[Federal Register Volume 68, Number 128 (Thursday, July 3, 2003)]
[Proposed Rules]
[Pages 40026-40074]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-16413]
[[Page 40025]]
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Part II
Nuclear Regulatory Commission
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10 CFR Part 2, et al.
Early Site Permits, Standard Design Certifications, and Combined
Licenses for Nuclear Power Plants; Proposed Rule
Federal Register / Vol. 68, No. 128 / Thursday, July 3, 2003 /
Proposed Rules
[[Page 40026]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 2, 20, 21, 50, 51, 52, 72, 73, 140, and 170
RIN 3150-AG24
Early Site Permits, Standard Design Certifications, and Combined
Licenses for Nuclear Power Plants
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC or Commission) is
proposing to amend its requirements for early site permits, standard
design certifications, combined licenses for nuclear power plants, and
for other licensing processes. The amendments are based on the NRC
staff's experience with the previous design certification reviews and
on discussions with stakeholders about the early site permit (ESP),
design certification, and combined license (COL) processes. This action
is expected to improve the effectiveness of the licensing processes for
future applicants.
DATES: Submit comments by September 16, 2003. Comments received after
this date will be considered, if it is practical to do so, but the
Commission is able to ensure consideration only for comments received
on or before this date.
ADDRESSES: You may submit comments by any one of the following methods.
Please include the following number RIN 3150-AG24 in the subject line
of your comments. Comments submitted in writing or in electronic form
will be made available to the public in their entirety on the NRC
rulemaking Web site. Personal information will not be removed from your
comments.
Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us
directly at (301) 415-1966. You may also submit comments via the NRC's
rulemaking Web site at http://ruleforum.llnl.gov. Address questions
about our rulemaking Web site to Carol Gallagher (301) 415-5905; email
[email protected].
Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays. (Telephone
(301) 415-1966).
Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at
(301) 415-1101.
Publicly available documents related to this rulemaking may be
examined and copied for a fee at the NRC's Public Document Room (PDR),
Public File Area O1 F21, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland. Selected documents, including comments, can be
viewed and downloaded electronically via the NRC rulemaking Web site at
http://ruleforum.llnl.gov.
Publicly available documents created or received at the NRC after
November 1, 1999, are available electronically at the NRC's Electronic
Reading Room at http://www.nrc.gov/NRC/ADAMS/index.html. From this
site, the public can gain entry into the NRC's Agencywide Document
Access and Management System (ADAMS), which provides text and image
files of NRC's public documents. If you do not have access to ADAMS or
if there are problems in accessing the documents located in ADAMS,
contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737
or by e-mail to [email protected].
FOR FURTHER INFORMATION CONTACT: Jerry N. Wilson, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone (301) 415-3145, email [email protected]; or Nanette V.
Gilles, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone (301) 415-1180, e-mail
[email protected].
SUPPLEMENTARY INFORMATION:
I. Background.
II. Reorganization of 10 CFR Part 52.
III. Discussion of Substantive Changes.
A. 10 CFR Part 52, Early Site Permits, Standard Design
Certifications, and Combined Licenses for Nuclear Power Plants.
General Provisions.
Early Site Permits.
Early Site Reviews.
Standard Design Certifications.
Design Certification Backfit Requirement.
Standard Design Approvals.
Combined Licenses.
Referencing an Early Site Permit.
Testing Requirements for Advanced Reactors.
Probabilistic Risk Assessments.
Resolution of ITAAC.
Commission Finding on Acceptance Criteria.
Combined License Change Process.
Design Certifications for ABWR, System 80+, and AP600.
B. 10 CFR Part 2, Rules of Practice for Domestic Licensing
Proceedings and Issuance of Orders.
C. 10 CFR Part 20, Standards for Protection Against Radiation.
D. 10 CFR Part 21, Reporting of Defects and Noncompliance.
E. 10 CFR Part 50, Domestic Licensing of Production and
Utilization Facilities.
F. 10 CFR Part 51, Environmental Protection Regulations for
Domestic Licensing and Related Regulatory Functions.
G. 10 CFR Part 72, Licensing Requirements for the Independent
Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.
H. 10 CFR Part 73, Physical Protection of Plants and Materials.
I. 10 CFR Part 140, Financial Protection Requirements and
Indemnity Agreements.
J. 10 CFR Part 170, Fees for Facilities, Materials, Import and
Export Licenses, and Other Regulatory Services Under the Atomic
Energy Act of 1954, as Amended.
IV. Specific Requests for Comments.
V. Availability of Documents.
VI. Plain Language.
VII. Voluntary Consensus Standards.
VIII.Environmental Impact'Categorical Exclusion.
IX. Paperwork Reduction Act Statement.
X. Regulatory Analysis.
XI. Regulatory Flexibility Certification.
XII. Backfit Analysis.
I. Background
The Commission promulgated 10 CFR part 52 on April 18, 1989 (54 FR
15386), to reform the licensing process for future nuclear power plant
applicants. The rule added alternative licensing processes in 10 CFR
part 52 for early site permits, standard design certifications, and
combined licenses. These were additions to the two-step licensing
process that already existed in 10 CFR part 50. The processes in 10 CFR
part 52 resolve safety and environmental issues early in licensing
proceedings and are intended to enhance the safety and reliability of
nuclear power plants through standardization. The rule also moved the
licensing processes in appendices M, N, O, and Q of 10 CFR part 50 to
10 CFR part 52. Subsequently, the NRC certified three nuclear plant
designs under subpart B of 10 CFR part 52--the U.S. Advanced Boiling
Water Reactor (ABWR) (62 FR 25827, May 12, 1997), System 80+ (62 FR
27867, May 21, 1997), and AP600 (64 FR 72015, December 23, 1999)
designs--and codified these designs in Appendices A, B, and C of 10 CFR
part 52, respectively.
The NRC had planned to update 10 CFR part 52 after using the design
certification process for these three certified standard plant designs.
In addition, discussions with stakeholders at public meetings and
comments on SECY-00-0092, ``Combined License Review Process,'' dated
April 20, 2000, identified licensing issues associated with subparts A
and C of 10 CFR part 52. As a result, the NRC initiated this proposed
rulemaking to (1) clarify and/
[[Page 40027]]
or correct 10 CFR parts 2, 20, 21, 50, 51, 52 (including appendices A,
B, and C), 72, 73, 140, and 170; (2) update 10 CFR part 52; and (3)
incorporate stakeholder comments.
This rulemaking action began with the issuance of SECY-98-282,
``part 52 Rulemaking Plan,'' on December 4, 1998. The Commission issued
a staff requirements memorandum on January 14, 1999, approving the NRC
staff's plan for revising 10 CFR part 52. A notice of the rulemaking
plan was added to the NRC's rulemaking Web site on June 16, 1999. On
September 3, 1999, letters were sent to 10 external stakeholders
alerting them to this proposed rulemaking. In addition, the NRC staff
held three public meetings with interested stakeholders on the 10 CFR
part 52 rulemaking on December 14, 2000, February 16, 2001, and March
7, 2001. Following those meetings, on April 3, 2001, the Nuclear Energy
Institute (NEI) submitted comments on issues discussed during the
meetings.
On September 27, 2001, the NRC staff posted draft rule language for
10 CFR part 52 on the NRC's rulemaking Web site. The NRC received
comments on the draft rule language in November 2001, from General
Electric, Entergy, NEI, Westinghouse Electric, and Exelon Generation.
The NRC staff has considered these comments in the development of this
proposed rule and posted revised draft rule language for 10 CFR part 52
on the NRC's rulemaking Web site on February 28, 2002.
II. Reorganization of 10 CFR Part 52
The NRC is proposing to reorganize 10 CFR part 52 to establish a
separate subpart for each of the seven licensing processes currently
described in 10 CFR part 52 (early site permits, early site reviews,
standard design certification, standard design approvals, combined
licenses, manufacturing licenses, and duplicate design licenses). The
purpose of this reorganization is to clarify that each licensing
process has equal standing. In addition, several subparts would be
reserved for future licensing processes. No substantive changes are
intended by the incorporation of current appendices M, N, O, and Q into
the new subparts in 10 CFR Part 52.
The NRC is also proposing to retitle 10 CFR part 52 as ``Additional
Licensing Processes for Nuclear Power Plants,'' to clarify that the
licensing processes in 10 CFR part 52 are in addition to and supplement
the two-step licensing process in 10 CFR part 50 and the license
renewal process in 10 CFR part 54, and are not limited to the early
site permit, standard design certification, and combined license
processes as the current title implies.
The proposed rule would amend Sec. 52.1 to clarify that all seven
licensing processes are within the scope of 10 CFR part 52. Paragraphs
within current Appendices M, N, O, and Q would also become new sections
of the revised part. In addition, the proposed rule would reserve
subparts for future licensing processes. In doing so, the NRC hopes to
convey that 10 CFR part 52 is the preferred location in 10 CFR for
nuclear power plant licensing processes.
The proposed rule would amend Sec. 52.19, the current Sec. 52.49
(proposed Sec. 52.111), and the current Sec. 52.83 (proposed Sec.
52.215) to provide a standard format in subparts A, D, and G. This
standard format would set forth the standards for review of
applications and the applicability of NRC requirements in a consistent
manner in each of these subparts. The references to the part 170 fee
requirements would be moved to be included in the sections on filing of
applications. This reorganization of 10 CFR part 52 will make the
subparts on early site permits and standard design certifications
consistent with the existing arrangement in the subpart for combined
licenses.
The proposed rule would also move the requirement on duration of a
combined license that is currently located in Sec. 52.83,
``Applicability of part 50 provisions,'' to paragraph (e) of proposed
Sec. 52.227, ``Issuance of combined licenses.'' Proposed Sec.
52.227(e) is a more appropriate location for this requirement.
The Commission has prepared the following table that cross-
references the new proposed provisions in 10 CFR part 52 to the
superseded provisions of 10 CFR part 52.
