiction of a presiding officer, administrative judge,
administrative law judge, Atomic Safety and Licensing Board, Atomic
Safety and Licensing Appeal Board, or the Commission acting as a
collegial body.
Sec. Appendix A to Subpart A of Part 51--Format for Presentation of
Material in Environmental Impact Statements
1. General
[[Page 58]]
2. Cover sheet
3. Summary
4. Purpose of and need for action
5. Alternatives including the proposed action
6. Affected environment
7. Environmental consequences and mitigating actions
8. List of preparers
9. Appendices
1. General.
(a) The Commission will use a format for environmental impact
statements which will encourage good analysis and clear presentation of
the alternatives including the proposed action. The following standard
format for environmental impact statements should be followed unless
there is a compelling reason to do otherwise:
(1) Cover sheet*
(2) Summary*
(3) Table of Contents
(4) Purpose of and Need for Action*
(5) Alternatives including the proposed action*
(6) Affected Environment*
(7) Environmental Consequences and Mitigating Actions*
(8) List of Preparers*
(9) List of Agencies, Organizations and Persons to Whom Copies of
the Statement are Sent
(10) Substantive Comments Received and NRC Staff Responses
(11) Index
(12) Appendices (if any)*
If a different format is used, it shall include paragraphs (1), (2),
(3), (8), (9), (10), and (11) of this section and shall include the
substance of paragraphs (4), (5), (6), (7), and (12) of this section, in
any appropriate format.
Additional guidance on the presentation of material under the format
headings identified by an asterisk is set out in sections 2.-9. of this
appendix.
(b) The techniques of tiering and incorporation by reference
described respectively in 40 CFR 1502.20 and 1508.28 and 40 CFR 1502.21
\1\ of CEQ's NEPA regulations may be used as appropriate to aid in the
presentation of issues, eliminate repetition or reduce the size of an
environmental impact statement. In appropriate circumstances, draft or
final environmental impact statements prepared by other Federal agencies
may be adopted in whole or in part in accordance with the procedures
outlined in 40 CFR 1506.3 \2\ of CEQ's NEPA regulations. In final
environmental impact statements, material under the following format
headings will normally be presented in less than 150 pages: Purpose of
and Need for Action, Alternatives Including the Proposed Action,
Affected Environment, and Environmental Consequences and Mitigating
Actions. For proposals of unusual scope or complexity, the material
presented under these format headings may extend to 300 pages.
---------------------------------------------------------------------------
\1\ Tiering--40 CFR 1502.20, 40 CFR 1508.28; Incorporation by
reference--40 CFR 1502.21.
\2\ Adoption--40 CFR 1506.3.
---------------------------------------------------------------------------
2. Cover sheet.
The cover sheet will not exceed one page. It will include:
(a) The name of the NRC office responsible for preparing the
statement and a list of any cooperating agencies.
(b) The title of the proposed action that is the subject of the
statement with a list of the states, counties or municipalities where
the facility or other subject of the action is located, as appropriate.
(c) The name, address, and telephone number of the individual in NRC
who can supply further information.
(d) A designation of the statement as a draft or final statement, or
a draft or final supplement.
(e) A one paragraph abstract of the statement.
(f) For draft environmental impact statements, the date by which
comments must be received. This date may be specified in the form of the
following or a substantially similar statement:
``Comments should be filed no later than \3\ days after the date on
which the Environmental Protection Agency notice stating that the draft
environmental impact statement has been filed with EPA is published in
the Federal Register. Comments received after the expiration of the
comment period will be considered if it is practical to do so but
assurance of consideration of late comments cannot be given.''
---------------------------------------------------------------------------
\3\ The number of days in the comment period should be inserted. The
minimum comment period is 45 days (see Sec. 51.73.)
---------------------------------------------------------------------------
3. Summary.
Each environmental impact statement will contain a summary which
adequately and accurately summarizes the statement. The summary will
stress the major issues considered. The summary will discuss the areas
of controversy, will identify any remaining issues to be resolved, and
will present the major conclusions and recommendations. The summary will
normally not exceed 15 pages.
4. Purpose of and need for action.
The statement will briefly describe and specify the need for the
proposed action. The alternative of no action will be discussed. In the
case of nuclear power plant construction or siting, consideration will
be given to the potential impact of conservation measures in determining
the demand for power and
[[Page 59]]
consequent need for additional generating capacity.
5. Alternatives including the proposed action.
This section is the heart of the environmental impact statement. It
will present the environmental impacts of the proposal and the
alternatives in comparative form. Where important to the comparative
evaluation of alternatives, appropriate mitigating measures of the
alternatives will be discussed. All reasonable alternatives will be
identified. The range of alternatives discussed will encompass those
proposed to be considered by the ultimate decisionmaker. An otherwise
reasonable alternative will not be excluded from discussion solely on
the ground that it is not within the jurisdiction of the NRC. \4\ The
discussion of alternatives will take into accounts, without duplicating,
the environmental information and analyses included in sections, 4., 6.
and 7. of this appendix.
---------------------------------------------------------------------------
\4\ With respect to limitations on NRC's NEPA authority and
responsibility imposed by the Federal Water Pollution Control Act
Amendments of 1972, see Sec. Sec. 51.10(c), 51.22(c)(17) and 51.71(d).
---------------------------------------------------------------------------
In the draft environmental impact statement, this section will
either include a preliminary recommendation on the action to be taken,
or identify the alternatives under consideration.
In the final environmental impact statement, this section will
include a final recommendation on the action to be taken.
6. Affected environment.
The environmental impact statement will succinctly describe the
environment to be affected by the proposed action. Data and analyses in
the statement will be commensurate with the importance of the impact,
with less important material summarized, consolidated, or simply
referenced. Effort and attention will be concentrated on important
issues; useless bulk will be eliminated.
7. Environmental consequences and mitigating actions.
This section discusses the environmental consequences of
alternatives, including the proposed actions and any mitigating actions
which may be taken. Alternatives eliminated from detailed study will be
identified and a discussion of those alternatives will be confined to a
brief statement of the reasons why the alternatives were eliminated. The
level of information for each alternative considered in detail will
reflect the depth of analysis required for sound decisionmaking.
The discussion will include any adverse environmental effects which
cannot be avoided should the alternative be implemented, the
relationship between short-term uses of man's environment and the
maintenance and enhancement of long-term productivity, and any
irreversible or irretrievable commitments of resources which would be
involved in the alternative should it be implemented. This section will
include discussions of:
(a) Direct effects and their significance.
(b) Indirect effects and their significance.
(c) Possible conflicts between the alternative and the objectives of
Federal, regional, State, and local (and in the case of a reservation,
Indian tribe) land use plans, policies and controls for the area
concerned.
(d) Means to mitigate adverse environmental impacts.
8. List of preparers.
The environmental impact statement will list the names and
qualifications (expertise, experience, professional disciplines), of the
persons who were primarily responsible for preparing the environmental
impact statement or significant background papers. Persons responsible
for making an independent evaluation of information submitted by the
applicant or petitioner for rulemaking or others will be included in the
list. Where possible, the persons who are responsible for a particular
analysis, including analyses in background papers, will be identified.
9. Appendices.
An appendix to an environmental impact statement will:
(a) Consist of material prepared in connection with an environmental
impact statement (as distinct from material which is not so prepared and
which is incorporated by reference (40 CFR 1502.21)).
(b) Normally consist of material which substantiates any analysis
fundamental to the impact statement. Discussion of methodology used may
be placed in an appendix.
(c) Normally be analytic.
(d) Be relevant to the decision to be made.
(e) Be circulated with the environmental impact statement or be
readily available on request.
Discussion of Footnotes
1. Tiering.
40 CFR 1502.20 states:
``Agencies are encouraged to tier their environmental impact
statements to eliminate repetitive discussions of the same issues and to
focus on the actual issues ripe for decision at each level of
environmental review (Sec. 1508.28). Whenever a broad environmental
impact statement has been prepared (such as a program or policy
statement) and a subsequent statement or environmental assessment is
then prepared on an action included within the entire program or policy
(such as a site specific action) the subsequent statement or
environmental assessment need only
[[Page 60]]
summarize the issues discussed in the broader statement and incorporate
discussions from the broader statement by reference and shall
concentrate on the issues specific to the subsequent action. The
subsequent document shall state where the earlier document is available.
Tiering may also be appropriate for different stages of actions. (Sec.
1508.28).''
40 CFR 1508.28 states:
`` `Tiering' refers to the coverage of general matters in broader
environmental impact statements (such as national program or policy
statements) with subsequent narrower statements or environmental
analyses (such as regional or basinwide program statements or ultimately
site-specific statements) incorporating by reference the general
discussions and concentrating solely on the issues specific to the
statement subsequently prepared. Tiering is appropriate when the
sequence of statements or analyses is:
``(a) From a program, plan, or policy environmental impact statement
to a program, plan, or policy statement or analysis of lesser scope or
to a site-specific statement or analysis.
``(b) From an environmental impact statement on a specific action at
an early stage (such as need and site selection) to a supplement (which
is preferred) or a subsequent statement or analysis at a later stage
(such as environmental mitigation). Tiering in such cases is appropriate
when it helps the lead agency to focus on the issues which are ripe for
decision and exclude from consideration issues already decided or not
yet ripe.''
Incorporation by reference. 40 CFR 1502.21 states:
``Agencies shall incorporate material into an environmental impact
statement by reference when the effect will be to cut down on bulk
without impeding agency and public review of the action. The
incorporated material shall be cited in the statement and its content
briefly described. No material may be incorporated by reference unless
it is reasonably available for inspection by potentially interested
persons within the time allowed for comment. Material based on
proprietary data which is itself not available for review and comment
shall not be incorporated by reference.''
2. Adoption.
40 CFR 1506.3 states:
``(a) An agency may adopt a Federal draft or final environmental
impact statement or portion thereof provided that the statement or
portion thereof meets the standards for an adequate statement under
these regulations.
``(b) If the actions covered by the original environmental impact
statement and the proposed action are substantially the same, the agency
adopting another agency's statement is not required to recirculate it
except as a final statement. Otherwise the adopting agency shall treat
the statement as a draft and recirculate it (except as provided in
paragraph (c) of this section).
``(c) A cooperating agency may adopt without recirculating the
environmental impact statement of a lead agency when, after an
independent review of the statement, the cooperating agency concludes
that its comments and suggestions have been satisfied.
``(d) When an agency adopts a statement which is not final within
the agency that prepared it, or when the action it assesses is the
subject of a referral under part 1504, or when the statement's adequacy
is the subject of a judicial action which is not final, the agency shall
so specify.''
[49 FR 9381, Mar. 12, 1984, as amended at 61 FR 28490, June 5, 1996; 61
FR 66546, Dec. 18, 1996]
Sec. Appendix B to Subpart A of Part 51--Environmental Effect of
Renewing the Operating License of a Nuclear Power Plant
The Commission has assessed the environmental impacts associated
with granting a renewed operating license for a nuclear power plant to a
licensee who holds either an operating license or construction permit as
of June 30, 1995. Table B-1 summarizes the Commission's findings on the
scope and magnitude of environmental impacts of renewing the operating
license for a nuclear power plant as required by section 102(2) of the
National Environmental Policy Act of 1969, as amended. Table B-1,
subject to an evaluation of those issues identified in Category 2 as
requiring further analysis and possible significant new information,
represents the analysis of the environmental impacts associated with
renewal of any operating license and is to be used in accordance with
Sec. 51.95(c). On a 10-year cycle, the Commission intends to review the
material in this appendix and update it if necessary. A scoping notice
must be published in the Federal Register indicating the results of the
NRC's review and inviting public comments and proposals for other areas
that should be updated.
[[Page 61]]
Table B-1--Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants \1\
----------------------------------------------------------------------------------------------------------------
Category
Issue \2\ Findings \3\
----------------------------------------------------------------------------------------------------------------
Surface Water Quality, Hydrology, and Use (for all plants)
----------------------------------------------------------------------------------------------------------------
Impacts of refurbishment on surface water 1 SMALL. Impacts are expected to be negligible during
quality. refurbishment because best management practices are
expected to be employed to control soil erosion and
spills.
Impacts of refurbishment on surface water 1 SMALL. Water use during refurbishment will not
use. increase appreciably or will be reduced during plant
outage.
Altered current patterns at intake and 1 SMALL. Altered current patterns have not been found to
discharge structures. be a problem at operating nuclear power plants and
are not expected to be a problem during the license
renewal term.
Altered salinity gradients................. 1 SMALL. Salinity gradients have not been found to be a
problem at operating nuclear power plants and are not
expected to be a problem during the license renewal
term.
Altered thermal stratification of lakes.... 1 SMALL. Generally, lake stratification has not been
found to be a problem at operating nuclear power
plants and is not expected to be a problem during the
license renewal term.
Temperature effects on sediment transport 1 SMALL. These effects have not been found to be a
capacity. problem at operating nuclear power plants and are not
expected to be a problem during the license renewal
term.
Scouring caused by discharged cooling water 1 SMALL. Scouring has not been found to be a problem at
most operating nuclear power plants and has caused
only localized effects at a few plants. It is not
expected to be a problem during the license renewal
term.
Eutrophication............................. 1 SMALL. Eutrophication has not been found to be a
problem at operating nuclear power plants and is not
expected to be a problem during the license renewal
term.
Discharge of chlorine or other biocides.... 1 SMALL. Effects are not a concern among regulatory and
resource agencies, and are not expected to be a
problem during the license renewal term.
Discharge of sanitary wastes and minor 1 SMALL. Effects are readily controlled through NPDES
chemical spills. permit and periodic modifications, if needed, and are
not expected to be a problem during the license
renewal term.
Discharge of other metals in waste water... 1 SMALL. These discharges have not been found to be a
problem at operating nuclear power plants with
cooling-tower-based heat dissipation systems and have
been satisfactorily mitigated at other plants. They
are not expected to be a problem during the license
renewal term.
Water use conflicts (plants with once- 1 SMALL. These conflicts have not been found to be a
through cooling systems). problem at operating nuclear power plants with once-
through heat dissipation systems.
Water use conflicts (plants with cooling 2 SMALL OR MODERATE. The issue has been a concern at
ponds or cooling towers using make-up nuclear power plants with cooling ponds and at plants
water from a small river with low flow). with cooling towers. Impacts on instream and riparian
communities near these plants could be of moderate
significance in some situations. See Sec.
51.53(c)(3)(ii)(A).
----------------------------------------------------------------------------------------------------------------
Aquatic Ecology (for all plants)
----------------------------------------------------------------------------------------------------------------
Refurbishment.............................. 1 SMALL. During plant shutdown and refurbishment there
will be negligible effects on aquatic biota because
of a reduction of entrainment and impingement of
organisms or a reduced release of chemicals.
Accumulation of contaminants in sediments 1 SMALL. Accumulation of contaminants has been a concern
or biota. at a few nuclear power plants but has been
satisfactorily mitigated by replacing copper alloy
condenser tubes with those of another metal. It is
not expected to be a problem during the license
renewal term.
Entrainment of phytoplankton and 1 SMALL. Entrainment of phytoplankton and zooplankton
zooplankton. has not been found to be a problem at operating
nuclear power plants and is not expected to be a
problem during the license renewal term.
Cold shock................................. 1 SMALL. Cold shock has been satisfactorily mitigated at
operating nuclear plants with once-through cooling
systems, has not endangered fish populations or been
found to be a problem at operating nuclear power
plants with cooling towers or cooling ponds, and is
not expected to be a problem during the license
renewal term.
Thermal plume barrier to migrating fish.... 1 SMALL. Thermal plumes have not been found to be a
problem at operating nuclear power plants and are not
expected to be a problem during the license renewal
term.
Distribution of aquatic organisms.......... 1 SMALL. Thermal discharge may have localized effects
but is not expected to effect the larger geographical
distribution of aquatic organisms.
Premature emergence of aquatic insects..... 1 SMALL. Premature emergence has been found to be a
localized effect at some operating nuclear power
plants but has not been a problem and is not expected
to be a problem during the license renewal term.
[[Page 62]]
Gas supersaturation (gas bubble disease)... 1 SMALL. Gas supersaturation was a concern at a small
number of operating nuclear power plants with once-
through cooling systems but has been satisfactorily
mitigated. It has not been found to be a problem at
operating nuclear power plants with cooling towers or
cooling ponds and is not expected to be a problem
during the license renewal term.
Low dissolved oxygen in the discharge...... 1 SMALL. Low dissolved oxygen has been a concern at one
nuclear power plant with a once-through cooling
system but has been effectively mitigated. It has not
been found to be a problem at operating nuclear power
plants with cooling towers or cooling ponds and is
not expected to be a problem during the license
renewal term.
Losses from predation, parasitism, and 1 SMALL. These types of losses have not been found to be
disease among organisms exposed to a problem at operating nuclear power plants and are
sublethal stresses. not expected to be a problem during the license
renewal term.
Stimulation of nuisance organisms (e.g., 1 SMALL. Stimulation of nuisance organisms has been
shipworms). satisfactorily mitigated at the single nuclear power
plant with a once-through cooling system where
previously it was a problem. It has not been found to
be a problem at operating nuclear power plants with
cooling towers or cooling ponds and is not expected
to be a problem during the license renewal term.
----------------------------------------------------------------------------------------------------------------
Aquatic Ecology (for plants with once-through and cooling pond heat dissipation systems)
----------------------------------------------------------------------------------------------------------------
Entrainment of fish and shellfish in early 2 SMALL, MODERATE, OR LARGE. The impacts of entrainment
life stages. are small at many plants but may be moderate or even
large at a few plants with once-through and cooling-
pond cooling systems. Further, ongoing efforts in the
vicinity of these plants to restore fish populations
may increase the numbers of fish susceptible to
intake effects during the license renewal period,
such that entrainment studies conducted in support of
the original license may no longer be valid. See Sec.
51.53(c)(3)(ii)(B).
Impingement of fish and shellfish.......... 2 SMALL, MODERATE, OR LARGE. The impacts of impingement
are small at many plants but may be moderate or even
large at a few plants with once-through and cooling-
pond cooling systems. See Sec. 51.53(c)(3)(ii)(B).
Heat shock................................. 2 SMALL, MODERATE, OR LARGE. Because of continuing
concerns about heat shock and the possible need to
modify thermal discharges in response to changing
environmental conditions, the impacts may be of
moderate or large significance at some plants. See
Sec. 51.53(c)(3)(ii)(B).
----------------------------------------------------------------------------------------------------------------
Aquatic Ecology (for plants with cooling-tower-based heat dissipation systems)
----------------------------------------------------------------------------------------------------------------
Entrainment of fish and shellfish in early 1 SMALL. Entrainment of fish has not been found to be a
life stages. problem at operating nuclear power plants with this
type of cooling system and is not expected to be a
problem during the license renewal term.
Impingement of fish and shellfish.......... 1 SMALL. The impingement has not been found to be a
problem at operating nuclear power plants with this
type of cooling system and is not expected to be a
problem during the license renewal term.
Heat shock................................. 1 SMALL. Heat shock has not been found to be a problem
at operating nuclear power plants with this type of
cooling system and is not expected to be a problem
during the license renewal term.
----------------------------------------------------------------------------------------------------------------
Ground-water Use and Quality
----------------------------------------------------------------------------------------------------------------
Impacts of refurbishment on ground-water 1 SMALL. Extensive dewatering during the original
use and quality. construction on some sites will not be repeated
during refurbishment on any sites. Any plant wastes
produced during refurbishment will be handled in the
same manner as in current operating practices and are
not expected to be a problem during the license
renewal term.
Ground-water use conflicts (potable and 1 SMALL. Plants using less than 100 gpm are not expected
service water; plants that use <100 gpm). to cause any ground-water use conflicts.
Ground-water use conflicts (potable and 2 SMALL, MODERATE, OR LARGE. Plants that use more than
service water, and dewatering; plants that 100 gpm may cause ground-water use conflicts with
use 100 gpm). nearby ground-water users. See Sec.
51.53(c)(3)(ii)(C).
Ground-water use conflicts (plants using 2 SMALL, MODERATE, OR LARGE. Water use conflicts may
cooling towers withdrawing make-up water result from surface water withdrawals from small
from a small river). water bodies during low flow conditions which may
affect aquifer recharge, especially if other ground-
water or upstream surface water users come on line
before the time of license renewal. See Sec.
51.53(c)(3)(ii)(A).
[[Page 63]]
Ground-water use conflicts (Ranney wells).. 2 SMALL, MODERATE, OR LARGE. Ranney wells can result in
potential ground-water depression beyond the site
boundary. Impacts of large ground-water withdrawal
for cooling tower makeup at nuclear power plants
using Ranney wells must be evaluated at the time of
application for license renewal. See Sec.
51.53(c)(3)(ii)(C).
Ground-water quality degradation (Ranney 1 SMALL. Ground-water quality at river sites may be
wells). degraded by induced infiltration of poor-quality
river water into an aquifer that supplies large
quantities of reactor cooling water. However, the
lower quality infiltrating water would not preclude
the current uses of ground water and is not expected
to be a problem during the license renewal term.
Ground-water quality degradation (saltwater 1 SMALL. Nuclear power plants do not contribute
intrusion). significantly to saltwater intrusion.
Ground-water quality degradation (cooling 1 SMALL. Sites with closed-cycle cooling ponds may
ponds in salt marshes). degrade ground-water quality. Because water in salt
marshes is brackish, this is not a concern for plants
located in salt marshes.
Ground-water quality degradation (cooling 2 SMALL, MODERATE, OR LARGE. Sites with closed-cycle
ponds at inland sites). cooling ponds may degrade ground-water quality. For
plants located inland, the quality of the ground
water in the vicinity of the ponds must be shown to
be adequate to allow continuation of current uses.
See Sec. 51.53(c)(3)(ii)(D).
----------------------------------------------------------------------------------------------------------------
Terrestrial Resources
----------------------------------------------------------------------------------------------------------------
Refurbishment impacts...................... 2 SMALL, MODERATE, OR LARGE. Refurbishment impacts are
insignificant if no loss of important plant and
animal habitat occurs. However, it cannot be known
whether important plant and animal communities may be
affected until the specific proposal is presented
with the license renewal application. See Sec.
51.53(c)(3)(ii)(E).
Cooling tower impacts on crops and 1 SMALL. Impacts from salt drift, icing, fogging, or
ornamental vegetation. increased humidity associated with cooling tower
operation have not been found to be a problem at
operating nuclear power plants and are not expected
to be a problem during the license renewal term.
Cooling tower impacts on native plants..... 1 SMALL. Impacts from salt drift, icing, fogging, or
increased humidity associated with cooling tower
operation have not been found to be a problem at
operating nuclear power plants and are not expected
to be a problem during the license renewal term.
Bird collisions with cooling towers........ 1 SMALL. These collisions have not been found to be a
problem at operating nuclear power plants and are not
expected to be a problem during the license renewal
term.
Cooling pond impacts on terrestrial 1 SMALL. Impacts of cooling ponds on terrestrial
resources. ecological resources are considered to be of small
significance at all sites.
Power line right-of-way management (cutting 1 SMALL. The impacts of right-of-way maintenance on
and herbicide application). wildlife are expected to be of small significance at
all sites.
Bird collision with power lines............ 1 SMALL. Impacts are expected to be of small
significance at all sites.
Impacts of electromagnetic fields on flora 1 SMALL. No significant impacts of electromagnetic
and fauna (plants, agricultural crops, fields on terrestrial flora and fauna have been
honeybees, wildlife, livestock). identified. Such effects are not expected to be a
problem during the license renewal term.
Floodplains and wetland on power line right 1 SMALL. Periodic vegetation control is necessary in
of way. forested wetlands underneath power lines and can be
achieved with minimal damage to the wetland. No
significant impact is expected at any nuclear power
plant during the license renewal term.
----------------------------------------------------------------------------------------------------------------
Threatened or Endangered Species (for all plants)
----------------------------------------------------------------------------------------------------------------
Threatened or endangered species........... 2 SMALL, MODERATE, OR LARGE. Generally, plant
refurbishment and continued operation are not
expected to adversely affect threatened or endangered
species. However, consultation with appropriate
agencies would be needed at the time of license
renewal to determine whether threatened or endangered
species are present and whether they would be
adversely affected. See Sec. 51.53(c)(3)(ii)(E).
----------------------------------------------------------------------------------------------------------------
Air Quality
----------------------------------------------------------------------------------------------------------------
Air quality during refurbishment (non- 2 SMALL, MODERATE, OR LARGE. Air quality impacts from
attainment and maintenance areas). plant refurbishment associated with license renewal
are expected to be small. However, vehicle exhaust
emissions could be cause for concern at locations in
or near nonattainment or maintenance areas. The
significance of the potential impact cannot be
determined without considering the compliance status
of each site and the numbers of workers expected to
be employed during the outage. See Sec.
51.53(c)(3)(ii)(F).
[[Page 64]]
Air quality effects of transmission lines.. 1 SMALL. Production of ozone and oxides of nitrogen is
insignificant and does not contribute measurably to
ambient levels of these gases.
----------------------------------------------------------------------------------------------------------------
Land Use
----------------------------------------------------------------------------------------------------------------
Onsite land use............................ 1 SMALL. Projected onsite land use changes required
during refurbishment and the renewal period would be
a small fraction of any nuclear power plant site and
would involve land that is controlled by the
applicant.
Power line right of way.................... 1 SMALL. Ongoing use of power line right of ways would
continue with no change in restrictions. The effects
of these restrictions are of small significance.
----------------------------------------------------------------------------------------------------------------
Human Health
----------------------------------------------------------------------------------------------------------------
Radiation exposures to the public during 1 SMALL. During refurbishment, the gaseous effluents
refurbishment. would result in doses that are similar to those from
current operation. Applicable regulatory dose limits
to the public are not expected to be exceeded.
Occupational radiation exposures during 1 SMALL. Occupational doses from refurbishment are
refurbishment. expected to be within the range of annual average
collective doses experienced for pressurized-water
reactors and boiling-water reactors. Occupational
mortality risk from all causes including radiation is
in the mid-range for industrial settings.
Microbiological organisms (occupational 1 SMALL. Occupational health impacts are expected to be
health). controlled by continued application of accepted
industrial hygiene practices to minimize worker
exposures.
Microbiological organisms (public 2 SMALL, MODERATE, OR LARGE. These organisms are not
health)(plants using lakes or canals, or expected to be a problem at most operating plants
cooling towers or cooling ponds that except possibly at plants using cooling ponds, lakes,
discharge to a small river). or canals that discharge to small rivers. Without
site-specific data, it is not possible to predict the
effects generically. See Sec. 51.53(c)(3)(ii)(G).
Noise...................................... 1 SMALL. Noise has not been found to be a problem at
operating plants and is not expected to be a problem
at any plant during the license renewal term.
Electromagnetic fields, acute effects 2 SMALL, MODERATE, OR LARGE. Electrical shock resulting
(electric shock). from direct access to energized conductors or from
induced charges in metallic structures have not been
found to be a problem at most operating plants and
generally are not expected to be a problem during the
license renewal term. However, site-specific review
is required to determine the significance of the
electric shock potential at the site. See Sec.
51.53(c)(3)(ii)(H).
Electromagnetic fields, chronic effects \5\ \4\ NA UNCERTAIN. Biological and physical studies of 60-Hz
electromagnetic fields have not found consistent
evidence linking harmful effects with field
exposures. However, research is continuing in this
area and a consensus scientific view has not been
reached. \5\
Radiation exposures to public (license 1 SMALL. Radiation doses to the public will continue at
renewal term). current levels associated with normal operations.
Occupational radiation exposures (license 1 SMALL. Projected maximum occupational doses during the
renewal term). license renewal term are within the range of doses
experienced during normal operations and normal
maintenance outages, and would be well below
regulatory limits.
----------------------------------------------------------------------------------------------------------------
Socioeconomics
----------------------------------------------------------------------------------------------------------------
Housing impacts............................ 2 SMALL, MODERATE, OR LARGE. Housing impacts are
expected to be of small significance at plants
located in a medium or high population area and not
in an area where growth control measures that limit
housing development are in effect. Moderate or large
housing impacts of the workforce associated with
refurbishment may be associated with plants located
in sparsely populated areas or in areas with growth
control measures that limit housing development. See
Sec. 51.53(c)(3)(ii)(I).
Public services: public safety, social 1 SMALL. Impacts to public safety, social services, and
services, and tourism and recreation. tourism and recreation are expected to be of small
significance at all sites.
Public services: public utilities.......... 2 SMALL OR MODERATE. An increased problem with water
shortages at some sites may lead to impacts of
moderate significance on public water supply
availability. See Sec. 51.53(c)(3)(ii)(I).
Public services, education (refurbishment). 2 SMALL, MODERATE, OR LARGE. Most sites would experience
impacts of small significance but larger impacts are
possible depending on site- and project-specific
factors. See Sec. 51.53(c)(3)(ii)(I).
Public services, education (license renewal 1 SMALL. Only impacts of small significance are
term). expected.
[[Page 65]]
Offsite land use (refurbishment)........... 2 SMALL OR MODERATE. Impacts may be of moderate
significance at plants in low population areas. See
Sec. 51.53(c)(3)(ii)(I).
Offsite land use (license renewal term).... 2 SMALL, MODERATE, OR LARGE. Significant changes in land
use may be associated with population and tax revenue
changes resulting from license renewal. See Sec.
51.53(c)(3)(ii)(I).
Public services, Transportation............ 2 SMALL, MODERATE, OR LARGE. Transportation impacts
(level of service) of highway traffic generated
during plant refurbishment and during the term of the
renewed license are generally expected to be of small
significance. However, the increase in traffic
associated with additional workers and the local road
and traffic control conditions may lead to impacts of
moderate or large significance at some sites. See
Sec. 51.53(c)(3)(ii)(J).
Historic and archaeological resources...... 2 SMALL, MODERATE, OR LARGE. Generally, plant
refurbishment and continued operation are expected to
have no more than small adverse impacts on historic
and archaeological resources. However, the National
Historic Preservation Act requires the Federal agency
to consult with the State Historic Preservation
Officer to determine whether there are properties
present that require protection. See Sec.
51.53(c)(3)(ii)(K).
Aesthetic impacts (refurbishment).......... 1 SMALL. No significant impacts are expected during
refurbishment.
Aesthetic impacts (license renewal term)... 1 SMALL. No significant impacts are expected during the
license renewal term.
Aesthetic impacts of transmission lines 1 SMALL. No significant impacts are expected during the
(license renewal term). license renewal term.
----------------------------------------------------------------------------------------------------------------
Postulated Accidents
----------------------------------------------------------------------------------------------------------------
Design basis accidents..................... 1 SMALL. The NRC staff has concluded that the
environmental impacts of design basis accidents are
of small significance for all plants.
Severe accidents........................... 2 SMALL. The probability weighted consequences of
atmospheric releases, fallout onto open bodies of
water, releases to ground water, and societal and
economic impacts from severe accidents are small for
all plants. However, alternatives to mitigate severe
accidents must be considered for all plants that have
not considered such alternatives. See Sec.
51.53(c)(3)(ii)(L).
----------------------------------------------------------------------------------------------------------------
Uranium Fuel Cycle and Waste Management
----------------------------------------------------------------------------------------------------------------
Offsite radiological impacts (individual 1 SMALL. Off-site impacts of the uranium fuel cycle have
effects from other than the disposal of been considered by the Commission in Table S-3 of
spent fuel and high level waste). this part. Based on information in the GEIS, impacts
on individuals from radioactive gaseous and liquid
releases including radon-222 and technetium-99 are
small.
Offsite radiological impacts (collective 1 The 100 year environmental dose commitment to the U.S.
effects). population from the fuel cycle, high level waste and
spent fuel disposal excepted, is calculated to be
about 14,800 person rem, or 12 cancer fatalities, for
each additional 20-year power reactor operating term.
Much of this, especially the contribution of radon
releases from mines and tailing piles, consists of
tiny doses summed over large populations. This same
dose calculation can theoretically be extended to
include many tiny doses over additional thousands of
years as well as doses outside the U. S. The result
of such a calculation would be thousands of cancer
fatalities from the fuel cycle, but this result
assumes that even tiny doses have some statistical
adverse health effect which will not ever be
mitigated (for example no cancer cure in the next
thousand years), and that these doses projected over
thousands of years are meaningful. However, these
assumptions are questionable. In particular, science
cannot rule out the possibility that there will be no
cancer fatalities from these tiny doses. For
perspective, the doses are very small fractions of
regulatory limits, and even smaller fractions of
natural background exposure to the same populations.
Nevertheless, despite all the uncertainty, some
judgement as to the regulatory NEPA implications of
these matters should be made and it makes no sense to
repeat the same judgement in every case. Even taking
the uncertainties into account, the Commission
concludes that these impacts are acceptable in that
these impacts would not be sufficiently large to
require the NEPA conclusion, for any plant, that the
option of extended operation under 10 CFR Part 54
should be eliminated. Accordingly, while the
Commission has not assigned a single level of
significance for the collective effects of the fuel
cycle, this issue is considered Category 1.
[[Page 66]]
Offsite radiological impacts (spent fuel 1 For the high level waste and spent fuel disposal
and high level waste disposal). component of the fuel cycle, there are no current
regulatory limits for offsite releases of
radionuclides for the current candidate repository
site. However, if we assume that limits are developed
along the lines of the 1995 National Academy of
Sciences (NAS) report, ``Technical Bases for Yucca
Mountain Standards,'' and that in accordance with the
Commission's Waste Confidence Decision, 10 CFR 51.23,
a repository can and likely will be developed at some
site which will comply with such limits, peak doses
to virtually all individuals will be 100 millirem per
year or less. However, while the Commission has
reasonable confidence that these assumptions will
prove correct, there is considerable uncertainty
since the limits are yet to be developed, no
repository application has been completed or
reviewed, and uncertainty is inherent in the models
used to evaluate possible pathways to the human
environment. The NAS report indicated that 100
millirem per year should be considered as a starting
point for limits for individual doses, but notes that
some measure of consensus exists among national and
international bodies that the limits should be a
fraction of the 100 millirem per year. The lifetime
individual risk from 100 millirem annual dose limit
is about 3 x 10-3.
Estimating cumulative doses to populations over
thousands of years is more problematic. The
likelihood and consequences of events that could
seriously compromise the integrity of a deep geologic
repository were evaluated by the Department of Energy
in the ``Final Environmental Impact Statement:
Management of Commercially Generated Radioactive
Waste,'' October 1980. The evaluation estimated the
70-year whole-body dose commitment to the maximum
individual and to the regional population resulting
from several modes of breaching a reference
repository in the year of closure, after 1,000 years,
after 100,000 years, and after 100,000,000 years.