Table 1.--Cross-References Between New and Old 10 CFR Part 52
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New section Old section
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General Provisions:
52.1.................................... 52.1
52.3.................................... 52.3
52.5.................................... 52.5
52.8.................................... 52.8
None.................................... 52.9
Subpart A--Early Site Permits:
52.11................................... 52.11
52.13................................... 52.13
52.15................................... 52.15
52.17................................... 52.17
52.18................................... 52.18
52.19................................... 52.19
52.21................................... 52.21
52.23................................... 52.23
52.24................................... 52.24
52.25................................... 52.25
52.27................................... 52.27
52.29................................... 52.29
52.31................................... 52.31
52.33................................... 52.33
52.35................................... 52.35
52.37................................... 52.37
52.39................................... 52.39
Subpart B--Early Site Reviews:
52.41................................... App. Q, Introduction
52.43(a)................................ App. Q, Paragraph 1
52.43(b)................................ App. Q, Paragraph 2
52.43(c)................................ App. Q, Paragraph 1
52.45................................... App. Q, Paragraph 3
52.46................................... N/A
52.47(a)................................ App. Q, Paragraph 4
52.47(b)................................ App. Q, Paragraph 5
52.47(c)................................ App. Q, Paragraph 6
52.49................................... App. Q, Paragraph 7
Subpart D--Standard Design Certification:
52.101.................................. 52.41
52.103.................................. 52.43
52.105.................................. 52.45
52.107.................................. 52.47
52.109.................................. 52.48
52.111.................................. 52.49
52.113.................................. 52.51
52.115.................................. 52.53
52.117.................................. 52.54
52.119.................................. 52.55
52.121.................................. 52.57
52.123.................................. 52.59
52.125.................................. 52.61
52.127.................................. 52.63
Subpart E--Standard Design Approvals:
52.131.................................. App. O, Introduction
52.133(a)............................... App. O, Paragraph 1
52.133(b)............................... App. O, Paragraph 2
52.135.................................. App. O, Paragraph 3
52.137.................................. App. O, Paragraph 4
52.139(a)............................... App. O, Paragraph 5
52.139(b)............................... None
52.141(a)............................... App. O, Paragraph 5
52.141(b)............................... App. O, Paragraph 6
52.143.................................. App. O, Paragraph 7
Subpart G--Combined Licenses:
52.201.................................. 52.71
52.203.................................. 52.73
52.205.................................. 52.75
52.207.................................. 52.77
52.209.................................. 52.78
52.211.................................. 52.79
52.213.................................. 52.81
52.215.................................. 52.83
52.217.................................. 52.85
52.219.................................. 52.87
52.221.................................. 52.89
52.223.................................. 52.91
52.225.................................. 52.93
52.227.................................. 52.97
52.229.................................. 52.99
52.231.................................. 52.103
Subpart H--Manufacturing Licenses:
52.241.................................. App. M, Introduction
[[Page 40028]]
52.243(a)............................... N/A
52.243(b)............................... App. M, Paragraph 7
52.243(c)............................... App. M, Paragraph 9
52.243(d)............................... App. M, Paragraph 10
52.243(e)............................... App. M, Paragraph 11
52.243(f)............................... App. M, Paragraph 8
52.245(a)............................... App. M, Paragraph 2
52.245(b)............................... App. M, Paragraph 3
52.245(c)............................... App. M, Paragraph 4(b)
52.247.................................. App. M, Paragraph 1
52.249.................................. App. M, Paragraph 4(a)
52.251.................................. N/A
52.253 (a) & (b)........................ App. M, Paragraph 5
52.253(c)............................... App. M, Paragraph 6
52.255.................................. N/A
52.257.................................. App. M, Paragraph 12
Subpart I--Duplicate Design Licenses:
52.261.................................. App. N, Introduction
52.263.................................. App. N, Paragraph 1
52.265.................................. App. N, Paragraph 2
52.265(c)............................... App. N, Paragraph 3
Subpart M--Enforcement:
52.401.................................. 52.111
52.403.................................. 52.113
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III. Discussion of Substantive Changes
A section-by-section analysis that explains the purpose and meaning
of all sections in 10 CFR part 52 will be provided in the supplementary
information for the final rule. The proposed rule makes the following
substantive changes:
A. 10 CFR Part 52, Early Site Permits, Standard Design Certifications,
and Combined Licenses for Nuclear Power Plants
General Provisions
The proposed rule would amend Sec. 52.3 to add definitions for
``modular design'' and ``prototype plant'' to the current 10 CFR part
52. A definition of modular design is added to explain the type of
modular reactor design to which the Commission intended to refer to in
the second sentence of the current Sec. 52.103(g) (proposed Sec.
52.231(g)). This special provision for modular designs was added to 10
CFR part 52 to facilitate the licensing of nuclear plants, such as the
Modular High Temperature Gas-Cooled Reactor (MHTGR) and Power Reactor
Innovative Small Module (PRISM) designs, that consisted of 3 or 4
nuclear reactors in a single power block with a shared power conversion
system. During the period that the power block is under construction,
the Commission could separately authorize operation for each nuclear
reactor when each reactor and all of its necessary support systems were
completed. In a letter dated November 13, 2001 (comment A), NEI stated
that ``Part 1 of the definition would need to be revised for this
purpose so that it does not describe typical multi-unit sites. The NRC
staff should reconsider the need to define this term at all.'' The
Commission disagrees with NEI's recommendation because the term
``modular design'' needs to be defined to aid future use of the current
Sec. 52.103(g) (proposed Sec. 52.231(g)) by distinguishing the
intended definition from other definitions for ``modular design.''
Currently licensed multi-unit sites would not be affected by the
proposed Sec. 52.231(g). However, future applicants for a combined
license for a multi-unit site similar in concept to current multi-unit
sites (where each unit is similar in design but independent of all
other units) could also use this provision.
A definition for prototype plant is added to explain the type of
nuclear reactor that the Commission intended in the current Sec.
52.47(b) (proposed Sec. 52.107(b)) and intends in the proposed Sec.
52.211(b)(3). A prototype plant is a licensed nuclear reactor test
facility that is similar to and representative of either the first-of-
a-kind or certified nuclear plant design in all features and size, but
may have additional safety features. The purpose of the prototype plant
is to perform testing of new or innovative design features for the
first-of-a-kind or certified, advanced nuclear plant design, as well as
being used as a commercial nuclear power facility.
The proposed rule would remove Sec. Sec. 52.5 and 52.9 and replace
them with a new Sec. 52.5 listing all of the licensing provisions in
10 CFR part 50 that also apply to all of the licensing processes in 10
CFR part 52. The purpose of this amendment is to clarify that these 10
CFR part 50 provisions are applicable to the licensing processes that
were formerly in 10 CFR part 50 (Appendices M, N, O, and Q) and are now
in 10 CFR part 52, as well as to the new licensing processes for early
site permits, standard design certifications, and combined licenses.
Although these provisions in 10 CFR part 50 may not refer to the
additional licensing processes in 10 CFR part 52, the new Sec. 52.5
makes it clear that a holder of or applicant for an approval,
certification, permit, site report, or license issued under 10 CFR part
52 must comply with all requirements in these provisions that are
otherwise applicable to applicants or licensees under 10 CFR part 50.
In a letter dated November 13, 2001 (comment G), NEI stated:
The industry proposes that additional General Provisions be
added to part 52 in addition to an appropriate provision on Written
Communications. This approach is preferable to including cross-
references in part 52 to part 50 general provisions because these
provisions typically must be tailored to apply appropriately to the
variety of licensing processes in part 52.
The Commission disagrees with the industry's proposal to create over 35
new general provisions that are tailored for 10 CFR part 52 because it
would appear to be an inefficient and burdensome addition. Therefore,
the Commission is proposing a new Sec. 52.5 that would make the
existing general provisions in 10 CFR part 50 applicable to the
licensing processes in 10 CFR part 52.
Early Site Permits
The proposed rule would amend Sec. 52.13 to state that an early
site permit can also be referenced in an application for a combined
license or a duplicate design license.
The proposed rule would amend Sec. 52.17(a)(1) to state that the
early site permit application should specify the range of facilities
that the applicant is requesting the site to be qualified for (e.g.
one, two, or three pressurized-water reactors). This new language is
consistent with the language in Paragraph 2 of current Appendix Q. The
Commission assumes that an applicant for an early site permit does not
know what type of nuclear plant it will build at the site. Therefore,
the application must specify the postulated design parameters for the
range of reactor types, the numbers of reactors, etc., to increase the
likelihood that the site will be qualified for the actual plant or
plants that the applicant decides to build. In a letter dated November
13, 2001 (comment 27), NEI stated, ``The proposed change is too
limited. To address the required assessment of major SSCs [structures,
systems, and components] that bear on radiological consequences and all
items 52.17(a)(1)(i-viii), industry recommends a new Sec. 52.17a.2.''
The Commission disagrees with NEI's proposal to have a separate
provision for applicants who have not determined the type of plant that
they plan to build at the proposed site. The Commission expects that
applicants for an early site permit will not have decided on a
particular type of nuclear power plant and Sec. 52.17(a)(1) was
revised to address this situation.
The Commission proposes to amend Sec. 52.17(a)(2) to clarify that
an ESP applicant has the flexibility of either addressing the matter of
alternative energy sources in the environmental
[[Page 40029]]
report supporting its ESP application, or deferring the consideration
of alternative energy sources to the time that the ESP is referenced in
a licensing proceeding. The Commission believes the current regulations
already afford the ESP applicant such flexibility, inasmuch as Sec.
52.17(a)(2) states that the environmental report submitted in support
of an ESP application must ``focus on the environmental effects of
construction and operation of a reactor, or reactors * * *.'' The
environmental report's discussion of alternative energy sources does
not, per se, address the ``environmental effects of construction and
operation of a reactor,'' which is one of the matters which must be
addressed in an environmental impact statement (EIS). See 10 CFR
51.71(d); National Environmental Policy Act of 1969 (NEPA), Sec.
102(2)(C) (i), (ii) and (v). Rather, alternative energy sources
constitutes part of the discussion of reasonable alternatives to the
proposed action, which is required by Sec. 102(2)(C)(iii) of NEPA. See
10 CFR 51.71(e) n.4; 46 FR 39440 (August 3, 1981) (proposed rule
eliminating consideration of need for power and alternative energy
sources at operating license stage), at 39441 (first column).
Accordingly, it is the Commission's view that Sec. 52.17(a)(2) already
provides the ESP applicant the flexibility of choosing to defer
consideration of alternative energy sources to the time (if ever) that
the ESP is referenced in a combined license or a construction permit
application. The proposed rule clarifies that the ESP applicant may
either include a discussion of alternative energy sources in its
environmental report, or defer consideration of the matter. The
Commission proposes to make a conforming amendment to Sec. Sec. 52.18
and 52.21 to make clear that the NRC's EIS need not address need for
power, or alternative energy sources (and therefore such matters may
not be litigated) if the ESP applicant chooses not to address either or
both of these matters in its environmental report. The Commission notes
that both the environmental report and EIS for an ESP must address the
benefits associated with issuance of the ESP (e.g., early resolution of
siting issues, early resolution of issues on the environmental impacts
of construction and operation of a reactor(s) that fall within the site
parameters, and ability of potential nuclear power plant licensees to
``bank'' sites on which nuclear power plants could be located, without
obtaining a full construction permit or combined license). The benefits
(and impacts) of issuing an ESP must always be addressed in the
environmental report and EIS for an ESP, regardless of whether the ESP
applicant chooses to defer, pursuant to Sec. 52.17(a)(2),
consideration of the benefits associated with the construction and
operation of a nuclear power plant that may be located at the ESP site.
This is because the ``benefits * * * of the proposed action'' for which
the discussion may be deferred under Sec. Sec. 52.17(a)(2) are the
benefits associated with the construction and operation of a nuclear
power plant that may be located at the ESP site; the benefits which may
be deferred under Sec. 52.17(a)(2) are entirely separate from the
benefits of issuing an ESP. To put it another way, the proposed action
of issuing an ESP is not the same as the ``proposed action'' of
constructing and operating a nuclear power plant for which the
discussion of benefits (including need for power) may be deferred under
Sec. 52.17(a)(2)\1\. With this clarification, the Commission does not
believe that further changes to the language of Sec. Sec. 52.17 and
52.18 are necessary.
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\1\ The Commission emphasizes that under Sec. 52.17(a)(2), only
the discussion of benefits (including need for power) of
constructing and operating a nuclear power reactor (or reactors),
and the discussion of alternative energy sources, may be deferred.
The ER must always address the ``environmental impacts of
construction and operation of a reactor, or reactors, which have
characteristics which fall within the postulated site parameters.''
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The proposed rule would amend Sec. Sec. 52.24 and 52.39 to
clarify: (1) The information that the NRC must include in the early
site permit when it is issued; (2) the matters accorded finality in any
subsequent NRC review and proceeding for an application referencing the
early site permit; and (3) the matters that may be challenged in a
contention to be resolved in an adjudication, versus those matters that
may be raised in a petition to be processed in accord with 10 CFR
2.206. Section 52.21 would be amended to clarify that an application
referencing an early site permit must, in addition to showing that the
design of the facility falls within the site parameters specified in
the early site permit, demonstrate that all terms and conditions of the
early site permit have been satisfied. Section 52.24 would also be
amended to provide that the early site permit must state the site
parameters, as well as the ``terms and conditions,'' of the early site
permit, rather than the ``conditions and limitations'' as is currently
provided. No substantive change in Sec. 52.24 is intended by the
proposed amendment; the change is proposed to provide consistency with
Sec. 52.39(a)(2) and paragraph (a)(2)(iii) of the current rule, which
also refer to ``site parameters'' and ``terms and conditions.''
The proposed rule would add Sec. 52.28 to state that transfer of
an early site permit from its existing holder to a new applicant will
be processed under 10 CFR 50.80, which contains provisions for transfer
of licenses. In a letter dated November 13, 2001 (comment 19), NEI
recommended that a new section be added to part 52 to clarify the
process for transfer of an early site permit. The Commission has
determined that a new section is not necessary because an early site
permit is a partial construction permit and, therefore, is considered
to be a license under the AEA. The Commission believes that the
procedures and criteria for transfer of utilization facility licenses
in 10 CFR 50.80 (and the procedures in subpart M of 10 CFR part 2 for
the conduct of any hearing) should apply to the transfer of an early
site permit.
Section 52.39(a) would be amended to uniformly refer to ``terms or
conditions'' of an early site permit. Section 52.39(a)(1) would also be
amended to remove the term, ``requirements,'' and clarify that the
Commission may not change or impose new site characteristics, terms, or
conditions on the early site permit, including emergency planning
requirements, unless the special backfitting criteria in Sec.
52.39(a)(1) are satisfied. No substantive change is intended by this
clarification; the Commission believes that ``site characteristics,
terms, or conditions'' of an early site permit more accurately describe
the existing scope of matters subject to the special backfitting
criteria in Sec. 52.39(a)(1).
Early Site Reviews
The proposed rule would amend certain paragraphs of the current
Appendix Q to 10 CFR part 52 (proposed Sec. Sec. 52.41, 52.43, and
52.47) to clarify that an early site review can also be used in an
application for a combined license or a duplicate design license.
Standard Design Certifications
The proposed rule would amend the current Sec. Sec. 52.41 and
52.45 (proposed Sec. 52.101 and Sec. 52.105) to clarify that a
certified design may be referenced in an application for a duplicate
design license, as well as a combined license application, filed under
part 52.
The proposed rule would remove the requirements currently located
in Sec. Sec. 52.43(c), 52.45(c), and 52.47(b)(2)(ii) because the
Commission has decided not to require a final design approval (FDA) as
a prerequisite for certification
[[Page 40030]]
of a standard plant design under the new subpart D of 10 CFR part 52.
This requirement was included in 10 CFR part 52 because, at the time of
the original rulemaking, the NRC had no experience with design
certification applications. By requiring an FDA as a prerequisite for
certification, the NRC indicated that the licensing processes for
design certifications and FDAs were similar, even though the
requirements for and finality of design certifications differ from that
of FDAs. The NRC has considerable experience with design certification
applications and the requirement to apply for an FDA as part of an
application for design certification is no longer needed.
In a letter dated April 3, 2001 (comment 2), NEI commented
``Industry prefers to retain modified provisions. We agree that an FDA
should be an option but not a prerequisite. Also, deletion recommended
for 52.47(b)(2)(ii).'' The Commission has decided not to retain these
provisions. The proposed processes in subparts D and E allow future
applicants for design certification the option to apply for an FDA for
the same design information.
The proposed rule would also amend the current Sec. 52.45(d)
(proposed Sec. 52.105(c)) to correct the reference to the filing
requirements in Sec. 50.30(a) and delete the reference to Sec. 50.4.
The applicability of the requirements in Sec. 50.4 is set forth in the
new Sec. 52.5. No substantive change in the filing requirements is
intended by this correction.
The proposed rule would amend the current Sec. 52.47 (proposed
Sec. 52.107) to conform the statement of the requirements for
acceptable inspections, tests, analyses, and acceptance criteria
(ITAAC) in Sec. 52.107 with the Atomic Energy Act (AEA) and the
requirements in the current Sec. 52.97(b) [proposed Sec. 52.227(b)].
This clarification of the previous regulatory text, which condensed the
language in Sec. 52.79(c) and Sec. 52.97(b), is intended to avoid any
future misunderstandings.
Design Certification Backfit Requirement
The proposed rule would amend the special backfit requirement in
the current Sec. 52.63(a)(1) (proposed Sec. 52.127(a)(1)) to provide
the Commission with the ability to make changes to the design
certification rules or the certification information in the generic
design control documents that reduce unnecessary regulatory burdens.
Section 52.63(a)(1) currently states that the Commission may not
modify, rescind, or impose new requirements on the certification unless
the change is: (1) Necessary for compliance with Commission regulations
applicable and in effect at the time the certification was issued, or
(2) necessary to provide adequate protection of the public health and
safety or common defense and security. The regulation does not appear,
on its face, to permit changes to the certification which reduce
unnecessary regulatory burdens, in circumstances where the change
continues to maintain protection to public health and safety and common
defense and security. An example of a change which may not be able to
be made under the current Sec. 52.63(a)(1) is a proposed change to the
three design certification rules in Appendices A, B and C of 10 CFR
part 52, to incorporate into the Tier 2 change process the revised
change criteria in 10 CFR 50.59. Section 50.59 was revised in 1999 to
provide new criteria for, inter alia, making changes to a facility, as
described in the final safety analysis report, without prior NRC
approval, in order to reduce unnecessary regulatory burden (64 FR
53582, October 4, 1999).
To allow the Commission to modify the design certification rules in
10 CFR part 52 to incorporate the revised Sec. 50.59 change criteria,
and to allow the Commission to make future changes to reduce
unnecessary regulatory burden, the Commission is proposing to amend
Sec. 52.127(a)(1) to include a new provision that explicitly allows
the Commission to change the design certification rules or
certification information if the change provides a reduction in
regulatory burden and maintains protection to public health and safety
and common defense and security. Maintaining protection generally
embodies the same safety principles used by the NRC in applying risk-
informed decision making, e.g., ensuring that adequate protection is
provided, applicable regulations are met, sufficient safety margins are
maintained, defense-in-depth is maintained, and that any changes in
risk are small and consistent with the Commission's Safety Goal Policy
Statement (refer to NRC's Regulatory Guide 1.174). Changes to the
design certification rules must be accomplished through rulemaking,
with opportunity for public comment. Once a design certification rule
is changed through rulemaking, under proposed Sec. 52.127(a)(2) the
provisions would apply to all future applications referencing the
design certification rule as well as all current plans referencing the
design certification, unless the change has been rendered ``technically
irrelevant'' through other action taken under paragraphs (a)(3) or
(b)(1) of Sec. 52.127. Thus, standardization is maintained by ensuring
that any changes to a design certification rule intended to reduce
regulatory burden are imposed upon all nuclear power plants referencing
the design certification rule.
In a letter dated November 13, 2001, NEI stated:
Furthermore, we do not think it is necessary to modify 10 CFR
52.63(a)(1) in order to make conforming, administrative or similar
changes to the DCRs, such as those needed to conform the DCRs to the
revised 10 CFR 50.59. Nor do we think the Commission intended the
DCR backfit provisions to inhibit these types of changes. Rather, we
believe 10 CFR 52.63(a)(1) is intended to apply to changes in the
standard design approved via the DCR. We recommend the Commission
clarify this intent and provide guidance to the NRC staff allowing
certain changes to the DCRs (such as those needed to conform to the
revised 10 CFR 50.59) within the existing DCR backfit provisions.
The Commission received similar comments from General Electric Company,
Entergy, and Exelon in November 2001. The Commission disagrees with
these comments and has concluded that it is necessary to amend Sec.
52.63(a)(1) to allow changes to the design certification rules that
reduce unnecessary regulatory burden, or do not constitute a backfit.
The current Sec. 52.63(a)(1) (proposed Sec. 52.127(a)(1)) was
also modified to replace ``a modification'' with ``the change,'' in
order to clarify that the three criteria for changes apply to
modifications, rescissions or imposition of new requirements. Also, the
Commission is clarifying the proposed Sec. 52.127 to be consistent
with its original intent (refer to 54 FR 15372; April 18, 1989) that
the special backfit requirements apply to the certification information
in the generic design control documents, not to the provisions in the
design certification rules, e.g., Section VI.E of Appendix A to 10 CFR
part 52. Any proposed changes to these provisions that set forth how
the design certification rules are to be used are controlled by the
normal backfit requirements in 10 CFR 50.109.
The proposed rule would amend the current Sec. 52.63(a)(2)
(proposed Sec. 52.127(a)(2)) to delete the reference to Sec.
52.63(a)(4) (proposed Sec. 52.127(a)(4)). The reference to Sec.
52.63(a)(4) was in error because this paragraph discusses the finality
of the findings required for issuance of a combined license or
operating license, whereas Sec. 52.63(a)(2) deals with modifications
that the NRC may impose on a design certification rule under Sec.
52.63(a)(3) or Sec. 52.63(b)(1)
[[Page 40031]]
(proposed Sec. 52.127(a)(3) or Sec. 52.127(b)(1)). No substantive
change is intended by the amendment which merely clarifies the original
intent of the rule.
Standard Design Approvals
The proposed rule would amend the current Section 3 of Appendix O
to 10 CFR part 52 (proposed Sec. 52.135) to clarify that applications
for standard design approvals should contain all of the applicable
technical information required by Sec. 50.34. The amendment would also
require applications for standard design approvals to provide the same
technical information required for applications for standard design
certifications (e.g., demonstration of compliance with any technically
relevant Three Mile Island requirements, proposed technical resolutions
of unresolved safety issues and medium- and high-priority generic
safety issues, and a design-specific probabilistic risk assessment
(PRA)). This clarification is consistent with past practice regarding
applications for future designs and would implement the Commission's
Policy Statements on Severe Reactor Accidents (50 FR 32138, August 8,
1985) and Nuclear Power Plant Standardization (52 FR 34884, September
15, 1987). This amendment would not require applicants to provide
proposed ITAAC because standard design approvals are referenced in
applications for construction permits and operating licenses under 10
CFR part 50, and the verification process used for 10 CFR Part 50
applications does not use ITAAC.
The proposed rule would amend the current Appendix O to 10 CFR Part
52 (proposed Sec. 52.139) to specify that the duration of a standard
design approval is for 15 years. In a letter dated November 13, 2001
(comment 18.a), NEI commented:
Industry recommends FDAs be valid for 15 years. This is
consistent with Commission direction in COMSECY-94-025 to update the
lead plant FDA to provide a 15 year duration instead of the five
years initially provided. The ABWR and System 80+ FDAs were so
revised in 1994; the designs were certified in 1997.
The Commission agrees with industry's recommendation. The final design
approvals (FDAs) for the three certified designs were originally issued
for a five year duration, in accordance with the Commission's Policy
Statement on Standardization of Nuclear Power Plants (43 FR 38954,
August 31, 1978). Only after design certifications were issued for the
ABWR and the System 80+ designs did the Commission direct, for
consistency, that the FDAs be revised to provide the same term as for
the design certification. These actions did not change the Commission's
policy for FDAs issued by themselves. The Commission has now decided
that the duration of standard design approvals should correspond to the
duration of design certifications. The Commission has not identified
any compelling technical or policy considerations that would lead the
Commission to maintain a shorter effective time period for an FDA as
compared to a design certification.
Combined Licenses
The proposed rule would amend the current Sec. 52.73 (proposed
Sec. 52.203(a)) to clarify that a site report issued under proposed
subpart B of 10 CFR part 52 may also be referenced in an application
for a combined license application filed under 10 CFR part 52. This
amendment would also add the requirements in the current Sec. 52.63(c)
(proposed Sec. 52.127(c)) to the new Sec. 52.203(b) to clarify that
this requirement applies to applicants for a combined license. This
provision requires that, prior to granting a combined license which
references a standard design certification, information normally
contained in certain procurement specifications and construction and
installation specifications be completed and available for audit if
such information is necessary for the Commission to make its safety
determinations, including the determination that the application is
consistent with the certified design. No substantive change is intended
by the restatement of this requirement. In a letter dated April 3, 2001
(comments 3 and 3.a), NEI agreed with the proposed change but
recommended that the last sentence of Sec. 52.63(c) be deleted and the
remaining provision be added to the current Sec. 52.79 rather than the
current Sec. 52.73. The Commission agrees with NEI that 10 CFR part 52
should be modified to clarify that the requirement in current Sec.
52.63(c) applies to applicants for a combined license, and that the
last sentence be deleted. However, the Commission is adding the
remaining provision to what was Sec. 52.73(b) (proposed Sec.
52.203(b)) and not to Sec. 52.79 (proposed Sec. 52.211) as
recommended by NEI.