Subsequently, the NRC and other federal agencies have
expended considerable effort to develop models for
the design and for the licensing of a high level
waste repository, especially for the candidate
repository at Yucca Mountain. More meaningful
estimates of doses to population may be possible in
the future as more is understood about the
performance of the proposed Yucca Mountain
repository. Such estimates would involve very great
uncertainty, especially with respect to cumulative
population doses over thousands of years. The
standard proposed by the NAS is a limit on maximum
individual dose. The relationship of potential new
regulatory requirements, based on the NAS report, and
cumulative population impacts has not been
determined, although the report articulates the view
that protection of individuals will adequately
protect the population for a repository at Yucca
Mountain. However, EPA's generic repository standards
in 40 CFR part 191 generally provide an indication of
the order of magnitude of cumulative risk to
population that could result from the licensing of a
Yucca Mountain repository, assuming the ultimate
standards will be within the range of standards now
under consideration. The standards in 40 CFR part 191
protect the population by imposing ``containment
requirements'' that limit the cumulative amount of
radioactive material released over 10,000 years.
Reporting performance standards that will be required
by EPA are expected to result in releases and
associated health consequences in the range between
10 and 100 premature cancer deaths with an upper
limit of 1,000 premature cancer deaths world-wide for
a 100,000 metric tonne (MTHM) repository.
Nevertheless, despite all the uncertainty, some
judgement as to the regulatory NEPA implications of
these matters should be made and it makes no sense to
repeat the same judgement in every case. Even taking
the uncertainties into account, the Commission
concludes that these impacts are acceptable in that
these impacts would not be sufficiently large to
require the NEPA conclusion, for any plant, that the
option of extended operation under 10 CFR part 54
should be eliminated. Accordingly, while the
Commission has not assigned a single level of
significance for the impacts of spent fuel and high
level waste disposal, this issue is considered
Category 1.
Nonradiological impacts of the uranium fuel 1 SMALL. The nonradiological impacts of the uranium fuel
cycle. cycle resulting from the renewal of an operating
license for any plant are found to be small.
[[Page 67]]
Low-level waste storage and disposal....... 1 SMALL. The comprehensive regulatory controls that are
in place and the low public doses being achieved at
reactors ensure that the radiological impacts to the
environment will remain small during the term of a
renewed license. The maximum additional on-site land
that may be required for low-level waste storage
during the term of a renewed license and associated
impacts will be small. Nonradiological impacts on air
and water will be negligible. The radiological and
nonradiological environmental impacts of long-term
disposal of low-level waste from any individual plant
at licensed sites are small. In addition, the
Commission concludes that there is reasonable
assurance that sufficient low-level waste disposal
capacity will be made available when needed for
facilities to be decommissioned consistent with NRC
decommissioning requirements.
Mixed waste storage and disposal........... 1 SMALL. The comprehensive regulatory controls and the
facilities and procedures that are in place ensure
proper handling and storage, as well as negligible
doses and exposure to toxic materials for the public
and the environment at all plants. License renewal
will not increase the small, continuing risk to human
health and the environment posed by mixed waste at
all plants. The radiological and nonradiological
environmental impacts of long-term disposal of mixed
waste from any individual plant at licensed sites are
small. In addition, the Commission concludes that
there is reasonable assurance that sufficient mixed
waste disposal capacity will be made available when
needed for facilities to be decommissioned consistent
with NRC decommissioning requirements.
On-site spent fuel......................... 1 SMALL. The expected increase in the volume of spent
fuel from an additional 20 years of operation can be
safely accommodated on site with small environmental
effects through dry or pool storage at all plants if
a permanent repository or monitored retrievable
storage is not available.
Nonradiological waste...................... 1 SMALL. No changes to generating systems are
anticipated for license renewal. Facilities and
procedures are in place to ensure continued proper
handling and disposal at all plants.
Transportation............................. 1 SMALL. The impacts of transporting spent fuel enriched
up to 5 percent uranium-235 with average burnup for
the peak rod to current levels approved by NRC up to
62,000 MWd/MTU and the cumulative impacts of
transporting high-level waste to a single repository,
such as Yucca Mountain, Nevada are found to be
consistent with the impact values contained in 10 CFR
51.52(c), Summary Table S-4--Environmental Impact of
Transportation of Fuel and Waste to and from One
Light-Water-Cooled Nuclear Power Reactor. If fuel
enrichment or burnup conditions are not met, the
applicant must submit an assessment of the
implications for the environmental impact values
reported in Sec. 51.52.
----------------------------------------------------------------------------------------------------------------
Decommissioning
----------------------------------------------------------------------------------------------------------------
Radiation doses............................ 1 SMALL. Doses to the public will be well below
applicable regulatory standards regardless of which
decommissioning method is used. Occupational doses
would increase no more than 1 man-rem caused by
buildup of long-lived radionuclides during the
license renewal term.
Waste management........................... 1 SMALL. Decommissioning at the end of a 20-year license
renewal period would generate no more solid wastes
than at the end of the current license term. No
increase in the quantities of Class C or greater than
Class C wastes would be expected.
Air quality................................ 1 SMALL. Air quality impacts of decommissioning are
expected to be negligible either at the end of the
current operating term or at the end of the license
renewal term.
Water quality.............................. 1 SMALL. The potential for significant water quality
impacts from erosion or spills is no greater whether
decommissioning occurs after a 20-year license
renewal period or after the original 40-year
operation period, and measures are readily available
to avoid such impacts.
Ecological resources....................... 1 SMALL. Decommissioning after either the initial
operating period or after a 20-year license renewal
period is not expected to have any direct ecological
impacts.
Socioeconomic impacts...................... 1 SMALL. Decommissioning would have some short-term
socioeconomic impacts. The impacts would not be
increased by delaying decommissioning until the end
of a 20-year relicense period, but they might be
decreased by population and economic growth.
----------------------------------------------------------------------------------------------------------------
[[Page 68]]
Environmental Justice
----------------------------------------------------------------------------------------------------------------
Environmental justice \6\.................. \4\ NA NONE. The need for and the content of an analysis of
environmental justice will be addressed in plant-
specific reviews. \6\
----------------------------------------------------------------------------------------------------------------
\1\ Data supporting this table are contained in NUREG-1437, ``Generic Environmental Impact Statement for License
Renewal of Nuclear Plants'' (May 1996) and NUREG-1437, Vol. 1, Addendum 1, ``Generic Environmental Impact
Statement for License Renewal of Nuclear Plants: Main Report Section 6.3--`Transportation,' Table 9.1 `Summary
of findings on NEPA issues for license renewal of nuclear power plants,' Final Report'' (August 1999).
\2\ The numerical entries in this column are based on the following category definitions:
Category 1: For the issue, the analysis reported in the Generic Environmental Impact Statement has shown:
(1) The environmental impacts associated with the issue have been determined to apply either to all plants or,
for some issues, to plants having a specific type of cooling system or other specified plant or site
characteristic;
(2) A single significance level (i.e., small, moderate, or large) has been assigned to the impacts (except for
collective off site radiological impacts from the fuel cycle and from high level waste and spent fuel
disposal); and
(3) Mitigation of adverse impacts associated with the issue has been considered in the analysis, and it has been
determined that additional plant-specific mitigation measures are likely not to be sufficiently beneficial to
warrant implementation.
The generic analysis of the issue may be adopted in each plant-specific review.
Category 2: For the issue, the analysis reported in the Generic Environmental Impact Statement has shown that
one or more of the criteria of Category 1 cannot be met, and therefore additional plant-specific review is
required.
\3\ The impact findings in this column are based on the definitions of three significance levels. Unless the
significance level is identified as beneficial, the impact is adverse, or in the case of ``small,'' may be
negligible. The definitions of significance follow:
SMALL--For the issue, environmental effects are not detectable or are so minor that they will neither
destabilize nor noticeably alter any important attribute of the resource. For the purposes of assessing
radiological impacts, the Commission has concluded that those impacts that do not exceed permissible levels in
the Commission's regulations are considered small as the term is used in this table.
MODERATE--For the issue, environmental effects are sufficient to alter noticeably, but not to destabilize,
important attributes of the resource.
LARGE--For the issue, environmental effects are clearly noticeable and are sufficient to destabilize important
attributes of the resource.
For issues where probability is a key consideration (i.e., accident consequences), probability was a factor in
determining significance.
\4\ NA (not applicable). The categorization and impact finding definitions do not apply to these issues.
\5\ If, in the future, the Commission finds that, contrary to current indications, a consensus has been reached
by appropriate Federal health agencies that there are adverse health effects from electromagnetic fields, the
Commission will require applicants to submit plant-specific reviews of these health effects as part of their
license renewal applications. Until such time, applicants for license renewal are not required to submit
information on this issue.
\6\ Environmental Justice was not addressed in NUREG-1437, ``Generic Environmental Impact Statement for License
Renewal of Nuclear Plants,'' because guidance for implementing Executive Order 12898 issued on February 11,
1994, was not available prior to completion of NUREG-1437. This issue will be addressed in individual license
renewal reviews.
[61 FR 66546, Dec. 18, 1996, as amended at 62 FR 59276, Nov. 3, 1997; 64
FR 48507, Sept. 3, 1999; 66 FR 39278, July 30, 2001]
Subpart B [Reserved]
PART 52_LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER
PLANTS--Table of Contents
General Provisions
Sec.
52.0 Scope; applicability of 10 CFR Chapter I provisions.
52.1 Definitions.
52.2 Interpretations.
52.3 Written communications.
52.4 Deliberate misconduct.
52.5 Employee protection.
52.6 Completeness and accuracy of information.
52.7 Specific exemptions.
52.8 Combining licenses; elimination of repetition.
52.9 Jurisdictional limits.
52.10 Attacks and destructive acts.
52.11 Information collection requirements: OMB approval.
Subpart A_Early Site Permits
52.12 Scope of subpart.
52.13 Relationship to other subparts.
52.15 Filing of applications.
52.16 Contents of applications; general information.
52.17 Contents of applications; technical information.
52.18 Standards for review of applications.
52.21 Administrative review of applications; hearings.
52.23 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
52.24 Issuance of early site permit.
52.25 Extent of activities permitted.
52.26 Duration of permit.
52.27 Limited work authorization after issuance of early site permit.
52.28 Transfer of early site permit.
52.29 Application for renewal.
52.31 Criteria for renewal.
52.33 Duration of renewal.
52.35 Use of site for other purposes.
52.39 Finality of early site permit determinations.
[[Page 69]]
Subpart B_Standard Design Certifications
52.41 Scope of subpart.
52.43 Relationship to other subparts.
52.45 Filing of applications.
52.46 Contents of applications; general information.
52.47 Contents of applications; technical information.
52.48 Standards for review of applications.
52.51 Administrative review of applications.
52.53 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
52.54 Issuance of standard design certification.
52.55 Duration of certification.
52.57 Application for renewal.
52.59 Criteria for renewal.
52.61 Duration of renewal.
52.63 Finality of standard design certifications.
Subpart C_Combined Licenses
52.71 Scope of subpart.
52.73 Relationship to other subparts.
52.75 Filing of applications.
52.77 Contents of applications; general information.
52.79 Contents of applications; technical information in final safety
analysis report.
52.80 Contents of applications; additional technical information.
52.81 Standards for review of applications.
52.83 Finality of referenced NRC approvals; partial initial decision on
site suitability.
52.85 Administrative review of applications; hearings.
52.87 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
52.89 [Reserved]
52.91 Authorization to conduct limited work authorization activities.
52.93 Exemptions and variances.
52.97 Issuance of combined licenses.
52.98 Finality of combined licenses; information requests.
52.99 Inspection during construction.
52.103 Operation under a combined license.
52.104 Duration of combined license.
52.105 Transfer of combined license.
52.107 Application for renewal.
52.109 Continuation of combined license.
52.110 Termination of license.
Subpart D [Reserved]
Subpart E_Standard Design Approvals
52.131 Scope of subpart.
52.133 Relationship to other subparts.
52.135 Filing of applications.
52.136 Contents of applications; general information.
52.137 Contents of applications; technical information.
52.139 Standards for review of applications.
52.141 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
52.143 Staff approval of design.
52.145 Finality of standard design approvals; information requests.
52.147 Duration of design approval.
Subpart F_Manufacturing Licenses
52.151 Scope of subpart.
52.153 Relationship to other subparts.
52.155 Filing of applications.
52.156 Contents of applications; general information.
52.157 Contents of applications; technical information in final safety
analysis report.
52.158 Contents of application; additional technical information.
52.159 Standards for review of application.
52.161 [Reserved]
52.163 Administrative review of applications; hearings.
52.165 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
52.167 Issuance of manufacturing license.
52.169 [Reserved]
52.171 Finality of manufacturing licenses; information requests.
52.173 Duration of manufacturing license.
52.175 Transfer of manufacturing license.
52.177 Application for renewal.
52.179 Criteria for renewal.
52.181 Duration of renewal.
Subpart G [Reserved]
Subpart H_Enforcement
52.301 Violations.
52.303 Criminal penalties.
Appendix A to Part 52--Design Certification Rule for the U.S. Advanced
Boiling Water Reactor
Appendix B to Part 52--Design Certification Rule for the System 80+
Design
Appendix C to Part 52--Design Certification Rule for the AP600 Design
Appendix D to Part 52--Design Certification Rule for the AP1000 Design
Appendixes E-M to Part 52 [Reserved]
Appendix N to Part 52--Standardization of Nuclear Power Plant Designs:
Combined Licenses to Construct and Operate Nuclear Power
Reactors of Identical Design at Multiple Sites
Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936,
948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, 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, 5846);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy Policy Act of
2005, Pub. L. No. 109-58, 119 Stat. 594 (2005), secs. 147 and 149 of the
Atomic Energy Act.
[[Page 70]]
Source: 72 FR 49517, Aug. 28, 2007, unless otherwise noted.
General Provisions
Sec. 52.0 Scope; applicability of 10 CFR Chapter I provisions.
(a) This part governs the issuance of early site permits, standard
design certifications, combined licenses, standard design approvals, and
manufacturing licenses for nuclear power facilities licensed under
Section 103 of the Atomic Energy Act of 1954, as amended (68 Stat. 919),
and Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242).
This part also gives notice to all persons who knowingly provide to any
holder of or applicant for an approval, certification, permit, or
license, or to a contractor, subcontractor, or consultant of any of
them, components, equipment, materials, or other goods or services that
relate to the activities of a holder of or applicant for an approval,
certification, permit, or license, subject to this part, that they may
be individually subject to NRC enforcement action for violation of the
provisions in 10 CFR 52.4.
(b) Unless otherwise specifically provided for in this part, the
regulations in 10 CFR Chapter I apply to a holder of or applicant for an
approval, certification, permit, or license. A holder of or applicant
for an approval, certification, permit, or license issued under this
part shall comply with all requirements in 10 CFR Chapter I that are
applicable. A license, approval, certification, or permit issued under
this part is subject to all requirements in 10 CFR Chapter I which, by
their terms, are applicable to early site permits, design
certifications, combined licenses, design approvals, or manufacturing
licenses.
Sec. 52.1 Definitions.
(a) As used in this part--
Combined license means a combined construction permit and operating
license with conditions for a nuclear power facility issued under
subpart C of this part.
Decommission means to remove a facility or site safely from service
and reduce residual radioactivity to a level that permits--
(i) Release of the property for unrestricted use and termination of
the license; or
(ii) Release of the property under restricted conditions and
termination of the license.
Design characteristics are the actual features of a reactor or
reactors. Design characteristics are specified in a standard design
approval, a standard design certification, a combined license
application, or a manufacturing license.
Design parameters are the postulated features of a reactor or
reactors that could be built at a proposed site. Design parameters are
specified in an early site permit.
Early site permit means a Commission approval, issued under subpart
A of this part, for a site or sites for one or more nuclear power
facilities. An early site permit is a partial construction permit.
License means a license, including an early site permit, combined
license or manufacturing license under this part or a renewed license
issued by the Commission under this part or part 54 of this chapter.
Licensee means a person who is authorized to conduct activities
under a license issued by the Commission.
Limited work authorization means the authorization provided by the
Director of New Reactors or the Director of Nuclear Reactor Regulation
under Sec. 50.10 of this chapter.
Major feature of the emergency plans means an aspect of those plans
necessary to:
(i) Address in whole or part one or more of the 16 standards in 10
CFR 50.47(b); or
(ii) Describe the emergency planning zones as required in 10 CFR
50.33(g).
Manufacturing license means a license, issued under subpart F of
this part, authorizing the manufacture of nuclear power reactors but not
their construction, installation, or operation at the sites on which the
reactors are to be operated.
Modular design means a nuclear power station that consists of two or
more essentially identical nuclear reactors (modules) and each module is
a separate nuclear reactor capable of being operated independent of the
state
[[Page 71]]
of completion or operating condition of any other module co-located on
the same site, even though the nuclear power station may have some
shared or common systems.
Prototype plant means a nuclear power plant that is used to test new
safety features, such as the testing required under 10 CFR 50.43(e). The
prototype plant is similar to a first-of-a-kind or standard plant design
in all features and size, but may include additional safety features to
protect the public and the plant staff from the possible consequences of
accidents during the testing period.
Site characteristics are the actual physical, environmental and
demographic features of a site. Site characteristics are specified in an
early site permit or in a final safety analysis report for a combined
license.
Site parameters are the postulated physical, environmental and
demographic features of an assumed site. Site parameters are specified
in a standard design approval, standard design certification, or
manufacturing license.
Standard design means a design which is sufficiently detailed and
complete to support certification or approval in accordance with subpart
B or E of this part, and which is usable for a multiple number of units
or at a multiple number of sites without reopening or repeating the
review.
Standard design approval or design approval means an NRC staff
approval, issued under subpart E of this part, of a final standard
design for a nuclear power reactor of the type described in 10 CFR
50.22. The approval may be for either the final design for the entire
reactor facility or the final design of major portions thereof.
Standard design certification or design certification means a
Commission approval, issued under subpart B of this part, of a final
standard design for a nuclear power facility. This design may be
referred to as a certified standard design.
(b) All other terms in this part have the meaning set out in 10 CFR
50.2, or Section 11 of the Atomic Energy Act, as applicable.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57446, Oct. 9, 2007]
Sec. 52.2 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission other than a written
interpretation by the General Counsel will be recognized to be binding
upon the Commission.
Sec. 52.3 Written communications.
(a) General requirements. All correspondence, reports, applications,
and other written communications from an applicant, licensee, or holder
of a standard design approval to the Nuclear Regulatory Commission
concerning the regulations in this part, individual license conditions,
or the terms and conditions of an early site permit or standard design
approval, must be sent either by mail addressed: ATTN: Document Control
Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by
hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville,
Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; or,
where practicable, by electronic submission, for example, via Electronic
Information Exchange, e-mail, or CD-ROM. Electronic submissions must be
made in a manner that enables the NRC to receive, read, authenticate,
distribute, and archive the submission, and process and retrieve it a
single page at a time. Detailed guidance on making electronic
submissions can be obtained by visiting the NRC's Web site at http://
www.nrc.gov/site-help/e-submittals.html; by e-mail to
[email protected]; or by writing the Office of Information Services,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The
guidance discusses, among other topics, the formats the NRC can accept,
the use of electronic signatures, and the treatment of nonpublic
information. If the communication is on paper, the signed original must
be sent. If a submission due date falls on a Saturday, Sunday, or
Federal holiday, the next Federal working day becomes the official due
date.
(b) Distribution requirements. Copies of all correspondence,
reports, and other written communications concerning
[[Page 72]]
the regulations in this part or individual license conditions, or the
terms and conditions of an early site permit or standard design
approval, must be submitted to the persons listed in paragraph (b)(1) of
this section (addresses for the NRC Regional Offices are listed in
appendix D to part 20 of this chapter).
(1) Applications for amendment of permits and licenses; reports; and
other communications. All written communications (including responses
to: generic letters, bulletins, information notices, regulatory
information summaries, inspection reports, and miscellaneous requests
for additional information) that are required of holders of early site
permits, standard design approvals, combined licenses, or manufacturing
licenses issued under this part must be submitted as follows, except as
otherwise specified in paragraphs (b)(2) through (b)(7) of this section:
to the NRC's Document Control Desk (if on paper, the signed original),
with a copy to the appropriate Regional Office, and a copy to the
appropriate NRC Resident Inspector, if one has been assigned to the site
of the facility or the place of manufacture of a reactor licensed under
subpart F of this part.
(2) Applications and amendments to applications. Applications for
early site permits, standard design approvals, combined licenses,
manufacturing licenses and amendments to any of these types of
applications must be submitted to the NRC's Document Control Desk, with
a copy to the appropriate Regional Office, and a copy to the appropriate
NRC Resident Inspector, if one has been assigned to the site of the
facility or the place of manufacture of a reactor licensed under subpart
F of this part, except as otherwise specified in paragraphs (b)(3)
through (b)(7) of this section. If the application or amendment is on
paper, the submission to the Document Control Desk must be the signed
original.
(3) Acceptance review application. Written communications required
for an application for determination of suitability for docketing must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(4) Security plan and related submissions. Written communications,
as defined in paragraphs (b)(4)(i) through (iv) of this section, must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(i) Physical security plan under Sec. 52.79 of this chapter;
(ii) Safeguards contingency plan under Sec. 52.79 of this chapter;
(iii) Change to security plan, guard training and qualification
plan, or safeguards contingency plan made without prior Commission
approval under Sec. 50.54(p) of this chapter;
(iv) Application for amendment of physical security plan, guard
training and qualification plan, or safeguards contingency plan under
Sec. 50.90 of this chapter.
(5) Emergency plan and related submissions. Written communications
as defined in paragraphs (b)(5)(i) through (iii) of this section must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original.
(i) Emergency plan under Sec. 52.17(b) or Sec. 52.79(a);
(ii) Change to an emergency plan under Sec. 50.54(q) of this
chapter;
(iii) Emergency implementing procedures under appendix E, Section V
of part 50 of this chapter.
(6) Updated FSAR. An updated final safety analysis report (FSAR) or
replacement pages under Sec. 50.71(e) of this chapter, or the
regulations in this part must be submitted to the NRC's Document Control
Desk, with a copy to the appropriate Regional Office, and a copy to the
appropriate NRC Resident Inspector if one has been assigned to the site
of the facility or the place of manufacture of a reactor licensed under
[[Page 73]]
subpart F of this part. Paper copy submissions may be made using
replacement pages; however, if a licensee chooses to use electronic
submission, all subsequent updates or submissions must be performed
electronically on a total replacement basis. If the communication is on
paper, the submission to the Document Control Desk must be the signed
original. If the communications are submitted electronically, see
Guidance for Electronic Submissions to the Commission.
(7) Quality assurance related submissions. (i) A change to the
safety analysis report quality assurance program description under Sec.
50.54(a)(3) or Sec. 50.55(f)(4) of this chapter, or a change to a
licensee's NRC-accepted quality assurance topical report under Sec.
50.54(a)(3) or Sec. 50.55(f)(4) of this chapter, must be submitted to
the NRC's Document Control Desk, with a copy to the appropriate Regional
Office, and a copy to the appropriate NRC Resident Inspector if one has
been assigned to the site of the facility. If the communication is on
paper, the submission to the Document Control Desk must be the signed
original.
(ii) A change to an NRC-accepted quality assurance topical report
from nonlicensees (i.e., architect/engineers, NSSS suppliers, fuel
suppliers, constructors, etc.) must be submitted to the NRC's Document
Control Desk. If the communication is on paper, the signed original must
be sent.
(8) Certification of permanent cessation of operations. The
licensee's certification of permanent cessation of operations under
Sec. 52.110(a)(1), must state the date on which operations have ceased
or will cease, and must be submitted to the NRC's Document Control Desk.
This submission must be under oath or affirmation.
(9) Certification of permanent fuel removal. The licensee's
certification of permanent fuel removal under Sec. 52.110(a)(1), must
state the date on which the fuel was removed from the reactor vessel and
the disposition of the fuel, and must be submitted to the NRC's Document
Control Desk. This submission must be under oath or affirmation.
(c) Form of communications. All paper copies submitted to meet the
requirements set forth in paragraph (b) of this section must be
typewritten, printed or otherwise reproduced in permanent form on
unglazed paper. Exceptions to these requirements imposed on paper
submissions may be granted for the submission of micrographic,
photographic, or similar forms.
(d) Regulation governing submission. Applicants, licensees, and
holders of standard design approvals submitting correspondence, reports,
and other written communications under the regulations of this part are
requested but not required to cite whenever practical, in the upper
right corner of the first page of the submission, the specific
regulation or other basis requiring submission.
[72 FR 49517, Aug. 28, 2007, as amended at 74 FR 62682, Dec. 1, 2009]
Sec. 52.4 Deliberate misconduct.
(a) Applicability. This section applies to any:
(1) Licensee;
(2) Holder of a standard design approval;
(3) Applicant for a standard design certification;
(4) Applicant for a license or permit;
(5) Applicant for a standard design approval;
(6) Employee of a licensee;
(7) Employee of an applicant for a license, a standard design
certification, or a standard design approval;
(8) Any contractor (including a supplier or consultant),
subcontractor, or employee of a contractor or subcontractor of any
licensee; or
(9) Any contractor (including a supplier or consultant),
subcontractor, or employee of a contractor or subcontractor of any
applicant for a license, a standard design certification, or a standard
design approval.
(b) Definitions. For purposes of this section:
Deliberate misconduct means an intentional act or omission that a
person or entity knows:
(i) Would cause a licensee or an applicant for a license, standard
design certification, or standard design approval to be in violation of
any rule, regulation, or order; or any term, condition, or limitation,
of any license, standard
[[Page 74]]
design certification, or standard design approval; or
(ii) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee, holder
of a standard design approval, applicant for a license, standard design
certification, or standard design approval, or contractor, or
subcontractor.
(c) Prohibition against deliberate misconduct. Any person or entity
subject to this section, who knowingly provides to any licensee, any
applicant for a license, standard design certification or standard
design approval, or a contractor, or subcontractor of a person or entity
subject to this section, any components, equipment, materials, or other
goods or services that relate to a licensee's or applicant's activities
under this part, may not:
(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee, holder of a standard design
approval, or applicant to be in violation of any rule, regulation, or
order; or any term, condition, or limitation of any license issued by
the Commission, any standard design approval, or standard design
certification; or
(2) Deliberately submit to the NRC; a licensee, an applicant for a
license, standard design certification or standard design approval; or a
licensee's, standard design approval holder's, or applicant's contractor
or subcontractor, information that the person submitting the information
knows to be incomplete or inaccurate in some respect material to the
NRC.
(d) A person or entity who violates paragraph (c)(1) or (c)(2) of
this section may be subject to enforcement action in accordance with the
procedures in 10 CFR part 2, subpart B.
Sec. 52.5 Employee protection.
(a) Discrimination by a Commission licensee, holder of a standard
design approval, an applicant for a license, standard design
certification, or standard design approval, a contractor or
subcontractor of a Commission licensee, holder of a standard design
approval, applicant for a license, standard design certification, or
standard design approval, against an employee for engaging in certain
protected activities is prohibited. Discrimination includes discharge
and other actions that relate to compensation, terms, conditions, or
privileges of employment. The protected activities are established in
Section 211 of the Energy Reorganization Act of 1974, as amended, and in
general are related to the administration or enforcement of a
requirement imposed under the Atomic Energy Act or the Energy
Reorganization Act.
(1) The protected activities include but are not limited to:
(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in the
introductory text of paragraph (a) of this section or possible
violations of requirements imposed under either of those statutes;
(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in the introductory text of paragraph (a) of this
section or under these requirements if the employee has identified the
alleged illegality to the employer;
(iii) Requesting the Commission to institute action against his or
her employer for the administration or enforcement of these
requirements;
(iv) Testifying in any Commission proceeding, or before Congress, or
at any Federal or State proceeding regarding any provision (or proposed
provision) of either of the statutes named in the introductory text of
paragraph (a) of this section; and
(v) Assisting or participating in, or is about to assist or
participate in, these activities.
(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee assistance or
participation.
(3) This section has no application to any employee alleging
discrimination prohibited by this section who, acting without direction
from his or her employer (or the employer's agent), deliberately causes
a violation of any requirement of the Energy Reorganization Act of 1974,
as amended, or the Atomic Energy Act of 1954, as amended.
[[Page 75]]
(b) Any employee who believes that he or she has been discharged or
otherwise discriminated against by any person for engaging in protected
activities specified in paragraph (a)(1) of this section may seek a
remedy for the discharge or discrimination through an administrative
proceeding in the Department of Labor. The administrative proceeding
must be initiated within 180 days after an alleged violation occurs. The
employee may do this by filing a complaint alleging the violation with
the Department of Labor, Employment Standards Administration, Wage and
Hour Division. The Department of Labor may order reinstatement, back
pay, and compensatory damages.
(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, a holder of a standard design approval, an
applicant for a Commission license, standard design certification, or a
standard design approval, or a contractor or subcontractor of a
Commission licensee, holder of a standard design approval, or any
applicant may be grounds for--
(1) Denial, revocation, or suspension of the license or standard
design approval;
(2) Withdrawal or revocation of a proposed or final standard design
certification;
(3) Imposition of a civil penalty on the licensee, holder of a
standard design approval, or applicant (including an applicant for a
standard design certification under this part following Commission
adoption of final design certification rule) or a contractor or
subcontractor of the licensee, holder of a standard design approval, or
applicant.
(4) Other enforcement action.
(d) Actions taken by an employer, or others, which adversely affect
an employee may be predicated upon nondiscriminatory grounds. The
prohibition applies when the adverse action occurs because the employee
has engaged in protected activities. An employee's engagement in
protected activities does not automatically render him or her immune
from discharge or discipline for legitimate reasons or from adverse
action dictated by nonprohibited considerations.
(e)(1) Each licensee, each holder of a standard design approval, and
each applicant for a license, standard design certification, or standard
design approval, shall prominently post the revision of NRC Form 3,
``Notice to Employees,'' referenced in 10 CFR 19.11(e). This form must
be posted at locations sufficient to permit employees protected by this
section to observe a copy on the way to or from their place of work.
Premises must be posted not later than thirty (30) days after an
application is docketed and remain posted while the application is
pending before the Commission, during the term of the license, standard
design certification, or standard design approval under 10 CFR part 52,
and for 30 days following license termination or the expiration or
termination of the standard design certification or standard design
approval under 10 CFR part 52.
(2) Copies of NRC Form 3 may be obtained by writing to the Regional
Administrator of the appropriate U.S. Nuclear Regulatory Commission
Regional Office listed in appendix D to part 20 of this chapter, by
calling (301) 415-7232, via e-mail to [email protected], or by
visiting the NRC's Web site at http://www.nrc.gov and selecting forms
from the index found on the NRC's home page.
(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a complaint
filed by an employee with the Department of Labor under Section 211 of
the Energy Reorganization Act of 1974, as amended, may contain any
provision which would prohibit, restrict, or otherwise discourage an
employee from participating in protected activity as defined in
paragraph (a)(1) of this section including, but not limited to,
providing information to the NRC or to his or her employer on potential
violations or other matters within NRC's regulatory responsibilities.
(g) Part 19 of this chapter sets forth requirements and regulatory
provisions applicable to licensees, holders of a standard design
approval, applicants for a license, standard design certification, or
standard design approval, and contractors or subcontractors of a
Commission licensee, or holder of a
[[Page 76]]
standard design approval, and are in addition to the requirements in
this section.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 63974, Nov. 14, 2007;
73 FR 30458, May 28, 2008]
Sec. 52.6 Completeness and accuracy of information.
(a) Information provided to the Commission by a licensee (including
an early site permit holder, a combined license holder, and a
manufacturing license holder), a holder of a standard design approval
under this part, and an applicant for a license or an applicant for a
standard design certification or a standard design approval under this
part, and information required by statute or by the Commission's
regulations, orders, license conditions, or terms and conditions of a
standard design approval to be maintained by the licensee, the holder of
a standard design approval under this part, the applicant for a standard
design certification under this part following Commission adoption of a
final design certification rule, and an applicant for a license, a
standard design certification, or a standard design approval under this
part shall be complete and accurate in all material respects.
(b) Each applicant or licensee, each holder of a standard design
approval under this part, and each applicant for a standard design
certification under this part following Commission adoption of a final
design certification regulation, shall notify the Commission of
information identified by the applicant or the licensee as having for
the regulated activity a significant implication for public health and
safety or common defense and security. An applicant, licensee, or holder
violates this paragraph only if the applicant, licensee, or holder fails
to notify the Commission of information that the applicant, licensee, or
holder has been identified as having a significant implication for
public health and safety or common defense and security. Notification
shall be provided to the Administrator of the appropriate Regional
Office within 2 working days of identifying the information. This
requirement is not applicable to information which is already required
to be provided to the Commission by other reporting or updating
requirements.
Sec. 52.7 Specific exemptions.
The Commission may, upon application by any interested person or
upon its own initiative, grant exemptions from the requirements of the
regulations of this part. The Commission's consideration will be
governed by Sec. 50.12 of this chapter, unless other criteria are
provided for in this part, in which case the Commission's consideration
will be governed by the criteria in this part. Only if those criteria
are not met will the Commission's consideration be governed by Sec.
50.12 of this chapter. The Commission's consideration of requests for
exemptions from requirements of the regulations of other parts in this
chapter, which are applicable by virtue of this part, shall be governed
by the exemption requirements of those parts.
Sec. 52.8 Combining licenses; elimination of repetition.
(a) An applicant for a license under this part may combine in its
application several applications for different kinds of licenses under
the regulations of this chapter.
(b) An applicant may incorporate by reference in its application
information contained in previous applications, statements or reports
filed with the Commission, provided, however, that such references are
clear and specific.
(c) The Commission may combine in a single license the activities of
an applicant which would otherwise be licensed separately.
Sec. 52.9 Jurisdictional limits.
No permit, license, standard design approval, or standard design
certification under this part shall be deemed to have been issued for
activities which are not under or within the jurisdiction of the United
States.
Sec. 52.10 Attacks and destructive acts.
Neither an applicant for a license to manufacture, construct, and
operate a utilization facility under this part, nor for an amendment to
this license, or an applicant for an early site permit, a
[[Page 77]]
standard design certification, or standard design approval under this
part, or for an amendment to the early site permit, standard design
certification, or standard design approval, is required to provide for
design features or other measures for the specific purpose of protection
against the effects of--
(a) Attacks and destructive acts, including sabotage, directed
against the facility by an enemy of the United States, whether a foreign
government or other person; or
(b) Use or deployment of weapons incident to U.S. defense
activities.