The proposed rule would amend the current Sec. 52.78 (proposed
Sec. 52.209) to clarify the requirements applicable to an applicant
for, and holder of, a combined license with respect to the training
program required by 10 CFR 50.120. As currently written, Sec. 52.78
simply indicates that the application must demonstrate compliance with
the training program requirements in Sec. 50.120. There is no explicit
requirement with respect to the applicant/licensee to implement the
training program. Furthermore, proposed Sec. 52.215(b) indicates that,
after a combined license is issued but before the Commission has
authorized operation under Sec. 52.231, the combined license holder
shall comply with all requirements in Title 10 of the Code of Federal
Regulations applicable to holders of construction permits for nuclear
power reactors. However, Sec. 50.120 refers to a ``nuclear power plant
applicant;'' therefore, Sec. 50.120 would not apply to a combined
license holder even under the language of proposed Sec. 52.215(b).
To remove any ambiguity in this matter, the Commission is proposing
to revise in its entirety the language in current Sec. 52.78, which is
being re-designated as Sec. 52.209. The proposed rule provides that
the application must ``describe'' the training program required by
Sec. 50.120. In addition, the proposed rule states that the training
program described in the application must be ``established,
implemented, and maintained'' no later than eighteen (18) months prior
to the scheduled date for initial loading of fuel, as provided for in
Sec. 52.231(a). By ``established [and] implemented'', the Commission
intends to distinguish between the requirement to merely ``describe''
the training program in the application, versus the requirement for the
combined license holder to establish (e.g., establish a training
organization, fill staff positions, write procedures, etc.) and
implement (i.e., perform training of applicable operating plant
personnel in accordance with Sec. 50.120) the training program. The
proposed rule also clarifies that the eighteen (18) month period by
which the training program must be established and implemented is
measured from the combined licensee's scheduled date for fuel load
under proposed Sec. 52.231(a) (current Sec. 52.103(a)).
Referencing an Early Site Permit
The proposed rule would amend current Sec. Sec. 52.39 and 52.79
(proposed Sec. 52.211) to require a license applicant referencing an
early site permit to update and correct the emergency preparedness
information provided under Sec. 52.17(b). The issue of updating an
early site permit was first raised by the Illinois Department of
Nuclear Safety, who suggested in a September 28, 1994 letter that
emergency plans and/or offsite certifications approved as part of an
early site permit review be kept up-to-date throughout the duration of
an early site permit and the
[[Page 40032]]
construction phase of a combined license. In SECY-95-090, ``Emergency
Planning Under 10 CFR part 52,'' (April 11, 1995), the NRC staff stated
that 10 CFR part 52 does not clearly require an applicant referencing
an early site permit to submit updated information on changes in
emergency preparedness information and any emergency plans that were
approved as part of the early site permit in accordance with Sec.
52.18. SECY-95-090 indicated (p. 4) that, in view of the lack of
industry interest in pursuing an early site permit, resolution of this
matter may be deferred until a ``lessons learned'' rulemaking updating
10 CFR part 52 is conducted after the first design certification
rulemakings are issued. Following public release of a draft SECY paper
setting forth the NRC staff's preliminary views on the licensing
process for a combined license, the Nuclear Energy Institute (NEI)
submitted a letter dated September 8, 1998 (comment 2.d), expressing
NEI's opposition to a requirement for updating emergency preparedness
information throughout the duration of an early site permit absent an
application referencing the early site permit. As an alternative to
updating throughout the duration of an early site permit, NEI proposed
that emergency planning information be updated when an application for
a license referencing the early site permit is filed; portions of the
emergency plans that are unchanged would continue to have finality
under 10 CFR 52.39. Thereafter, in a September 3, 1999 letter, the NRC
staff identified updating of emergency preparedness information in
early site permits as a possible subject for the part 52 rulemaking.
The Commission agrees with the Illinois Department of Nuclear
Safety that the emergency preparedness information approved when the
early site permit was issued must be updated if there is new
information which may materially affect the Commission's earlier
determination on emergency preparedness, or if the new information is
needed to correct inaccuracies in the emergency preparedness
information approved in the early site permit. In the absence of such
an updating requirement, the NRC would bear the responsibility of
identifying whether there is new information on emergency preparedness
that necessitates a re-examination of the Commission's earlier
emergency preparedness determinations for the early site permit, and
the early site permit holder or applicant referencing the early site
permit would be under no obligation to correct inaccurate emergency
preparedness information in the early site permit or approved emergency
plan. However, the Commission also agrees with NEI that a
``continuous'' early site permit update requirement would impose
burdens upon the early site permit holder without any commensurate
benefit if the early site permit is not subsequently referenced.
Accordingly, the Commission has decided that Sec. 52.39 and current
Sec. 52.79 (proposed Sec. 52.211) should contain an updating
requirement to be imposed upon the applicant referencing an early site
permit.
The proposed rule redesignates paragraph (b) of current Sec. 52.39
as paragraph (c), and adds a new paragraph (b) requiring an applicant
for a construction permit, operating license, duplicate design license,
or combined license whose application references an early site permit
to update and correct the emergency preparedness information provided
under Sec. 52.17(b), and to discuss whether the new information may
materially change the bases for compliance with the applicable NRC
requirements. A parallel requirement is included in proposed Sec.
52.211(d)(1) to ensure that applicants for combined licenses
referencing an early site permit will submit the updated emergency
preparedness information. New information which materially changes the
bases for compliance includes: (1) Information which substantially
alters the bases for a previous NRC conclusion with respect to the
acceptability of a material aspect of emergency preparedness or an
emergency preparedness plan, as well as (2) information which would
constitute a sufficient basis for the Commission to modify or impose
new terms and conditions related to emergency preparedness in
accordance with Sec. 52.39(a)(1). New information which materially
changes the Commission's determination of the matters in Sec.
52.17(b), or results in modifications of existing terms and conditions
under Sec. 52.39(a)(1) would be subject to litigation during the
construction permit, operating license, duplicate design license, or
combined license proceedings in accordance with Sec. 52.39(a)(2)(ii).
Not all new information on emergency preparedness would be subject
to challenge in a hearing under Sec. 52.39(a)(2)(ii). For example, an
emergency plan may have to be updated to reflect current telephone
numbers, the names of governmental officials whose positions and
responsibilities are defined in the plan (e.g., the name of the current
police chief for a municipality), or the current name of a hospital
facility. Such corrections do not materially change the NRC's
previously-stated bases for accepting the early site permit emergency
plan; therefore, a hearing contention would not be admitted under Sec.
52.39(a)(2)(ii) (or any other provision of Sec. 52.39) in a proceeding
for a license referencing the early site permit. By contrast, if an
emergency plan submitted as part of an early site permit relies upon a
bridge to provide the primary path of evacuation, and that bridge no
longer exists, the change could materially affect the NRC's previous
determination that the emergency plan complied with the Commission's
emergency preparedness regulations in effect at the time of the
issuance of the early site permit. Thus, such information may be the
basis for a change in the early site permit's terms and conditions
related to emergency preparedness under Sec. 52.39(a)(1), as well as
the basis for a hearing contention under Sec. 52.39(a)(2)(ii)--
assuming that the requirements in 10 CFR part 2 for admission of a
contention are met.
An updating requirement for early site permit information other
than emergency preparedness information does not appear to be
necessary, inasmuch as it is unlikely that there would be changes to
the information previously submitted on the site, such that a
significant change to the site characteristics, terms, and conditions
would be necessary if requested under the provisions of Sec.
52.39(a)(2). If the site does not conform to the characteristics of the
early site permit, an interested person may submit a petition under
Sec. 52.39(a)(2)(ii) alleging that the site does not conform to the
early site permit. Accordingly, the proposed rule does not include an
updating requirement for other early site permit information.
The proposed rule would amend Sec. 52.79(a)(1) (proposed Sec.
52.211(a)(1)), which currently requires a combined license application
referencing an early site permit to contain information demonstrating
that the design of the facility falls within the parameters specified
in the early site permit, and information needed to resolve any other
significant environmental issue not considered in the proceeding on the
referenced early site permit. Currently, Sec. 52.79(a)(1) requires a
combined license application referencing an early site permit to
contain information demonstrating that the design of the facility falls
within the site parameters specified in the early site permit. However,
Sec. 52.79(a) does not explicitly require the application to address
whether the terms and conditions specified in the early site permit
under
[[Page 40033]]
Sec. 52.24 have been met by the combined license holder, although this
is implicit by the inclusion of any terms and conditions in the early
site permit. To remove any ambiguity in this matter, the Commission is
proposing to include a proposed Sec. 52.211(a)(1)(iii) by requiring
the application to address whether the terms and conditions specified
in the early site permit under Sec. 52.24 have been met (the
Commission also proposes to rearrange paragraph (a)(1) by dividing the
criteria to be met by an application referencing an early site permit
into separate subdivisions (i), (ii), and (iii)). The Commission's
intent, as reflected in the words, ``have been met,'' is that all terms
and conditions will be met prior to issuance of the combined license.
Testing Requirements for Advanced Reactors
The proposed rule would amend the current Sec. 52.79(b) (proposed
Sec. 52.211(b)) to revise the requirements for combined license
applications that do not reference a design certification rule by
adding the current Sec. 52.47(b)(2) (proposed Sec. 52.107(b)(2)) to
the list of requirements in the proposed Sec. 52.211(b)(1) that a
combined license applicant must comply with. This amendment will
provide consistency between the current advanced reactor testing
requirements in subpart B of part 52 (Sec. 52.47(b)(2)) and the
proposed testing requirements in the proposed subpart G of part 52
(Sec. 52.211(b)). This amendment will require a combined license
applicant that references a custom advanced reactor design to also
perform the design qualification testing required by the current Sec.
52.47(b)(2) for design certification applicants. If a combined license
application references a certified advanced reactor design, the
qualification testing required by Sec. 52.47(b)(2) will have been
performed. The amendment also requires (proposed Sec. 52.211(b)(3))
that if a licensed prototype plant (see definition in proposed Sec.
52.3) is used to meet the qualification testing requirements in the
current Sec. 52.47(b)(2), additional requirements on siting, safety
features, or operational conditions may be required for licensing, in
order to compensate for uncertainties associated with the performance
of new or innovative safety features in the prototype plant.
The codification of testing requirements in the current Sec.
52.47(b)(2) was a principal issue in the development of 10 CFR part 52
(see Section II of 54 FR 15372; April 18, 1989). The testing
requirements in Sec. 52.47(b)(2), to demonstrate the performance of
safety features for nuclear power plants that differ significantly from
evolutionary light-water reactors or utilize simplified, inherent,
passive, or other innovative means to accomplish their safety functions
(advanced reactors), were included in 10 CFR part 52 to ensure that
these safety features will perform as predicted in the applicant's
safety analysis report, that the effects of systems interactions are
acceptable, and to provide sufficient data to validate analytical
codes. The design qualification testing requirements may be met with
either separate effects or integral system tests; prototype tests; or a
combination of tests, analyses, and operating experience. These
requirements implement the Commission's policy on proof-of-performance
testing for all advanced reactors (see 51 FR 24643; July 8, 1986) and
the Commission's goal of resolving all design issues before authorizing
construction.
During the development of 10 CFR part 52, the focus of the nuclear
industry and the NRC staff was on applications for design
certification. That is why the testing requirements to qualify new or
innovative safety features was only included in subpart B of 10 CFR
part 52, ``Standard Design Certifications.'' The tests to qualify a
design feature are different than verification tests, which are
required by Sec. 52.79(c) and performed in accordance with section XI,
``Test Control,'' of Appendix B to 10 CFR part 50. Verification tests
are used to provide assurance that construction and installation of
equipment (as-built) in the facility has been accomplished in
accordance with the approved design.
Exelon Generation and NEI commented on the addition of testing
requirements for combined license applications, in letters dated
November 13, 2001. NEI stated:
COL application requirements in Sec. 52.79(b)(1) have been
modified to include a reference to the design certification
application requirements of Sec. 52.47(b)(2)(i). Under this
proposal, an applicant seeking a COL for a non-certified design that
differs significantly from typical light water reactors would have
to demonstrate safety feature performance through either (A)
analysis, testing, or experience, or (B) full-scale prototype
testing. This requirement is entirely appropriate for design
certification applicants. However, as discussed below, we believe it
is unnecessary to apply these requirements to COL applicants, and
that the potential requirement for full-scale prototype testing is
particularly inappropriate.
First, part 52 should not be modified to open the door to
requiring a COL applicant, who does not reference a certified
design, to build and complete testing of a full-scale prototype
before the granting of the license. The potential to require
prototype testing to support issuance of a COL is contrary to
Commission guidance in the part 52 Statements of Consideration. The
Commission clearly recognized ``licensing the prototype for
commercial operation'' as a path open to applicants under subpart C
of part 52 that could lessen the burden of having to demonstrate
innovative designs through full scale prototype testing. We agree
with the further statement by the Commission that, ``[i]t is well to
remember also that, under the rule, prototype testing is required
only for certification or an unconditional design approval, if at
all.'' * * * In sum, through its existing requirements and
regulatory authority, the NRC is assured of (1) Adequate information
to support required COL reviews and safety determinations, and (2)
satisfactory demonstration of innovative design features during
startup and power ascension testing. The proposed new COL
application requirements are unnecessary and should not be carried
forward into the part 52 NOPR (Notice of Proposed Rulemaking).
The Commission disagrees with NEI and Exelon regarding the need to
perform qualification testing for new or innovative safety features in
all advanced reactor designs. The Commission reformed the licensing
process for new nuclear plants with the issuance of 10 CFR part 52 in
1989 and required applicants to demonstrate that safety features will
perform as predicted in their final safety analysis report. Although
the focus of the NRC staff in 1989 was on applications for design
certification, the Commission intended that testing to qualify design
features (proof-of-performance testing) would be required for all
advanced reactors, including custom designs (see Question 6 at 51 FR
24646; July 8, 1986). Furthermore, it would make no sense for the
Commission to require testing for design certification (paper designs)
and not require testing for applications to build and operate an actual
advanced nuclear reactor.
Although the Commission has stated that it favors the use of
prototypical demonstration facilities and that prototype testing is
likely to be required for certification of advanced non-light-water
designs (see policy at 51 FR 24646; July 8, 1986 and Section II of 54
FR 15372 on 10 CFR part 52; April 18, 1989), the proposed rule does not
mandate the use of a prototype plant. Rather, the proposed rule
provides that if a prototype plant is used to qualify an advanced
reactor design, then additional requirements may be required for
licensing of the prototype to compensate for any uncertainties with the
unproven safety features. Also, the prototype plant could be used for
[[Page 40034]]
commercial operation. Therefore, the Commission proposes to amend Sec.
52.79(b) (proposed Sec. 52.211(b)) to implement its original intent in
adopting 10 CFR Part 52 and its policy on advanced reactors that it is
necessary to demonstrate the performance of new or innovative safety
features through design qualification testing for all advanced nuclear
reactors.
Probabilistic Risk Assessments
The proposed rule would also amend the current Sec. 52.79(b)
(proposed Sec. 52.211(b)) to adopt a requirement to submit a plant-
specific PRA as part of an application for a combined license. The
current Sec. 52.79(b) references Sec. 52.47(a)(1)(v), which requires
a design-specific PRA within a design certification application. This
amendment (Sec. 52.211(b)(2)) would require an application for a
combined license to contain a plant-specific PRA that covers all of the
nuclear plant design, including site-specific design features (e.g.,
the ultimate heat sink). If the combined license application referenced
a certified design, this amendment (Sec. 52.211(b)(5)) would require
the design-specific PRA to be updated to include site-specific design
features and to account for any design changes. In a letter dated April
3, 2001 (comment 11.1a), NEI stated ``we agree on the NRC vision for a
plant-specific PRA at COL that supplements the DC PRA with any changes
that affect the DC PRA plus site-specific (interface) design
information.''
The purpose of the requirement for a plant-specific PRA is to
identify and address potential design and operational vulnerabilities,
gain insights about the risk of the design, assess the balance between
preventive and mitigative features in the design, to determine
quantitatively whether the design represents a reduction in risk over
current operating plants, and to determine how the risk associated with
the new design relates to the Commission's safety goals. Accordingly,
the Commission proposes to amend Sec. 52.211(b) to require an
application for a combined license to contain a plant-specific PRA.
Resolution of ITAAC
The proposed rule would amend the current Sec. 52.79(c) (proposed
Sec. 52.211(c)), current Sec. 52.97(a) (proposed Sec. 52.227(a)),
current Sec. 52.99 (proposed Sec. 52.229(e)), and current Sec. Sec.
52.103(a) and (g) (proposed Sec. Sec. 52.231(a) and (g)) to provide an
applicant for a combined license with a process for resolving certain
acceptance criteria in one or more of the ITAAC required by the
proposed Sec. 52.211(c) before issuance of the combined license. In a
letter dated November 13, 2001 (comment 20), NEI recommended that
Subpart C be revised to allow for completion of design acceptance
criteria (DAC) at the COL application stage. NEI made this
recommendation because applicants might want to complete certain DAC
before construction. DAC are special design certification rule ITAAC.
DAC set forth processes and criteria for completing certain design
information, such as information about the digital instrumentation and
control system. DAC were originally written to be verified as part of
the normal, post-combined license, ITAAC verification process.
The Commission agrees with NEI's recommendation that combined
license applicants be permitted to demonstrate DAC completion as part
of the combined license application, for several reasons. First,
completion of the design matters covered by DAC before the issuance of
a combined license is consistent with the Commission's original concept
for design certification and issuance of a combined license. When it
adopted 10 CFR part 52, the Commission intended that a design
certification contain final and complete design information. Allowing a
finding of acceptable completion of DAC before issuance of a combined
license is, therefore, consistent with the Commission's original
intent. Second, completion of DAC before issuance of the combined
license is consistent with the Commission's goal of resolving issues
before construction. Determining whether DAC have been successfully
completed before issuance of the combined license avoids the
possibility that improperly completed DAC will result in the
construction of improperly designed structures, systems, and
components. Finally, the Commission believes that completion of DAC
before issuance of the combined license will enhance public confidence
in the overall licensing process because the public will have an
opportunity to challenge whether the design has been properly completed
before construction begins. Accordingly, the Commission proposes that a
finding of successful completion of DAC may be made when a combined
license is issued, if the combined license applicant demonstrates that
the DAC have been successfully completed. This new process would also
allow findings on successful completion of inspections or tests of
components procured before the issuance of the combined license.
The proposed rule would also amend the current Sec. 52.99
(proposed Sec. 52.229 (b), (c) and (d)) and the current Sec. 52.103
(proposed Sec. 52.231(h)) to incorporate rule language from the design
certification rules in 10 CFR part 52 regarding the completion of ITAAC
(see paragraphs IX.A and IX.B.3 of Appendix A to part 52). During the
preparation of the design certification rules for the ABWR and System
80+ designs, the NRC staff and nuclear industry representatives agreed
on certain requirements for the performance and completion of the
inspections, tests, or analyses in ITAAC. In the design certification
rulemakings, the Commission codified these ITAAC requirements into
Section IX of the rules. The purpose of the requirement in paragraph
(b) of proposed Sec. 52.229 is to make it clear that an applicant may
proceed at its own risk with design and procurement activities subject
to ITAAC, and that a licensee may proceed at its own risk with design,
procurement, construction, and preoperational testing activities
subject to an ITAAC, even though the NRC may not have found that any
particular ITAAC has been successfully completed. Paragraph (c) of
proposed Sec. 52.229 requires the licensee to notify the NRC that the
required inspections, tests, and analyses in the ITAAC have been
completed and that the acceptance criteria have been met. Paragraph (d)
simply states the options that a licensee will have in the event that
it is determined that any of the acceptance criteria in the ITAAC have
not been met. Finally, paragraph (h) of Sec. 52.231 states that ITAAC
do not, by virtue of their inclusion in the DCD, constitute regulatory
requirements after the licensee has received authorization to load fuel
or for renewal of the license. However, subsequent modifications must
comply with the design descriptions in the design control document
unless the applicable requirements in the current Sec. 52.97 and
Section VIII of the design certification rules have been complied with.
In a letter dated April 3, 2001 (comment 23), NEI stated ``consider
incorporating DCR general provisions into subpart C as appropriate.''
The Commission has decided to add these ITAAC requirements to proposed
Sec. 52.229 because it believes that these provisions embody general
principles that are applicable to all holders of combined licenses.
Commission Finding on Acceptance Criteria
The proposed rule would amend the current Sec. 52.83 (proposed
Sec. 52.215) and the current Sec. 52.99 (proposed Sec. 52.229(e)) to
clearly state the
[[Page 40035]]
Commission's determination that the NRC staff should be responsible for
ensuring (through its inspection and audit activities) that the
combined license holder performs and documents the completion of
inspections, tests and analyses in the ITAAC. Currently, Sec. 52.99
states that ``the Commission shall ensure that the required
inspections, tests, and analyses are performed and, prior to operation
of the facility, shall find that the prescribed acceptance criteria are
met.'' When part 52 was first adopted by the Commission in 1989 (54 FR
15372, April 18, 1989), Sec. 52.99 provided that the NRC staff shall
ensure that the inspections, tests and analyses in the ITAAC are
performed, and did not refer to the Commission finding on acceptance
criteria being met. The requirement for a Commission finding on
acceptance criteria was contained in Sec. 52.103(g). The Commission
adopted the current language of Sec. 52.99 in 1992 (57 FR 60975,
December 23, 1992) to reflect changes to Section 185 of the AEA made by
Congress in the Energy Policy Act of 1992 (1992 EPA), which states:
Following issuance of the combined license, the Commission shall
ensure that the prescribed inspections, tests, and analyses are
performed and, prior to operation of the facility, shall find that
the prescribed acceptance criteria are met.
Thus, the revisions to Sec. 52.99 adopted by the Commission in 1992
simply reflect the language of the 1992 EPA. However, the Commission
does not believe that Congress, by adopting language in section 185
stating that the Commission shall ensure that the ITAAC are performed,
intended to alter the Commission's determination that the NRC staff is
responsible for ensuring that ``the required inspections, tests and
analyses in the ITAAC are performed,'' and by doing so alter the
Commission's long-standing delegation of inspection and oversight
activities to the NRC staff. For these reasons, the Commission proposes
that Sec. 52.99 (proposed Sec. 52.229(e)) state that the NRC staff
shall be responsible for ensuring that inspections, tests and analyses
in the ITAAC have been performed. The requirement for a Commission
finding on acceptance criteria will continue to be addressed separately
in Sec. 52.103(g) (proposed Sec. 52.231(g)).
In a letter dated February 22, 1993, the Nuclear Management and
Resources Council, Inc. (NUMARC) stated:
There is nothing in Title XXVIII or its legislative history
which compels a change in the Staff responsibilities from that
reflected in prior Sec. 52.99. Indeed, any other implementation of
Sec. 52.99 would be wholly unworkable. Accordingly, it is our
understanding that the reference to ``the Commission'' in amended
Sec. 52.99 is to be read as authorizing the Commission to delegate
to the Staff the responsibility for overseeing ITAAC performance
during the period of facility construction; and further that this is
the Commission's intention. Responsibility for the pre-operational
finding of acceptance criteria conformance would, of course, be the
responsibility of the Commission, as reflected in both amended
Sec. Sec. 52.99 and 52.103(g).
The proposed rule is consistent with NUMARC's recommendation.
The requirements in the proposed Sec. 52.229(e) will be limited to
the responsibilities of the NRC staff. The staff will ensure that the
inspections, tests, and analyses in the ITAAC have been performed and
will publish notices in the Federal Register of the successful
completion of inspections, tests, and analyses. The NRC staff will
perform periodic inspections during construction of the facility and
implementation of the licensee's operational programs, e.g., emergency
planning and training. The NRC staff will issue reports on these
inspections and will make these reports publically available. At the
conclusion of construction, the staff will make a recommendation to the
Commission on its assessment of the licensee's completion of ITAAC. If
the Commission determines that all of the acceptance criteria in the
ITAAC for the combined license have been met, it will make the finding
required under proposed Sec. 52.231(g).
Consistent with the language in proposed Sec. 52.229(e), the
proposed rule would also amend the current Sec. 52.83 (proposed Sec.