Sec. 52.11 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number. OMB
has approved the information collection requirements contained in this
part under Control Number 3150-0151.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 52.7, 52.15, 52.16, 52.17, 52.29, 52.35,
52.39, 52.45, 52.46, 52.47, 52.57, 52.63, 52.75, 52.77, 52.79, 52.80,
52.93, 52.99, 52.110, 52.135, 52.136, 52.137, 52.155, 52.156, 52.157,
52.158, 52.171, 52.177, and appendices A, B, C, D, and N of part 52.
Subpart A_Early Site Permits
Sec. 52.12 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of an early site permit for approval of a site for
one or more nuclear power facilities separate from the filing of an
application for a construction permit or combined license for the
facility.
Sec. 52.13 Relationship to other subparts.
This subpart applies when any person who may apply for a
construction permit under 10 CFR part 50, or for a combined license
under this part seeks an early site permit from the Commission
separately from an application for a construction permit or a combined
license.
Sec. 52.15 Filing of applications.
(a) Any person who may apply for a construction permit under 10 CFR
part 50, or for a combined license under this part, may file an
application for an early site permit with the Director, Office of New
Reactors, or the Director, Office of Nuclear Reactor Regulation, as
appropriate. An application for an early site permit may be filed
notwithstanding the fact that an application for a construction permit
or a combined license has not been filed in connection with the site for
which a permit is sought.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of an application
for the initial issuance or renewal of an early site permit are set
forth in 10 CFR part 170.
Sec. 52.16 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d) and (j) of this chapter.
Sec. 52.17 Contents of applications; technical information.
(a) For applications submitted before September 27, 2007, the rule
provisions in effect at the date of docketing apply unless otherwise
requested by the applicant in writing. The application must contain:
(1) A site safety analysis report. The site safety analysis report
shall include the following:
(i) The specific number, type, and thermal power level of the
facilities, or range of possible facilities, for which the site may be
used;
(ii) The anticipated maximum levels of radiological and thermal
effluents each facility will produce;
(iii) The type of cooling systems, intakes, and outflows that may be
associated with each facility;
(iv) The boundaries of the site;
[[Page 78]]
(v) The proposed general location of each facility on the site;
(vi) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site with appropriate consideration of
the most severe of the natural phenomena that have been historically
reported for the site and surrounding area and with sufficient margin
for the limited accuracy, quantity, and period of time in which the
historical data have been accumulated;
(vii) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(viii) The existing and projected future population profile of the
area surrounding the site;
(ix) A description and safety assessment of the site on which a
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site under
the radiological consequence evaluation factors identified in paragraphs
(a)(1)(ix)(A) and (a)(1)(ix)(B) of this section. In performing this
assessment, an applicant shall assume a fission product release \1\ from
the core into the containment assuming that the facility is operated at
the ultimate power level contemplated. The applicant shall perform an
evaluation and analysis of the postulated fission product release, using
the expected demonstrable containment leak rate and any fission product
cleanup systems intended to mitigate the consequences of the accidents,
together with applicable site characteristics, including site
meteorology, to evaluate the offsite radiological consequences. Site
characteristics must comply with part 100 of this chapter. The
evaluation must determine that:
---------------------------------------------------------------------------
\1\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. Such accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(A) An individual located at any point on the boundary of the
exclusion area for any 2 hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \2\ total effective dose equivalent (TEDE).
---------------------------------------------------------------------------
\2\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose for
radiation workers which, according to NCRP recommendations at the time
could be disregarded in the determination of their radiation exposure
status (see NBS Handbook 69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an acceptable limit for
an emergency dose to the public under accident conditions. Rather, this
dose value has been set forth in this section as a reference value,
which can be used in the evaluation of plant design features with
respect to postulated reactor accidents, to assure that these designs
provide assurance of low risk of public exposure to radiation, in the
event of an accidents.
---------------------------------------------------------------------------
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period of
its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(x) Information demonstrating that site characteristics are such
that adequate security plans and measures can be developed;
(xi) For applications submitted after September 27, 2007, a
description of the quality assurance program applied to site-related
activities for the future design, fabrication, construction, and testing
of the structures, systems, and components of a facility or facilities
that may be constructed on the site. Appendix B to 10 CFR part 50 sets
forth the requirements for quality assurance programs for nuclear power
plants. The description of the quality assurance program for a nuclear
power plant site shall include a discussion of how the applicable
requirements of appendix B to part 50 of this chapter will be satisfied;
and
(xii) An evaluation of the site against applicable sections of the
Standard Review Plan (SRP) revision in effect 6 months before the docket
date of the application. The evaluation required by this section shall
include an identification and description of all differences
[[Page 79]]
in analytical techniques and procedural measures proposed for a site and
those corresponding techniques and measures given in the SRP acceptance
criteria. Where such a difference exists, the evaluation shall discuss
how the proposed alternative provides an acceptable method of complying
with the Commission's regulations, or portions thereof, that underlie
the corresponding SRP acceptance criteria. The SRP is not a substitute
for the regulations, and compliance is not a requirement.
(2) A complete environmental report as required by 10 CFR 51.50(b).
(b)(1) The site safety analysis report must identify physical
characteristics of the proposed site, such as egress limitations from
the area surrounding the site, that could pose a significant impediment
to the development of emergency plans. If physical characteristics are
identified that could pose a significant impediment to the development
of emergency plans, the application must identify measures that would,
when implemented, mitigate or eliminate the significant impediment.
(2) The site safety analysis report may also:
(i) Propose major features of the emergency plans, in accordance
with the pertinent standards of 10 CFR 50.47, and the requirements of
appendix E to 10 CFR part 50, such as the exact size and configuration
of the emergency planning zones, for review and approval by NRC, in
consultation with the Department of Homeland Security (DHS) in the
absence of complete and integrated emergency plans; or
(ii) Propose complete and integrated emergency plans for review and
approval by the NRC, in consultation with DHS, in accordance with the
applicable standards of 10 CFR 50.47, and the requirements of appendix E
to 10 CFR part 50. To the extent approval of emergency plans is sought,
the application must contain the information required by Sec. Sec.
50.33(g) and (j) of this chapter.
(3) Emergency plans submitted under paragraph (b)(2)(ii) of this
section must include the proposed inspections, tests, and analyses that
the holder of a combined license referencing the early site permit shall
perform, and the acceptance criteria that are necessary and sufficient
to provide reasonable assurance that, if the inspections, tests, and
analyses are performed and the acceptance criteria met, the facility has
been constructed and will be operated in conformity with the emergency
plans, the provisions of the Act, and the Commission's rules and
regulations. Major features of an emergency plan submitted under
paragraph (b)(2)(i) of this section may include proposed inspections,
tests, analyses, and acceptance criteria.
(4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the site
safety analysis report must include a description of contacts and
arrangements made with Federal, State, and local governmental agencies
with emergency planning responsibilities. The site safety analysis
report must contain any certifications that have been obtained. If these
certifications cannot be obtained, the site safety analysis report must
contain information, including a utility plan, sufficient to show that
the proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site. Under the option set forth in paragraph (b)(2)(ii) of this
section, the applicant shall make good faith efforts to obtain from the
same governmental agencies certifications that:
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations,
and
(iii) That these agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(c) An applicant may request that a limited work authorization under
10 CFR 50.10 be issued in conjunction with the early site permit. The
application must include the information otherwise required by 10 CFR
50.10(d)(3). Applications submitted before, and pending as of November
8, 2007, must include the information required by Sec. 52.17(c)
effective on the date of docketing.
(d) Each applicant for an early site permit under this part shall
protect
[[Page 80]]
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007; 73
FR 63571, Oct. 24, 2008]
Sec. 52.18 Standards for review of applications.
Applications filed under this subpart will be reviewed according to
the applicable standards set out in 10 CFR part 50 and its appendices
and 10 CFR part 100. In addition, the Commission shall prepare an
environmental impact statement during review of the application, in
accordance with the applicable provisions of 10 CFR part 51. The
Commission shall determine, after consultation with DHS, whether the
information required of the applicant by Sec. 52.17(b)(1) shows that
there is no significant impediment to the development of emergency plans
that cannot be mitigated or eliminated by measures proposed by the
applicant, whether any major features of emergency plans submitted by
the applicant under Sec. 52.17(b)(2)(i) are acceptable in accordance
with the applicable standards of 10 CFR 50.47 and the requirements of
appendix E to 10 CFR part 50, and whether any emergency plans submitted
by the applicant under Sec. 52.17(b)(2)(ii) provide reasonable
assurance that adequate protective measures can and will be taken in the
event of a radiological emergency.
Sec. 52.21 Administrative review of applications; hearings.
An early site permit is subject to all procedural requirements in 10
CFR part 2, including the requirements for docketing in Sec.
2.101(a)(1) through (4) of this chapter, and the requirements for
issuance of a notice of hearing in Sec. Sec. 2.104(a) and (d) of this
chapter, provided that the designated sections may not be construed to
require that the environmental report, or draft or final environmental
impact statement include an assessment of the benefits of construction
and operation of the reactor or reactors, or an analysis of alternative
energy sources. The presiding officer in an early site permit hearing
shall not admit contentions proffered by any party concerning an
assessment of the benefits of construction and operation of the reactor
or reactors, or an analysis of alternative energy sources if those
issues were not addressed by the applicant in the early site permit
application. All hearings conducted on applications for early site
permits filed under this part are governed by the procedures contained
in subparts C, G, L, and N of 10 CFR part 2, as applicable.
Sec. 52.23 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
The Commission shall refer a copy of the application for an early
site permit to the ACRS. The ACRS shall report on those portions of the
application which concern safety.
Sec. 52.24 Issuance of early site permit.
(a) After conducting a hearing under Sec. 52.21 and receiving the
report to be submitted by the ACRS under Sec. 52.23, the Commission may
issue an early site permit, in the form the Commission deems
appropriate, if the Commission finds that:
(1) An application for an early site permit meets the applicable
standards and requirements of the Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the site is in conformity
with the provisions of the Act, and the Commission's regulations;
(4) The applicant is technically qualified to engage in any
activities authorized;
(5) The proposed inspections, tests, analyses and acceptance
criteria, including any on emergency planning, are necessary and
sufficient, within the scope of the early site permit, to provide
reasonable assurance that the facility has been constructed and will be
operated in conformity with the license, the provisions of the Act, and
the Commission's regulations;
(6) Issuance of the permit will not be inimical to the common
defense and security or to the health and safety of the public;
[[Page 81]]
(7) Any significant adverse environmental impact resulting from
activities requested under Sec. 52.17(c) can be redressed; and
(8) The findings required by subpart A of 10 CFR part 51 have been
made.
(b) The early site permit must specify the site characteristics,
design parameters, and terms and conditions of the early site permit the
Commission deems appropriate. Before issuance of either a construction
permit or combined license referencing an early site permit, the
Commission shall find that any relevant terms and conditions of the
early site permit have been met. Any terms or conditions of the early
site permit that could not be met by the time of issuance of the
construction permit or combined license, must be set forth as terms or
conditions of the construction permit or combined license.
(c) The early site permit shall specify those 10 CFR 50.10
activities requested under Sec. 52.17(c) that the permit holder is
authorized to perform.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007]
Sec. 52.25 Extent of activities permitted.
If the activities authorized by Sec. 52.24(c) are performed and the
site is not referenced in an application for a construction permit or a
combined license issued under subpart C of this part while the permit
remains valid, then the early site permit remains in effect solely for
the purpose of site redress, and the holder of the permit shall redress
the site in accordance with the terms of the site redress plan required
by Sec. 52.17(c). If, before redress is complete, a use not envisaged
in the redress plan is found for the site or parts thereof, the holder
of the permit shall carry out the redress plan to the greatest extent
possible consistent with the alternate use.
Sec. 52.26 Duration of permit.
(a) Except as provided in paragraph (b) of this section, an early
site permit issued under this subpart may be valid for not less than 10,
nor more than 20 years from the date of issuance.
(b) An early site permit continues to be valid beyond the date of
expiration in any proceeding on a construction permit application or a
combined license application that references the early site permit and
is docketed before the date of expiration of the early site permit, or,
if a timely application for renewal of the permit has been docketed,
before the Commission has determined whether to renew the permit.
(c) An applicant for a construction permit or combined license may,
at its own risk, reference in its application a site for which an early
site permit application has been docketed but not granted.
(d) Upon issuance of a construction permit or combined license, a
referenced early site permit is subsumed, to the extent referenced, into
the construction permit or combined license.
[72 FR 49517, Aug. 28, 2007. Redesignated at 72 FR 57447, Oct. 9, 2007]
Sec. 52.27 Limited work authorization after issuance of early site permit.
A holder of an early site permit may request a limited work
authorization in accordance with Sec. 50.10 of this chapter.
[72 FR 57447, Oct. 9, 2007]
Sec. 52.28 Transfer of early site permit.
An application to transfer an early site permit will be processed
under 10 CFR 50.80.
Sec. 52.29 Application for renewal.
(a) Not less than 12, nor more than 36 months before the expiration
date stated in the early site permit, or any later renewal period, the
permit holder may apply for a renewal of the permit. An application for
renewal must contain all information necessary to bring up to date the
information and data contained in the previous application.
(b) Any person whose interests may be affected by renewal of the
permit may request a hearing on the application for renewal. The request
for a hearing must comply with 10 CFR 2.309. If a hearing is granted,
notice of the hearing will be published in accordance with 10 CFR 2.309.
(c) An early site permit, either original or renewed, for which a
timely application for renewal has been filed, remains in effect until
the Commission
[[Page 82]]
has determined whether to renew the permit. If the permit is not
renewed, it continues to be valid in certain proceedings in accordance
with the provisions of Sec. 52.27(b).
(d) The Commission shall refer a copy of the application for renewal
to the ACRS. The ACRS shall report on those portions of the application
which concern safety and shall apply the criteria set forth in Sec.
52.31.
Sec. 52.31 Criteria for renewal.
(a) The Commission shall grant the renewal if it determines that:
(1) The site complies with the Act, the Commission's regulations,
and orders applicable and in effect at the time the site permit was
originally issued; and
(2) Any new requirements the Commission may wish to impose are:
(i) Necessary for adequate protection to public health and safety or
common defense and security;
(ii) Necessary for compliance with the Commission's regulations, and
orders applicable and in effect at the time the site permit was
originally issued; or
(iii) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
(b) A denial of renewal for failure to comply with the provisions of
Sec. 52.31(a) does not bar the permit holder or another applicant from
filing a new application for the site which proposes changes to the site
or the way that it is used to correct the deficiencies cited in the
denial of the renewal.
Sec. 52.33 Duration of renewal.
Each renewal of an early site permit may be for not less than 10,
nor more than 20 years, plus any remaining years on the early site
permit then in effect before renewal.
Sec. 52.35 Use of site for other purposes.
A site for which an early site permit has been issued under this
subpart may be used for purposes other than those described in the
permit, including the location of other types of energy facilities. The
permit holder shall inform the Director, Office of New Reactors or
Director, Office of Nuclear Reactor Regulation, as appropriate,
(Director) of any significant uses for the site which have not been
approved in the early site permit. The information about the activities
must be given to the Director at least 30 days in advance of any actual
construction or site modification for the activities. The information
provided could be the basis for imposing new requirements on the permit,
in accordance with the provisions of Sec. 52.39. If the permit holder
informs the Director that the holder no longer intends to use the site
for a nuclear power plant, the Director may terminate the permit.
[73 FR 5724, Jan. 31, 2008]
Sec. 52.39 Finality of early site permit determinations.
(a) Commission finality. (1) Notwithstanding any provision in 10 CFR
50.109, while an early site permit is in effect under Sec. Sec. 52.27
or 52.33, the Commission may not change or impose new site
characteristics, design parameters, or terms and conditions, including
emergency planning requirements, on the early site permit unless the
Commission:
(i) Determines that a modification is necessary to bring the permit
or the site into compliance with the Commission's regulations and orders
applicable and in effect at the time the permit was issued;
(ii) Determines the modification is necessary to assure adequate
protection of the public health and safety or the common defense and
security;
(iii) Determines that a modification is necessary based on an update
under paragraph (b) of this section; or
(iv) Issues a variance requested under paragraph (d) of this
section.
(2) In making the findings required for issuance of a construction
permit or combined license, or the findings required by Sec. 52.103, or
in any enforcement hearing other than one initiated by the Commission
under paragraph (a)(1) of this section, if the application for the
construction permit or combined license references an early site permit,
the Commission shall treat as
[[Page 83]]
resolved those matters resolved in the proceeding on the application for
issuance or renewal of the early site permit, except as provided for in
paragraphs (b), (c), and (d) of this section.
(i) If the early site permit approved an emergency plan (or major
features thereof) that is in use by a licensee of a nuclear power plant,
the Commission shall treat as resolved changes to the early site permit
emergency plan (or major features thereof) that are identical to changes
made to the licensee's emergency plans in compliance with Sec. 50.54(q)
of this chapter occurring after issuance of the early site permit.
(ii) If the early site permit approved an emergency plan (or major
features thereof) that is not in use by a licensee of a nuclear power
plant, the Commission shall treat as resolved changes that are
equivalent to those that could be made under Sec. 50.54(q) of this
chapter without prior NRC approval had the emergency plan been in use by
a licensee.
(b) Updating of early site permit-emergency preparedness. An
applicant for a construction permit, operating license, or combined
license who has filed an application referencing an early site permit
issued under this subpart shall update the emergency preparedness
information that was provided under Sec. 52.17(b), and discuss whether
the updated information materially changes the bases for compliance with
applicable NRC requirements.
(c) Hearings and petitions. (1) In any proceeding for the issuance
of a construction permit, operating license, or combined license
referencing an early site permit, contentions on the following matters
may be litigated in the same manner as other issues material to the
proceeding:
(i) The nuclear power reactor proposed to be built does not fit
within one or more of the site characteristics or design parameters
included in the early site permit;
(ii) One or more of the terms and conditions of the early site
permit have not been met;
(iii) A variance requested under paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is provided in the application
that substantially alters the bases for a previous NRC conclusion or
constitutes a sufficient basis for the Commission to modify or impose
new terms and conditions related to emergency preparedness; or
(v) Any significant environmental issue that was not resolved in the
early site permit proceeding, or any issue involving the impacts of
construction and operation of the facility that was resolved in the
early site permit proceeding for which significant new information has
been identified.
(2) Any person may file a petition requesting that the site
characteristics, design parameters, or terms and conditions of the early
site permit should be modified, or that the permit should be suspended
or revoked. The petition will be considered in accordance with Sec.
2.206 of this chapter. Before construction commences, the Commission
shall consider the petition and determine whether any immediate action
is required. If the petition is granted, an appropriate order will be
issued. Construction under the construction permit or combined license
will not be affected by the granting of the petition unless the order is
made immediately effective. Any change required by the Commission in
response to the petition must meet the requirements of paragraph (a)(1)
of this section.
(d) Variances. An applicant for a construction permit, operating
license, or combined license referencing an early site permit may
include in its application a request for a variance from one or more
site characteristics, design parameters, or terms and conditions of the
early site permit, or from the site safety analysis report. In
determining whether to grant the variance, the Commission shall apply
the same technically relevant criteria applicable to the application for
the original or renewed early site permit. Once a construction permit or
combined license referencing an early site permit is issued, variances
from the early site permit will not be granted for that construction
permit or combined license.
(e) Early site permit amendment. The holder of an early site permit
may not make changes to the early site permit,
[[Page 84]]
including the site safety analysis report, without prior Commission
approval. The request for a change to the early site permit must be in
the form of an application for a license amendment, and must meet the
requirements of 10 CFR 50.90 and 50.92.
(f) Information requests. Except for information requests seeking to
verify compliance with the current licensing basis of the early site
permit, information requests to the holder of an early site permit must
be evaluated before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance of
the issue to be addressed in the requested information. Each evaluation
performed by the NRC staff must be in accordance with 10 CFR 50.54(f),
and must be approved by the Executive Director for Operations or his or
her designee before issuance of the request.
Subpart B_Standard Design Certifications
Sec. 52.41 Scope of subpart.
(a) This subpart sets forth the requirements and procedures
applicable to Commission issuance of rules granting standard design
certifications for nuclear power facilities separate from the filing of
an application for a construction permit or combined license for such a
facility.
(b)(1) Any person may seek a standard design certification for an
essentially complete nuclear power plant design which is an evolutionary
change from light water reactor designs of plants which have been
licensed and in commercial operation before April 18, 1989.
(2) Any person may also seek a standard design certification for a
nuclear power plant design which differs significantly from the light
water reactor designs described in paragraph (b)(1) of this section or
uses simplified, inherent, passive, or other innovative means to
accomplish its safety functions.
Sec. 52.43 Relationship to other subparts.
(a) This subpart applies to a person that requests a standard design
certification from the NRC separately from an application for a combined
license filed under subpart C of this part for a nuclear power facility.
An applicant for a combined license may reference a standard design
certification.
(b) Subpart E of this part governs the NRC staff review and approval
of a final standard design. Subpart E may be used independently of the
provisions in this subpart.
(c) Subpart F of this part governs the issuance of licenses to
manufacture nuclear power reactors to be installed and operated at sites
not identified in the manufacturing license application. Subpart F may
be used independently of the provisions in this subpart. However, an
applicant for a manufacturing license under subpart F may reference a
design certification.
Sec. 52.45 Filing of applications.
(a) An application for design certification may be filed
notwithstanding the fact that an application for a construction permit,
combined license, or manufacturing license for such a facility has not
been filed.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and Sec. Sec. 2.811 through 2.819 of
this chapter.
(c) The fees associated with the review of an application for the
initial issuance or renewal of a standard design certification are set
forth in 10 CFR part 170.
Sec. 52.46 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (c) and (j).
Sec. 52.47 Contents of applications; technical information.
The application must contain a level of design information
sufficient to enable the Commission to judge the applicant's proposed
means of assuring that construction conforms to the design and to reach
a final conclusion on all safety questions associated with the design
before the certification is granted. The information submitted for a
design certification must include performance requirements and design
information sufficiently detailed to permit the preparation of
acceptance and inspection requirements by the NRC,
[[Page 85]]
and procurement specifications and construction and installation
specifications by an applicant. The Commission will require, before
design certification, that information normally contained in certain
procurement specifications and construction and installation
specifications be completed and available for audit if the information
is necessary for the Commission to make its safety determination.
(a) The application must contain a final safety analysis report
(FSAR) that describes the facility, presents the design bases and the
limits on its operation, and presents a safety analysis of the
structures, systems, and components and of the facility as a whole, and
must include the following information:
(1) The site parameters postulated for the design, and an analysis
and evaluation of the design in terms of those site parameters;
(2) A description and analysis of the structures, systems, and
components (SSCs) of the facility, with emphasis upon performance
requirements, the bases, with technical justification therefor, upon
which these requirements have been established, and the evaluations
required to show that safety functions will be accomplished. It is
expected that the standard plant will reflect through its design,
construction, and operation an extremely low probability for accidents
that could result in the release of significant quantities of
radioactive fission products. The description shall be sufficient to
permit understanding of the system designs and their relationship to the
safety evaluations. Such items as the reactor core, reactor coolant
system, instrumentation and control systems, electrical systems,
containment system, other engineered safety features, auxiliary and
emergency systems, power conversion systems, radioactive waste handling
systems, and fuel handling systems shall be discussed insofar as they
are pertinent. The following power reactor design characteristics will
be taken into consideration by the Commission:
(i) Intended use of the reactor including the proposed maximum power
level and the nature and inventory of contained radioactive materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials; and
(iv) The safety features that are to be engineered into the facility
and those barriers that must be breached as a result of an accident
before a release of radioactive material to the environment can occur.
Special attention must be directed to plant design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \3\ from
the core into the containment assuming that the facility is operated at
the ultimate power level contemplated. The applicant shall perform an
evaluation and analysis of the postulated fission product release, using
the expected demonstrable containment leak rate and any fission product
cleanup systems intended to mitigate the consequences of the accidents,
together with applicable postulated site parameters, including site
meteorology, to evaluate the offsite radiological consequences. The
evaluation must determine that:
---------------------------------------------------------------------------
\3\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. These accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \4\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------
\4\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose for
radiation workers which, according to NCRP recommendations at the time
could be disregarded in the determination of their radiation exposure
status (see NBS Handbook 69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an acceptable limit for
an emergency dose to the public under accident conditions. This dose
value has been set forth in this section as a reference value, which can
be used in the evaluation of plant design features with respect to
postulated reactor accidents, to assure that these designs provide
assurance of low risk of public exposure to radiation, in the event of
an accident.
---------------------------------------------------------------------------
[[Page 86]]
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period of
its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(3) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to 10
CFR part 50, general design criteria (GDC), establishes minimum
requirements for the principal design criteria for water-cooled nuclear
power plants similar in design and location to plants for which
construction permits have previously been issued by the Commission and
provides guidance to applicants in establishing principal design
criteria for other types of nuclear power units;
(ii) The design bases and the relation of the design bases to the
principal design criteria;
(iii) Information relative to materials of construction, general
arrangement, and approximate dimensions, sufficient to provide
reasonable assurance that the design will conform to the design bases
with an adequate margin for safety;
(4) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and components
provided for the prevention of accidents and the mitigation of the
consequences of accidents. Analysis and evaluation of emergency core
cooling system (ECCS) cooling performance and the need for high-point
vents following postulated loss-of-coolant accidents shall be performed
in accordance with the requirements of Sec. Sec. 50.46 and 50.46a of
this chapter;
(5) The kinds and quantities of radioactive materials expected to be
produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(6) The information required by Sec. 20.1406 of this chapter;
(7) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(8) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and
(f)(3)(v);
(9) For applications for light-water-cooled nuclear power plants, an
evaluation of the standard plant design against the Standard Review Plan
(SRP) revision in effect 6 months before the docket date of the
application. The evaluation required by this section shall include an
identification and description of all differences in design features,
analytical techniques, and procedural measures proposed for the design
and those corresponding features, techniques, and measures given in the
SRP acceptance criteria. Where a difference exists, the evaluation shall
discuss how the proposed alternative provides an acceptable method of
complying with the Commission's regulations, or portions thereof, that
underlie the corresponding SRP acceptance criteria. The SRP is not a
substitute for the regulations, and compliance is not a requirement.
(10) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations described in 10 CFR
50.34a(e);
(11) Proposed technical specifications prepared in accordance with
the requirements of Sec. Sec. 50.36 and 50.36a of this chapter;
[[Page 87]]
(12) An analysis and description of the equipment and systems for
combustible gas control as required by 10 CFR 50.44;
(13) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(14) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in 10 CFR
50.60 and 50.61;
(15) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram events in Sec. 50.62;
(16) A coping analysis, and any design features necessary to address
station blackout, as required by 10 CFR 50.63;
(17) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(18) A description and analysis of the fire protection design
features for the standard plant necessary to comply with 10 CFR part 50,
appendix A, GDC 3, and Sec. 50.48 of this chapter;
(19) A description of the quality assurance program applied to the
design of the structures, systems, and components of the facility.
Appendix B to 10 CFR part 50, ``Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants,'' sets forth the requirements
for quality assurance programs for nuclear power plants. The description
of the quality assurance program for a nuclear power plant shall include
a discussion of how the applicable requirements of appendix B to 10 CFR
part 50 were satisfied;
(20) The information necessary to demonstrate that the standard
plant complies with the earthquake engineering criteria in 10 CFR part
50, appendix S;
(21) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(22) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection, hydrogen
combustion, and containment bypass;
(24) A representative conceptual design for those portions of the
plant for which the application does not seek certification, to aid the
NRC in its review of the FSAR and to permit assessment of the adequacy
of the interface requirements in paragraph (a)(25) of this section;
(25) The interface requirements to be met by those portions of the
plant for which the application does not seek certification. These
requirements must be sufficiently detailed to allow completion of the
FSAR;
(26) Justification that compliance with the interface requirements
of paragraph (a)(25) of this section is verifiable through inspections,
tests, or analyses. The method to be used for verification of interface
requirements must be included as part of the proposed ITAAC required by
paragraph (b)(1) of this section; and
(27) A description of the design-specific probabilistic risk
assessment (PRA) and its results.
(28) For applications for standard design certifications which are
subject to 10 CFR 50.150(a), the information required by 10 CFR
50.150(b).
(b) The application must also contain:
(1) The proposed inspections, tests, analyses, and acceptance
criteria that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met, a facility that incorporates the design
certification has been constructed and will be operated in conformity
with the design certification, the provisions of the Act, and the
Commission's rules and regulations; and
(2) An environmental report as required by 10 CFR 51.55.
[[Page 88]]
(c) This paragraph applies, according to its provisions, to
particular applications:
(1) An application for certification of a nuclear power reactor
design that is an evolutionary change from light-water reactor designs
of plants that have been licensed and in commercial operation before
April 18, 1989, must provide an essentially complete nuclear power plant
design except for site-specific elements such as the service water
intake structure and the ultimate heat sink;
(2) An application for certification of a nuclear power reactor
design that differs significantly from the light-water reactor designs
described in paragraph (c)(1) of this section or uses simplified,
inherent, passive, or other innovative means to accomplish its safety
functions must provide an essentially complete nuclear power reactor
design except for site-specific elements such as the service water
intake structure and the ultimate heat sink, and must meet the
requirements of 10 CFR 50.43(e); and
(3) An application for certification of a modular nuclear power
reactor design must describe and analyze the possible operating
configurations of the reactor modules with common systems, interface
requirements, and system interactions. The final safety analysis must
also account for differences among the configurations, including any
restrictions that will be necessary during the construction and startup
of a given module to ensure the safe operation of any module already
operating.
(d) Each applicant for a standard design certification under this
part shall protect Safeguards Information against unauthorized
disclosure in accordance with the requirements in Sec. Sec. 73.21 and
73.22 of this chapter, as applicable.
[72 FR 49517, Aug. 28, 2007, as amended at 73 FR 63571, Oct. 24, 2008;
74 FR 28147, June 12, 2009]
Sec. 52.48 Standards for review of applications.
Applications filed under this subpart will be reviewed for
compliance with the standards set out in 10 CFR parts 20, 50 and its
appendices, 51, 73, and 100.
Sec. 52.51 Administrative review of applications.
(a) A standard design certification is a rule that will be issued in
accordance with the provisions of subpart H of 10 CFR part 2, as
supplemented by the provisions of this section. The Commission shall
initiate the rulemaking after an application has been filed under Sec.
52.45 and shall specify the procedures to be used for the rulemaking.
The notice of proposed rulemaking published in the Federal Register must
provide an opportunity for the submission of comments on the proposed
design certification rule. If, at the time a proposed design
certification rule is published in the Federal Register under this
paragraph (a), the Commission decides that a legislative hearing should
be held, the information required by 10 CFR 2.1502(c) must be included
in the Federal Register document for the proposed design certification.
(b) Following the submission of comments on the proposed design
certification rule, the Commission may, at its discretion, hold a
legislative hearing under the procedures in subpart O of part 2 of this
chapter. The Commission shall publish a document in the Federal Register
of its decision to hold a legislative hearing. The document shall
contain the information specified in paragraph (c) of this section, and
specify whether the Commission or a presiding officer will conduct the
legislative hearing.
(c) Notwithstanding anything in 10 CFR 2.390 to the contrary,
proprietary information will be protected in the same manner and to the
same extent as proprietary information submitted in connection with
applications for licenses, provided that the design certification shall
be published in Chapter I of this title.
Sec. 52.53 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which concern
safety.
[[Page 89]]
Sec. 52.54 Issuance of standard design certification.
(a) After conducting a rulemaking proceeding under Sec. 52.51 on an
application for a standard design certification and receiving the report
to be submitted by the Advisory Committee on Reactor Safeguards under
Sec. 52.53, the Commission may issue a standard design certification in
the form of a rule for the design which is the subject of the
application, if the Commission determines that:
(1) The application meets the applicable standards and requirements
of the Atomic Energy Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the standard design conforms
with the provisions of the Act, and the Commission's regulations;
(4) The applicant is technically qualified;
(5) The proposed inspections, tests, analyses, and acceptance
criteria are necessary and sufficient, within the scope of the standard
design, to provide reasonable assurance that, if the inspections, tests,
and analyses are performed and the acceptance criteria met, the facility
has been constructed and will be operated in accordance with the design
certification, the provisions of the Act, and the Commission's
regulations;
(6) Issuance of the standard design certification will not be
inimical to the common defense and security or to the health and safety
of the public;
(7) The findings required by subpart A of part 51 of this chapter
have been made; and
(8) The applicant has implemented the quality assurance program
described or referenced in the safety analysis report.
(b) The design certification rule must specify the site parameters,
design characteristics, and any additional requirements and restrictions
of the design certification rule.
(c) After the Commission has adopted a final design certification
rule, the applicant shall not permit any individual to have access to or
any facility to possess restricted data or classified National Security
Information until the individual and/or facility has been approved for
access under the provisions of 10 CFR parts 25 and/or 95, as applicable.
Sec. 52.55 Duration of certification.
(a) Except as provided in paragraph (b) of this section, a standard
design certification issued under this subpart is valid for 15 years
from the date of issuance.
(b) A standard design certification continues to be valid beyond the
date of expiration in any proceeding on an application for a combined
license or an operating license that references the standard design
certification and is docketed either before the date of expiration of
the certification, or, if a timely application for renewal of the
certification has been filed, before the Commission has determined
whether to renew the certification. A design certification also
continues to be valid beyond the date of expiration in any hearing held
under Sec. 52.103 before operation begins under a combined license that
references the design certification.
(c) An applicant for a construction permit or a combined license
may, at its own risk, reference in its application a design for which a
design certification application has been docketed but not granted.
Sec. 52.57 Application for renewal.
(a) Not less than 12 nor more than 36 months before the expiration
of the initial 15-year period, or any later renewal period, any person
may apply for renewal of the certification. An application for renewal
must contain all information necessary to bring up to date the
information and data contained in the previous application. The
Commission will require, before renewal of certification, that
information normally contained in certain procurement specifications and
construction and installation specifications be completed and available
for audit if this information is necessary for the Commission to make
its safety determination. Notice and comment procedures must be used for
a rulemaking
[[Page 90]]
proceeding on the application for renewal. The Commission, in its
discretion, may require the use of additional procedures in individual
renewal proceedings.
(b) A design certification, either original or renewed, for which a
timely application for renewal has been filed remains in effect until
the Commission has determined whether to renew the certification. If the
certification is not renewed, it continues to be valid in certain
proceedings, in accordance with the provisions of Sec. 52.55.
(c) The Commission shall refer a copy of the application for renewal
to the Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall
report on those portions of the application which concern safety and
shall apply the criteria set forth in Sec. 52.59.
Sec. 52.59 Criteria for renewal.
(a) The Commission shall issue a rule granting the renewal if the
design, either as originally certified or as modified during the
rulemaking on the renewal, complies with the Atomic Energy Act and the
Commission's regulations applicable and in effect at the time the
certification was issued, provided, however, that the first time the
Commission issues a rule granting the renewal for a standard design
certification in effect on July 13, 2009, the Commission shall, in
addition, find that the renewed design complies with the applicable
requirements of 10 CFR 50.150.
(b) The Commission may impose other requirements if it determines
that:
(1) They are necessary for adequate protection to public health and
safety or common defense and security;
(2) They are necessary for compliance with the Commission's
regulations and orders applicable and in effect at the time the design
certification was issued; or
(3) There is a substantial increase in overall protection of the
public health and safety or the common defense and security to be
derived from the new requirements, and the direct and indirect costs of
implementing those requirements are justified in view of this increased
protection.
(c) In addition, the applicant for renewal may request an amendment
to the design certification. The Commission shall grant the amendment
request if it determines that the amendment will comply with the Atomic
Energy Act and the Commission's regulations in effect at the time of
renewal. If the amendment request entails such an extensive change to
the design certification that an essentially new standard design is
being proposed, an application for a design certification must be filed
in accordance with this subpart.
(d) Denial of renewal does not bar the applicant, or another
applicant, from filing a new application for certification of the
design, which proposes design changes that correct the deficiencies
cited in the denial of the renewal.
[72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]
Sec. 52.61 Duration of renewal.
Each renewal of certification for a standard design will be for not
less than 10, nor more than 15 years.
Sec. 52.63 Finality of standard design certifications.
(a)(1) Notwithstanding any provision in 10 CFR 50.109, while a
standard design certification rule is in effect under Sec. Sec. 52.55
or 52.61, the Commission may not modify, rescind, or impose new
requirements on the certification information, whether on its own
motion, or in response to a petition from any person, unless the
Commission determines in a rulemaking that the change:
(i) Is necessary either to bring the certification information or
the referencing plants into compliance with the Commission's regulations
applicable and in effect at the time the certification was issued;
(ii) Is necessary to provide adequate protection of the public
health and safety or the common defense and security;
(iii) Reduces unnecessary regulatory burden and maintains protection
to public health and safety and the common defense and security;
[[Page 91]]
(iv) Provides the detailed design information to be verified under
those inspections, tests, analyses, and acceptance criteria (ITAAC)
which are directed at certification information (i.e., design acceptance
criteria);
(v) Is necessary to correct material errors in the certification
information;
(vi) Substantially increases overall safety, reliability, or
security of facility design, construction, or operation, and the direct
and indirect costs of implementation of the rule change are justified in
view of this increased safety, reliability, or security; or
(vii) Contributes to increased standardization of the certification
information.
(2)(i) In a rulemaking under Sec. 52.63(a)(1), except for Sec.
52.63(a)(1)(ii), the Commission will give consideration to whether the
benefits justify the costs for plants that are already licensed or for
which an application for a permit or license is under consideration.
(ii) The rulemaking procedures for changes under Sec. 52.63(a)(1)
must provide for notice and opportunity for public comment.
(3) Any modification the NRC imposes on a design certification rule
under paragraph (a)(1) of this section will be applied to all plants
referencing the certified design, except those to which the modification
has been rendered technically irrelevant by action taken under
paragraphs (a)(4) or (b)(1) of this section.
(4) The Commission may not impose new requirements by plant-specific
order on any part of the design of a specific plant referencing the
design certification rule if that part was approved in the design
certification while a design certification rule is in effect under Sec.
52.55 or Sec. 52.61, unless:
(i) A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time the
certification was issued, or to assure adequate protection of the public
health and safety or the common defense and security; and
(ii) Special circumstances as defined in 10 CFR 52.7 are present. In
addition to the factors listed in Sec. 52.7, the Commission shall
consider whether the special circumstances which Sec. 52.7 requires to
be present outweigh any decrease in safety that may result from the
reduction in standardization caused by the plant-specific order.
(5) Except as provided in 10 CFR 2.335, in making the findings
required for issuance of a combined license, construction permit,
operating license, or manufacturing license, or for any hearing under
Sec. 52.103, the Commission shall treat as resolved those matters
resolved in connection with the issuance or renewal of a design
certification rule.
(b)(1) An applicant or licensee who references a design
certification rule may request an exemption from one or more elements of
the certification information. The Commission may grant such a request
only if it determines that the exemption will comply with the
requirements of Sec. 52.7. In addition to the factors listed in Sec.
52.7, the Commission shall consider whether the special circumstances
that Sec. 52.7 requires to be present outweigh any decrease in safety
that may result from the reduction in standardization caused by the
exemption. The granting of an exemption on request of an applicant is
subject to litigation in the same manner as other issues in the
operating license or combined license hearing.
(2) Subject to Sec. 50.59 of this chapter, a licensee who
references a design certification rule may make departures from the
design of the nuclear power facility, without prior Commission approval,
unless the proposed departure involves a change to the design as
described in the rule certifying the design. The licensee shall maintain
records of all departures from the facility and these records must be
maintained and available for audit until the date of termination of the
license.
(c) The Commission will require, before granting a construction
permit, combined license, operating license, or manufacturing license
which references a design certification rule, that information normally
contained in certain procurement specifications and construction and
installation specifications be completed and available for audit if the
information is necessary for the Commission to make its safety
[[Page 92]]
determinations, including the determination that the application is
consistent with the certification information. This information may be
acquired by appropriate arrangements with the design certification
applicant.
Subpart C_Combined Licenses
Sec. 52.71 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of combined licenses for nuclear power facilities.
Sec. 52.73 Relationship to other subparts.
(a) An application for a combined license under this subpart may,
but need not, reference a standard design certification, standard design
approval, or manufacturing license issued under subparts B, E, or F of
this part, respectively, or an early site permit issued under subpart A
of this part. In the absence of a demonstration that an entity other
than the one originally sponsoring and obtaining a design certification
is qualified to supply a design, the Commission will entertain an
application for a combined license that references a standard design
certification issued under subpart B of this part only if the entity
that sponsored and obtained the certification supplies the design for
the applicant's use.
(b) The Commission will require, before granting a combined license
that references a standard design certification, that information
normally contained in certain procurement specifications and
construction and installation specifications be completed and available
for audit if the information is necessary for the Commission to make its
safety determinations, including the determination that the application
is consistent with the certification information.
Sec. 52.75 Filing of applications.
(a) Any person except one excluded by Sec. 50.38 of this chapter
may file an application for a combined license for a nuclear power
facility with the Director, Office of New Reactors or Director, Office
of Nuclear Reactor Regulation, as appropriate.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
[72 FR 49517, Aug. 28, 2007, as amended at 73 FR 5724, Jan. 31, 2008]
Sec. 52.77 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33.
Sec. 52.79 Contents of applications; technical information in final
safety analysis report.
(a) The application must contain a final safety analysis report that
describes the facility, presents the design bases and the limits on its
operation, and presents a safety analysis of the structures, systems,
and components of the facility as a whole. The final safety analysis
report shall include the following information, at a level of
information sufficient to enable the Commission to reach a final
conclusion on all safety matters that must be resolved by the Commission
before issuance of a combined license:
(1)(i) The boundaries of the site;
(ii) The proposed general location of each facility on the site;
(iii) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site with appropriate consideration of
the most severe of the natural phenomena that have been historically
reported for the site and surrounding area and with sufficient margin
for the limited accuracy, quantity, and time in which the historical
data have been accumulated;
(iv) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(v) The existing and projected future population profile of the area
surrounding the site;
(vi) A description and safety assessment of the site on which the
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of
[[Page 93]]
the site under the radiological consequence evaluation factors
identified in paragraphs (a)(1)(vi)(A) and (a)(1)(vi)(B) of this
section. In performing this assessment, an applicant shall assume a
fission product release \5\ from the core into the containment assuming
that the facility is operated at the ultimate power level contemplated.
The applicant shall perform an evaluation and analysis of the postulated
fission product release, using the expected demonstrable containment
leak rate and any fission product cleanup systems intended to mitigate
the consequences of the accidents, together with applicable site
characteristics, including site meteorology, to evaluate the offsite
radiological consequences. Site characteristics must comply with part
100 of this chapter. The evaluation must determine that:
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\5\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. These accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
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(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \6\ total effective dose equivalent (TEDE).
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\6\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose for
radiation workers which, according to NCRP recommendations at the time
could be disregarded in the determination of their radiation exposure
status (see NBS Handbook 69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an acceptable limit for
an emergency dose to the public under accident conditions. Rather, this
dose value has been set forth in this section as a reference value,
which can be used in the evaluation of plant design features with
respect to postulated reactor accidents, to assure that these designs
provide assurance of low risk of public exposure to radiation, in the
event of an accident.
---------------------------------------------------------------------------
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period of
its passage) would not receive a radiation dose in excess of 25 rem
TEDE; and
(2) A description and analysis of the structures, systems, and
components of the facility with emphasis upon performance requirements,
the bases, with technical justification therefor, upon which these
requirements have been established, and the evaluations required to show
that safety functions will be accomplished. It is expected that reactors
will reflect through their design, construction, and operation an
extremely low probability for accidents that could result in the release
of significant quantities of radioactive fission products. The
descriptions shall be sufficient to permit understanding of the system
designs and their relationship to safety evaluations. Items such as the
reactor core, reactor coolant system, instrumentation and control
systems, electrical systems, containment system, other engineered safety
features, auxiliary and emergency systems, power conversion systems,
radioactive waste handling systems, and fuel handling systems shall be
discussed insofar as they are pertinent. The following power reactor
design characteristics and proposed operation will be taken into
consideration by the Commission:
(i) Intended use of the reactor including the proposed maximum power
level and the nature and inventory of contained radioactive materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials;
(iv) The safety features that are to be engineered into the facility
and those barriers that must be breached as a result of an accident
before a release of radioactive material to the environment can occur.
Special attention must be directed to plant design features intended to
mitigate the radiological consequences of accidents. In
[[Page 94]]
performing this assessment, an applicant shall assume a fission product
release \7\ from the core into the containment assuming that the
facility is operated at the ultimate power level contemplated;
---------------------------------------------------------------------------
\7\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. These accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(3) The kinds and quantities of radioactive materials expected to be
produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(4) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to
part 50 of this chapter, ``General Design Criteria for Nuclear Power
Plants,'' establishes minimum requirements for the principal design
criteria for water-cooled nuclear power plants similar in design and
location to plants for which construction permits have previously been
issued by the Commission and provides guidance to applicants in
establishing principal design criteria for other types of nuclear power
units;
(ii) The design bases and the relation of the design bases to the
principal design criteria;
(iii) Information relative to materials of construction,
arrangement, and dimensions, sufficient to provide reasonable assurance
that the design will conform to the design bases with adequate margin
for safety.
(5) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and components
provided for the prevention of accidents and the mitigation of the
consequences of accidents. Analysis and evaluation of ECCS cooling
performance and the need for high-point vents following postulated loss-
of-coolant accidents shall be performed in accordance with the
requirements of Sec. Sec. 50.46 and 50.46a of this chapter;
(6) A description and analysis of the fire protection design
features for the reactor necessary to comply with 10 CFR part 50,
appendix A, GDC 3, and Sec. 50.48 of this chapter;
(7) A description of protection provided against pressurized thermal
shock events, including projected values of the reference temperature
for reactor vessel beltline materials as defined in Sec. Sec. 50.60 and
50.61(b)(1) and (b)(2) of this chapter;
(8) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(9) The coping analyses, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(10) A description of the program, and its implementation, required
by Sec. 50.49(a) of this chapter for the environmental qualification of
electric equipment important to safety and the list of electric
equipment important to safety that is required by 10 CFR 50.49(d);
(11) A description of the program(s), and their implementation,
necessary to ensure that the systems and components meet the
requirements of the ASME Boiler and Pressure Vessel Code and the ASME
Code for Operation and Maintenance of Nuclear Power Plants in accordance
with 50.55a of this chapter;
(12) A description of the primary containment leakage rate testing
program, and its implementation, necessary to ensure that the
containment meets the requirements of appendix J to 10 CFR part 50;
(13) A description of the reactor vessel material surveillance
program required by appendix H to 10 CFR part 50 and its implementation;
(14) A description of the operator training program, and its
implementation, necessary to meet the requirements of 10 CFR part 55;
[[Page 95]]
(15) A description of the program, and its implementation, for
monitoring the effectiveness of maintenance necessary to meet the
requirements of Sec. 50.65 of this chapter;
(16)(i) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations, as described in
Sec. 50.34a(d) of this chapter;
(ii) A description of the process and effluent monitoring and
sampling program required by appendix I to 10 CFR part 50 and its
implementation.
(17) The information with respect to compliance with technically
relevant positions of the Three Mile Island requirements in Sec.
50.34(f) of this chapter, with the exception of Sec. Sec.
50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v);
(18) If the applicant seeks to use risk-informed treatment of SSCs
in accordance with Sec. 50.69 of this chapter, the information required
by Sec. 50.69(b)(2) of this chapter;
(19) Information necessary to demonstrate that the plant complies
with the earthquake engineering criteria in 10 CFR part 50, appendix S;
(20) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(21) Emergency plans complying with the requirements of Sec. 50.47
of this chapter, and 10 CFR part 50, appendix E;
(22)(i) All emergency plan certifications that have been obtained
from the State and local governmental agencies with emergency planning
responsibilities must state that:
(A) The proposed emergency plans are practicable;
(B) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(C) These agencies are committed to executing their responsibilities
under the plans in the event of an emergency;
(ii) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(23) [Reserved]
(24) If the application is for a nuclear power reactor design which
differs significantly from light-water reactor designs that were
licensed before 1997 or use simplified, inherent, passive, or other
innovative means to accomplish their safety functions, the application
must describe how the design meets the requirements in Sec. 50.43(e) of
this chapter;
(25) A description of the quality assurance program, applied to the
design, and to be applied to the fabrication, construction, and testing,
of the structures, systems, and components of the facility. Appendix B
to 10 CFR part 50 sets forth the requirements for quality assurance
programs for nuclear power plants. The description of the quality
assurance program for a nuclear power plant must include a discussion of
how the applicable requirements of appendix B to 10 CFR part 50 have
been and will be satisfied, including a discussion of how the quality
assurance program will be implemented;
(26) The applicant's organizational structure, allocations or
responsibilities and authorities, and personnel qualifications
requirements for operation;
(27) Managerial and administrative controls to be used to assure
safe operation. Appendix B to 10 CFR part 50 sets forth the requirements
for these controls for nuclear power plants. The information on the
controls to be used for a nuclear power plant shall include a discussion
of how the applicable requirements of appendix B to 10 CFR part 50 will
be satisfied;
(28) Plans for preoperational testing and initial operations;
(29)(i) Plans for conduct of normal operations, including
maintenance, surveillance, and periodic testing of structures, systems,
and components;
(ii) Plans for coping with emergencies, other than the plans
required by Sec. 52.79(a)(21);
[[Page 96]]
(30) Proposed technical specifications prepared in accordance with
the requirements of Sec. Sec. 50.36 and 50.36a of this chapter;
(31) For nuclear power plants to be operated on multi-unit sites, an
evaluation of the potential hazards to the structures, systems, and
components important to safety of operating units resulting from
construction activities, as well as a description of the managerial and
administrative controls to be used to provide assurance that the
limiting conditions for operation are not exceeded as a result of
construction activities at the multi-unit sites;
(32) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(33) A description of the training program required by Sec. 50.120
of this chapter and its implementation;
(34) A description and plans for implementation of an operator
requalification program. The operator requalification program must as a
minimum, meet the requirements for those programs contained in Sec.
55.59 of this chapter;
(35)(i) A physical security plan, describing how the applicant will
meet the requirements of 10 CFR part 73 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
The plan must list tests, inspections, audits, and other means to be
used to demonstrate compliance with the requirements of 10 CFR parts 11
and 73, if applicable;
(ii) A description of the implementation of the physical security
plan;
(36)(i) A safeguards contingency plan in accordance with the
criteria set forth in appendix C to 10 CFR part 73. The safeguards
contingency plan shall include plans for dealing with threats, thefts,
and radiological sabotage, as defined in part 73 of this chapter,
relating to the special nuclear material and nuclear facilities licensed
under this chapter and in the applicant's possession and control. Each
application for this type of license shall include the information
contained in the applicant's safeguards contingency plan. \8\
(Implementing procedures required for this plan need not be submitted
for approval.)
---------------------------------------------------------------------------
\8\ A physical security plan that contains all the information
required in both Sec. 73.55 of this chapter and appendix C to 10 CFR
part 73 satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------
(ii) A training and qualification plan in accordance with the
criteria set forth in appendix B to 10 CFR part 73.
(iii) A cyber security plan in accordance with the criteria set
forth in Sec. 73.54 of this chapter;
(iv) A description of the implementation of the safeguards
contingency plan, training and qualification plan, and cyber security
plan; and
(v) Each applicant who prepares a physical security plan, a
safeguards contingency plan, a training and qualification plan, or a
cyber security plan, shall protect the plans and other related
Safeguards Information against unauthorized disclosure in accordance
with the requirements of Sec. 73.21 of this chapter.
(37) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(38) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection, hydrogen
combustion, and containment bypass;
(39) A description of the radiation protection program required by
Sec. 20.1101 of this chapter and its implementation.
(40) A description of the fire protection program required by Sec.
50.48 of this chapter and its implementation.
(41) For applications for light-water-cooled nuclear power plant
combined licenses, an evaluation of the facility against the Standard
Review Plan (SRP) revision in effect 6 months before the docket date of
the application. The evaluation required by this section shall include
an identification and description of all differences in design features,
analytical techniques, and
[[Page 97]]
procedural measures proposed for a facility and those corresponding
features, techniques, and measures given in the SRP acceptance criteria.
Where a difference exists, the evaluation shall discuss how the proposed
alternative provides an acceptable method of complying with the
Commission's regulations, or portions thereof, that underlie the
corresponding SRP acceptance criteria. The SRP is not a substitute for
the regulations, and compliance is not a requirement;
(42) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62 of this chapter;
(43) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68 of this chapter;
(44) A description of the fitness-for-duty program required by 10
CFR part 26 and its implementation.
(45) The information required by Sec. 20.1406 of this chapter.
(46) A description of the plant-specific probabilistic risk
assessment (PRA) and its results.
(47) For applications for combined licenses which are subject to 10
CFR 50.150(a), the information required by 10 CFR 50.150(b).
(b) If the combined license application references an early site
permit, then the following requirements apply:
(1) The final safety analysis report need not contain information or
analyses submitted to the Commission in connection with the early site
permit, provided, however, that the final safety analysis report must
either include or incorporate by reference the early site permit site
safety analysis report and must contain, in addition to the information
and analyses otherwise required, information sufficient to demonstrate
that the design of the facility falls within the site characteristics
and design parameters specified in the early site permit.
(2) If the final safety analysis report does not demonstrate that
design of the facility falls within the site characteristics and design
parameters, the application shall include a request for a variance that
complies with the requirements of Sec. Sec. 52.39 and 52.93.
(3) The final safety analysis report must demonstrate that all terms
and conditions that have been included in the early site permit, other
than those imposed under Sec. 50.36b, will be satisfied by the date of
issuance of the combined license. Any terms or conditions of the early
site permit that could not be met by the time of issuance of the
combined license, must be set forth as terms or conditions of the
combined license.
(4) If the early site permit approves complete and integrated
emergency plans, or major features of emergency plans, then the final
safety analysis report must include any new or additional information
that updates and corrects the information that was provided under Sec.
52.17(b), and discuss whether the new or additional information
materially changes the bases for compliance with the applicable
requirements. The application must identify changes to the emergency
plans or major features of emergency plans that have been incorporated
into the proposed facility emergency plans and that constitute or would
constitute a decrease in effectiveness under Sec. 50.54(q) of this
chapter.
(5) If complete and integrated emergency plans are approved as part
of the early site permit, new certifications meeting the requirements of
paragraph (a)(22) of this section are not required.
(c) If the combined license application references a standard design
approval, then the following requirements apply:
(1) The final safety analysis report need not contain information or
analyses submitted to the Commission in connection with the design
approval, provided, however, that the final safety analysis report must
either include or incorporate by reference the standard design approval
final safety analysis report and must contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the characteristics of the site fall within the site
parameters specified in the design approval. In addition, the plant-
specific PRA information must use the PRA information for
[[Page 98]]
the design approval and must be updated to account for site-specific
design information and any design changes or departures.
(2) The final safety analysis report must demonstrate that all terms
and conditions that have been included in the final design approval will
be satisfied by the date of issuance of the combined license.
(d) If the combined license application references a standard design
certification, then the following requirements apply:
(1) The final safety analysis report need not contain information or
analyses submitted to the Commission in connection with the design
certification, provided, however, that the final safety analysis report
must either include or incorporate by reference the standard design
certification final safety analysis report and must contain, in addition
to the information and analyses otherwise required, information
sufficient to demonstrate that the site characteristics fall within the
site parameters specified in the design certification. In addition, the
plant-specific PRA information must use the PRA information for the
design certification and must be updated to account for site-specific
design information and any design changes or departures.
(2) The final safety analysis report must demonstrate that the
interface requirements established for the design under Sec. 52.47 have
been met.
(3) The final safety analysis report must demonstrate that all
requirements and restrictions set forth in the referenced design
certification rule, other than those imposed under Sec. 50.36b, must be
satisfied by the date of issuance of the combined license. Any
requirements and restrictions set forth in the referenced design
certification rule that could not be satisfied by the time of issuance
of the combined license, must be set forth as terms or conditions of the
combined license.
(e) If the combined license application references the use of one or
more manufactured nuclear power reactors licensed under subpart F of
this part, then the following requirements apply:
(1) The final safety analysis report need not contain information or
analyses submitted to the Commission in connection with the
manufacturing license, provided, however, that the final safety analysis
report must either include or incorporate by reference the manufacturing
license final safety analysis report and must contain, in addition to
the information and analyses otherwise required, information sufficient
to demonstrate that the site characteristics fall within the site
parameters specified in the manufacturing license. In addition, the
plant-specific PRA information must use the PRA information for the
manufactured reactor and must be updated to account for site-specific
design information and any design changes or departures.
(2) The final safety analysis report must demonstrate that the
interface requirements established for the design have been met.
(3) The final safety analysis report must demonstrate that all terms
and conditions that have been included in the manufacturing license,
other than those imposed under Sec. 50.36b, will be satisfied by the
date of issuance of the combined license. Any terms or conditions of the
manufacturing license that could not be met by the time of issuance of
the combined license, must be set forth as terms or conditions of the
combined license.
(f) Each applicant for a combined license under this subpart shall
protect Safeguards Information against unauthorized disclosure in
accordance with the requirements in Sec. Sec. 73.21 and 73.22 of this
chapter, as applicable.
[72 FR 49517, Aug. 28, 2007, as amended at 73 FR 63571, Oct. 24, 2008;
74 FR 13970, Mar. 27, 2009; 74 FR 28147, June 12, 2009]
Sec. 52.80 Contents of applications; additional technical information.
The application must contain:
(a) The proposed inspections, tests, and analyses, including those
applicable to emergency planning, that the licensee shall perform, and
the acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the combined
license,
[[Page 99]]
the provisions of the Act, and the Commission's rules and regulations.
(1) If the application references an early site permit with ITAAC,
the early site permit ITAAC must apply to those aspects of the combined
license which are approved in the early site permit.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those portions
of the facility design which are approved in the design certification.
(3) If the application references an early site permit with ITAAC or
a standard design certification or both, the application may include a
notification that a required inspection, test, or analysis in the ITAAC
has been successfully completed and that the corresponding acceptance
criterion has been met. The Federal Register notification required by
Sec. 52.85 must indicate that the application includes this
notification.
(b) An environmental report, either in accordance with 10 CFR
51.50(c) if a limited work authorization under 10 CFR 50.10 is not
requested in conjunction with the combined license application, or in
accordance with Sec. Sec. 51.49 and 51.50(c) of this chapter if a
limited work authorization is requested in conjunction with the combined
license application.
(c) If the applicant wishes to request that a limited work
authorization under 10 CFR 50.10 be issued before issuance of the
combined license, the application must include the information otherwise
required by 10 CFR 50.10, in accordance with either 10 CFR 2.101(a)(1)
through (a)(4), or 10 CFR 2.101(a)(9).
(d) A description and plans for implementation of the guidance and
strategies intended to maintain or restore core cooling, containment,
and spent fuel pool cooling capabilities under the circumstances
associated with the loss of large areas of the plant due to explosions
or fire as required by Sec. 50.54(hh)(2) of this chapter.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007; 74
FR 13970, Mar. 27, 2009]
Sec. 52.81 Standards for review of applications.
Applications filed under this subpart will be reviewed according to
the standards set out in 10 CFR parts 20, 50, 51, 54, 55, 73, 100, and
140.
Sec. 52.83 Finality of referenced NRC approvals; partial initial
decision on site suitability.
(a) If the application for a combined license under this subpart
references an early site permit, design certification rule, standard
design approval, or manufacturing license, the scope and nature of
matters resolved for the application and any combined license issued are
governed by the relevant provisions addressing finality, including
Sec. Sec. 52.39, 52.63, 52.98, 52.145, and 52.171.
(b) While a partial decision on site suitability is in effect under
10 CFR 2.617(b)(2), the scope and nature of matters resolved in the
proceeding are governed by the finality provisions in 10 CFR 2.629.
Sec. 52.85 Administrative review of applications; hearings.
A proceeding on a combined license is subject to all applicable
procedural requirements contained in 10 CFR part 2, including the
requirements for docketing (Sec. 2.101 of this chapter) and issuance of
a notice of hearing (Sec. 2.104 of this chapter). If an applicant
requests a Commission finding on certain ITAAC with the issuance of the
combined license, then those ITAAC will be identified in the notice of
hearing. All hearings on combined licenses are governed by the
procedures contained in 10 CFR part 2.
Sec. 52.87 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application that concern
safety and shall apply the standards referenced in Sec. 52.81, in
accordance with the finality provisions in Sec. 52.83.
[[Page 100]]
Sec. 52.89 [Reserved]
Sec. 52.91 Authorization to conduct limited work authorization activities.
(a) If the application does not reference an early site permit which
authorizes the holder to perform the activities under 10 CFR 50.10(d),
the applicant may not perform those activities without obtaining the
separate authorization required by 10 CFR 50.10(d). Authorization may be
granted only after the presiding officer in the proceeding on the
application has made the findings and determination required by 10 CFR
50.10(e), and the Director of New Reactors or the Director of Nuclear
Reactor Regulation makes the determination required by 10 CFR 50.10(e).
(b) If, after an applicant has performed the activities permitted by
paragraph (a) of this section, the application for the combined license
is withdrawn or denied, then the applicant shall implement the approved
site redress plan.
[72 FR 57447, Oct. 9, 2007]
Sec. 52.93 Exemptions and variances.
(a) Applicants for a combined license under this subpart, or any
amendment to a combined license, may include in the application a
request for an exemption from one or more of the Commission's
regulations.
(1) If the request is for an exemption from any part of a referenced
design certification rule, the Commission may grant the request if it
determines that the exemption complies with any exemption provisions of
the referenced design certification rule, or with Sec. 52.63 if there
are no applicable exemption provisions in the referenced design
certification rule.
(2) For all other requests for exemptions, the Commission may grant
a request if it determines that the exemption complies with Sec. 52.7.
(b) An applicant for a combined license who has filed an application
referencing an early site permit issued under subpart A of this part may
include in the application a request for a variance from one or more
site characteristics, design parameters, or terms and conditions of the
permit, or from the site safety analysis report. In determining whether
to grant the variance, the Commission shall apply the same technically
relevant criteria as were applicable to the application for the original
or renewed site permit. Once a construction permit or combined license
referencing an early site permit is issued, variances from the early
site permit will not be granted for that construction permit or combined
license.
(c) An applicant for a combined license who has filed an application
referencing a nuclear power reactor manufactured under a manufacturing
license issued under subpart F of this part may include in the
application a request for a departure from one or more design
characteristics, site parameters, terms and conditions, or approved
design of the manufactured reactor. The Commission may grant a request
only if it determines that the departure will comply with the
requirements of 10 CFR 52.7, and that the special circumstances outweigh
any decrease in safety that may result from the reduction in
standardization caused by the departure.
(d) Issuance of a variance under paragraph (b) or a departure under
paragraph (c) of this section is subject to litigation during the
combined license proceeding in the same manner as other issues material
to that proceeding.
Sec. 52.97 Issuance of combined licenses.
(a)(1) After conducting a hearing in accordance with Sec. 52.85 and
receiving the report submitted by the ACRS, the Commission may issue a
combined license if the Commission finds that:
(i) The applicable standards and requirements of the Act and the
Commission's regulations have been met;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) There is reasonable assurance that the facility will be
constructed and will operate in conformity with the license, the
provisions of the Act, and the Commission's regulations.
(iv) The applicant is technically and financially qualified to
engage in the activities authorized; and
[[Page 101]]
(v) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public; and
(vi) The findings required by subpart A of part 51 of this chapter
have been made.
(2) The Commission may also find, at the time it issues the combined
license, that certain acceptance criteria in one or more of the
inspections, tests, analyses, and acceptance criteria (ITAAC) in a
referenced early site permit or standard design certification have been
met. This finding will finally resolve that those acceptance criteria
have been met, those acceptance criteria will be deemed to be excluded
from the combined license, and findings under Sec. 52.103(g) with
respect to those acceptance criteria are unnecessary.
(b) The Commission shall identify within the combined license the
inspections, tests, and analyses, including those applicable to
emergency planning, that the licensee shall perform, and the acceptance
criteria that, if met, are necessary and sufficient to provide
reasonable assurance that the facility has been constructed and will be
operated in conformity with the license, the provisions of the Act, and
the Commission's rules and regulations.
(c) A combined license shall contain the terms and conditions,
including technical specifications, as the Commission deems necessary
and appropriate.
Sec. 52.98 Finality of combined licenses; information requests.
(a) After issuance of a combined license, the Commission may not
modify, add, or delete any term or condition of the combined license,
the design of the facility, the inspections, tests, analyses, and
acceptance criteria contained in the license which are not derived from
a referenced standard design certification or manufacturing license,
except in accordance with the provisions of Sec. 52.103 or Sec. 50.109
of this chapter, as applicable.
(b) If the combined license does not reference a design
certification or a reactor manufactured under a subpart F of this part
manufacturing license, then 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) under the applicable
change processes in 10 CFR part 50 (e.g., Sec. Sec. 50.54, 50.59, or
50.90 of this chapter).
(c) If the combined license references a certified design, then--
(1) Changes to or departures from information within the scope of
the referenced design certification rule are subject to the applicable
change processes in that rule; and
(2) Changes that are not within the scope of the referenced design
certification rule are subject to the applicable change processes in 10
CFR part 50, unless they also involve changes to or noncompliance with
information within the scope of the referenced design certification
rule. In these cases, the applicable provisions of this section and the
design certification rule apply.
(d) If the combined license references a reactor manufactured under
a subpart F of this part manufacturing license, then--
(1) Changes to or departures from information within the scope of
the manufactured reactor's design are subject to the change processes in
Sec. 52.171; and
(2) Changes that are not within the scope of the manufactured
reactor's design are subject to the applicable change processes in 10
CFR part 50.
(e) The Commission may issue and make immediately effective any
amendment to a combined license upon a determination by the Commission
that the amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person. The amendment may be issued and made
immediately effective in advance of the holding and completion of any
required hearing. The amendment will be processed in accordance with the
procedures specified in 10 CFR 50.91.
(f) Any modification to, addition to, or deletion from the terms and
conditions of a combined license, including
[[Page 102]]
any modification to, addition to, or deletion from the inspections,
tests, analyses, or related acceptance criteria contained in the license
is a proposed amendment to the license. There must be an opportunity for
a hearing on the amendment.
(g) Except for information sought to verify licensee compliance with
the current licensing basis for that facility, information requests to
the holder of a combined license must be evaluated before issuance to
ensure that the burden to be imposed on the licensee is justified in
view of the potential safety significance of the issue to be addressed
in the requested information. Each evaluation performed by the NRC staff
must be in accordance with 10 CFR 50.54(f) and must be approved by the
Executive Director for Operations or his or her designee before issuance
of the request.
Sec. 52.99 Inspection during construction.
(a) The licensee shall submit to the NRC, no later that 1 year after
issuance of the combined license or at the start of construction as
defined in 10 CFR 50.10(a), whichever is later, its schedule for
completing the inspections, tests, or analyses in the ITAAC. The
licensee shall submit updates to the ITAAC schedules every 6 months
thereafter and, within 1 year of its scheduled date for initial loading
of fuel, the licensee shall submit updates to the ITAAC schedule every
30 days until the final notification is provided to the NRC under
paragraph (c)(1) of this section.
(b) With respect to activities subject to an ITAAC, an applicant for
a combined license may proceed at its own risk with design and
procurement activities, and a licensee may proceed at its own risk with
design, procurement, construction, and pre-operational activities, even
though the NRC may not have found that any one of the prescribed
acceptance criteria have been met.
(c)(1) The licensee shall notify the NRC that the prescribed
inspections, tests, and analyses have been performed and that the
prescribed acceptance criteria have been met. The notification must
contain sufficient information to demonstrate that the prescribed
inspections, tests, and analyses have been performed and that the
prescribed acceptance criteria have been met.
(2) If the licensee has not provided, by the date 225 days before
the scheduled date for initial loading of fuel, the notification
required by paragraph (c)(1) of this section for all ITAAC, then the
licensee shall notify the NRC that the prescribed inspections, tests, or
analyses for all uncompleted ITAAC will be performed and that the
prescribed acceptance criteria will be met prior to operation. The
notification must be provided no later than the date 225 days before the
scheduled date for initial loading of fuel, and must provide sufficient
information to demonstrate that the prescribed inspections, tests, or
analyses will be performed and the prescribed acceptance criteria for
the uncompleted ITAAC will be met, including, but not limited to, a
description of the specific procedures and analytical methods to be used
for performing the prescribed inspections, tests, and analyses and
determining that the prescribed acceptance criteria have been met.
(d)(1) In the event that an activity is subject to an ITAAC derived
from a referenced standard design certification and the licensee has not
demonstrated that the ITAAC has been met, the licensee may take
corrective actions to successfully complete that ITAAC or request an
exemption from the standard design certification ITAAC, as applicable. A
request for an exemption must also be accompanied by a request for a
license amendment under Sec. 52.98(f).
(2) In the event that an activity is subject to an ITAAC not derived
from a referenced standard design certification and the licensee has not
demonstrated that the ITAAC has been met, the licensee may take
corrective actions to successfully complete that ITAAC or request a
license amendment under Sec. 52.98(f).