52.215(c)) to state that the requirements in 10 CFR part 50 that are
applicable to holders of operating licenses become applicable to
holders of combined licenses after the Commission's finding of
successful ITAAC completion under current Sec. 52.103(g) (proposed
Sec. 52.231(g)), rather than referring to the Commission finding under
the current Sec. 52.99. As discussed above, the Commission's 1992
rulemaking amended Sec. 52.99 to refer to the Commission's finding of
ITAAC completion, and amended Sec. 52.83 to refer to the Commission's
finding under Sec. 52.99. Inasmuch as the Commission finding and
authorization of operation would be addressed in proposed Sec.
52.231(g), it follows that proposed Sec. 52.215(c) should refer to the
Commission's authorization of operation under Sec. 52.231(g) rather
than the NRC staff's activities under proposed Sec. 52.229(e).
Combined License Change Process
The proposed rule would amend the current Sec. 52.97 (proposed
Sec. 52.227) to clarify the applicability of the change processes in
10 CFR part 50 and Section VIII of the design certification rules in 10
CFR part 52 to a combined license. This amendment will add Sec.
52.227(c), which states that the change processes in 10 CFR part 50
apply to a combined license that does not reference a design
certification rule. This amendment will also add Sec. 52.227(d), which
states that the change processes in Section VIII of the design
certification rules apply to changes within the scope of the referenced
certified design. However, if the proposed change affects the design
information that is outside of the scope of the design certification
rule, the part 50 change processes apply unless the change also affects
the design certification information. For that situation, both change
processes may apply.
In a letter dated November 13, 2001 (comment 21(a)(2)), NEI
recommended that proposed Sec. Sec. 52.227(c) and (d)(2) state that
changes outside the scope of a certified design are subject to ``the
applicable change control requirements in 10 CFR part 50, e.g., 10 CFR
50.59, 50.54 or 50.90.'' The Commission has decided to propose this
amendment to clarify which change processes are applicable to a
combined license and this amendment is consistent with NEI's
recommendation.
Design Certifications for ABWR, System 80+, and AP600
The proposed rule would amend paragraphs VI.B.4, 5, and 6 of the
three design certification rules in 10 CFR part 52, Appendices A, B,
and C (for U.S. ABWR, System 80+, and AP600 designs, respectively), by
substituting the phrase ``but only for that plant'' for the erroneous
phrase ``but only for that proceeding'' (emphasis added). The new
phrase correctly characterizes the scope of issue resolution in three
situations. Paragraph VI.B.4 describes how issues associated with a
design certification rule are resolved when an exemption has been
granted for a plant referencing the design certification rule.
Paragraph VI.B.5 describes how issues are resolved when a plant
referencing the design certification rule obtains a license amendment
for a departure from Tier 2 information. Paragraph VI.B.6 describes how
issues are resolved when the applicant or licensee departs from the
Tier 2 information on the basis of paragraph VIII.B.5, which waives the
requirement to get NRC approval. Thus, once a matter (e.g., an
exemption in the
[[Page 40036]]
case of paragraph VI.B.4) was addressed for a specific plant
referencing a design certification rule, the adequacy of that matter
for that plant would not ordinarily be subject to challenge in any
subsequent proceeding or action (such as an enforcement action) listed
in the introductory portion of paragraph IV.B, but there would not be
any issue resolution on that subject matter for any other plant.
Unfortunately, the three design certification rules use the phrase
``but only for that proceeding,'' which may lead to the erroneous
conclusion that issue resolution exists only in the proceeding in which
the matter was approved and/or adjudicated, and not in all subsequent
proceedings for that plant.
In letters dated November 12, 2001, and November 13, 2001,
respectively, General Electric Company and Westinghouse Electric
Company reiterated earlier recommendations the two companies had made
that Sections VI.B.4 and 5 of the design certification rules state that
exemptions and license amendments have finality ``but only for that
plant.'' For the reasons discussed above, the Commission agrees, and
the Commission proposes to substitute the phrase ``but only for that
plant,'' in order to clarify that issue resolution on a matter applies
in subsequent proceedings for that plant.
Each of the design certification rules in 10 CFR part 52
(Appendices A, B, and C) includes a Section VIII on change processes.
These processes apply to changes depending upon the category of design
information affected. For plant-specific tier 2 information, the change
process established in the rules mirrors, in large part, that in the
former 10 CFR 50.59. The proposed rule would amend paragraph VIII.B.5
of the design certification rules to conform the terminology in the
50.59-like change process to that used in the revised Sec. 50.59. This
amendment deletes references to unreviewed safety question and safety
evaluation, and conforms the evaluation criteria concerning when prior
NRC approval is needed. Also, a definition has been added (paragraph
II.G) for ``departure from a method of evaluation'' to support the
evaluation criterion in VIII.B.5.b(8).
In an earlier rulemaking (see 64 FR 53582; October 4, 1999), the
Commission revised Sec. 50.59 to incorporate new thresholds for
permitting changes to a plant as described in the final safety analysis
report without NRC approval. For consistency and clarity, similar
changes are now being proposed for 10 CFR part 52 applicants or
licensees. Because of some differences in how the change control
requirements are structured in the design certification rules, certain
definitions contained in Sec. 50.59 are not necessary for or
applicable to 10 CFR part 52 and are not being included in this
proposed rule. One definition that the Commission is including is the
definition from the new Sec. 50.59 for a ``departure from a method of
evaluation,'' which is appropriate to include in this rulemaking so
that the eighth criterion in Section VIII.B.5.b of the design
certification rules will be implemented as intended.
B. 10 CFR Part 2, Rules of Practice for Domestic Licensing Proceedings
and Issuance of Orders
The proposed rule would amend Sec. Sec. 2.110, 2.400, 2.401,
2.402, 2.403, 2.404, 2.406, 2.500, 2.501, and 2.502 to correct
references to former 10 CFR part 52 appendices that have been
redesignated as subparts.
C. 10 CFR Part 20, Standards for Protection Against Radiation
The proposed rule would amend Sec. 20.1002 to clarify that the
regulations in 10 CFR part 20 also apply to licenses issued under 10
CFR part 52. This conforming change was inadvertently overlooked when
the Commission originally promulgated 10 CFR part 52.
D. 10 CFR Part 21, Reporting of Defects and Noncompliance
The proposed rule would amend Sec. Sec. 21.2, 21.3, and 21.21 to
clarify the applicability of 10 CFR part 21 to individuals,
corporations, partnerships, or other entities doing business within the
United States, and directors and responsible officers of such
organizations, that hold a permit or license under 10 CFR part 52.
These conforming changes would correct an oversight when the Commission
first adopted 10 CFR part 52, to ensure that the requirements in 10 CFR
part 21 apply to applicants for, and holders of licenses under 10 CFR
part 52, as well as to suppliers of basic components to such licensees.
Combined Licenses, Manufacturing Licenses, Duplicate Design Licenses
The proposed rule would make 10 CFR part 21 applicable to
applicants for, and holders of combined licenses, manufacturing
licenses, and duplicate design licenses under 10 CFR part 52, and
suppliers of basic components to such applicants and holders, by
amending paragraphs (a), (b), and (c) of Sec. 21.2 regarding the scope
of 10 CFR part 21 and amending the definitions of basic component,
commercial grade item, critical characteristics, dedicating entity,
dedication, defect, and substantial safety hazard in Sec. 21.3. In
addition, the proposed rule would amend Sec. 21.21 to clearly state
when a director or responsible officer subject to 10 CFR part 21 must
notify the Commission that the director or officer has information
reasonably indicating a failure to comply or a defect affecting the
construction or operation of a facility or an activity that is subject
to the licensing requirements under 10 CFR part 52 or affecting a basic
component supplied for a facility or an activity that is subject to the
licensing requirements under 10 CFR part 52. The Commission notes that
a supplier of safety-related analyses and services to a licensee under
part 52 is subject to part 21, inasmuch as such services constitute
``basic components;'' this is no different than the applicability of
part 21 to a supplier of such analyses and services to a licensee under
part 50.
Early Site Permits
With respect to early site permits, the Commission proposes to use
a different approach, such that the requirements of part 21 do not
apply to applicants for early site permits, or holders of early site
permits so long as the early site permit is not referenced in any
license application. During the pendency of the early site permit
application before the NRC, the applicant would be required by 10 CFR
50.9, ``Completeness and accuracy of information,'' to notify the
Commission of any information having a ``significant implication for
public health and safety or the common defense and security'' with
respect to the matters covered in the application, pursuant to proposed
Sec. 52.111. Failure to abide by the completeness and accuracy
requirements in Sec. 50.9 would subject the applicant to potential
criminal liability under Sec. 52.113 (proposed Sec. 52.403). In
addition, under current Sec. 52.9, the early site permit applicant
would be subject to penalties for deliberate misconduct, including
submission to the NRC of information known to be incomplete or
inaccurate in some material aspect. Finally, during the pendency of an
early site permit application, the application has no operative effect
with respect to issue resolution under Sec. 52.39; consequently, an
early site permit application itself could not result in a
``substantial safety hazard'' by virtue of the application being
referenced in a nuclear power plant licensing proceeding. Therefore,
the Commission does not believe that adopting the regulatory overlay of
part 21 during the pendency of an early site permit application is
necessary to effectuate the Commission's regulatory
[[Page 40037]]
responsibilities under the AEA, as amended, including providing
reasonable assurance of adequate protection of public health and safety
or common defense and security.
The Commission does not believe that part 21 should apply to the
early site permit holder after the early site permit has been issued,
but before the holder has referenced the permit in a license
application.\2\ With one exception, the early site permit does not
authorize any action by the holder with respect to the construction or
operation of a nuclear power plant. The exception is when the early
site permit authorizes the holder to conduct the site preparation
activities permitted under 10 CFR 50.10(e)(1) (commonly referred to as
limited work authorization-1, or LWA-1, activities). However, these
activities are related to site clearing and preparation, and do not
permit any construction (including subsurface preparation) for
``structures, systems and components which prevent or mitigate the
consequences of postulated accidents that could cause undue risk to the
health and safety of the public.'' Thus, the conduct of LWA-1
activities do not appear to have any reasonable possibility of
resulting in a ``substantial safety hazard.'' Furthermore, the inherent
nature of an early site permit is site-specific and not susceptible to
generic or wide-ranging applicability. For these reasons, the
Commission proposes that part 21 should not apply to an early site
permit holder until the permit is referenced by a license applicant.
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\2\ The Commission would not permit a license applicant to
reference an early site permit which it does not hold (or has rights
to the permit contingent upon a NRC decision to issue a license
whose application references the early site permit). To otherwise
permit referencing of an early site permit by a non-holder would
destroy the commercial value of the permit, and would prevent any
entity from seeking an early site permit. This would frustrate the
Commission's regulatory objective of providing early regulatory
approval of siting, emergency preparedness, and environmental
matters. Since the early site permit is a license, the relevant
requirements of part 21 are those applicable to a licensee.
---------------------------------------------------------------------------
Once an early site permit holder references the permit in a license
application, the Commission believes that the holder should be subject
to part 21. The Commission's safety review of a license application
referencing an early site permit is limited in accordance with
Sec. Sec. 52.39 and 52.79 (proposed Sec. 52.211), under the precept
that the site parameters, terms, and conditions which define the
envelope for safe siting of a nuclear power plant have been determined
by the NRC in the early site permit proceeding. If the early site
permit holder discovers a significant safety concern with respect to
its site (e.g., that the specified site parameter for seismic
acceleration is less than the projected acceleration due to new
information), the concern should be reported to the NRC so that it may
be considered in the review of the application referencing the early
site permit. This reporting attains special importance given the
Commission's proposal (see discussion in Section III.A.8 on referencing
an early site permit) not to impose an updating requirement for early
site permit information other than that related to emergency
preparedness. Accordingly, the Commission concludes that the early site
permit holder should be subject to part 21 once it references the
permit in a license application.