(e) The NRC shall ensure that the prescribed inspections, tests, and
analyses in the ITAAC are performed.
(1) At appropriate intervals until the last date for submission of
requests for hearing under Sec. 52.103(a), the NRC shall
[[Page 103]]
publish notices in the Federal Register of the NRC staff's determination
of the successful completion of inspections, tests, and analyses.
(2) The NRC shall make publicly available the licensee notifications
under paragraph (c)(1), and, no later than the date of publication of
the notice of intended operation required by Sec. 52.103(a), make
available all licensee notifications under paragraphs (c)(1) and (c)(2)
of this section.
[72 FR 49517, Aug. 28, 2007, as amended at 72 FR 57447, Oct. 9, 2007]
Sec. 52.103 Operation under a combined license.
(a) The licensee shall notify the NRC of its scheduled date for
initial loading of fuel no later than 270 days before the scheduled date
and shall notify the NRC of updates to its schedule every 30 days
thereafter. Not less than 180 days before the date scheduled for initial
loading of fuel into a plant by a licensee that has been issued a
combined license under this part, the Commission shall publish notice of
intended operation in the Federal Register. The notice must provide that
any person whose interest may be affected by operation of the plant may,
within 60 days, request that the Commission hold a hearing on whether
the facility as constructed complies, or on completion will comply, with
the acceptance criteria in the combined license, except that a hearing
shall not be granted for those ITAAC which the Commission found were met
under Sec. 52.97(a)(2).
(b) A request for hearing under paragraph (a) of this section must
show, prima facie, that--
(1) One or more of the acceptance criteria of the ITAAC in the
combined license have not been, or will not be, met; and
(2) The specific operational consequences of nonconformance that
would be contrary to providing reasonable assurance of adequate
protection of the public health and safety.
(c) The Commission, acting as the presiding officer, shall determine
whether to grant or deny the request for hearing in accordance with the
applicable requirements of 10 CFR 2.309. If the Commission grants the
request, the Commission, acting as the presiding officer, shall
determine whether during a period of interim operation there will be
reasonable assurance of adequate protection to the public health and
safety. The Commission's determination must consider the petitioner's
prima facie showing and any answers thereto. If the Commission
determines there is such reasonable assurance, it shall allow operation
during an interim period under the combined license.
(d) The Commission, in its discretion, shall determine appropriate
hearing procedures, whether informal or formal adjudicatory, for any
hearing under paragraph (a) of this section, and shall state its reasons
therefore.
(e) The Commission shall, to the maximum possible extent, render a
decision on issues raised by the hearing request within 180 days of the
publication of the notice provided by paragraph (a) of this section or
by the anticipated date for initial loading of fuel into the reactor,
whichever is later.
(f) A petition to modify the terms and conditions of the combined
license will be processed as a request for action in accordance with 10
CFR 2.206. The petitioner shall file the petition with the Secretary of
the Commission. Before the licensed activity allegedly affected by the
petition (fuel loading, low power testing, etc.) commences, the
Commission shall determine whether any immediate action is required. If
the petition is granted, then an appropriate order will be issued. Fuel
loading and operation under the combined license will not be affected by
the granting of the petition unless the order is made immediately
effective.
(g) The licensee shall not operate the facility until the Commission
makes a finding that the acceptance criteria in the combined license are
met, except for those acceptance criteria that the Commission found were
met under Sec. 52.97(a)(2). If the combined license is for a modular
design, each reactor module may require a separate finding as
construction proceeds.
(h) After the Commission has made the finding in paragraph (g) of
this section, the ITAAC do not, by virtue of their inclusion in the
combined license,
[[Page 104]]
constitute regulatory requirements either for licensees or for renewal
of the license; except for the specific ITAAC for which the Commission
has granted a hearing under paragraph (a) of this section, all ITAAC
expire upon final Commission action in the proceeding. However,
subsequent changes to the facility or procedures described in the final
safety analysis report (as updated) must comply with the requirements in
Sec. Sec. 52.98(e) or (f), as applicable.
Sec. 52.104 Duration of combined license.
A combined license is issued for a specified period not to exceed 40
years from the date on which the Commission makes a finding that
acceptance criteria are met under Sec. 52.103(g) or allowing operation
during an interim period under the combined license under Sec.
52.103(c).
Sec. 52.105 Transfer of combined license.
A combined license may be transferred in accordance with Sec. 50.80
of this chapter.
Sec. 52.107 Application for renewal.
The filing of an application for a renewed license must be in
accordance with 10 CFR part 54.
Sec. 52.109 Continuation of combined license.
Each combined license for a facility that has permanently ceased
operations, continues in effect beyond the expiration date to authorize
ownership and possession of the production or utilization facility,
until the Commission notifies the licensee in writing that the license
is terminated. During this period of continued effectiveness the
licensee shall--
(1) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control and maintenance of the spent fuel, in a
safe condition; and
(2) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the combined license for the facility.
Sec. 52.110 Termination of license.
(a)(1) When a licensee has determined to permanently cease
operations the licensee shall, within 30 days, submit a written
certification to the NRC, consistent with the requirements of Sec.
52.3(b)(8);
(2) Once fuel has been permanently removed from the reactor vessel,
the licensee shall submit a written certification to the NRC that meets
the requirements of Sec. 52.3(b)(9); and
(3) For licensees whose licenses have been permanently modified to
allow possession but not operation of the facility, before September 27,
2007, the certification required in paragraph (a)(1) of this section
shall be deemed to have been submitted.
(b) Upon docketing of the certifications for permanent cessation of
operations and permanent removal of fuel from the reactor vessel, or
when a final legally effective order to permanently cease operations has
come into effect, the 10 CFR part 52 license no longer authorizes
operation of the reactor or emplacement or retention of fuel into the
reactor vessel.
(c) Decommissioning will be completed within 60 years of permanent
cessation of operations. Completion of decommissioning beyond 60 years
will be approved by the Commission only when necessary to protect public
health and safety. Factors that will be considered by the Commission in
evaluating an alternative that provides for completion of
decommissioning beyond 60 years of permanent cessation of operations
include unavailability of waste disposal capacity and other site-
specific factors affecting the licensee's capability to carry out
decommissioning, including presence of other nuclear facilities at the
site.
(d)(1) Before or within 2 years following permanent cessation of
operations, the licensee shall submit a post-shutdown decommissioning
activities report (PSDAR) to the NRC, and a copy to the affected
State(s). The report must include a description of the planned
decommissioning activities along with a schedule for their
accomplishment, an estimate of expected costs, and a discussion that
provides the reasons for concluding that the environmental impacts
associated with
[[Page 105]]
site-specific decommissioning activities will be bounded by appropriate
previously issued environmental impact statements.
(2) The NRC shall notice receipt of the PSDAR and make the PSDAR
available for public comment. The NRC shall also schedule a public
meeting in the vicinity of the licensee's facility upon receipt of the
PSDAR. The NRC shall publish a document in the Federal Register and in a
forum, such as local newspapers, that is readily accessible to
individuals in the vicinity of the site, announcing the date, time and
location of the meeting, along with a brief description of the purpose
of the meeting.
(e) Licensees shall not perform any major decommissioning
activities, as defined in Sec. 50.2 of this chapter, until 90 days
after the NRC has received the licensee's PSDAR submittal and until
certifications of permanent cessation of operations and permanent
removal of fuel from the reactor vessel, as required under Sec.
52.110(a)(1), have been submitted.
(f) Licensees shall not perform any decommissioning activities, as
defined in Sec. 52.1, that--
(1) Foreclose release of the site for possible unrestricted use;
(2) Result in significant environmental impacts not previously
reviewed; or
(3) Result in there no longer being reasonable assurance that
adequate funds will be available for decommissioning.
(g) In taking actions permitted under Sec. 50.59 of this chapter
following submittal of the PSDAR, the licensee shall notify the NRC in
writing and send a copy to the affected State(s), before performing any
decommissioning activity inconsistent with, or making any significant
schedule change from, those actions and schedules described in the
PSDAR, including changes that significantly increase the decommissioning
cost.
(h)(1) Decommissioning trust funds may be used by licensees if--
(i) The withdrawals are for expenses for legitimate decommissioning
activities consistent with the definition of decommissioning in Sec.
52.1;
(ii) The expenditure would not reduce the value of the
decommissioning trust below an amount necessary to place and maintain
the reactor in a safe storage condition if unforeseen conditions or
expenses arise and;
(iii) The withdrawals would not inhibit the ability of the licensee
to complete funding of any shortfalls in the decommissioning trust
needed to ensure the availability of funds to ultimately release the
site and terminate the license.
(2) Initially, 3 percent of the generic amount specified in Sec.
50.75 of this chapter may be used for decommissioning planning. For
licensees that have submitted the certifications required under Sec.
52.110(a) and commencing 90 days after the NRC has received the PSDAR,
an additional 20 percent may be used. A site-specific decommissioning
cost estimate must be submitted to the NRC before the licensee may use
any funding in excess of these amounts.
(3) Within 2 years following permanent cessation of operations, if
not already submitted, the licensee shall submit a site-specific
decommissioning cost estimate.
(4) For decommissioning activities that delay completion of
decommissioning by including a period of storage or surveillance, the
licensee shall provide a means of adjusting cost estimates and
associated funding levels over the storage or surveillance period.
(i) All power reactor licensees must submit an application for
termination of license. The application for termination of license must
be accompanied or preceded by a license termination plan to be submitted
for NRC approval.
(1) The license termination plan must be a supplement to the FSAR or
equivalent and must be submitted at least 2 years before termination of
the license date.
(2) The license termination plan must include--
(i) A site characterization;
(ii) Identification of remaining dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final radiation survey;
(v) A description of the end use of the site, if restricted;
[[Page 106]]
(vi) An updated site-specific estimate of remaining decommissioning
costs;
(vii) A supplement to the environmental report, under Sec. 51.53 of
this chapter, describing any new information or significant
environmental change associated with the licensee's proposed termination
activities; and
(viii) Identification of parts, if any, of the facility or site that
were released for use before approval of the license termination plan.
(3) The NRC shall notice receipt of the license termination plan and
make the license termination plan available for public comment. The NRC
shall also schedule a public meeting in the vicinity of the licensee's
facility upon receipt of the license termination plan. The NRC shall
publish a document in the Federal Register and in a forum, such as local
newspapers, which is readily accessible to individuals in the vicinity
of the site, announcing the date, time and location of the meeting,
along with a brief description of the purpose of the meeting.
(j) If the license termination plan demonstrates that the remainder
of decommissioning activities will be performed in accordance with the
regulations in this chapter, will not be inimical to the common defense
and security or to the health and safety of the public, and will not
have a significant effect on the quality of the environment and after
notice to interested persons, the Commission shall approve the plan, by
license amendment, subject to terms and conditions as it deems
appropriate and necessary and authorize implementation of the license
termination plan.
(k) The Commission shall terminate the license if it determines
that--
(1) The remaining dismantlement has been performed in accordance
with the approved license termination plan; and
(2) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E to 10 CFR part 20.
(l) For a facility that has permanently ceased operation before the
expiration of its license, the collection period for any shortfall of
funds will be determined, upon application by the licensee, on a case-
by-case basis taking into account the specific financial situation of
each licensee.
Subpart D [Reserved]
Subpart E_Standard Design Approvals
Sec. 52.131 Scope of subpart.
This subpart sets out procedures for the filing, NRC staff review,
and referral to the Advisory Committee on Reactor Safeguards of standard
designs for a nuclear power reactor of the type described in Sec. 50.22
of this chapter or major portions thereof.
Sec. 52.133 Relationship to other subparts.
(a) This subpart applies to a person that requests a standard design
approval from the NRC staff separately from an application for a
construction permit filed under 10 CFR part 50 or a combined license
filed under subpart C of this part. An applicant for a construction
permit or combined license may reference a standard design approval.
(b) Subpart B of this part governs the certification by rulemaking
of the design of a nuclear power plant. Subpart B may be used
independently of the provisions in this subpart.
(c) Subpart F of this part governs the issuance of licenses to
manufacture nuclear power reactors to be installed and operated at sites
not identified in the manufacturing license application. Subpart F of
this part may be used independently of the provisions in this subpart.
Sec. 52.135 Filing of applications.
(a) Any person may submit a proposed standard design for a nuclear
power reactor of the type described in 10 CFR 50.22 to the NRC staff for
its review. The submittal may consist of either the final design for the
entire facility or the final design of major portions thereof.
(b) The submittal for review of the proposed standard design must be
made
[[Page 107]]
in the same manner and in the same number of copies as provided in 10
CFR 50.30 and 52.3 for license applications.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.136 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d) and (j).
Sec. 52.137 Contents of applications; technical information.
If the applicant seeks review of a major portion of a standard
design, the application need only contain the information required by
this section to the extent the requirements are applicable to the major
portion of the standard design for which NRC staff approval is sought.
(a) The application must contain a final safety analysis report that
describes the facility, presents the design bases and the limits on its
operation, and presents a safety analysis of the structures, systems,
and components and of the facility, or major portion thereof, and must
include the following information:
(1) The site parameters postulated for the design, and an analysis
and evaluation of the design in terms of those site parameters;
(2) A description and analysis of the SSCs of the facility, with
emphasis upon performance requirements, the bases, with technical
justification, upon which the requirements have been established, and
the evaluations required to show that safety functions will be
accomplished. It is expected that the standard plant will reflect
through its design, construction, and operation an extremely low
probability for accidents that could result in the release of
significant quantities of radioactive fission products. The description
shall be sufficient to permit understanding of the system designs and
their relationship to the safety evaluations. Items such as the reactor
core, reactor coolant system, instrumentation and control systems,
electrical systems, containment system, other engineered safety
features, auxiliary and emergency systems, power conversion systems,
radioactive waste handling systems, and fuel handling systems shall be
discussed insofar as they are pertinent. The following power reactor
design characteristics will be taken into consideration by the
Commission:
(i) Intended use of the reactor including the proposed maximum power
level and the nature and inventory of contained radioactive materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials; and
(iv) The safety features that are to be engineered into the facility
and those barriers that must be breached as a result of an accident
before a release of radioactive material to the environment can occur.
Special attention must be directed to plant design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \9\ from
the core into the containment assuming that the facility is operated at
the ultimate power level contemplated. The applicant shall perform an
evaluation and analysis of the postulated fission product release, using
the expected demonstrable containment leak rate and any fission product
cleanup systems intended to mitigate the consequences of the accidents,
together with applicable postulated site parameters, including site
meteorology, to evaluate the offsite radiological consequences. The
evaluation must determine that:
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\9\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. These accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
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(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission
[[Page 108]]
product release, would not receive a radiation dose in excess of 25 rem
\10\ total effective dose equivalent (TEDE); and
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\10\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose for
radiation workers which, according to NCRP recommendations at the time
could be disregarded in the determination of their radiation exposure
status (see NBS Handbook 69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an acceptable limit for
an emergency dose to the public under accident conditions. Rather, this
dose value has been set forth in this section as a reference value,
which can be used in the evaluation of plant design features with
respect to postulated reactor accidents, to assure that these designs
provide assurance of low risk of public exposure to radiation, in the
event of an accident.
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(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period of
its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(3) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to 10
CFR part 50, general design criteria (GDC), establishes minimum
requirements for the principal design criteria for water-cooled nuclear
power plants similar in design and location to plants for which
construction permits have previously been issued by the Commission and
provides guidance to applicants in establishing principal design
criteria for other types of nuclear power units;
(ii) The design bases and the relation of the design bases to the
principal design criteria; and
(iii) Information relative to materials of construction, general
arrangement, and approximate dimensions, sufficient to provide
reasonable assurance that the design will conform to the design bases
with adequate margin for safety;
(4) An analysis and evaluation of the design and performance of SSC
with the objective of assessing the risk to public health and safety
resulting from operation of the facility and including determination of
the margins of safety during normal operations and transient conditions
anticipated during the life of the facility, and the adequacy of SSCs
provided for the prevention of accidents and the mitigation of the
consequences of accidents. Analysis and evaluation of ECCS cooling
performance and the need for high-point vents following postulated loss-
of-coolant accidents shall be performed in accordance with the
requirements of 10 CFR 50.46 and 50.46a;
(5) The kinds and quantities of radioactive materials expected to be
produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(6) The information required by Sec. 20.1406 of this chapter;
(7) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(8) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and
(f)(3)(v) of 10 CFR 50.34(f);
(9) For applications for light-water-cooled nuclear power plants, an
evaluation of the standard plant design against the Standard Review Plan
(SRP) revision in effect 6 months before the docket date of the
application. The evaluation required by this section shall include an
identification and description of all differences in design features,
analytical techniques, and procedural measures proposed for the design
and those corresponding features, techniques, and measures given in the
SRP acceptance criteria. Where a difference exists, the evaluation shall
discuss how the proposed alternative provides an acceptable method of
complying with the Commission's regulations, or portions thereof, that
underlie the corresponding SRP acceptance criteria. The SRP is not a
substitute for the regulations, and compliance is not a requirement;
(10) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced
[[Page 109]]
during normal reactor operations described in 10 CFR 50.34a(e);
(11) The information pertaining to design features that affect plans
for coping with emergencies in the operation of the reactor facility or
a major portion thereof;
(12) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(13) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(14) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in 10 CFR
50.60 and 50.61;
(15) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62;
(16) The coping analysis, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(17) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(18) A description and analysis of the fire protection design
features for the standard plant necessary to comply with part 50,
appendix A, GDC 3, and Sec. 50.48 of this chapter;
(19) A description of the quality assurance program applied to the
design of the SSCs of the facility. Appendix B to 10 CFR part 50,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth the requirements for quality assurance
programs for nuclear power plants. The description of the quality
assurance program for a nuclear power plant shall include a discussion
of how the applicable requirements of appendix B to 10 CFR part 50 were
satisfied;
(20) The information necessary to demonstrate that the standard
plant complies with the earthquake engineering criteria in 10 CFR part
50, appendix S;
(21) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(22) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection, hydrogen
combustion, and containment bypass;
(24) A description, analysis, and evaluation of the interfaces
between the standard design and the balance of the nuclear power plant;
and
(25) A description of the design-specific probabilistic risk
assessment and its results.
(26) For applications for standard design approvals which are
subject to 10 CFR 50.150(a), the information required by 10 CFR
50.150(b).
(b) An application for approval of a standard design, which differs
significantly from the light-water reactor designs of plants that have
been licensed and in commercial operation before April 18, 1989, or uses
simplified, inherent, passive, or other innovative means to accomplish
its safety functions, must meet the requirements of 10 CFR 50.43(e).
[72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]
Sec. 52.139 Standards for review of applications.
Applications filed under this subpart will be reviewed for
compliance with the standards set out in 10 CFR parts 20, 50 and its
appendices, and 10 CFR parts 73 and 100.
Sec. 52.141 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which concern
safety.
[[Page 110]]
Sec. 52.143 Staff approval of design.
Upon completion of its review of a submittal under this subpart and
receipt of a report by the Advisory Committee on Reactor Safeguards
under Sec. 52.141 of this subpart, the NRC staff shall publish a
determination in the Federal Register as to whether or not the design is
acceptable, subject to appropriate terms and conditions, and make an
analysis of the design in the form of a report available at the NRC Web
site, http://www.nrc.gov.
Sec. 52.145 Finality of standard design approvals; information requests.
(a) An approved design must be used by and relied upon by the NRC
staff and the ACRS in their review of any individual facility license
application that incorporates by reference a standard design approved in
accordance with this paragraph unless there exists significant new
information that substantially affects the earlier determination or
other good cause.
(b) The determination and report by the NRC staff do not constitute
a commitment to issue a permit or license, or in any way affect the
authority of the Commission, Atomic Safety and Licensing Board Panel, or
presiding officers in any proceeding under part 2 of this chapter.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of the standard design approval,
information requests to the holder of a standard design approval must be
evaluated before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance of
the issue to be addressed in the requested information. Each evaluation
performed by the NRC staff must be in accordance with 10 CFR 50.54(f)
and must be approved by the Executive Director for Operations or his or
her designee before issuance of the request.
Sec. 52.147 Duration of design approval.
A standard design approval issued under this subpart is valid for 15
years from the date of issuance and may not be renewed. A design
approval continues to be valid beyond the date of expiration in any
proceeding on an application for a construction permit or an operating
license under part 50 or a combined license or manufacturing license
under part 52 that references the final design approval and is docketed
before the date of expiration of the design approval.
Subpart F_Manufacturing Licenses
Sec. 52.151 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of a license authorizing manufacture of nuclear
power reactors to be installed at sites not identified in the
manufacturing license application.
Sec. 52.153 Relationship to other subparts.
(a) A nuclear power reactor manufactured under a manufacturing
license issued under this subpart may only be transported to and
installed at a site for which either a construction permit under part 50
of this chapter or a combined license under subpart C of this part has
been issued.
(b) Subpart B of this part governs the certification by rulemaking
of the design of standard nuclear power facilities. Subpart E of this
part governs the NRC staff review and approval of standard designs for a
nuclear power facility. A manufacturing license applicant may reference
a standard design certification or a standard design approval in its
application. These subparts may also be used independently of the
provisions in this subpart.
Sec. 52.155 Filing of applications.
(a) Any person, except one excluded by 10 CFR 50.38, may file an
application for a manufacturing license under this subpart with the
Director of New Reactors or the Director of Nuclear Reactor Regulation,
as appropriate.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
[[Page 111]]
Sec. 52.156 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d), and (j).
Sec. 52.157 Contents of applications; technical information in final
safety analysis report.
The application must contain a final safety analysis report
containing the information set forth below, with a level of design
information sufficient to enable the Commission to judge the applicant's
proposed means of assuring that the manufacturing conforms to the design
and to reach a final conclusion on all safety questions associated with
the design, permit the preparation of construction and installation
specifications by an applicant who seeks to use the manufactured
reactor, and permit the preparation of acceptance and inspection
requirements by the NRC:
(a) The principal design criteria for the reactor to be
manufactured. Appendix A of 10 CFR part 50, ``General Design Criteria
for Nuclear Power Plants,'' establishes minimum requirements for the
principal design criteria for water-cooled nuclear power plants similar
in design and location to plants for which construction permits have
previously been issued by the Commission and provides guidance to
applicants in establishing principal design criteria for other types of
nuclear power units;
(b) The design bases and the relation of the design bases to the
principal design criteria;
(c) A description and analysis of the structures, systems, and
components of the reactor to be manufactured, with emphasis upon the
materials of manufacture, performance requirements, the bases, with
technical justification therefor, upon which the performance
requirements have been established, and the evaluations required to show
that safety functions will be accomplished. The description shall be
sufficient to permit understanding of the system designs and their
relationship to safety evaluations. Items such as the reactor core,
reactor coolant system, instrumentation and control systems, electrical
systems, containment system, other engineered safety features, auxiliary
and emergency systems, power conversion systems, radioactive waste
handling systems, and fuel handling systems shall be discussed insofar
as they are pertinent. The following power reactor design
characteristics will be taken into consideration by the Commission:
(1) Intended use of the manufactured reactor including the proposed
maximum power level and the nature and inventory of contained
radioactive materials;
(2) The extent to which generally accepted engineering standards are
applied to the design of the reactor; and
(3) The extent to which the reactor incorporates unique, unusual or
enhanced safety features having a significant bearing on the probability
or consequences of accidental release of radioactive materials;
(d) The safety features that are engineered into the reactor and
those barriers that must be breached as a result of an accident before a
release of radioactive material to the environment can occur. Special
attention must be directed to reactor design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \11\
from the core into the containment assuming that the facility is
operated at the ultimate power level contemplated. The applicant shall
perform an evaluation and analysis of the postulated fission product
release, using the expected demonstrable containment leak rate and any
fission product cleanup systems intended to mitigate the consequences of
the accidents, together with applicable postulated site parameters,
including site meteorology, to evaluate the offsite radiological
consequences. The evaluation must determine that:
---------------------------------------------------------------------------
\11\ The fission product release assumed for this evaluation should
be based upon a major accident, hypothesized for purposes of site
analysis or postulated from considerations of possible accidental
events. These accidents have generally been assumed to result in
substantial meltdown of the core with subsequent release into the
containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(1) An individual located at any point on the boundary of the
exclusion area
[[Page 112]]
for any 2 hour period following the onset of the postulated fission
product release, would not receive a radiation dose in excess of 25 rem
\12\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------
\12\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose for
radiation workers which, according to NCRP recommendations at the time
could be disregarded in the determination of their radiation exposure
status (see NBS Handbook 69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an acceptable limit for
an emergency dose to the public under accident conditions. Rather, this
dose value has been set forth in this section as a reference value,
which can be used in the evaluation of plant design features with
respect to postulated reactor accidents, to assure that these designs
provide assurance of low risk of public exposure to radiation, in the
event of an accident.
---------------------------------------------------------------------------
(2) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period of
its passage) would not receive a radiation dose in excess of 25 rem
TEDE; and
(e) The kinds and quantities of radioactive materials expected to be
produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter.
(f) Information necessary to establish that the design of the
reactor to be manufactured complies with the technical requirements in
10 CFR Chapter I, including:
(1) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and components
provided for the prevention of accidents and the mitigation of the
consequences of accidents. Analysis and evaluation of ECCS cooling
performance and the need for high-point vents following postulated loss-
of-coolant accidents shall be performed in accordance with the
requirements of Sec. Sec. 50.46 and 50.46a of this chapter;
(2) A description and analysis of the fire protection design
features for the reactor necessary to comply with 10 CFR part 50,
appendix A, GDC 3 and Sec. 50.48 of this chapter;
(3) A description of protection provided against pressurized thermal
shock events, including projected values of the reference temperature
for reactor vessel beltline materials as defined in Sec. Sec. 50.60 and
50.61 of this chapter;
(4) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(5) The coping analysis, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(6) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(7) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62;
(8) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(9) The information required by Sec. 20.1406 of this chapter;
(10) [Reserved]
(11) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations, as described in
Sec. 50.34a(e) of this chapter;
(12) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in Sec. 50.34(f) of this chapter, except paragraphs (f)(1)(xii),
(f)(2)(ix), and (f)(3)(v);
(13) If the applicant seeks to use risk-informed treatment of SSCs
in accordance with Sec. 50.69 of this chapter, the information required
by Sec. 50.69(b)(2) of this chapter;
[[Page 113]]
(14) The information necessary to demonstrate that the manufactured
reactor complies with the earthquake engineering criteria in appendix S
to 10 CFR part 50;
(15) Information sufficient to demonstrate compliance with the
applicable requirements regarding testing, analysis, and prototypes as
set forth in Sec. 50.43(e) of this chapter;
(16) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(17) A description of the quality assurance program applied to the
design, and to be applied to the manufacture of, the structures,
systems, and components of the reactor. Appendix B to 10 CFR part 50,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth the requirements for quality assurance
programs for nuclear power plants. The description of the quality
assurance program must include a discussion of how the applicable
requirements of appendix B to 10 CFR part 50 have been and will be
satisfied; and
(18) Proposed technical specifications applicable to the reactor
being manufactured, prepared in accordance with the requirements of
Sec. Sec. 50.36 and 50.36a of this chapter;
(19) The site parameters postulated for the design, and an analysis
and evaluation of the reactor design in terms of those site parameters;
(20) The interface requirements between the manufactured reactor and
the remaining portions of the nuclear power plant. These requirements
must be sufficiently detailed to allow for completion of the final
safety analysis;
(21) Justification that compliance with the interface requirements
of paragraph (f)(20) of this section is verifiable through inspections,
testing, or analysis. The method to be used for verification of
interface requirements must be included as part of the proposed ITAAC
required by Sec. 52.158(a);
(22) A representative conceptual design for a nuclear power facility
using the manufactured reactor, to aid the NRC in its review of the
final safety analysis required by this section and to permit assessment
of the adequacy of the interface requirements in paragraph (f)(20) of
this section;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection, hydrogen
combustion, and containment bypass;
(24) [Reserved]
(25) If the reactor is to be used in modular plant design, a
description of the possible operating configurations of the reactor
modules with common systems, interface requirements, and system
interactions. The final safety analysis must also account for
differences among the configurations, including any restrictions that
will be necessary during the construction and startup of a given module
to ensure the safe operation of any module already operating;
(26) A description of the management plan for design and
manufacturing activities, including:
(i) The organizational and management structure singularly
responsible for direction of design and manufacture of the reactor;
(ii) Technical resources directed by the applicant, and the
qualifications requirements;
(iii) Details of the interaction of design and manufacture within
the applicant's organization and the manner by which the applicant will
ensure close integration of the architect engineer and the nuclear steam
supply vendor, as applicable;
(iv) Proposed procedures governing the preparation of the
manufactured reactor for shipping to the site where it is to be
operated, the conduct of shipping, and verifying the condition of the
manufactured reactor upon receipt at the site; and
(v) The degree of top level management oversight and technical
control to be exercised by the applicant during design and manufacture,
including the preparation and implementation of procedures necessary to
guide the effort;
(27) Necessary parameters to be used in developing plans for
preoperational testing and initial operation;
[[Page 114]]
(28) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(29) The information necessary to demonstrate how operating
experience insights have been incorporated into the manufactured reactor
design;
(30) For applications for light-water-cooled nuclear power plants,
an evaluation of the design to be manufactured against the Standard
Review Plan (SRP) revision in effect 6 months before the docket date of
the application. The evaluation required by this section shall include
an identification and description of all differences in design features,
analytical techniques, and procedural measures proposed for the design
and those corresponding features, techniques, and measures given in the
SRP acceptance criteria. Where a difference exists, the evaluation shall
discuss how the proposed alternative provides an acceptable method of
complying with the Commission's regulations, or portions thereof, that
underlie the corresponding SRP acceptance criteria. The SRP is not a
substitute for the regulations, and compliance is not a requirement; and
(31) A description of the design-specific probabilistic risk
assessment and its results.
(32) For applications for manufacturing licenses which are subject
to 10 CFR 50.150(a), the information required by 10 CFR 50.150(b).
[72 FR 49517, Aug. 28, 2007, as amended at 74 FR 28147, June 12, 2009]
Sec. 52.158 Contents of application; additional technical information.
The application must contain:
(a)(1) Inspections, tests, analyses, and acceptance criteria
(ITAAC). The proposed inspections, tests, and analyses that the licensee
who will be operating the reactor shall perform, and the acceptance
criteria that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met:
(i) The reactor has been manufactured in conformity with the
manufacturing license; the provisions of the Act, and the Commission's
rules and regulations; and
(ii) The manufactured reactor will be operated in conformity with
the approved design and any license authorizing operation of the
manufactured reactor.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those portions
of the facility design which are covered by the design certification.
(3) If the application references a standard design certification,
the application may include a notification that a required inspection,
test, or analysis in the design certification ITAAC has been
successfully completed and that the corresponding acceptance criterion
has been met. The Federal Register notification required by Sec. 52.163
must indicate that the application includes this notification.
(b)(1) An environmental report as required by 10 CFR 51.54.
(2) If the manufacturing license application references a standard
design certification, the environmental report need not contain a
discussion of severe accident mitigation design alternatives for the
reactor.
Sec. 52.159 Standards for review of application.
Applications filed under this subpart will be reviewed according to
the applicable standards set out in 10 CFR parts 20, 50 and its
appendices, 51, 73, and 100 and its appendices.
Sec. 52.161 [Reserved]
Sec. 52.163 Administrative review of applications; hearings.
A proceeding on a manufacturing license is subject to all applicable
procedural requirements contained in 10 CFR part 2, including the
requirements for docketing in Sec. 2.101(a)(1) through (4) of this
chapter, and the requirements for issuance of a notice of proposed
action in Sec. 2.105 of this chapter, provided, however, that the
designated sections may not be construed to require that the
environmental report or draft or
[[Page 115]]
final environmental impact statement include an assessment of the
benefits of constructing and/or operating the manufactured reactor or an
evaluation of alternative energy sources. All hearings on manufacturing
licenses are governed by the hearing procedures contained in 10 CFR part
2, subparts C, G, L, and N.
Sec. 52.165 Referral to the Advisory Committee on Reactor Safeguards (ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which concern
safety.
Sec. 52.167 Issuance of manufacturing license.
(a) After completing any hearing under Sec. 52.163, and receiving
the report submitted by the ACRS, the Commission may issue a
manufacturing license if the Commission finds that:
(1) Applicable standards and requirements of the Act and the
Commission's regulations have been met;
(2) There is reasonable assurance that the reactor(s) will be
manufactured, and can be transported, incorporated into a nuclear power
plant, and operated in conformity with the manufacturing license, the
provision of the Act, and the Commission's regulations;
(3) The proposed reactor(s) can be incorporated into a nuclear power
plant and operated at sites having characteristics that fall within the
site parameters postulated for the design of the manufactured reactor(s)
without undue risk to the health and safety of the public;
(4) The applicant is technically qualified to design and manufacture
the proposed nuclear power reactor(s);
(5) The proposed inspections, tests, analyses and acceptance
criteria are necessary and sufficient, within the scope of the
manufacturing license, to provide reasonable assurance that the
manufactured reactor has been manufactured and will be operated in
conformity with the license, the provisions of the Act, and the
Commission's regulations;
(6) The issuance of a license to the applicant will not be inimical
to the common defense and security or to the health and safety of the
public; and
(7) The findings required by subpart A of part 51 of this chapter
have been made.
(b) Each manufacturing license issued under this subpart shall
specify:
(1) Terms and conditions as the Commission deems necessary and
appropriate;
(2) Technical specifications for operation of the manufactured
reactor, as the Commission deems necessary and appropriate;
(3) Site parameters and design characteristics for the manufactured
reactor; and
(4) The interface requirements to be met by the site-specific
elements of the facility, such as the service water intake structure and
the ultimate heat sink, not within the scope of the manufactured
reactor.
(c)(1) A holder of a manufacturing license may not transport or
allow to be removed from the place of manufacture the manufactured
reactor except to the site of a licensee with either a construction
permit under part 50 of this chapter or a combined license under subpart
C of this part. The construction permit or combined license must
authorize the construction of a nuclear power facility using the
manufactured reactor(s).
(2) A holder of a manufacturing license shall include, in any
contract governing the transport of a manufactured reactor from the
place of manufacture to any other location, a provision requiring that
the person or entity transporting the manufactured reactor to comply
with all NRC-approved shipping requirements in the manufacturing
license.
Sec. 52.169 [Reserved]
Sec. 52.171 Finality of manufacturing licenses; information requests.