The Commission believes that changes to part 21 are unnecessary to
reflect these determinations with respect to early site permit
applicants and holders. A licensee's reporting requirements in part 21
apply only with respect to ``basic components'' used or to be used in
an NRC-licensed or otherwise regulated facility. The safety-related
analyses and consulting services supplied to an applicant for an early
site permit appear to fall within the definition of ``basic
component,'' in that they constitute ``safety-related design [and]
analyses * * * associated with component hardware'' (See 10 CFR 21.3,
``Basic component,'' paragraph (3)). Thus, part 21 could be interpreted
as applying to the early site permit holder immediately upon the
permit's issuance. However, there appears to be little reasonable
likelihood of a ``substantial safety hazard'' unless and until the
early site permit has been referenced by the permit holder in a license
application. Once the early site permit has been referenced, the
potential for a substantial safety hazard clearly exists if a known
defect in site parameters, terms, or conditions defining the envelope
for safe plant operation is not disclosed, and a plant is designed,
constructed, and allowed to operate which does not reflect the actual
limiting parameters and conditions of the site. Thus, no changes to
part 21 are necessary to reflect the Commission's intent.
The Commission also proposes that part 21 apply to suppliers of
safety-related analyses and services to an early site permit holder in
the same manner and extent as part 21 applies to the early site permit
holder. Such suppliers would be subject to part 21 only after the early
site permit holder references the permit in a license application.
Design Certification Rules
Similar to the approach for early site permit applicants and
holders, the Commission proposes that the requirements in part 21
should not apply to the applicant/vendor for a design certification
(and/or its successors) during the pendency of its design certification
application. During the pendency of the design certification
application, the applicant/vendor would be required by 10 CFR 50.9,
``Completeness and accuracy of information,'' to notify the Commission
of any information having a ``significant implication for public health
and safety or the common defense and security'' with respect to the
matters covered in the application, pursuant to proposed Sec. 52.111.
Failure to abide by the completeness and accuracy requirements in Sec.
50.9 would subject the applicant/vendor to potential criminal liability
under Sec. 52.113 (proposed Sec. 52.403). In addition, under current
Sec. 52.9, the applicant for a design certification is subject to
penalties for deliberate misconduct, including submission to the NRC of
information known to be incomplete or inaccurate in some material
aspect. Finally, during the pendency of a design certification
application, the application has no operative effect with respect to
issue resolution under current Sec. 52.63 (proposed Sec. 52.127);
consequently, a design certification application itself could not
result in a ``substantial safety hazard'' by virtue of the application
being referenced in a nuclear power plant licensing proceeding.
Therefore, the Commission does not believe that adopting the regulatory
overlay of part 21 during the pendency of a design certification
application is necessary to effectuate the Commission's regulatory
responsibilities under the AEA, as amended, including providing
reasonable assurance of adequate protection to public health and safety
or common defense and security.
The Commission also believes that the reporting requirements in
part 21 should not apply to the design certification applicant/vendor
after the Commission issuance of a final design certification rule but
before the design certification rule is referenced by at least one
applicant/licensee (nor should either Sec. Sec. 52.9 or 52.111 be
modified to make them applicable to the design certification applicant/
vendor). The Commission does not believe that a design certification
rule would reasonably result in a ``substantial safety hazard'' so long
as the design certification rule is not actually referenced in a
license application (and
[[Page 40038]]
thereafter incorporated by reference into a license). It is true that,
unlike an early site permit, a design certification rule is of general
applicability and that a complete nuclear power plant design could be
provided by an entity other than the original design certification
applicant/vendor (see Sec. 52.73 (proposed Sec. 52.203)).
Nonetheless, unless the other entity provides a design which is
subsequently referenced in an NRC license application, there is no
``substantial safety hazard'' created (although the Commission
acknowledges that the entity may incur significant redesign costs if
the entity completes substantial parts of the design before submission
of the application, only to find upon submission of the application
that there were significant defects in the certified design). Upon
weighing of all relevant factors, the Commission proposes that part 21
should not apply to the design certification applicant/vendor until a
final, Commission-approved design certification rule is referenced by
at least one applicant/licensee.
However, the Commission believes that once a design certification
rule is referenced by an applicant, the design certification applicant/
vendor should be subject to part 21. The Commission's safety review of
a license application referencing a design certification rule is
limited in accordance with Sec. 52.63 (proposed Sec. 52.127) and
Sec. 52.79 (proposed Sec. 52.211). If the design certification
applicant/vendor has discovered a significant safety concern with
respect to its certified design, it should be reported to the NRC so
that it may be considered in the review of the application referencing
the design certification rule. While this places a continuing
obligation on the design certification applicant/vendor to monitor
whether its design has been referenced in a license application, as a
practical matter it is likely that the license applicant will have
contractually engaged the design certification applicant/vendor prior
to submitting the application. In any event, the Commission concludes
that the design certification applicant/vendor should be subject to
part 21 after its design certification has been referenced by an
applicant for a license.
The Commission believes that, with one exception, changes to part
21 are unnecessary to reflect these determinations with respect to
design certification applicants/vendors. Designs submitted for
certification are ``basic components,'' as defined in Sec. 21.3, as
are any supporting analyses inasmuch as they constitute ``safety-
related design [and] analysis * * * associated with component hardware
whether these services are performed by the component supplier or
not.'' If the design certification applicant/vendor provides the
certified design to a license applicant pursuant to contract or
agreement, the design certification applicant/vendor ``supplies'' the
basic component, see Sec. 21.3. However, there is a possibility that
an entity other than the applicant/vendor of a design which was
certified in a design certification rule may supply the complete plant
design to a referencing license applicant. See Sec. 52.73 (proposed
Sec. 52.203). For these reasons, the Commission is considering a
change to the definition of ``supplying or supplies'' in Sec. 21.3 to
ensure that a design certification applicant/vendor who does not
pursuant to contract supply to a license applicant the complete design
for the design certification, is also subject to part 21 for this
special situation.
For the reasons discussed earlier, the Commission believes that it
is reasonable and appropriate to limit the applicability of part 21
such that it is applicable once the design certification rule has been
referenced by an applicant, permit holder, or licensee. Therefore,
although the potential ambit of part 21 extends to an applicant/vendor
of a design certification after issuance of a design certification
rule, the Commission has decided not to extend the applicability of
part 21 in such a fashion. By contrast, once the design certification
rule has been referenced, the potential for a substantial safety hazard
exists if a known defect in a design certification rule is not
disclosed, the remainder of the plant is designed, the plant
constructed, and subsequently allowed to operate. Accordingly, the
Commission concludes that part 21 should apply to the design
certification applicant/vendor after the design certification rule has
been referenced by a license applicant. Finally, the Commission
concludes that part 21 should apply to suppliers of safety-related
analyses and services to a design certification applicant/vendor in the
same manner and extent as part 21 applies to the design certification
applicant.
E. 10 CFR Part 50, Domestic Licensing of Production and Utilization
Facilities
The proposed rule would amend paragraph (a)(1) of Sec. 50.109
(backfit rule) to clearly state the applicability of the backfit rule
to some of the licensing processes 10 CFR part 52 and the date that
backfit protection commences for those licensing processes. The
licensing processes to which the backfitting provisions in Sec. 50.109
apply are standard design approvals, combined licenses, manufacturing
licenses, and duplication design licenses issued under subparts E, G,
H, and I of 10 CFR part 52, respectively. The backfitting requirement
in Sec. 50.109 does not apply to early site permits, early site
reviews, and standard design certifications issued under subparts A, B,
and D, respectively, in as much as these licensing processes have their
own special backfitting provisions (the special backfit requirements
set forth in Sec. 52.39, current sections 5 and 6 of Appendix Q
(proposed Sec. 52.47), and current Sec. 52.63(a) (proposed Sec.
52.127(a)) apply to early site permits, early site reviews, and
standard design certifications, respectively). Section
50.109(a)(1)(vii) sets forth the applicability of these special
backfitting provisions for a combined license that references an early
site permit, early site review, or design certification rule.
The proposed rule would also remove appendices M, N, O, and Q from
10 CFR part 50. These appendices were transferred to 10 CFR part 52
when it was first promulgated (54 FR 15372; April 18, 1989). However,
the Commission failed to remove those appendices from 10 CFR part 50,
though the Commission intended to do so (see 54 FR 15385; April 18,
1989).
F. 10 CFR Part 51, Environmental Protection Regulations for Domestic
Licensing and Related Regulatory Functions
The proposed rule would amend paragraph (b)(6) of Sec. 51.20,
``Criteria for and identification of licensing and regulatory actions
requiring environmental impact statements,'' to make clear that
issuance of a manufacturing license requires preparation of an
environmental impact statement or a supplement to an environmental
impact statement. Paragraph (b), which defines types of actions that
require an environmental impact statement or a supplement to an
environmental impact statement would replace the current reference to
Appendix M with a reference to subpart H of 10 CFR part 52 which is the
proposed subpart that sets forth the process for manufacturing
licenses, formerly contained in Appendix M.
G. 10 CFR Part 72, Licensing Requirements for the Independent Storage
of Spent Nuclear Fuel and High-Level Radioactive Waste
The proposed rule would amend Sec. 72.210 to indicate that a
general license would be issued for the storage
[[Page 40039]]
of spent fuel in an independent spent fuel storage installation at
power reactor sites to persons authorized to possess or operate nuclear
power reactors under a combined license or duplicate design license
under 10 CFR part 52. The proposed rule would also amend the
requirements in Sec. 72.218(b) regarding an application for
termination of a reactor operating license and the removal of the spent
fuel stored at the reactor site to indicate that this provision also
applies to applications for termination of a combined license or
duplicate design license.
H. 10 CFR Part 73, Physical Protection of Plants and Materials
The proposed rule would amend Sec. 73.1(b) to clarify that the
regulations in 10 CFR part 73 also apply to licenses issued under 10
CFR part 52.
I. 10 CFR Part 140, Financial Protection Requirements and Indemnity
Agreements
The proposed rule would amend Sec. Sec. 140.2, 140.10, 140.11, and
140.13 to correct the language to note that holders of combined
licenses issued under 10 CFR part 52 are required to conform with the
Commission's financial protection requirements implementing the Price-
Anderson Act (Section 170 of the Atomic Energy Act of 1954). The
proposed rule would also add new Sec. Sec. 140.11(c) and 140.13(b).
Section 140.11(c) would specify that a holder of a combined license
must have and maintain financial protection when the Commission
authorizes operation under Sec. 52.231(g). Section 140.13(b) would
require that each holder of a combined license who is also the holder
of a license under 10 CFR part 70 authorizing ownership, possession,
and storage only of special nuclear material at the site of the nuclear
reactor have and maintain financial protection in the amount of
$1,000,000. Proof of financial protection would be required to be filed
with the Commission in the manner specified prior to issuance of the
license under 10 CFR part 70.
J. 10 CFR Part 170, Fees for Facilities, Materials, Import and Export
Licenses, and Other Regulatory Services Under the Atomic Energy Act of
1954, as Amended
The proposed rule would amend Sec. 170.2 to clarify the
applicability of the regulations in 10 CFR part 170 to the licensing
processes in 10 CFR parts 50 and 52.
IV. Specific Requests for Comments
In addition to the general invitation to submit comments on the
proposed rule, the Commission also requests comments on the following
questions:
1. Should the final rule include an updating requirement for other
than emergency preparedness information and what portions of the early
site permit (ESP) should be subject to the updating requirement? Also,
if an updating requirement is adopted, in what manner could an
interested person challenge the updated information? (refer to Sec.
52.39(a))
2. Should the final rule include revisions to 10 CFR part 52 to:
(1) Distinguish between site characteristics, site parameters, design
characteristics, and design parameters; (2) require the Commission to
specify the site characteristics and design parameters when issuing
early site permits; (3) require the design certification rule to
specify the site parameters and design characteristics for the design;
(4) require a combined license applicant referencing an early site
permit to demonstrate that either the design of the nuclear power plant
or the site parameters and design characteristics of a referenced
design certification rule fall within the design parameters and site
characteristics of the early site permit; and (5) require a combined
license applicant referencing a design certification rule to
demonstrate that the site parameters and design characteristics of the
design certification rule fall within either: (i) The site
characteristics of a site, or (ii) the site characteristics and design
parameters of a referenced early site permit?