(a)(1) Notwithstanding any provision in 10 CFR 50.109, during the
term of a manufacturing license the Commission may not modify, rescind,
or impose new requirements on the design of the nuclear power reactor
being manufactured, or the requirements for the manufacture of the
nuclear power reactor, unless the Commission determines that
[[Page 116]]
a modification is necessary to bring the design of the reactor or its
manufacture into compliance with the Commission's requirements
applicable and in effect at the time the manufacturing license was
issued, or to provide reasonable assurance of adequate protection to
public health and safety or common defense and security.
(2) Any modification to the design of a manufactured nuclear power
reactor which is imposed by the Commission under paragraph (a)(1) of
this section will be applied to all reactors manufactured under the
license, including those that have already been transported and sited,
except those reactors to which the modification has been rendered
technically irrelevant by action taken under paragraph (b) of this
section.
(3) In making the findings required for issuance of a construction
permit, operating license, combined license, in any hearing under Sec.
52.103, or in any enforcement hearing other than one initiated by the
Commission under paragraph (a)(1) of this section, for which a nuclear
power reactor manufactured under this subpart is referenced or used, the
Commission shall treat as resolved those matters resolved in the
proceeding on the application for issuance or renewal of the
manufacturing license, including the adequacy of design of the
manufactured reactor, the costs and benefits of severe accident
mitigation design alternatives, and the bases for not incorporating
severe accident mitigation design alternatives into the design of the
reactor to be manufactured.
(b)(1) The holder of a manufacturing license may not make changes to
the design of the nuclear power reactor authorized to be manufactured
without prior Commission approval. The request for a change to the
design must be in the form of an application for a license amendment,
and must meet the requirements of 10 CFR 50.90 and 50.92.
(2) An applicant or licensee who references or uses a nuclear power
reactor manufactured under a manufacturing license under this subpart
may request a departure from the design characteristics, site
parameters, terms and conditions, or approved design of the manufactured
reactor. The Commission may grant a request only if it determines that
the departure will comply with the requirements of 10 CFR 52.7, and that
the special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
The granting of a departure on request of an applicant is subject to
litigation in the same manner as other issues in the construction permit
or combined license hearing.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of either the manufacturing license or
the manufactured reactor, information requests to the holder of a
manufacturing license or an applicant or licensee using a manufactured
reactor must be evaluated before issuance to ensure that the burden to
be imposed on respondents is justified in view of the potential safety
significance of the issue to be addressed in the requested information.
Each evaluation performed by the NRC staff must be in accordance with 10
CFR 50.54(f) and must be approved by the Executive Director for
Operations or his or her designee before issuance of the request.
Sec. 52.173 Duration of manufacturing license.
A manufacturing license issued under this subpart may be valid for
not less than 5, nor more than 15 years from the date of issuance. A
holder of a manufacturing license may not initiate the manufacture of a
reactor less than 3 years before the expiration of the license even
though a timely application for renewal has been docketed with the NRC.
Upon expiration of the manufacturing license, the manufacture of any
uncompleted reactors must cease unless a timely application for renewal
has been docketed with the NRC.
Sec. 52.175 Transfer of manufacturing license.
A manufacturing license may be transferred in accordance with Sec.
50.80 of this chapter.
Sec. 52.177 Application for renewal.
(a) Not less than 12 months, nor more than 5 years before the
expiration of the manufacturing license, or any later
[[Page 117]]
renewal period, the holder of the manufacturing license may apply for a
renewal of the license. An application for renewal must contain all
information necessary to bring up to date the information and data
contained in the previous application.
(b) The filing of an application for a renewed license must be in
accordance with subpart A of 10 CFR part 2 and 10 CFR 52.3 and 50.30.
(c) A manufacturing license, either original or renewed, for which a
timely application for renewal has been filed, remains in effect until
the Commission has made a final determination on the renewal
application, provided, however, that in accordance with Sec. 52.173,
the holder of a manufacturing license may not begin manufacture of a
reactor less than 3 years before the expiration of the license.
(d) Any person whose interest may be affected by renewal of the
permit may request a hearing on the application for renewal. The request
for a hearing must comply with 10 CFR 2.309. If a hearing is granted,
notice of the hearing will be published in accordance with 10 CFR 2.104.
(e) The Commission shall refer a copy of the application for renewal
to the Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall
report on those portions of the application which concern safety and
shall apply the criteria set forth in Sec. 52.159.
Sec. 52.179 Criteria for renewal.
The Commission may grant the renewal if the Commission determines:
(a) The manufacturing license complies with the Atomic Energy Act
and the Commission's regulations and orders applicable and in effect at
the time the manufacturing license was originally issued; and
(b) Any new requirements the Commission may wish to impose are:
(1) Necessary for adequate protection to public health and safety or
common defense and security;
(2) Necessary for compliance with the Commission's regulations and
orders applicable and in effect at the time the manufacturing license
was originally issued; or
(3) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
Sec. 52.181 Duration of renewal.
A renewed manufacturing license may be issued for a term of not less
than 5, nor more than 15 years, plus any remaining years on the
manufacturing license then in effect before renewal. The renewed license
shall be subject to the requirements of Sec. Sec. 52.171 and 52.175.
Subpart G [Reserved]
Subpart H_Enforcement
Sec. 52.301 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of--
(1) The Atomic Energy Act of 1954, as amended;
(2) Title II of the Energy Reorganization Act of 1974, as amended;
or
(3) A regulation or order issued under those Acts.
(b) The Commission may obtain a court order for the payment of a
civil penalty imposed under Section 234 of the Atomic Energy Act:
(1) For violations of--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the Atomic Energy Act of 1954, as amended;
(ii) Section 206 of the Energy Reorganization Act;
(iii) Any regulation, or order issued under the sections specified
in paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation of any license issued under
the sections specified in paragraph (b)(1)(i) of this section.
(2) For any violation for which a license may be revoked under
Section 186 of the Atomic Energy Act of 1954, as amended.
Sec. 52.303 Criminal penalties.
(a) Section 223 of the Atomic Energy Act of 1954, as amended,
provides for criminal sanctions for willful violation of, attempted
violation of, or conspiracy to violate, any regulation
[[Page 118]]
issued under Sections 161b, 161i, or 161o of the Act. For purposes of
Section 223, all the regulations in part 52 are issued under one or more
of Sections 161b, 161i, or 160o, except for the sections listed in
paragraph (b) of this section.
(b) The regulations in part 52 that are not issued under Sections
161b, 161i, or 161o for the purposes of Section 223 are as follows:
Sec. Sec. 52.0, 52.1, 52.2, 52.3, 52.7, 52.8, 52.9, 52.10, 52.11,
52.12, 52.13, 52.15, 52.16, 52.17, 52.18, 52.21, 52.23, 52.24, 52.27,
52.28, 52.29, 52.31, 52.33, 52.39, 52.41, 52.43, 52.45, 52.46, 52.47,
52.48, 52.51, 52.53, 52.54, 52.55, 52.57, 52.59, 52.61, 52.63, 52.71,
52.73, 52.75, 52.77, 52.79, 52.80, 52.81, 52.83, 52.85, 52.87, 52.93,
52.97, 52.98, 52.103, 52.104, 52.105, 52.107, 52.109, 52.131, 52.133,
52.135, 52.136, 52.137, 52.139, 52.141, 52.143, 52.145, 52.147, 52.151,
52.153, 52.155, 52.156, 52.157, 52.158, 52.159, 52.161, 52.163, 52.165,
52.167, 52.171, 52.173, 52.175, 52.177, 52.179, 52.181, 52.301, and
52.303.
Sec. Appendix A to Part 52--Design Certification Rule for the U.S.
Advanced Boiling Water Reactor
I. Introduction
Appendix A constitutes the standard design certification for the
U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10
CFR part 52, subpart B. The applicant for certification of the U.S. ABWR
design was GE Nuclear Energy.
II. Definitions
A. Generic design control document (generic DCD) means the document
containing the Tier 1 and Tier 2 information and generic technical
specifications that is incorporated by reference into this appendix.
B. Generic technical specifications means the information, required
by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within
the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an applicant
or licensee who references this appendix, consisting of the information
in the generic DCD, as modified and supplemented by the plant-specific
departures and exemptions made under Section VIII of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by this
appendix (Tier 2 information). Compliance with Tier 2 is required, but
generic changes to and plant-specific departures from Tier 2 are
governed by Section VIII of this appendix. Compliance with Tier 2
provides a sufficient, but not the only acceptable, method for complying
with Tier 1. Compliance methods differing from Tier 2 must satisfy the
change process in Section VIII of this appendix. Regardless of these
differences, an applicant or licensee must meet the requirement in
Section III.B of this appendix to reference Tier 2 when referencing Tier
1. Tier 2 information includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c), with
the exception of generic technical specifications and conceptual design
information;
2. Supporting information on the inspections, tests, and analyses
that will be performed to demonstrate that the acceptance criteria in
the ITAAC have been met; and
3. Combined license (COL) action items (COL license information),
which identify certain matters that must be addressed in the site-
specific portion of the final safety analysis report (FSAR) by an
applicant who references this appendix. These items constitute
information requirements but are not the only acceptable set of
information in the FSAR. An applicant may depart from or omit these
items, provided that the departure or omission is identified and
justified in the FSAR. After issuance of a construction permit or COL,
these items are not requirements for the licensee unless such items are
restated in the FSAR.
F. Tier 2* means the portion of the Tier 2 information, designated
as such in the generic DCD, which is subject to the change process in
Section VIII.B.6 of this appendix. This designation expires for some
Tier 2* information under Section VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are conservative
or essentially the same; or
(2) Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by NRC for the
intended application.
H. All other terms in this appendix have the meaning set out in 10
CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as
amended, as applicable.
[[Page 119]]
III. Scope and Contents
A. Tier 1, Tier 2, and the generic technical specifications in the
U.S. ABWR Design Control Document, GE Nuclear Energy, Revision 4 dated
March 1997, are approved for incorporation by reference by the Director
of the Office of the Federal Register in accordance with 5 U.S.C. 552(a)
and 1 CFR part 51. Copies of the generic DCD may be obtained from the
National Technical Information Service, 5285 Port Royal Road,
Springfield, Virginia 22161. A copy is available for examination and
copying at the NRC Public Document Room located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Copies are also available for examination at the NRC Library located at
Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20582
and the Office of the Federal Register, 800 North Capitol Street, NW.,
Suite 700, Washington, DC.
B. An applicant or licensee referencing this appendix, in accordance
with Section IV of this appendix, shall incorporate by reference and
comply with the requirements of this appendix, including Tier 1, Tier 2,
and the generic technical specifications except as otherwise provided in
this appendix. Conceptual design information, as set forth in the
generic DCD, and the ``Technical Support Document for the ABWR'' are not
part of this appendix. Tier 2 references to the probabilistic risk
assessment (PRA) in the ABWR standard safety analysis report do not
incorporate the PRA into Tier 2.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then
Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the U.S. ABWR design or NUREG-
1503, ``Final Safety Evaluation Report related to the Certification of
the Advanced Boiling Water Reactor Design'' (FSER), and Supplement No.
1, then the generic DCD controls.
E. Design activities for structures, systems, and components that
are wholly outside the scope of this appendix may be performed using
site characteristics, provided the design activities do not affect the
DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10 CFR
52.77, 52.79, and 52.80, comply with the following requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information and
using the same organization and numbering as the generic DCD for the
U.S. ABWR design, as modified and supplemented by the applicant's
exemptions and departures;
b. The reports on departures from and updates to the plant-specific
DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are required by
10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters and
interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47 that is not within the scope
of this appendix.
3. Include, in the plant-specific DCD, the proprietary information
and safeguards information referenced in the U.S. ABWR DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the U.S. ABWR design are in 10 CFR parts 20,
50, 73, and 100, codified as of May 2, 1997, that are applicable and
technically relevant, as described in the FSER (NUREG-1503) and
Supplement No. 1.
B. The U.S. ABWR design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety
Parameter Display Console;
2. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident Sampling
for Boron, Chloride, and Dissolved Gases; and
3. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment
Penetration.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the U.S. ABWR design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of the
public. A conclusion that a matter is resolved includes the finding that
additional or alternative structures, systems, components, design
features, design criteria, testing, analyses, acceptance criteria, or
justifications are not necessary for the U.S. ABWR design.
B. The Commission considers the following matters resolved within
the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance
of a combined license, amendment of a combined license, or renewal of a
combined license, proceedings held under 10 CFR
[[Page 120]]
52.103, and enforcement proceedings involving plants referencing this
appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with the
information in the FSER and Supplement No. 1, Tier 1, Tier 2 (including
referenced information which the context indicates is intended as
requirements), and the rulemaking record for certification of the U.S.
ABWR design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and in
context, are intended as requirements in the generic DCD for the U.S.
ABWR design;
3. All generic changes to the DCD under and in compliance with the
change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but
only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix, all
departures from Tier 2 pursuant to and in compliance with the change
processes in paragraph VIII.B.5 of this appendix that do not require
prior NRC approval, but only for that plant;
7. All environmental issues concerning severe accident mitigation
design alternatives associated with the information in the NRC's final
environmental assessment for the U.S. ABWR design and Revision 1 of the
technical support document for the U.S. ABWR, dated December 1994, for
plants referencing this appendix whose site parameters are within those
specified in the technical support document.
C. The Commission does not consider operational requirements for an
applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an applicant
or licensee who references this appendix by rule, regulation, order, or
license condition.
D. Except in accordance with the change processes in Section VIII of
this appendix, the Commission may not require an applicant or licensee
who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures, systems,
components, or design features discussed in the generic DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the DCD for the U.S. ABWR
design, in order to request or participate in the hearing required by 10
CFR 52.85 or the hearing provided under 10 CFR 52.103, or to request or
participate in any other hearing relating to this appendix in which
interested persons have adjudicatory hearing rights, shall first request
access to such information from GE Nuclear Energy. The request must
state with particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the public
at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public
Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to prepare a
request for hearing, the request must be filed no later than 15 days
after publication in the Federal Register of the notice required either
by 10 CFR 52.85 or 10 CFR 52.103. If GE Nuclear Energy declines to
provide the information sought, GE Nuclear Energy shall send a written
response within 10 days of receiving the request to the requesting
person setting forth with particularity the reasons for its refusal. The
person may then request the Commission (or presiding officer, if a
proceeding has been established) to order disclosure. The person shall
include copies of the original request (and any subsequent clarifying
information provided by the requesting party to the applicant) and the
applicant's response. The Commission and presiding officer shall base
their decisions solely on the person's original request (including any
clarifying information provided by the requesting person to GE Nuclear
Energy), and GE Nuclear Energy's response. The Commission and presiding
officer may order GE Nuclear Energy to provide access to some or all of
the requested information, subject to an appropriate non-disclosure
agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from June
11, 1997, except as provided for in 10 CFR 52.55(b) and 52.57(b). This
appendix remains valid for an applicant or licensee who references this
appendix until the application is withdrawn or the license expires,
including any period of extended operation under a renewed license.
[[Page 121]]
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the design
change will result in a significant decrease in the level of safety
otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix, or to
assure adequate protection of the public health and safety or the common
defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may request
an exemption from Tier 2 information. The Commission may grant such a
request only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The Commission will deny a request for an
exemption from Tier 2, if it finds that the design change will result in
a significant decrease in the level of safety otherwise provided by the
design. The grant of an exemption to an applicant must be subject to
litigation in the same manner as other issues material to the license
hearing. The grant of an exemption to a licensee must be subject to an
opportunity for a hearing in the same manner as license amendments.
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
requires a license amendment under paragraphs B.5.b or 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 frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component (SSC)
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 a SSC important to safety previously evaluated in the
plant-specific DCD;
(5) Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier as
described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described in
the plant-specific DCD used in establishing the design bases or in the
safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an ex-
vessel severe accident design feature identified in the plant-specific
DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe accident
previously reviewed and determined to be not credible could become
credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously reviewed.
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.
e. A departure from Tier 2 information that is made under paragraph
B.5 of this section does not require an exemption from this appendix.
[[Page 122]]
f. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an applicant or licensee who references
this appendix has not complied with paragraph VIII.B.5 of this appendix
when departing from Tier 2 information, may petition the NRC to admit
into the proceeding such a contention. In addition to compliance with
the general requirements of 10 CFR 2.309, the petition must demonstrate
that the departure does not comply with paragraph VIII.B.5 of this
appendix. Further, the petition must demonstrate that the change bears
on an asserted noncompliance with an ITAAC acceptance criterion in the
case of a 10 CFR 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a license
amendment. Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall certify
the matter directly to the Commission for determination of the
admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart from
Tier 2* information, which is designated with italicized text or
brackets and an asterisk in the generic DCD, without NRC approval. The
departure will not be considered a resolved issue, within the meaning of
Section VI of this appendix and 10 CFR 52.63(a)(5).
b. A licensee who references this appendix may not depart from the
following Tier 2* matters without prior NRC approval. A request for a
departure will be treated as a request for a license amendment under 10
CFR 50.90.
(1) Fuel burnup limit (4.2).
(2) Fuel design evaluation (4.2.3).
(3) Fuel licensing acceptance criteria (appendix 4B).
c. A licensee who references this appendix may not, before the plant
first achieves full power following the finding required by 10 CFR
52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant first
achieves full power, the following Tier 2* matters revert to Tier 2
status and are thereafter subject to the departure provisions in
paragraph B.5 of this section.
(1) ASME Boiler & Pressure Vessel Code, Section III.
(2) ACI 349 and ANSI/AISC-690.
(3) Motor-operated valves.
(4) Equipment seismic qualification methods.
(5) Piping design acceptance criteria.
(6) Fuel system and assembly design (4.2), except burnup limit.
(7) Nuclear design (4.3).
(8) Equilibrium cycle and control rod patterns (App. 4A).
(9) Control rod licensing acceptance criteria (App. 4C).
(10) Instrument setpoint methodology.
(11) EMS performance specifications and architecture.
(12) SSLC hardware and software qualification.
(13) Self-test system design testing features and commitments.
(14) Human factors engineering design and implementation process.
d. Departures from Tier 2* information that are made under paragraph
B.6 of this section do not require an exemption from this appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved in
the design certification rulemaking and do not require a change to a
design feature in the generic DCD are governed by the requirements in 10
CFR 50.109. Generic changes that do require a change to a design feature
in the generic DCD are governed by the requirements in paragraphs A or B
of this section.
2. Generic changes to generic TS and other operational requirements
are applicable to all applicants who reference this appendix, except
those for which the change has been rendered technically irrelevant by
action taken under paragraphs C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on generic
technical specifications and other operational requirements that were
completely reviewed and approved, provided a change to a design feature
in the generic DCD is not required and special circumstances as defined
in 10 CFR 2.335 are present. The Commission may modify or supplement
generic technical specifications and other operational requirements that
were not completely reviewed and approved or require additional
technical specifications and other operational requirements on a plant-
specific basis, provided a change to a design feature in the generic DCD
is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other operational
requirements. The Commission may grant such a request only if it
determines that the exemption will comply with the requirements of 10
CFR 52.7. The grant of an exemption must be subject to litigation in the
same manner as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10
[[Page 123]]
CFR 52.103(a), who believes that an operational requirement approved in
the DCD or a technical specification derived from the generic technical
specifications must be changed may petition to admit into the proceeding
such a contention. Such petition must comply with the general
requirements of 10 CFR 2.309 and must demonstrate why special
circumstances as defined in 10 CFR 2.335 are present, or for compliance
with the Commission's regulations in effect at the time this appendix
was approved, as set forth in Section V of this appendix. Any other
party may file a response thereto. If, on the basis of the petition and
any response, the presiding officer determines that a sufficient showing
has been made, the presiding officer shall certify the matter directly
to the Commission for determination of the admissibility of the
contention. All other issues with respect to the plant-specific
technical specifications or other operational requirements are subject
to a hearing as part of the license proceeding.
6. After issuance of a license, the generic technical specifications
have no further effect on the plant-specific technical specifications
and changes to the plant-specific technical specifications will be
treated as license amendments under 10 CFR 50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1. An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even though
the NRC may not have found that any particular ITAAC has been met.
2. The licensee who references this appendix shall notify the NRC
that the required inspections, tests, and analyses in the ITAAC have
been successfully completed and that the corresponding acceptance
criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not demonstrated
that the ITAAC has been met, the applicant or licensee may either take
corrective actions to successfully complete that ITAAC, request an
exemption from the ITAAC in accordance with Section VIII of this
appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this
appendix by changing the requirements of the ITAAC, under 10 CFR 2.802
and 52.97(b). Such rulemaking changes to the ITAAC must meet the
requirements of paragraph VIII.A.1 of this appendix.
B.1. The NRC shall ensure that the required inspections, tests, and
analyses in the ITAAC are performed. The NRC shall verify that the
inspections, tests, and analyses referenced by the licensee have been
successfully completed and, based solely thereon, find the prescribed
acceptance criteria have been met. At appropriate intervals during
construction, the NRC shall publish notices of the successful completion
of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall find
that the acceptance criteria in the ITAAC for the license are met before
fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the subject
of a Sec. 52.103(a) hearing, their expiration will occur upon final
Commission action in such proceeding. However, subsequent modifications
must comply with the Tier 1 and Tier 2 design descriptions in the plant-
specific DCD unless the licensee has complied with the applicable
requirements of 10 CFR 52.98 and Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and the
generic TS and other operational requirements. The applicant shall
maintain the proprietary and safeguards information referenced in the
generic DCD for the period that this appendix may be referenced, as
specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application and
for the term of the license (including any period of renewal).
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).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any plant-
specific departures from the DCD, including a summary of the evaluation
of each. This report must be filed in accordance with the filing
requirements applicable to reports in 10 CFR 52.3.
[[Page 124]]
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes and the
plant-specific departures from the generic DCD made under Section VIII
of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in 10
CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and X.B.2
must be submitted as follows:
a. On the date that an application for a license referencing this
appendix is submitted, the application must include the report and any
updates to the generic DCD.
b. During the interval from the date of application for a license to
the date the Commission makes the finding required by 10 CFR 52.103(g),
the report must be submitted semiannually. Updates to the plant-specific
DCD must be submitted annually and may be submitted along with
amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the final
safety analysis report for the facility, at the intervals required by 10
CFR 50.59(d)(2) and 10 CFR 50.71(e)(4), respectively, or at shorter
intervals as specified in the license.
Sec. Appendix B to Part 52--Design Certification Rule for the System 80+
Design
I. Introduction
Appendix B constitutes design certification for the System 80+ \1\
standard plant design, in accordance with 10 CFR part 52, subpart B. The
applicant for certification of the System 80+ design was Combustion
Engineering, Inc. (ABB-CE), which is now Westinghouse Electric Company
LLC.
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\1\ ``System 80+'' is a trademark of Westinghouse Electric Company
LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the document
containing the Tier 1 and Tier 2 information and generic technical
specifications that is incorporated by reference into this appendix.
B. Generic technical specifications means the information, required
by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within
the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an applicant
or licensee who references this appendix, consisting of the information
in the generic DCD, as modified and supplemented by the plant-specific
departures and exemptions made under Section VIII of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by this
appendix (Tier 2 information). Compliance with Tier 2 is required, but
generic changes to and plant-specific departures from Tier 2 are
governed by Section VIII of this appendix. Compliance with Tier 2
provides a sufficient, but not the only acceptable, method for complying
with Tier 1. Compliance methods differing from Tier 2 must satisfy the
change process in Section VIII of this appendix. Regardless of these
differences, an applicant or licensee must meet the requirement in
Section III.B of this appendix to reference Tier 2 when referencing Tier
1. Tier 2 information includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c), with
the exception of generic technical specifications and conceptual design
information;
2. Supporting information on the inspections, tests, and analyses
that will be performed to demonstrate that the acceptance criteria in
the ITAAC have been met; and
3. Combined license (COL) action items (COL license information),
which identify certain matters that must be addressed in the site-
specific portion of the final safety analysis report (FSAR) by an
applicant who references this appendix. These items constitute
information requirements but are not the only acceptable set of
information in the FSAR. An applicant may depart from or omit these
items, provided that the departure or omission is identified and
justified in the FSAR. After issuance of a construction permit or COL,
these items are not requirements for the licensee unless such items are
restated in the FSAR.
F. Tier 2* means the portion of the Tier 2 information, designated
as such in the generic DCD, which is subject to the change process in
Section VIII.B.6 of this appendix. This designation expires for some
Tier 2* information under Section VIII.B.6 of this appendix.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
[[Page 125]]
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are conservative
or essentially the same; or
(2) Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by NRC for the
intended application.
H. All other terms in this appendix have the meaning set out in 10
CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as
amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2, and the generic technical specifications in the
System 80+ Design Control Document, ABB-CE, with revisions dated January
1997, are approved for incorporation by reference by the Director of the
Office of the Federal Register in accordance with 5 U.S.C. 552(a) and 1
CFR part 51. Copies of the generic DCD may be obtained from the National
Technical Information Service, 5285 Port Royal Road, Springfield,
Virginia 22161. A copy is available for examination and copying at the
NRC Public Document Room located at One White Flint North, 11555
Rockville Pike (first floor), Rockville, Maryland 20852. Copies are also
available for examination at the NRC Library located at Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland 20582 and the Office of
the Federal Register, 800 North Capitol Street, NW., Suite 700,
Washington, DC.
B. An applicant or licensee referencing this appendix, in accordance
with Section IV of this appendix, shall incorporate by reference and
comply with the requirements of this appendix, including Tier 1, Tier 2,
and the generic technical specifications except as otherwise provided in
this appendix. Conceptual design information, as set forth in the
generic DCD, and the Technical Support Document for the System 80+
design are not part of this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then
Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the System 80+ design or NUREG-
1462, ``Final Safety Evaluation Report Related to the Certification of
the System 80+ Design,'' (FSER) and Supplement No. 1, then the generic
DCD controls.
E. Design activities for structures, systems, and components that
are wholly outside the scope of this appendix may be performed using
site characteristics, provided the design activities do not affect the
DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10 CFR
52.77, 52.79, and 52.80, comply with the following requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information and
using the same organization and numbering as the generic DCD for the
System 80+ design, as modified and supplemented by the applicant's
exemptions and departures;
b. The reports on departures from and updates to the plant-specific
DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are required by
10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters and
interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47 that is not within the scope
of this appendix.
3. Include, in the plant-specific DCD, the proprietary information
referenced in the System 80+ DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the System 80+ design are in 10 CFR parts 20,
50, 73, and 100, codified as of May 9, 1997, that are applicable and
technically relevant, as described in the FSER (NUREG-1462) and
Supplement No. 1.
B. The System 80+ design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety
Parameter Display Console;
2. Paragraphs (f)(2) (vii), (viii), (xxvi), and (xxviii) of 10 CFR
50.34--Accident Source Terms;
3. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident Sampling
for Hydrogen, Boron, Chloride, and Dissolved Gases;
4. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment
Penetration; and
5. Paragraphs III.A.1(a) and III.C.3(b) of Appendix J to 10 CFR 50--
Containment Leakage Testing.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the System 80+ design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this
[[Page 126]]
appendix; and therefore, provide adequate protection to the health and
safety of the public. A conclusion that a matter is resolved includes
the finding that additional or alternative structures, systems,
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for the System
80+ design.
B. The Commission considers the following matters resolved within
the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance
of a combined license, amendment of a combined license, or renewal of a
combined license, proceedings held under 10 CFR 52.103, and enforcement
proceedings involving plants referencing this appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with the
information in the FSER and Supplement No. 1, Tier 1, Tier 2 (including
referenced information which the context indicates is intended as
requirements), and the rulemaking record for certification of the System
80+ design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and in
context, are intended as requirements in the generic DCD for the System
80+ design;
3. All generic changes to the DCD under and in compliance with the
change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but
only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix, all
departures from Tier 2 under and in compliance with the change processes
in paragraph VIII.B.5 of this appendix that do not require prior NRC
approval, but only for that plant;
7. All environmental issues concerning severe accident mitigation
design alternatives associated with the information in the NRC's final
environmental assessment for the System 80+ design and the technical
support document for the System 80+ design, dated January 1995, for
plants referencing this appendix whose site parameters are within those
specified in the technical support document.
C. The Commission does not consider operational requirements for an
applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an applicant
or licensee who references this appendix by rule, regulation, order, or
license condition.
D. Except in accordance with the change processes in Section VIII of
this appendix, the Commission may not require an applicant or licensee
who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures, systems,
components, or design features discussed in the generic DCD.
E.1. Persons who wish to review proprietary information or other
secondary references in the DCD for the System 80+ design, in order to
request or participate in the hearing required by 10 CFR 52.85 or the
hearing provided under 10 CFR 52.103, or to request or participate in
any other hearing relating to this appendix in which interested persons
have adjudicatory hearing rights, shall first request access to such
information from Westinghouse. The request must state with
particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the public
at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public
Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to prepare a
request for hearing, the request must be filed no later than 15 days
after publication in the Federal Register of the notice required either
by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide
the information sought, Westinghouse shall send a written response
within ten (10) days of receiving the request to the requesting person
setting forth with particularity the reasons for its refusal. The person
may then request the Commission (or presiding officer, if a proceeding
has been established) to order disclosure. The person shall include
copies of the original request (and any subsequent clarifying
information provided by the requesting party to the applicant) and the
applicant's response. The Commission and presiding officer shall base
their decisions solely on the person's original request (including any
clarifying information provided by the requesting person to
Westinghouse), and Westinghouse's response. The Commission and presiding
officer may order Westinghouse to provide access to
[[Page 127]]
some or all of the requested information, subject to an appropriate non-
disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from June
20, 1997, except as provided for in 10 CFR 52.55(b) and 52.57(b). This
appendix remains valid for an applicant or licensee who references this
appendix until the application is withdrawn or the license expires,
including any period of extended operation under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the design
change will result in a significant decrease in the level of safety
otherwise provided by the design.
B. Tier 2 Information
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix, or to
assure adequate protection of the public health and safety or the common
defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may request
an exemption from Tier 2 information. The Commission may grant such a
request only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The Commission will deny a request for an
exemption from Tier 2, if it finds that the design change will result in
a significant decrease in the level of safety otherwise provided by the
design. The grant of an exemption to an applicant must be subject to
litigation in the same manner as other issues material to the license
hearing. The grant of an exemption to a licensee must be subject to an
opportunity for a hearing in the same manner as license amendments.
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
requires a license amendment under paragraphs B.5.b or 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 frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component (SSC)
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 an SSC important to safety previously evaluated in the
plant-specific DCD;
(5) Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier as
described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described in
the plant-specific DCD used in establishing the design bases or in the
safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an ex-
vessel severe accident design feature identified in the plant-specific
DCD, requires a license amendment if:
[[Page 128]]
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe accident
previously reviewed and determined to be not credible could become
credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously reviewed.
d. If a departure requires a license amendment under paragraph B.5.b
or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under paragraph
B.5 of this section does not require an exemption from this appendix.
f. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an applicant or licensee who references
this appendix has not complied with paragraph VIII.B.5 of this appendix
when departing from Tier 2 information, may petition the NRC to admit
into the proceeding such a contention. In addition to compliance with
the general requirements of 10 CFR 2.309, the petition must demonstrate
that the departure does not comply with paragraph VIII.B.5 of this
appendix. Further, the petition must demonstrate that the change bears
on an asserted noncompliance with an ITAAC acceptance criterion in the
case of a 10 CFR 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a license
amendment. Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall certify
the matter directly to the Commission for determination of the
admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart from
Tier 2* information, which is designated with italicized text or
brackets and an asterisk in the generic DCD, without NRC approval. The
departure will not be considered a resolved issue, within the meaning of
Section VI of this appendix and 10 CFR 52.63(a)(5).
b. A licensee who references this appendix may not depart from the
following Tier 2* matters without prior NRC approval. A request for a
departure will be treated as a request for a license amendment under 10
CFR 50.90.
(1) Maximum fuel rod average burnup.
(2) Control room human factors engineering.
c. A licensee who references this appendix may not, before the plant
first achieves full power following the finding required by 10 CFR
52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant first
achieves full power, the following Tier 2* matters revert to Tier 2
status and are thereafter subject to the departure provisions in
paragraph B.5 of this section.
(1) ASME Boiler & Pressure Vessel Code, Section III.
(2) ACI 349 and ANSI/AISC-690.
(3) Motor-operated valves.
(4) Equipment seismic qualification methods.
(5) Piping design acceptance criteria.
(6) Fuel and control rod design, except burnup limit.
(7) Instrumentation and controls setpoint methodology.
(8) Instrumentation and controls hardware and software changes.
(9) Instrumentation and controls environmental qualification.
(10) Seismic design criteria for non-seismic Category I structures.
d. Departures from Tier 2* information that are made under paragraph
B.6 of this section do not require an exemption from this appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved in
the design certification rulemaking and do not require a change to a
design feature in the generic DCD are governed by the requirements in 10
CFR 50.109. Generic changes that do require a change to a design feature
in the generic DCD are governed by the requirements in paragraphs A or B
of this section.
2. Generic changes to generic TS and other operational requirements
are applicable to all applicants who reference this appendix, except
those for which the change has been rendered technically irrelevant by
action taken under paragraphs C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on generic
technical specifications and other operational requirements that were
completely reviewed and approved, provided a change to a design feature
in the generic DCD is not required and special circumstances as defined
in 10 CFR 2.335 are present. The Commission may modify or supplement
generic technical specifications and other operational requirements that
were not completely reviewed and approved or require additional
technical specifications and other operational requirements on a plant-
specific basis, provided a change to a design feature in the generic DCD
is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other operational
requirements. The Commission may grant such a request only if it
determines
[[Page 129]]
that the exemption will comply with the requirements of 10 CFR 52.7. The
grant of an exemption must be subject to litigation in the same manner
as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an operational requirement approved in the
DCD or a technical specification derived from the generic technical
specifications must be changed may petition to admit into the proceeding
such a contention. Such a petition must comply with the general
requirements of 10 CFR 2.309 and must demonstrate why special
circumstances as defined in 10 CFR 2.335 are present, or for compliance
with the Commission's regulations in effect at the time this appendix
was approved, as set forth in Section V of this appendix. Any other
party may file a response thereto. If, on the basis of the petition and
any response, the presiding officer determines that a sufficient showing
has been made, the presiding officer shall certify the matter directly
to the Commission for determination of the admissibility of the
contention. All other issues with respect to the plant-specific
technical specifications or other operational requirements are subject
to a hearing as part of the license proceeding.
6. After issuance of a license, the generic technical specifications
have no further effect on the plant-specific technical specifications
and changes to the plant-specific technical specifications will be
treated as license amendments under 10 CFR 50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1 An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even though
the NRC may not have found that any particular ITAAC has been met.
2. The licensee who references this appendix shall notify the NRC
that the required inspections, tests, and analyses in the ITAAC have
been successfully completed and that the corresponding acceptance
criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not demonstrated
that the ITAAC has been met, the applicant or licensee may either take
corrective actions to successfully complete that ITAAC, request an
exemption from the ITAAC in accordance with Section VIII of this
appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this
appendix by changing the requirements of the ITAAC, under 10 CFR 2.802
and 52.97(b). Such rulemaking changes to the ITAAC must meet the
requirements of Section VIII.A.1 of this appendix.
B.1 The NRC shall ensure that the required inspections, tests, and
analyses in the ITAAC are performed. The NRC shall verify that the
inspections, tests, and analyses referenced by the licensee have been
successfully completed and, based solely thereon, find the prescribed
acceptance criteria have been met. At appropriate intervals during
construction, the NRC shall publish notices of the successful completion
of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall find
that the acceptance criteria in the ITAAC for the license are met before
fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the subject
of a Sec. 52.103(a) hearing, their expiration will occur upon final
Commission action in such proceeding. However, subsequent modifications
must comply with the Tier 1 and Tier 2 design descriptions in the plant-
specific DCD unless the licensee has complied with the applicable
requirements of 10 CFR 52.98 and Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and the
generic TS and other operational requirements. The applicant shall
maintain the proprietary and safeguards information referenced in the
generic DCD for the period that this appendix may be referenced, as
specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application and
for the term of the license (including any period of renewal).
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).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the
[[Page 130]]
NRC containing a brief description of any plant-specific departures from
the DCD, including a summary of the evaluation of each. This report must
be filed in accordance with the filing requirements applicable to
reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes to and
plant-specific departures from the generic DCD made under Section VIII
of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in 10
CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and X.B.2
must be submitted as follows:
a. On the date that an application for a license referencing this
appendix is submitted, the application must include the report and any
updates to the generic DCD.
b. During the interval from the date of application for a license to
the date the Commission makes the finding required by 10 CFR 52.103(g),
the report must be submitted semi-annually. Updates to the plant-
specific DCD must be submitted annually and may be submitted along with
amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the final
safety analysis report for the facility, at the intervals required by 10
CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals
as specified in the license.
Sec. Appendix C to Part 52--Design Certification Rule for the AP600
Design
I. Introduction
Appendix C constitutes the standard design certification for the
AP600 \1\ design, in accordance with 10 CFR part 52, subpart B. The
applicant for certification of the AP600 design is Westinghouse Electric
Company LLC.
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\1\ AP600 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the document
containing the Tier 1 and Tier 2 information and generic technical
specifications that is incorporated by reference into this appendix.
B. Generic technical specifications means the information, required
by 10 CFR 50.36 and 50.36a, for the portion of the plant that is within
the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an applicant
or licensee who references this appendix, consisting of the information
in the generic DCD, as modified and supplemented by the plant-specific
departures and exemptions made under Section VIII of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by this
appendix (Tier 2 information). Compliance with Tier 2 is required, but
generic changes to and plant-specific departures from Tier 2 are
governed by Section VIII of this appendix. Compliance with Tier 2
provides a sufficient, but not the only acceptable, method for complying
with Tier 1. Compliance methods differing from Tier 2 must satisfy the
change process in Section VIII of this appendix. Regardless of these
differences, an applicant or licensee must meet the requirement in
Section III.B of this appendix to reference Tier 2 when referencing Tier
1. Tier 2 information includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c), with
the exception of generic technical specifications and conceptual design
information;
2. Supporting information on the inspections, tests, and analyses
that will be performed to demonstrate that the acceptance criteria in
the ITAAC have been met; and
3. Combined license (COL) action items (COL license information),
which identify certain matters that must be addressed in the site-
specific portion of the final safety analysis report (FSAR) by an
applicant who references this appendix. These items constitute
information requirements but are not the only acceptable set of
information in the FSAR. An applicant may depart from or omit these
items, provided that the departure or omission is identified and
justified in the FSAR. After issuance of a construction permit or COL,
these items are not requirements for the licensee unless such items are
restated in the FSAR.
4. The investment protection short-term availability controls in
Section 16.3 of the DCD.
F. Tier 2* means the portion of the Tier 2 information, designated
as such in the generic DCD, which is subject to the change process in
Section VIII.B.6 of this appendix.
[[Page 131]]
This designation expires for some Tier 2* information under Section
VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are conservative
or essentially the same; or
(2) Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by NRC for the
intended application.
H. All other terms in this appendix have the meaning set out in 10
CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, as
amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2 (including the investment protection short-term
availability controls in Section 16.3), and the generic technical
specifications in the AP600 DCD (12/99 revision) are approved for
incorporation by reference by the Director of the Office of the Federal
Register on January 24, 2000, in accordance with 5 U.S.C. 552(a) and 1
CFR part 51. Copies of the generic DCD may be obtained from Ronald P.
Vijuk, Manager, Passive Plant Engineering, Westinghouse Electric
Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355. A copy of
the generic DCD is available for examination and copying at the NRC
Public Document Room located at One White Flint North, 11555 Rockville
Pike (first floor), Rockville, Maryland 20852. Copies are also available
for examination at the NRC Library located at Two White Flint North,
11545 Rockville Pike, Rockville, Maryland 20582; and the Office of the
Federal Register, 800 North Capitol Street, NW., Suite 700, Washington,
DC.
B. An applicant or licensee referencing this appendix, in accordance
with Section IV of this appendix, shall incorporate by reference and
comply with the requirements of this appendix, including Tier 1, Tier 2
(including the investment protection short-term availability controls in
Section 16.3), and the generic technical specifications except as
otherwise provided in this appendix. Conceptual design information in
the generic DCD and the evaluation of severe accident mitigation design
alternatives in Appendix 1B of the generic DCD are not part of this
appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then
Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the AP600 design or NUREG-1512,
``Final Safety Evaluation Report Related to Certification of the AP600
Standard Design,'' (FSER), then the generic DCD controls.
E. Design activities for structures, systems, and components that
are wholly outside the scope of this appendix may be performed using
site characteristics, provided the design activities do not affect the
DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10 CFR
52.77, 52.79, and 52.80, comply with the following requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information and
utilizing the same organization and numbering as the generic DCD for the
AP600 design, as modified and supplemented by the applicant's exemptions
and departures;
b. The reports on departures from and updates to the plant-specific
DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are required by
10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters and
interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47 that is not within the scope
of this appendix.
3. Include, in the plant-specific DCD, the proprietary information
and safeguards information referenced in the AP600 DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the AP600 design are in 10 CFR parts 20, 50,
73, and 100, codified as of December 16, 1999, that are applicable and
technically relevant, as described in the FSER (NUREG-1512) and the
supplementary information for this section.
B. The AP600 design is exempt from portions of the following
regulations:
1. Paragraph (a)(1) of 10 CFR 50.34--whole body dose criterion;
2. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter
Display Console;
3. Paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10 CFR
50.34--Accident Source Term in TID 14844;
4. Paragraph (a)(2) of 10 CFR 50.55a--ASME Boiler and Pressure
Vessel Code;
5. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency)
feedwater system;
[[Page 132]]
6. Appendix A to 10 CFR part 50, GDC 17--Offsite Power Sources; and
7. Appendix A to 10 CFR part 50, GDC 19--whole body dose criterion.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the AP600 design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of the
public. A conclusion that a matter is resolved includes the finding that
additional or alternative structures, systems, components, design
features, design criteria, testing, analyses, acceptance criteria, or
justifications are not necessary for the AP600 design.
B. The Commission considers the following matters resolved within
the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance
of a combined license, amendment of a combined license, or renewal of a
combined license, proceedings held under 10 CFR 52.103, and enforcement
proceedings involving plants referencing this appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with the
information in the FSER and Supplement No. 1, Tier 1, Tier 2 (including
referenced information which the context indicates is intended as
requirements and the investment protection short-term availability
controls in Section 16.3), and the rulemaking record for certification
of the AP600 design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and in
context, are intended as requirements in the generic DCD for the AP600
design;
3. All generic changes to the DCD under and in compliance with the
change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but
only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix, all
departures from Tier 2 under and in compliance with the change processes
in paragraph VIII.B.5 of this appendix that do not require prior NRC
approval, but only for that plant;
7. All environmental issues concerning severe accident mitigation
design alternatives associated with the information in the NRC's
environmental assessment for the AP600 design and appendix 1B of the
generic DCD, for plants referencing this appendix whose site parameters
are within those specified in the severe accident mitigation design
alternatives evaluation.
C. The Commission does not consider operational requirements for an
applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an applicant
or licensee who references this appendix by rule, regulation, order, or
license condition.
D. Except in accordance with the change processes in Section VIII of
this appendix, the Commission may not require an applicant or licensee
who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures, systems,
components, or design features discussed in the generic DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the AP600 DCD, in order to
request or participate in the hearing required by 10 CFR 52.85 or the
hearing provided under 10 CFR 52.103, or to request or participate in
any other hearing relating to this appendix in which interested persons
have adjudicatory hearing rights, shall first request access to such
information from Westinghouse. The request must state with
particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the public
at the NRC Web site, http://www.nrc.gov, and/or at the NRC Public
Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to prepare a
request for hearing, the request must be filed no later than 15 days
after publication in the Federal Register of the notice required either
by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide
the information sought, Westinghouse shall send a written response
within 10 days of receiving the request to the requesting person setting
forth with particularity the reasons for its refusal. The person may
then request the Commission (or presiding officer, if a proceeding has
been established) to order disclosure. The person shall include copies
of the original request (and
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any subsequent clarifying information provided by the requesting party
to the applicant) and the applicant's response. The Commission and
presiding officer shall base their decisions solely on the person's
original request (including any clarifying information provided by the
requesting person to Westinghouse), and Westinghouse's response. The
Commission and presiding officer may order Westinghouse to provide
access to some or all of the requested information, subject to an
appropriate non-disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
January 24, 2000, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee who
references this appendix until the application is withdrawn or the
license expires, including any period of extended operation under a
renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the design
change will result in a significant decrease in the level of safety
otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix, or to
assure adequate protection of the public health and safety or the common
defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may request
an exemption from Tier 2 information. The Commission may grant such a
request only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The Commission will deny a request for an
exemption from Tier 2, if it finds that the design change will result in
a significant decrease in the level of safety otherwise provided by the
design. The grant of an exemption to an applicant must be subject to
litigation in the same manner as other issues material to the license
hearing. The grant of an exemption to a licensee must be subject to an
opportunity for a hearing in the same manner as license amendments.
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
requires a license amendment under paragraphs B.5.b or 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 frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component (SSC)
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 an SSC important to safety previously evaluated in the
plant-specific DCD;
(5) Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier as
described in the plant-specific DCD being exceeded or altered; or
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(8) Result in a departure from a method of evaluation described in
the plant-specific DCD used in establishing the design bases or in the
safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an ex-
vessel severe accident design feature identified in the plant-specific
DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe accident
previously reviewed and determined to be not credible could become
credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously reviewed.
d. If a departure requires a license amendment under paragraphs
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under paragraph
B.5 of this section does not require an exemption from this appendix.
f. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an applicant or licensee who references
this appendix has not complied with paragraph VIII.B.5 of this appendix
when departing from Tier 2 information, may petition the NRC to admit
into the proceeding such a contention. In addition to compliance with
the general requirements of 10 CFR 2.309, the petition must demonstrate
that the departure does not comply with paragraph VIII.B.5 of this
appendix. Further, the petition must demonstrate that the change bears
on an asserted noncompliance with an ITAAC acceptance criterion in the
case of a 10 CFR 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a license
amendment. Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall certify
the matter directly to the Commission for determination of the
admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6a. An applicant who references this appendix may not depart from
Tier 2* information, which is designated with italicized text or
brackets and an asterisk in the generic DCD, without NRC approval. The
departure will not be considered a resolved issue, within the meaning of
Section VI of this appendix and 10 CFR 52.63(a)(5).
b. A licensee who references this appendix may not depart from the
following Tier 2* matters without prior NRC approval. A request for a
departure will be treated as a request for a license amendment under 10
CFR 50.90.
(1) Maximum fuel rod average burn-up.
(2) Fuel principal design requirements.
(3) Fuel criteria evaluation process.
(4) Fire areas.
(5) Human factors engineering.
c. A licensee who references this appendix may not, before the plant
first achieves full power following the finding required by 10 CFR
52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant first
achieves full power, the following Tier 2* matters revert to Tier 2
status and are thereafter subject to the departure provisions in
paragraph B.5 of this section.
(1) Nuclear Island structural dimensions.
(2) ASME Boiler and Pressure Vessel Code, Section III, and Code
Case--284.
(3) Design Summary of Critical Sections.
(4) ACI 318, ACI 349, and ANSI/AISC--690.
(5) Definition of critical locations and thicknesses.
(6) Seismic qualification methods and standards.
(7) Nuclear design of fuel and reactivity control system, except
burn-up limit.
(8) Motor-operated and power-operated valves.
(9) Instrumentation and control system design processes, methods,
and standards.
(10) PRHR natural circulation test (first plant only).
(11) ADS and CMT verification tests (first three plants only).
d. Departures from Tier 2* information that are made under paragraph
B.6 of this section do not require an exemption from this appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved in
the design certification rulemaking and do not require a change to a
design feature in the generic DCD are governed by the requirements in 10
CFR 50.109. Generic changes that do require a change to a design feature
in the generic DCD are governed by the requirements in paragraphs A or B
of this section.
2. Generic changes to generic TS and other operational requirements
are applicable to all applicants who reference this appendix, except
those for which the change has been rendered technically irrelevant by
action taken under paragraphs C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on generic
technical specifications and other operational requirements that were
completely reviewed and approved, provided a change to a design feature
in the generic DCD is not required and special circumstances as defined
in 10 CFR 2.335 are
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present. The Commission may modify or supplement generic technical
specifications and other operational requirements that were not
completely reviewed and approved or require additional technical
specifications and other operational requirements on a plant-specific
basis, provided a change to a design feature in the generic DCD is not
required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other operational
requirements. The Commission may grant such a request only if it
determines that the exemption will comply with the requirements of 10
CFR 52.7. The grant of an exemption must be subject to litigation in the
same manner as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an operational requirement approved in the
DCD or a technical specification derived from the generic technical
specifications must be changed may petition to admit into the proceeding
such a contention. Such petition must comply with the general
requirements of 10 CFR 2.309 and must demonstrate why special
circumstances as defined in 10 CFR 2.335 are present, or for compliance
with the Commission's regulations in effect at the time this appendix
was approved, as set forth in Section V of this appendix. Any other
party may file a response thereto. If, on the basis of the petition and
any response, the presiding officer determines that a sufficient showing
has been made, the presiding officer shall certify the matter directly
to the Commission for determination of the admissibility of the
contention. All other issues with respect to the plant-specific
technical specifications or other operational requirements are subject
to a hearing as part of the license proceeding.
6. After issuance of a license, the generic technical specifications
have no further effect on the plant-specific technical specifications
and changes to the plant-specific technical specifications will be
treated as license amendments under 10 CFR 50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1 An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even though
the NRC may not have found that any particular ITAAC has been met.
2. The licensee who references this appendix shall notify the NRC
that the required inspections, tests, and analyses in the ITAAC have
been successfully completed and that the corresponding acceptance
criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not demonstrated
that the ITAAC has been met, the applicant or licensee may either take
corrective actions to successfully complete that ITAAC, request an
exemption from the ITAAC in accordance with Section VIII of this
appendix and 10 CFR 52.97(b), or petition for rulemaking to amend this
appendix by changing the requirements of the ITAAC, under 10 CFR 2.802
and 52.97(b). Such rulemaking changes to the ITAAC must meet the
requirements of paragraph VIII.A.1 of this appendix.
B.1. The NRC shall ensure that the required inspections, tests, and
analyses in the ITAAC are performed. The NRC shall verify that the
inspections, tests, and analyses referenced by the licensee have been
successfully completed and, based solely thereon, find the prescribed
acceptance criteria have been met. At appropriate intervals during
construction, the NRC shall publish notices of the successful completion
of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall find
that the acceptance criteria in the ITAAC for the license are met before
fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the subject
of a Sec. 52.103(a) hearing, their expiration will occur upon final
Commission action in such proceeding. However, subsequent modifications
must comply with the Tier 1 and Tier 2 design descriptions in the plant-
specific DCD unless the licensee has complied with the applicable
requirements of 10 CFR 52.98 and Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and the
generic TS and other operational requirements. The applicant shall
maintain the proprietary and safeguards information referenced in the
generic DCD for the period that this appendix may be referenced, as
specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of
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this appendix throughout the period of application and for the term of
the license (including any period of renewal).
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).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any plant-
specific departures from the DCD, including a summary of the evaluation
of each. This report must be filed in accordance with the filing
requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes to and
plant-specific departures from the generic DCD made under Section VIII
of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in 10
CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and X.B.2
must be submitted as follows:
a. On the date that an application for a license referencing this
appendix is submitted, the application must include the report and any
updates to the generic DCD.
b. During the interval from the date of application for a license to
the date the Commission makes the finding required by 10 CFR 52.103(g),
the report must be submitted semi-annually. Updates to the plant-
specific DCD must be submitted annually and may be submitted along with
amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the final
safety analysis report for the facility, at the intervals required by 10
CFR 50.59(d)(2) and 50.71(e), respectively, or at shorter intervals as
specified in the license.
Sec. Appendix D to Part 52--Design Certification Rule for the AP1000
Design
I. Introduction
Appendix D constitutes the standard design certification for the
AP1000 \1\ design, in accordance with 10 CFR part 52, subpart B. The
applicant for certification of the AP1000 design is Westinghouse
Electric Company LLC.
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\1\ AP1000 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the document
containing the Tier 1 and Tier 2 information and generic technical
specifications that is incorporated by reference into this appendix.
B. Generic technical specifications means the information required
by 10 CFR 50.36 and 50.36a for the portion of the plant that is within
the scope of this appendix.
C. Plant-specific DCD means the document maintained by an applicant
or licensee who references this appendix consisting of the information
in the generic DCD as modified and supplemented by the plant-specific
departures and exemptions made under Section VIII of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2 information.
Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria (ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by this
appendix (Tier 2 information). Compliance with Tier 2 is required, but
generic changes to and plant-specific departures from Tier 2 are
governed by Section VIII of this appendix. Compliance with Tier 2
provides a sufficient, but not the only acceptable, method for complying
with Tier 1. Compliance methods differing from Tier 2 must satisfy the
change process in Section VIII of this appendix. Regardless of these
differences, an applicant or licensee must meet the requirement in
Section III.B of this appendix to reference Tier 2 when referencing Tier
1. Tier 2 information includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c), with
the exception of generic technical specifications and conceptual design
information;
2. Supporting information on the inspections, tests, and analyses
that will be performed to demonstrate that the acceptance criteria in
the ITAAC have been met; and
3. Combined license (COL) action items (COL license information),
which identify certain matters that must be addressed in the site-
specific portion of the final safety analysis report (FSAR) by an
applicant who references this appendix. These items constitute
information requirements but are not the only acceptable set of
information in the FSAR. An applicant may depart from or
[[Page 137]]
omit these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit or
COL, these items are not requirements for the licensee unless such items
are restated in the FSAR.
4. The investment protection short-term availability controls in
Section 16.3 of the DCD.
F. Tier 2* means the portion of the Tier 2 information, designated
as such in the generic DCD, which is subject to the change process in
Section VIII.B.6 of this appendix. This designation expires for some
Tier 2* information under paragraph VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
1. Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are conservative
or essentially the same; or
2. Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by the NRC for the
intended application.
H. All other terms in this appendix have the meaning set out in 10
CFR 50.2, or 52.1, or Section 11 of the Atomic Energy Act of 1954, as
amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2 (including the investment protection short-term
availability controls in Section 16.3), and the generic TS in the AP1000
DCD (Revision 15, dated December 8, 2005) are approved for incorporation
by reference by the Director of the Office of the Federal Register on
February 27, 2006, under 5 U.S.C. 552(a) and 1 CFR part 51. Copies of
the generic DCD may be obtained from Ronald P. Vijuk, Manager, Passive
Plant Engineering, Westinghouse Electric Company, P.O. Box 355,
Pittsburgh, Pennsylvania 15230-0355. A copy of the generic DCD is also
available for examination and copying at the NRC Public Document Room,
One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.
Copies are available for examination at the NRC Library, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland, telephone (301) 415-
5610, e-mail [email protected] or at the National Archives and Records
Administration (NARA). For information on the availability of this
material at NARA, call (202) 741-6030 or go to http://www.archives.gov/
federal--register/code--of--federal--regulations/ibr--locations.html.
B. An applicant or licensee referencing this appendix, in accordance
with Section IV of this appendix, shall incorporate by reference and
comply with the requirements of this appendix, including Tier 1, Tier 2
(including the investment protection short-term availability controls in
Section 16.3 of the DCD), and the generic TS except as otherwise
provided in this appendix. Conceptual design information in the generic
DCD and the evaluation of severe accident mitigation design alternatives
in appendix 1B of the generic DCD are not part of this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then
Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the AP1000 design or NUREG-1793,
``Final Safety Evaluation Report Related to Certification of the AP1000
Standard Design,'' (FSER) and Supplement No. 1, then the generic DCD
controls.
E. Design activities for structures, systems, and components that
are wholly outside the scope of this appendix may be performed using
site characteristics, provided the design activities do not affect the
DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10 CFR
52.77, 52.79, and 52.80, comply with the following requirements:
1. Incorporate by reference, as part of its application, this
appendix.
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information and
using the same organization and numbering as the generic DCD for the
AP1000 design, as modified and supplemented by the applicant's
exemptions and departures;
b. The reports on departures from and updates to the plant-specific
DCD required by paragraph X.B of this appendix;
c. Plant-specific TS, consisting of the generic and site-specific TS
that are required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters and
interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47(a) that is not within the
scope of this appendix.
3. Include, in the plant-specific DCD, the proprietary information
and safeguards information referenced in the AP1000 DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the AP1000 design are in 10 CFR parts 20, 50,
73, and 100, codified as of January 23, 2006, that are applicable and
technically relevant,
[[Page 138]]
as described in the FSER (NUREG-1793) and Supplement No. 1.
B. The AP1000 design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter
Display Console;
2. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency)
feedwater system; and
3. Appendix A to 10 CFR part 50, GDC 17--Second offsite power supply
circuit.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the AP1000 design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of the
public. A conclusion that a matter is resolved includes the finding that
additional or alternative structures, systems, components, design
features, design criteria, testing, analyses, acceptance criteria, or
justifications are not necessary for the AP1000 design.
B. The Commission considers the following matters resolved within
the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings for issuance
of a COL, amendment of a COL, or renewal of a COL, proceedings held
under 10 CFR 52.103, and enforcement proceedings involving plants
referencing this appendix:
1. All nuclear safety issues, except for the generic TS and other
operational requirements, associated with the information in the FSER
and Supplement No. 1, Tier 1, Tier 2 (including referenced information,
which the context indicates is intended as requirements, and the
investment protection short-term availability controls in Section 16.3
of the DCD), and the rulemaking record for certification of the AP1000
design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and in
context, are intended as requirements in the generic DCD for the AP1000
design;
3. All generic changes to the DCD under and in compliance with the
change processes in Sections VIII.A.1 and VIII.B.1 of this appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, but
only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix, all
departures from Tier 2 under and in compliance with the change processes
in paragraph VIII.B.5 of this appendix that do not require prior NRC
approval, but only for that plant;
7. All environmental issues concerning severe accident mitigation
design alternatives associated with the information in the NRC's EA for
the AP1000 design and Appendix 1B of the generic DCD, for plants
referencing this appendix whose site parameters are within those
specified in the severe accident mitigation design alternatives
evaluation.
C. The Commission does not consider operational requirements for an
applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an applicant
or licensee who references this appendix by rule, regulation, order, or
license condition.
D. Except under the change processes in Section VIII of this
appendix, the Commission may not require an applicant or licensee who
references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures, systems,
components, or design features discussed in the generic DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the AP1000 DCD, in order to
request or participate in the hearing required by 10 CFR 52.85 or the
hearing provided under 10 CFR 52.103, or to request or participate in
any other hearing relating to this appendix in which interested persons
have adjudicatory hearing rights, shall first request access to such
information from Westinghouse. The request must state with
particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the public
in the NRC's public document room is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to prepare a
request for hearing, the request must be filed no later than 15 days
after publication in the Federal Register of the notice required either
by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide
the information sought, Westinghouse shall send a written response
within 10 days of receiving the request to the requesting person setting
forth with particularity the reasons for its refusal. The person
[[Page 139]]
may then request the Commission (or presiding officer, if a proceeding
has been established) to order disclosure. The person shall include
copies of the original request (and any subsequent clarifying
information provided by the requesting party to the applicant) and the
applicant's response. The Commission and presiding officer shall base
their decisions solely on the person's original request (including any
clarifying information provided by the requesting person to
Westinghouse), and Westinghouse's response. The Commission and presiding
officer may order Westinghouse to provide access to some or all of the
requested information, subject to an appropriate non-disclosure
agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
February 27, 2006, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee who
references this appendix until the application is withdrawn or the
license expires, including any period of extended operation under a
renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the design
change will result in a significant decrease in the level of safety
otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those for
which the change has been rendered technically irrelevant by action
taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under 10 CFR 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix, or to
ensure adequate protection of the public health and safety or the common
defense and security; and
b. Special circumstances as defined in 10 CFR 50.12(a) are present.
4. An applicant or licensee who references this appendix may request
an exemption from Tier 2 information. The Commission may grant such a
request only if it determines that the exemption will comply with the
requirements of 10 CFR 50.12(a). The Commission will deny a request for
an exemption from Tier 2, if it finds that the design change will result
in a significant decrease in the level of safety otherwise provided by
the design. The grant of an exemption to an applicant must be subject to
litigation in the same manner as other issues material to the license
hearing. The grant of an exemption to a licensee must be subject to an
opportunity for a hearing in the same manner as license amendments.
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 TS, or requires a license
amendment under paragraphs B.5.b or 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 frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component (SSC)
important to safety and 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 an SSC important to safety previously evaluated in the
plant-specific DCD;
(5) Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
[[Page 140]]
(7) Result in a design basis limit for a fission product barrier as
described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described in
the plant-specific DCD used in establishing the design bases or in the
safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an ex-
vessel severe accident design feature identified in the plant-specific
DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe accident
previously reviewed and determined to be not credible could become
credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously reviewed.
d. If a departure requires a license amendment under paragraph B.5.b
or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under paragraph
B.5 of this section does not require an exemption from this appendix.
f. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license or for operation under 10 CFR
52.103(a), who believes that an applicant or licensee who references
this appendix has not complied with paragraph VIII.B.5 of this appendix
when departing from Tier 2 information, may petition to admit into the
proceeding such a contention. In addition to compliance with the general
requirements of 10 CFR 2.309, the petition must demonstrate that the
departure does not comply with paragraph VIII.B.5 of this appendix.
Further, the petition must demonstrate that the change bears on an
asserted noncompliance with an ITAAC acceptance criterion in the case of
a 10 CFR 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a license
amendment. Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall certify
the matter directly to the Commission for determination of the
admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart from
Tier 2* information, which is designated with italicized text or
brackets and an asterisk in the generic DCD, without NRC approval. The
departure will not be considered a resolved issue, within the meaning of
Section VI of this appendix and 10 CFR 52.63(a)(5).
b. A licensee who references this appendix may not depart from the
following Tier 2* matters without prior NRC approval. A request for a
departure will be treated as a request for a license amendment under 10
CFR 50.90.
(1) Maximum fuel rod average burn-up.
(2) Fuel principal design requirements.
(3) Fuel criteria evaluation process.
(4) Fire areas.
(5) Human factors engineering.
(6) Small-break loss-of-coolant accident (LOCA) analysis
methodology.
c. A licensee who references this appendix may not, before the plant
first achieves full power following the finding required by 10 CFR
52.103(g), depart from the following Tier 2* matters except under
paragraph B.6.b of this section. After the plant first achieves full
power, the following Tier 2* matters revert to Tier 2 status and are
subject to the departure provisions in paragraph B.5 of this section.
(1) Nuclear Island structural dimensions.
(2) American Society of Mechanical Engineers Boiler & Pressure
Vessel Code (ASME Code), Section III, and Code Case-284.
(3) Design Summary of Critical Sections.
(4) American Concrete Institute (ACI) 318, ACI 349, American
National Standards Institute/American Institute of Steel Construction
(ANSI/AISC)-690, and American Iron and Steel Institute (AISI),
``Specification for the Design of Cold Formed Steel Structural Members,
Part 1 and 2,'' 1996 Edition and 2000 Supplement.
(5) Definition of critical locations and thicknesses.
(6) Seismic qualification methods and standards.
(7) Nuclear design of fuel and reactivity control system, except
burn-up limit.
(8) Motor-operated and power-operated valves.
(9) Instrumentation and control system design processes, methods,
and standards.
(10) Passive residual heat removal (PRHR) natural circulation test
(first plant only).
(11) Automatic depressurization system (ADS) and core make-up tank
(CMT) verification tests (first three plants only).
(12) Polar crane parked orientation.
(13) Piping design acceptance criteria.
(14) Containment vessel design parameters.
d. Departures from Tier 2* information that are made under paragraph
B.6 of this section do not require an exemption from this appendix.
C. Operational requirements.
1. Generic changes to generic TS and other operational requirements
that were completely reviewed and approved in the design certification
rulemaking and do not require a change to a design feature in the
generic DCD are governed by the requirements in 10 CFR 50.109. Generic
changes that require a change to a design feature in the generic
[[Page 141]]
DCD are governed by the requirements in paragraphs A or B of this
section.
2. Generic changes to generic TS and other operational requirements
are applicable to all applicants who reference this appendix, except
those for which the change has been rendered technically irrelevant by
action taken under paragraphs C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on generic
TS and other operational requirements that were completely reviewed and
approved, provided a change to a design feature in the generic DCD is
not required and special circumstances as defined in 10 CFR 2.335 are
present. The Commission may modify or supplement generic TS and other
operational requirements that were not completely reviewed and approved
or require additional TS and other operational requirements on a plant-
specific basis, provided a change to a design feature in the generic DCD
is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other operational
requirements. The Commission may grant such a request only if it
determines that the exemption will comply with the requirements of 10
CFR 52.7. The grant of an exemption must be subject to litigation in the
same manner as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for either the issuance,
amendment, or renewal of a license, or for operation under 10 CFR
52.103(a), who believes that an operational requirement approved in the
DCD or a TS derived from the generic TS must be changed may petition to
admit such a contention into the proceeding. The petition must comply
with the general requirements of 10 CFR 2.309 and must demonstrate why
special circumstances as defined in 10 CFR 2.335 are present, or
demonstrate compliance with the Commission's regulations in effect at
the time this appendix was approved, as set forth in Section V of this
appendix. Any other party may file a response to the petition. If, on
the basis of the petition and any response, the presiding officer
determines that a sufficient showing has been made, the presiding
officer shall certify the matter directly to the Commission for
determination of the admissibility of the contention. All other issues
with respect to the plant-specific TS or other operational requirements
are subject to a hearing as part of the license proceeding.
6. After issuance of a license, the generic TS have no further
effect on the plant-specific TS. Changes to the plant-specific TS will
be treated as license amendments under 10 CFR 50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1. An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities. A licensee may also proceed at its own risk with design,
procurement, construction, and preoperational activities, even though
the NRC may not have found that any particular ITAAC has been met.
2. The licensee who references this appendix shall notify the NRC
that the required inspections, tests, and analyses in the ITAAC have
been successfully completed and that the corresponding acceptance
criteria have been met.
3. If an activity is subject to an ITAAC and the applicant or
licensee who references this appendix has not demonstrated that the
ITAAC has been met, the applicant or licensee may either take corrective
actions to successfully complete that ITAAC, request an exemption from
the ITAAC under Section VIII of this appendix and 10 CFR 52.97(b), or
petition for rulemaking to amend this appendix by changing the
requirements of the ITAAC, under 10 CFR 2.802 and 52.97(b). Such
rulemaking changes to the ITAAC must meet the requirements of paragraph
VIII.A.1 of this appendix.
B.1. The NRC shall ensure that the required inspections, tests, and
analyses in the ITAAC are performed. The NRC shall verify that the
inspections, tests, and analyses referenced by the licensee have been
successfully completed and, based solely thereon, find that the
prescribed acceptance criteria have been met. At appropriate intervals
during construction, the NRC shall publish notices of the successful
completion of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall find
that the acceptance criteria in the ITAAC for the license are met before
fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the subject
of a Sec. 52.103(a) hearing, their expiration will occur upon final
Commission action in such a proceeding. However, subsequent
modifications must comply with the Tier 1 and Tier 2 design descriptions
in the plant-specific DCD unless the licensee has complied with the
applicable requirements of 10 CFR 52.98 and Section VIII of this
appendix.
[[Page 142]]
X. Records and Reporting
A. Records
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and the
generic TS and other operational requirements. The applicant shall
maintain the proprietary and safeguards information referenced in the
generic DCD for the period that this appendix may be referenced, as
specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application and
for the term of the license (including any period of renewal).
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).
B. Reporting
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any plant-
specific departures from the DCD, including a summary of the evaluation
of each. This report must be filed in accordance with the filing
requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes to and
plant-specific departures from the generic DCD made under Section VIII
of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in 10
CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and X.B.2
must be submitted as follows:
a. On the date that an application for a license referencing this
appendix is submitted, the application must include the report and any
updates to the generic DCD.
b. During the interval from the date of application for a license to
the date the Commission makes its findings required by 10 CFR 52.103(g),
the report must be submitted semi-annually. Updates to the plant-
specific DCD must be submitted annually and may be submitted along with
amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the final
safety analysis report for the facility, at the intervals required by 10
CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals
as specified in the license.
Sec. Appendixes E-M to Part 52 [Reserved]
Sec. Appendix N to Part 52--Standardization of Nuclear Power Plant
Designs: Combined Licenses To Construct and Operate Nuclear Power
Reactors of Identical Design at Multiple Sites
The Commission's regulations in part 2 of this chapter specifically
provide for the holding of hearings on particular issues separately from
other issues involved in hearings in licensing proceedings, and for the
consolidation of adjudicatory proceedings and of the presentations of
parties in adjudicatory proceedings such as licensing proceedings
(Sec. Sec. 2.316 and 2.317 of this chapter).
This appendix sets out the particular requirements and provisions
applicable to situations in which applications for combined licenses
under subpart C of this part are filed by one or more applicants for
licenses to construct and operate nuclear power reactors of identical
design (``common design'') to be located at multiple sites. \1\
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\1\ If the design for the power reactor(s) proposed in a particular
application is not identical to the others, that application may not be
processed under this appendix and subpart D of part 2 of this chapter.
---------------------------------------------------------------------------
1. Except as otherwise specified in this appendix or as the context
otherwise indicates, the provisions of subpart C of this part and
subpart D of part 2 of this chapter apply to combined license
applications subject to this appendix.
2. Each combined license application submitted pursuant to this
appendix must be submitted as specified in Sec. 52.75 and 10 CFR 2.101.
Each application must state that the applicant wishes to have the
application considered under 10 CFR part 52, appendix N, and must list
each of the applications to be treated together under this appendix.
3. Each application must include the information required by
Sec. Sec. 52.77, 52.79, and 52.80(a), provided however, that the
application must identify the common design, and, if applicable,
reference a standard design certification under subpart B of this part,
or the use of a reactor manufactured under subpart F of this part. The
final safety analysis report for each application must either
incorporate by reference or include the final safety analysis of the
common design, including, if applicable, the final safety analysis
report for the
[[Page 143]]
referenced design certification or the manufactured reactor. \2\
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\2\ As used in this appendix, the design of a nuclear power reactor
included in a single referenced safety analysis report means the design
of those structures, systems, and components important to radiological
health and safety and the common defense and security.
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4. Each combined license application submitted pursuant to this
appendix must contain an environmental report as required by Sec.
52.80(b), and which complies with the applicable provisions of 10 CFR
part 51, provided, however, that the application may incorporate by
reference a single environmental report on the environmental impacts of
the common design.
5. Upon a determination that each application is acceptable for
docketing under 10 CFR 2.101, each application will be docketed and a
notice of docketing for each application will be published in the
Federal Register, in accordance with 10 CFR 2.104, provided, however,
that the notice must state that the application will be processed under
the provisions of 10 CFR part 52, appendix N, and subpart D of part 2 of
this chapter. As the discretion of the Commission, a single notice of
docketing for multiple applications may be published in the Federal
Register.
6. The NRC staff shall prepare draft and final environmental impact
statements for each of the applications under part 51 of this chapter.
Scoping under 10 CFR 51.28 and 51.29 for each of the combined license
applications may be conducted simultaneously and joint scoping may be
conducted with respect to the environmental issues relevant to the
common design.
If the applications reference a standard design certification, then
the environmental impact statement for each of the applications must
incorporate by reference the design certification environmental
assessment. If the applications do not reference a standard design
certification, then the NRC staff shall prepare draft and final
supplemental environmental impact statements which address severe
accident mitigation design alternatives for the common design, which
must be incorporated by reference into the environmental impact
statement prepared for each application. Scoping under 10 CFR 51.28 and
51.29 for the supplemental environmental impact statement may be
conducted simultaneously, and may be part of the scoping for each of the
combined license applications.
7. The ACRS shall report on each of the applications as required by
Sec. 52.87. Each report must be limited to those safety matters for
each application which are not relevant to the common design. In
addition, the ACRS shall separately report on the safety of the common
design, provided, however, that the report need not address the safety
of a referenced standard design certification or reactor manufactured
under subpart F of this part.
8. The Commission shall designate a presiding officer to conduct the
proceeding with respect to the health and safety, common defense and
security, and environmental matters relating to the common design. The
hearing will be governed by the applicable provisions of subparts A, C,
G, L, N, and O of part 2 of this chapter relating to applications for
combined licenses. The presiding officer shall issue a partial initial
decision on the common design.
PART 53 [RESERVED]
PART 54_REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR
POWER PLANTS--Table of Contents
General Provisions
Sec.
54.1 Purpose.
54.3 Definitions.
54.4 Scope.
54.5 Interpretations.
54.7 Written communications.
54.9 Information collection requirements: OMB approval.
54.11 Public inspection of applications.
54.13 Completeness and accuracy of information.
54.15 Specific exemptions.
54.17 Filing of application.
54.19 Contents of application--general information.
54.21 Contents of application--technical information.
54.22 Contents of application--technical specifications.
54.23 Contents of application--environmental information.
54.25 Report of the Advisory Committee on Reactor Safeguards.
54.27 Hearings.
54.29 Standards for issuance of a renewed license.
54.30 Matters not subject to a renewal review.
54.31 Issuance of a renewed license.
54.33 Continuation of CLB and conditions of renewed license.
54.35 Requirements during term of renewed license.
54.37 Additional records and recordkeeping requirements.
54.41 Violations.
54.43 Criminal penalties.
Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83 Stat.
1244, as
[[Page 144]]
amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239,
2282); secs 201, 202, 206, 88 Stat. 1242, 1244, as amended (42 U.S.C.
5841, 5842).
Section 54.17 also issued under E.O.12829, 3 CFR, 1993 Comp., p.
570; E.O. 13526, as amended, 3 CFR, 1995 Comp., p. 333; E.O. 12968, 3
CFR, 1995 Comp., p. 391.
Source: 60 FR 22491, May 8, 1995, unless otherwise noted.
General Provisions
Sec. 54.1 Purpose.
This part governs the issuance of renewed operating licenses and
renewed combined licenses for nuclear power plants licensed pursuant to
Sections 103 or 104b of the Atomic Energy Act of 1954, as amended, and
Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242).
[72 FR 49560, Aug. 28, 2007]
Sec. 54.3 Definitions.
(a) As used in this part,
Current licensing basis (CLB) is the set of NRC requirements
applicable to a specific plant and a licensee's written commitments for
ensuring compliance with and operation within applicable NRC
requirements and the plant-specific design basis (including all
modifications and additions to such commitments over the life of the
license) that are docketed and in effect. The CLB includes the NRC
regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51,
52, 54, 55, 70, 72, 73, 100 and appendices thereto; orders; license
conditions; exemptions; and technical specifications. It also includes
the plant-specific design-basis information defined in 10 CFR 50.2 as
documented in the most recent final safety analysis report (FSAR) as
required by 10 CFR 50.71 and the licensee's commitments remaining in
effect that were made in docketed licensing correspondence such as
licensee responses to NRC bulletins, generic letters, and enforcement
actions, as well as licensee commitments documented in NRC safety
evaluations or licensee event reports.
Integrated plant assessment (IPA) is a licensee assessment that
demonstrates that a nuclear power plant facility's structures and
components requiring aging management review in accordance with Sec.
54.21(a) for license renewal have been identified and that the effects
of aging on the functionality of such structures and components will be
managed to maintain the CLB such that there is an acceptable level of
safety during the period of extended operation.
Nuclear power plant means a nuclear power facility of a type
described in 10 CFR 50.21(b) or 50.22.
Renewed combined license means a combined license originally issued
under part 52 of this chapter for which an application for renewal is
filed in accordance with 10 CFR 52.107 and issued under this part.
Time-limited aging analyses, for the purposes of this part, are
those licensee calculations and analyses that:
(1) Involve systems, structures, and components within the scope of
license renewal, as delineated in Sec. 54.4(a);
(2) Consider the effects of aging;
(3) Involve time-limited assumptions defined by the current
operating term, for example, 40 years;
(4) Were determined to be relevant by the licensee in making a
safety determination;
(5) Involve conclusions or provide the basis for conclusions related
to the capability of the system, structure, and component to perform its
intended functions, as delineated in Sec. 54.4(b); and
(6) Are contained or incorporated by reference in the CLB.
(b) All other terms in this part have the same meanings as set out
in 10 CFR 50.2 or Section 11 of the Atomic Energy Act, as applicable.
[60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]
Sec. 54.4 Scope.
(a) Plant systems, structures, and components within the scope of
this part are--
(1) Safety-related systems, structures, and components which are
those relied upon to remain functional during and following design-basis
events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following
functions--
(i) The integrity of the reactor coolant pressure boundary;
(ii) The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
[[Page 145]]
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec.
100.11 of this chapter, as applicable.
(2) All nonsafety-related systems, structures, and components whose
failure could prevent satisfactory accomplishment of any of the
functions identified in paragraphs (a)(1) (i), (ii), or (iii) of this
section.
(3) All systems, structures, and components relied on in safety
analyses or plant evaluations to perform a function that demonstrates
compliance with the Commission's regulations for fire protection (10 CFR
50.48), environmental qualification (10 CFR 50.49), pressurized thermal
shock (10 CFR 50.61), anticipated transients without scram (10 CFR
50.62), and station blackout (10 CFR 50.63).
(b) The intended functions that these systems, structures, and
components must be shown to fulfill in Sec. 54.21 are those functions
that are the bases for including them within the scope of license
renewal as specified in paragraphs (a) (1)-(3) of this section.
[60 FR 22491, May 8, 1995, as amended at 61 FR 65175, Dec. 11, 1996; 64
FR 72002, Dec. 23, 1999]
Sec. 54.5 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission other than a written
interpretation by the General Counsel will be recognized to be binding
upon the Commission.
Sec. 54.7 Written communications.
All applications, correspondence, reports, and other written
communications shall be filed in accordance with applicable portions of
10 CFR 50.4.
Sec. 54.9 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501, et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number. OMB
has approved the information collection requirements contained in this
part under control number 3150-0155.
(b) The approved information requirements contained in this part
appear in Sec. Sec. 54.13, 54.15, 54.17, 54.19, 54.21, 54.22, 54.23,
54.33, and 54.37.
[60 FR 22491, May 8, 1995, as amended at 62 FR 52188, Oct. 6, 1997; 67
FR 67100, Nov. 4, 2002]
Sec. 54.11 Public inspection of applications.
Applications and documents submitted to the Commission in connection
with renewal applications may be made available for public inspection in
accordance with the provisions of the regulations contained in 10 CFR
part 2.
Sec. 54.13 Completeness and accuracy of information.
(a) Information provided to the Commission by an applicant for a
renewed license or information required by statute or by the
Commission's regulations, orders, or license conditions to be maintained
by the applicant must be complete and accurate in all material respects.
(b) Each applicant shall notify the Commission of information
identified by the applicant as having, for the regulated activity, a
significant implication for public health and safety or common defense
and security. An applicant violates this paragraph only if the applicant
fails to notify the Commission of information that the applicant has
identified as having a significant implication for public health and
safety or common defense and security. Notification must be provided to
the Administrator of the appropriate regional office within 2 working
days of identifying the information. This requirement is not applicable
to information that is already required to be provided to the Commission
by other reporting or updating requirements.
[[Page 146]]
Sec. 54.15 Specific exemptions.
Exemptions from the requirements of this part may be granted by the
Commission in accordance with 10 CFR 50.12.
Sec. 54.17 Filing of application.
(a) The filing of an application for a renewed license must be in
accordance with subpart A of 10 CFR part 2 and 10 CFR 50.4 and 50.30.
(b) Any person who is a citizen, national, or agent of a foreign
country, or any corporation, or other entity which the Commission knows
or has reason to know is owned, controlled, or dominated by an alien, a
foreign corporation, or a foreign government, is ineligible to apply for
and obtain a renewed license.
(c) An application for a renewed license may not be submitted to the
Commission earlier than 20 years before the expiration of the operating
license or combined license currently in effect.
(d) An applicant may combine an application for a renewed license
with applications for other kinds of licenses.
(e) An application may incorporate by reference information
contained in previous applications for licenses or license amendments,
statements, correspondence, or reports filed with the Commission,
provided that the references are clear and specific.
(f) If the application contains Restricted Data or other defense
information, it must be prepared in such a manner that all Restricted
Data and other defense information are separated from unclassified
information in accordance with 10 CFR 50.33(j).
(g) As part of its application, and in any event before the receipt
of Restricted Data or classified National Security Information or the
issuance of a renewed license, the applicant shall agree in writing that
it will not permit any individual to have access to or any facility to
possess Restricted Data or classified National Security Information
until the individual and/or facility has been approved for such access
under the provisions of 10 CFR parts 25 and/or 95. The agreement of the
applicant in this regard shall be deemed part of the renewed license,
whether so stated therein or not.
[60 FR 22491, May 8, 1995, as amended at 62 FR 17690, Apr. 11, 1997; 72
FR 49560, Aug. 28, 2007]
Sec. 54.19 Contents of application--general information.
(a) Each application must provide the information specified in 10
CFR 50.33 (a) through (e), (h), and (i). Alternatively, the application
may incorporate by reference other documents that provide the
information required by this section.
(b) Each application must include conforming changes to the standard
indemnity agreement, 10 CFR 140.92, Appendix B, to account for the
expiration term of the proposed renewed license.
Sec. 54.21 Contents of application--technical information.
Each application must contain the following information:
(a) An integrated plant assessment (IPA). The IPA must--
(1) For those systems, structures, and components within the scope
of this part, as delineated in Sec. 54.4, identify and list those
structures and components subject to an aging management review.
Structures and components subject to an aging management review shall
encompass those structures and components--
(i) That perform an intended function, as described in Sec. 54.4,
without moving parts or without a change in configuration or properties.
These structures and components include, but are not limited to, the
reactor vessel, the reactor coolant system pressure boundary, steam
generators, the pressurizer, piping, pump casings, valve bodies, the
core shroud, component supports, pressure retaining boundaries, heat
exchangers, ventilation ducts, the containment, the containment liner,
electrical and mechanical penetrations, equipment hatches, seismic
Category I structures, electrical cables and connections, cable trays,
and electrical cabinets, excluding, but not limited to, pumps (except
casing), valves (except body), motors, diesel generators, air
compressors, snubbers, the control rod drive, ventilation
[[Page 147]]
dampers, pressure transmitters, pressure indicators, water level
indicators, switchgears, cooling fans, transistors, batteries, breakers,
relays, switches, power inverters, circuit boards, battery chargers, and
power supplies; and
(ii) That are not subject to replacement based on a qualified life
or specified time period.
(2) Describe and justify the methods used in paragraph (a)(1) of
this section.
(3) For each structure and component identified in paragraph (a)(1)
of this section, demonstrate that the effects of aging will be
adequately managed so that the intended function(s) will be maintained
consistent with the CLB for the period of extended operation.
(b) CLB changes during NRC review of the application. Each year
following submittal of the license renewal application and at least 3
months before scheduled completion of the NRC review, an amendment to
the renewal application must be submitted that identifies any change to
the CLB of the facility that materially affects the contents of the
license renewal application, including the FSAR supplement.
(c) An evaluation of time-limited aging analyses. (1) A list of
time-limited aging analyses, as defined in Sec. 54.3, must be provided.
The applicant shall demonstrate that--
(i) The analyses remain valid for the period of extended operation;
(ii) The analyses have been projected to the end of the period of
extended operation; or
(iii) The effects of aging on the intended function(s) will be
adequately managed for the period of extended operation.
(2) A list must be provided of plant-specific exemptions granted
pursuant to 10 CFR 50.12 and in effect that are based on time-limited
aging analyses as defined in Sec. 54.3. The applicant shall provide an
evaluation that justifies the continuation of these exemptions for the
period of extended operation.
(d) An FSAR supplement. The FSAR supplement for the facility must
contain a summary description of the programs and activities for
managing the effects of aging and the evaluation of time-limited aging
analyses for the period of extended operation determined by paragraphs
(a) and (c) of this section, respectively.
Sec. 54.22 Contents of application--technical specifications.
Each application must include any technical specification changes or
additions necessary to manage the effects of aging during the period of
extended operation as part of the renewal application. The justification
for changes or additions to the technical specifications must be
contained in the license renewal application.
Sec. 54.23 Contents of application--environmental information.
Each application must include a supplement to the environmental
report that complies with the requirements of subpart A of 10 CFR part
51.
Sec. 54.25 Report of the Advisory Committee on Reactor Safeguards.
Each renewal application will be referred to the Advisory Committee
on Reactor Safeguards for a review and report. Any report will be made
part of the record of the application and made available to the public,
except to the extent that security classification prevents disclosure.
Sec. 54.27 Hearings.
A notice of an opportunity for a hearing will be published in the
Federal Register in accordance with 10 CFR 2.105. In the absence of a
request for a hearing filed within 30 days by a person whose interest
may be affected, the Commission may issue a renewed operating license or
renewed combined license without a hearing upon 30-day notice and
publication in the Federal Register of its intent to do so.
[72 FR 49560, Aug. 28, 2007]
Sec. 54.29 Standards for issuance of a renewed license.
A renewed license may be issued by the Commission up to the full
term authorized by Sec. 54.31 if the Commission finds that:
(a) Actions have been identified and have been or will be taken with
respect to the matters identified in paragraphs (a)(1) and (a)(2) of
this section, such that there is reasonable assurance that
[[Page 148]]
the activities authorized by the renewed license will continue to be
conducted in accordance with the CLB, and that any changes made to the
plant's CLB in order to comply with this paragraph are in accord with
the Act and the Commission's regulations. These matters are:
(1) managing the effects of aging during the period of extended
operation on the functionality of structures and components that have
been identified to require review under Sec. 54.21(a)(1); and
(2) time-limited aging analyses that have been identified to require
review under Sec. 54.21(c).
(b) Any applicable requirements of subpart A of 10 CFR part 51 have
been satisfied.
(c) Any matters raised under Sec. 2.335 have been addressed.
[60 FR 22491, May 8, 1995, as amended at 69 FR 2279, Jan. 14, 2004]
Sec. 54.30 Matters not subject to a renewal review.
(a) If the reviews required by Sec. 54.21 (a) or (c) show that
there is not reasonable assurance during the current license term that
licensed activities will be conducted in accordance with the CLB, then
the licensee shall take measures under its current license, as
appropriate, to ensure that the intended function of those systems,
structures or components will be maintained in accordance with the CLB
throughout the term of its current license.
(b) The licensee's compliance with the obligation under Paragraph
(a) of this section to take measures under its current license is not
within the scope of the license renewal review.
Sec. 54.31 Issuance of a renewed license.
(a) A renewed license will be of the class for which the operating
license or combined license currently in effect was issued.
(b) A renewed license will be issued for a fixed period of time,
which is the sum of the additional amount of time beyond the expiration
of the operating license or combined license (not to exceed 20 years)
that is requested in a renewal application plus the remaining number of
years on the operating license or combined license currently in effect.
The term of any renewed license may not exceed 40 years.
(c) A renewed license will become effective immediately upon its
issuance, thereby superseding the operating license or combined license
previously in effect. If a renewed license is subsequently set aside
upon further administrative or judicial appeal, the operating license or
combined license previously in effect will be reinstated unless its term
has expired and the renewal application was not filed in a timely
manner.
(d) A renewed license may be subsequently renewed in accordance with
all applicable requirements.
[60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]
Sec. 54.33 Continuation of CLB and conditions of renewed license.
(a) Whether stated therein or not, each renewed license will contain
and otherwise be subject to the conditions set forth in 10 CFR 50.54.
(b) Each renewed license will be issued in such form and contain
such conditions and limitations, including technical specifications, as
the Commission deems appropriate and necessary to help ensure that
systems, structures, and components subject to review in accordance with
Sec. 54.21 will continue to perform their intended functions for the
period of extended operation. In addition, the renewed license will be
issued in such form and contain such conditions and limitations as the
Commission deems appropriate and necessary to help ensure that systems,
structures, and components associated with any time-limited aging
analyses will continue to perform their intended functions for the
period of extended operation.
(c) Each renewed license will include those conditions to protect
the environment that were imposed pursuant to 10 CFR 50.36b and that are
part of the CLB for the facility at the time of issuance of the renewed
license. These conditions may be supplemented or amended as necessary to
protect the environment during the term of the renewed license and will
be derived from information contained in the supplement to the
environmental report submitted pursuant to 10 CFR part 51, as
[[Page 149]]
analyzed and evaluated in the NRC record of decision. The conditions
will identify the obligations of the licensee in the environmental area,
including, as appropriate, requirements for reporting and recordkeeping
of environmental data and any conditions and monitoring requirements for
the protection of the nonaquatic environment.
(d) The licensing basis for the renewed license includes the CLB, as
defined in Sec. 54.3(a); the inclusion in the licensing basis of
matters such as licensee commitments does not change the legal status of
those matters unless specifically so ordered pursuant to paragraphs (b)
or (c) of this section.
Sec. 54.35 Requirements during term of renewed license.
During the term of a renewed license, licensees shall be subject to
and shall continue to comply with all Commission regulations contained
in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72,
73, and 100, and the appendices to these parts that are applicable to
holders of operating licenses or combined licenses, respectively.
[72 FR 49560, Aug. 28, 2007]
Sec. 54.37 Additional records and recordkeeping requirements.
(a) The licensee shall retain in an auditable and retrievable form
for the term of the renewed operating license or renewed combined
license all information and documentation required by, or otherwise
necessary to document compliance with, the provisions of this part.
(b) After the renewed license is issued, the FSAR update required by
10 CFR 50.71(e) must include any systems, structures, and components
newly identified that would have been subject to an aging management
review or evaluation of time-limited aging analyses in accordance with
Sec. 54.21. This FSAR update must describe how the effects of aging
will be managed such that the intended function(s) in Sec. 54.4(b) will
be effectively maintained during the period of extended operation.
[60 FR 22491, May 8, 1995, as amended at 72 FR 49560, Aug. 28, 2007]
Sec. 54.41 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of the following acts--
(1) The Atomic Energy Act of 1954, as amended.
(2) Title II of the Energy Reorganization Act of 1974, as amended or
(3) A regulation or order issued pursuant to those acts.
(b) The Commission may obtain a court order for the payment of a
civil penalty imposed under Section 234 of the Atomic Energy Act--
(1) For violations of the following--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the Atomic Energy Act of 1954, as amended;
(ii) Section 206 of the Energy Reorganization Act;
(iii) Any rule, regulation, or order issued pursuant to the sections
specified in paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation of any license issued under
the sections specified in paragraph (b)(1)(i) of this section.
(2) For any violation for which a license may be revoked under
Section 186 of the Atomic Energy Act of 1954, as amended.
Sec. 54.43 Criminal penalties.
(a) Section 223 of the Atomic Energy Act of 1954, as amended,
provides for criminal sanctions for willful violations of, attempted
violation of, or conspiracy to violate, any regulation issued under
sections 161b, 161i, or 161o of the Act. For purposes of section 223,
all the regulations in part 54 are issued under one or more of sections
161b, 161i, or 161o, except for the sections listed in paragraph (b) of
this section.
(b) The regulations in part 54 that are not issued under Sections
161b, 161i, or 161o for the purposes of Section 223 are as follows:
Sec. Sec. 54.1, 54.3, 54.4, 54.5, 54.7, 54.9, 54.11, 54.15, 54.17,
54.19, 54.21, 54.22, 54.23, 54.25, 54.27, 54.29, 54.31, 54.41, and
54.43.
[[Page 150]]
PART 55_OPERATORS' LICENSES--Table of Contents
Subpart A_General Provisions
Sec.
55.1 Purpose.
55.2 Scope.
55.3 License requirements.
55.4 Definitions.
55.5 Communications.
55.6 Interpretations.
55.7 Additional requirements.
55.8 Information collection requirements: OMB approval.
55.9 Completeness and accuracy of information.
Subpart B_Exemptions
55.11 Specific exemptions.
55.13 General exemptions.
Subpart C_Medical Requirements
55.21 Medical examination.
55.23 Certification.
55.25 Incapacitation because of disability or illness.
55.27 Documentation.
Subpart D_Applications
55.31 How to apply.
55.33 Disposition of an initial application.
55.35 Re-applications.
Subpart E_Written Examinations and Operating Tests
55.40 Implementation.
55.41 Written examination: Operators.
55.43 Written examination: Senior operators.
55.45 Operating tests.
55.46 Simulation facilities.
55.47 Waiver of examination and test requirements.
55.49 Integrity of examinations and tests.
Subpart F_Licenses
55.51 Issuance of licenses.
55.53 Conditions of licenses.
55.55 Expiration.
55.57 Renewal of licenses.
55.59 Requalification.
Subpart G_Modification and Revocation of Licenses
55.61 Modification and revocation of licenses.
Subpart H_Enforcement
55.71 Violations.
55.73 Criminal penalties.
Authority: Secs. 107, 161, 182, 68 Stat. 939, 948, 953, as amended,
sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2137, 2201, 2232, 2282);
secs. 201, as amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C.
5841, 5842); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
Sections 55.41, 55.43, 55.45, and 55.59 also issued under sec. 306,
Pub. L. 97-425, 96 Stat. 2262 (42 U.S.C. 10226).
Section 55.61 also issued under secs. 186, 187, 68 Stat. 955 (42
U.S.C. 2236, 2237).
Source: 52 FR 9460, Mar. 25, 1987, unless otherwise noted.
Subpart A_General Provisions
Sec. 55.1 Purpose.
The regulations in this part:
(a) Establish procedures and criteria for the issuance of licenses
to operators and senior operators of utilization facilities licensed
under the Atomic Energy Act of 1954, as amended, or Section 202 of the
Energy Reorganization Act of 1974, as amended, and part 50, part 52, or
part 54 of this chapter,
(b) Provide for the terms and conditions upon which the Commission
will issue or modify these licenses, and
(c) Provide for the terms and conditions to maintain and renew these
licenses.
[52 FR 9460, Mar. 25, 1987, as amended at 72 FR 49560, Aug. 28, 2007]
Sec. 55.2 Scope.
The regulations in this part apply to--
(a) Any individual who manipulates the controls of any utilization
facility licensed under parts 50, 52, or 54 of this chapter,
(b) Any individual designated by a facility licensee to be
responsible for directing any licensed activity of a licensed operator.
(c) Any facility license.
[52 FR 9460, Mar. 25, 1987, as amended at 59 FR 5938, Feb. 9, 1994; 72
FR 49560, Aug. 28, 2007]
Sec. 55.3 License requirements.
A person must be authorized by a license issued by the Commission to
perform the function of an operator or a senior operator as defined in
this part.
Sec. 55.4 Definitions.
As used in this part:
[[Page 151]]
Act means the Atomic Energy Act of 1954, including any amendments to
the Act.
Actively performing the functions of an operator or senior operator
means that an individual has a position on the shift crew that requires
the individual to be licensed as defined in the facility's technical
specifications, and that the individual carries out and is responsible
for the duties covered by that position.
Commission means the Nuclear Regulatory Commission or its duly
authorized representatives.
Controls when used with respect to a nuclear reactor means apparatus
and mechanisms the manipulation of which directly affects the reactivity
or power level of the reactor.
Facility means any utilization facility as defined in part 50 of
this chapter. In cases for which a license is issued for operation of
two or more facilities, facility means all facilities identified in the
license.
Facility licensee means an applicant for or holder of a license for
a facility.
Licensee means an individual licensed operator or senior operator.
Operator means any individual licensed under this part to manipulate
a control of a facility.
Performance testing means testing conducted to verify a simulation
facility's performance as compared to actual or predicted reference
plant performance.
Physician means an individual licensed by a State or territory of
the United States, the District of Columbia or the Commonwealth of
Puerto Rico to dispense drugs in the practice of medicine.
Plant-referenced simulator means a simulator modeling the systems of
the reference plant with which the operator interfaces in the control
room, including operating consoles, and which permits use of the
reference plant's procedures.
Reference plant means the specific nuclear power plant from which a
simulation facility's control room configuration, system control
arrangement, and design data are derived.
Senior operator means any individual licensed under this part to
manipulate the controls of a facility and to direct the licensed
activities of licensed operators.
Simulation facility means one or more of the following components,
alone or in combination: used for either the partial conduct of
operating tests for operators, senior operators, and license applicants,
or to establish on-the-job training and experience prerequisites for
operator license eligibility:
(1) A plant-referenced simulator;
(2) A Commission-approved simulator under Sec. 55.46(b); or
(3) Another simulation device, including part-task and limited scope
simulation devices, approved under Sec. 55.46(b).
Systems approach to training means a training program that includes
the following five elements:
(1) Systematic analysis of the jobs to be performed.
(2) Learning objectives derived from the analysis which describe
desired performance after training.
(3) Training design and implementation based on the learning
objectives.
(4) Evaluation of trainee mastery of the objectives during training.
(5) Evaluation and revision of the training based on the performance
of trained personnel in the job setting.
United States, when used in a geographical sense, includes Puerto
Rico and all territories and possessions of the United States.
[52 FR 9460, Mar. 25, 1987, as amended at 66 FR 52667, Oct. 17, 2001]
Sec. 55.5 Communications.
(a) Except as provided under a regional licensing program identified
in paragraph (b) of this section, an applicant or licensee or facility
licensee shall submit any communication or report concerning the
regulations in this part and shall submit any application filed under
these regulations to the Commission as follows:
(1) By mail addressed to--Director, Office of Nuclear Reactor
Regulation or Director, Office of New Reactors, as appropriate, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; or
(2) By delivery in person to the NRC's offices at 11555 Rockville
Pike, Rockville, Maryland, or
[[Page 152]]
(3) Where practicable, by electronic submission, for example, via
Electronic Information Exchange, or CD-ROM. Electronic submissions must
be made in a manner that enables the NRC to receive, read, authenticate,
distribute, and archive the submission, and process and retrieve it a
single page at a time. Detailed guidance on making electronic
submissions can be obtained by visiting the NRC's Web site at http://
www.nrc.gov/site-help/e-submittals.html; by e-mail to
[email protected]; or by writing the Office of Information Services,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The
guidance discusses, among other topics, the formats the NRC can accept,
the use of electronic signatures, and the treatment of nonpublic
information.
(b)(1) Except for test and research reactor facilities, the
Director, Office of Nuclear Reactor Regulation or Director, Office of
New Reactors, as appropriate, has delegated to the Regional
Administrators of Regions I, II, III, and IV authority and
responsibility under the regulations in this part for the issuance and
renewal of licenses for operators and senior operators of nuclear power
reactors licensed under 10 CFR part 50 or part 52 of this chapter and
located in these regions.
(2) Any application for a license or license renewal filed under the
regulations in this part involving a nuclear power reactor licensed
under 10 CFR part 50 or part 52 of this chapter and any related inquiry,
communication, information, or report must be submitted to the Regional
Administrator by an appropriate method listed in paragraph (a) of this
section. The Regional Administrator or the Administrator's designee will
transmit to the Director, Office of New Reactors or Director, Office of
Nuclear Reactor Regulation, as appropriate, any matter that is not
within the scope of the Regional Administrator's delegated authority.
(i) If the nuclear power reactor is located in Region I, submissions
must be made to the Regional Administrator of Region I. Submissions by
mail or hand delivery must be addressed to the Administrator at U.S.
Nuclear Regulatory Commission, 475 Allendale Road, King of Prussia,
Pennsylvania 19406-1415; where e-mail is appropriate it should be
addressed to [email protected].
(ii) If the nuclear power reactor is located in Region II,
submissions must be made to the Regional Administrator of Region II.
Submissions by mail or hand delivery must be addressed to the Regional
Administrator at U.S. Nuclear Regulatory Commission, 245 Peachtree
Center Avenue, NE., Suite 1200, Atlanta, Georgia 30303-1257. Where e-
mail is appropriate, it should be addressed to
[email protected].
(iii) If the nuclear power reactor is located in Region III,
submissions must be made to the Regional Administrator of Region III.
Submissions by mail or hand delivery must be addressed to the
Administrator at U.S. Nuclear Regulatory Commission, 2443 Warrenville
Road, Suite 210, Lisle, IL 60532-4352; where e-mail is appropriate it
should be addressed to [email protected].
(iv) If the nuclear power reactor is located in Region IV,
submissions must be made to the Regional Administrator of Region IV.
Submission by mail or hand delivery must be addressed to the
Administrator at U.S. Nuclear Regulatory Commission, 612 E. Lamar Blvd.,
Suite 400, Arlington, Texas 76011-4125; where e-mail is appropriate, it
should be addressed to [email protected].
(3)(i) Any application for a license or license renewal filed under
the regulations in this part involving a test and research reactor
facility licensed under 10 CFR part 50 and any related inquiry,
communication, information, or report must be submitted to the Office of
Nuclear Reactor Regulation, Division of Regulatory Improvement Programs
at the NRC's headquarters, by an appropriate method listed in paragraph
(a) of this section.
(ii) For all test and research reactor facilities located in the
NRC's Regions, submissions must be made to the Office of Nuclear Reactor
Regulation, Director of the Division of Regulatory Improvement Programs
at the NRC's
[[Page 153]]
headquarters, by an appropriate method listed in paragraph (a) of this
section.
[52 FR 9460, Mar. 25, 1987, as amended at 53 FR 6139, Mar. 1, 1988; 53
FR 43421, Oct. 27, 1988; 55 FR 41335, Oct. 11, 1990; 59 FR 17466, Apr.
13, 1994; 61 FR 9902, Mar. 12, 1996; 67 FR 77653, Dec. 19, 2002; 68 FR
58812, Oct. 10, 2003; 71 FR 15011, Mar. 27, 2006; 72 FR 49560, Aug. 28,
2007; 73 FR 5724, Jan. 31, 2008; 73 FR 30458, May 28, 2008; 74 FR 62682,
Dec. 1, 2009; 75 FR 21980, Apr. 27, 2010]
Sec. 55.6 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission other than a written
interpretation by the General Counsel will be recognized to be binding
upon the Commission.
Sec. 55.7 Additional requirements.
The Commission may, by rule, regulation, or order, impose upon any
licensee such requirements, in addition to those established in the
regulations in this part, as it deems appropriate or necessary to
protect health and to minimize danger to life or property.
Sec. 55.8 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number. OMB
has approved the information collection requirements contained in this
part under control number 3150-0018.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 55.11, 55.25, 55.27, 55.31, 55.35, 55.40,
55.41, 55.43, 55.45, 55.47, 55.53, 55.57, and 55.59.
(c) This part contains information collection requirements in
addition to those approved under the control number specified in
paragraph (a) of this section. These information collection requirements
and the control numbers under which they are approved are as follows:
(1) In Sec. Sec. 55.23, 55.25, 55.27, 55.31, NRC Form 396 is
approved under control number 3150-0024.
(2) In Sec. Sec. 55.31, 55.35, 55.47, and 55.57, NRC Form 398 is
approved under control number 3150-0090.
[62 FR 52188, Oct. 6, 1997, as amended at 64 FR 19878, Apr. 23, 1999; 66
FR 52667, Oct. 17, 2001; 67 FR 67100, Nov. 4, 2002]
Sec. 55.9 Completeness and accuracy of information.
Information provided to the Commission by an applicant for a license
or by a licensee or information required by statute or by the
Commission's regulations, orders, or license conditions to be maintained
by the applicant or the licensee shall be complete and accurate in all
material respects.
[52 FR 49372, Dec. 31, 1987]
Subpart B_Exemptions
Sec. 55.11 Specific exemptions.
The Commission may, upon application by an interested person, or
upon its own initiative, grant such exemptions from the requirements of
the regulations in this part as it determines are authorized by law and
will not endanger life or property and are otherwise in the public
interest.