Currently, 10 CFR art 52 uses the various terms, ``site
parameters,'' ``postulated site parameters,'' ``site characteristics,''
``physical characteristics,'' and ``the parameters specified in the
early site permit'' See, e.g., Sec. Sec. 52.17, 52.18, 52.21, 52.47
(proposed Sec. 52.107), Sec. 52.79 (proposed Sec. 52.211). In some
cases, it appears that different terms are used to apply to the same
concept, e.g., ``site parameters,'' and ``postulated site parameters.''
In other cases, information which would appear to constitute ``site
parameters'' as used in the current rule is not characterized as such,
e.g. Sec. 52.17(a)(1)(i) through (viii).
To address these inconsistencies, the Commission is considering
amending 10 CFR part 52, including proposed subparts A, D, and G, to
use the terms: ``site characteristics,'' ``site parameters,'' ``design
characteristics,'' and ``design parameters,'' to set forth in clear and
unambiguous terms the Commission's requirements on early site permits,
design certifications, and combined licenses. ``site characteristics''
would be the actual physical and demographic values for the site, e.g.,
the ground force acceleration of a defined earthquake, flood level, or
the atmospheric dispersion value. The ``design parameters'' for an
early site permit would include the postulated values for thermal power
level, radiological effluents, and type of cooling system for the
facility. ``Design characteristics'' for a design certification would
be the actual values for the design, e.g., thermal power level or
building height. ``Site parameters'' for a design certification would
include the postulated values for floods, ground force acceleration of
a postulated earthquake, and tornado wind speeds.
3. Are there terms and conditions for an ESP that can only be
fulfilled after issuance of the referencing combined license, such that
``have been met'' should be changed to ``will be met,'' or ``have been
and will be met''? (refer to proposed Sec. 52.211(a)(1))
4. Should the final rule include a requirement in Sec. 50.34(a)
for a construction permit application that references an ESP to
demonstrate that the design of the facility falls within the site
parameters of the ESP? (refer to proposed Sec. 52.211(a)(1))
5. Should the final rule include a requirement in 10 CFR part 50 to
perform testing to qualify advanced reactor designs before licensing?
The purpose of this testing requirement would be to demonstrate that
new or innovative safety features will perform as predicted in an
applicant's safety analysis report, that effects of systems
interactions have been found acceptable, and to provide sufficient data
for analytical code validation, as required by proposed Sec. Sec.
52.107(b) and 52.211(b).
6. Should the final rule include a revision to the current Sec.
52.63 (proposed Sec. 52.127) to allow the original design
certification applicant to petition the Commission for rulemaking to
amend the design certification rule to incorporate ``beneficial
changes,'' including improvements in safety, and/or design changes that
would ``significantly improve efficiency, reliability and economics.''
Refer to letters from Steven A. Hucik, GE Nuclear Energy (March 30,
2002) and Ronald L. Simard, Nuclear Energy Institute (March 22, 2002).
7. Should 10 CFR part 21 apply to: (a) A holder of an early site
permit, but only after the holder references the permit in a license
application, and (b) an applicant/vendor of a design which is the
subject of a design certification rule, but only after the design
certification rule is first referenced in a license application. In
both cases, the
[[Page 40040]]
Commission believes that there is no reasonable possibility of a
``substantial safety hazard'' until either the early site permit or
design certification rule is referenced. The Commission seeks public
comment on the Commission's proposed basis for this proposal, and
whether there are other factors and policy considerations, either in
support of, or in opposition to, the Commission's proposal.
V. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at 11555 Rockville Pike, Rockville, Maryland.
Rulemaking Website (Web). The NRC's interactive rulemaking Website
is located at http://ruleforum.llnl.gov. These documents may be viewed
and downloaded electronically via this Website.
NRC's Public Electronic Reading Room (PERR). The NRC's public
electronic reading room is located at www.nrc.gov/reading-rm.html.
----------------------------------------------------------------------------------------------------------------
Document PDR Web PERR
----------------------------------------------------------------------------------------------------------------
Comments on the draft rule language:
General Electric.......................... X X ML013180207
Entergy................................... X X ML013200006
Nuclear Energy Institute.................. X X ML013200158
Westinghouse.............................. X X ML013200173
Exelon........................................ X X ML020040187
Regulatory History of Design Certification X .......... ML003761550
\3\.
----------------------------------------------------------------------------------------------------------------
VI. Plain Language
The Presidential memorandum dated June 1, 1998, entitled ``Plain
Language in Government Writing'' directed that the Government's writing
be in plain language. This memorandum was published on June 10, 1998
(63 FR 31883). In complying with this directive, the NRC made editorial
changes to improve the organization and readability of the existing
language of the paragraphs being revised. These types of changes are
not discussed further in this document. The NRC requests comments on
the proposed rule specifically with respect to the clarity and
effectiveness of the language used. Comments should be sent to the
address listed under the ADDRESSES caption of the preamble.
---------------------------------------------------------------------------
\3\ The regulatory history of the NRC's design certification
reviews is a package of 100 documents that is available in NRC's
PERR and in the PDR. This history spans a 15-year period during
which the NRC simultaneously developed the regulatory standards for
reviewing these designs and the form and content of the rules that
certified the designs.
---------------------------------------------------------------------------
VII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless using such a standard is inconsistent with
applicable law or is otherwise impractical. In this proposed rule, the
NRC is revising the procedural requirements for early site permits,
standard design certifications, and combined licenses for nuclear power
plants to make certain corrections and changes based on the experience
of the previous design certification reviews and on discussions with
stakeholders on these licensing processes. In addition, this proposed
rule would amend certain portions of the three design certification
rules in 10 CFR part 52, appendices A, B, and C (for U.S. ABWR, System
80+, and AP600 designs, respectively) Design certifications are not
generic rulemakings in the sense that design certifications do not
establish standards or requirements with which all licensees must
comply. Rather, design certifications are Commission approvals of
specific nuclear power plant designs by rulemaking. Furthermore, design
certification rulemakings are initiated by an applicant for a design
certification, rather than the NRC. For these reasons, the Commission
concludes that this action does not constitute the establishment of a
standard that contains generally applicable requirements.
VIII. Environmental Impact: Categorical Exclusion
The NRC has determined that the changes made in this proposed rule
fall within the types of action described in categorical exclusions 10
CFR 51.22(c)(1), (c)(2), and (c)(3). Therefore, neither an
environmental impact statement nor an environmental assessment has been
prepared for this proposed regulation.\4\
---------------------------------------------------------------------------
\4\ When 10 CFR part 52 was promulgated in 1989, the NRC
determined that the regulation met the eligibility criteria for the
categorical exclusion set forth in 10 CFR 51.22(c)(3). As stated in
the Federal Register notice for the final rule (54 FR 15384, April
18, 1989), ``It makes no substantive difference for the purpose of
the categorical exclusion that the amendments are in a new 10 CFR
part 52 rather than in 10 CFR part 50. The amendments are, in fact,
amendments to the 10 CFR part 50 procedures and could have been
placed in that part.'' The categorical exclusion for the current
proposed change to 10 CFR part 52 is consistent with the original
categorical exclusion determination.
---------------------------------------------------------------------------
IX. Paperwork Reduction Act Statement
This proposed rule amends information collection requirements
contained in 10 CFR Part 52 that are subject to the Paperwork Reduction
Act of 1995 (44 U.S.C. 3501 et seq). These information collection
requirements have been submitted to the Office of Management and Budget
for review and approval. The proposed changes to 10 CFR parts 2, 20,
21, 50, 51, 72, 73, 140, and 170 do not contain new or amended
information collection requirements. Existing requirements were
approved by the Office of Management and Budget, approval number(s)
3150-0014, 3150-0035, 3150-0011, 3150-0021, 3150-0132, 3150-0039, and
3150-0002.
The burden to the public for the information collections in 10 CFR
part 52 is estimated to average 3,429 hours per response. This includes
the time for reviewing instructions, searching existing data sources,
gathering and maintaining the data needed, and completing and reviewing
the information collection. The U.S. Nuclear Regulatory Commission is
seeking public comment on the potential impact of the information
collections contained in the proposed rule and on the following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
2. Is the estimate of burden accurate?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the information collection be minimized,
including the use of automated collection techniques?
[[Page 40041]]
Send comments on any aspect of these proposed information
collections, including suggestions for reducing the burden, to the
Records Management Branch (T-6 E6), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by Internet electronic mail to
[email protected]; and to the Desk Officer, Office of Information
and Regulatory Affairs, NEOB-10202, (3150-0151, 3150-0011, and 3150-
0039), Office of Management and Budget, Washington, DC 20503.
Comments to OMB on the information collections or on the above
issues should be submitted by August 4, 2003. Comments received after
this date will be considered if it is practical to do so, but assurance
of consideration cannot be given to comments received after this date.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
X. Regulatory Analysis
The Commission has prepared the following draft regulatory analysis
on the substantive changes in this proposed regulation that could
impose regulatory burdens. The majority of the changes in this proposed
rule involve formatting, reorganization, or process changes that do not
affect regulatory burden. These types of changes are not addressed in
this regulatory analysis, as they would not affect the burden on future
applicants.
The proposed rule contains two amendments that appear to impose
regulatory burdens on future applicants for construction permits,
combined licenses, and duplicate design licenses who may file an
application referencing an early site permit or a certified design.
There are no current applicants who would be burdened by the proposed
amendments.
The first of these changes requires applicants who reference an
early site permit to update and correct emergency planning information
and discuss whether the new information materially alters the bases for
compliance with the applicable requirements. The second change requires
applicants who reference a certified design to include a plant-specific
probabilistic risk assessment (PRA) that uses the design-specific PRA
and is updated to account for site-specific design information and any
design changes.
The Commission believes that, practically speaking, there would be
no change in the burden on future applicants resulting from these
amendments. This is because the information required by the proposed
rule would, in all likelihood, be requested by the NRC staff during the
review of the application if these requirements were not adopted. The
staff could not perform an adequate review of an application
referencing an early site permit without reviewing the most up-to-date
emergency planning information. Therefore, if this updated information
was not required in the application, the staff would be compelled to
request the information from the applicant in order to make a finding
that there is reasonable assurance that adequate protective measures
can and will be taken in the event of a radiological emergency.
Likewise, if the Commission did not require an updated PRA in an
application for a combined license referencing a certified design, the
staff would be compelled to request the information from the applicant.
The Commission would need this information in order to assist it in
finding that the applicable requirements of 10 CFR part 50 have been
met, and in reviewing the licensee's proposed inspections, tests, and
analyses that the licensee must 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 Atomic Energy
Act, and the Commission's rules and regulations.
For these reasons, the Commission believes it is prudent to proceed
with this proposed rulemaking. The addition of these requirements for
applicants for construction permits, combined licenses, and duplicate
design licenses is necessary to ensure the NRC staff can meet its
regulatory obligations. In addition, giving future applicants
notification up front that the staff requires this information in the
application will relieve them of a larger burden of having to compile
the information during the application review process when the
Commission requests the information to complete its review. The need to
compile the information during the review process could impact the
review schedule and result in other unnecessary burdens on the
applicant.
The Commission requests public comment on the draft regulatory
analysis. Comments on the draft analysis may be submitted to the NRC as
indicated under the ADDRESSES heading.
XI. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the Commission certifies that this rule will not, if
promulgated, have a significant economic impact on a substantial number
of small entities. This proposed rule affects only the licensing of
nuclear power plants. The companies that will apply for an approval,
certification, permit, site report, or license in accordance with the
regulations affected by this proposed rule do not fall within the scope
of the definition of ``small entities'' set forth in the Regulatory
Flexibility Act or the size standards established by the NRC (10 CFR
2.810).
XII. Backfit Analysis
The NRC has determined that the backfit rule does not apply to this
proposed rule; therefore, a backfit analysis is not required for this
proposed rule because these amendments do not involve any provisions
that would impose backfits as defined in 10 CFR 50.109. The proposed
rule would revise the requirements for early site permits, standard
design certifications, and combined licenses for nuclear power plants,
so it would affect a potential applicant who might, in the future,
apply for an early site permit, design certification, or combined
license. However, the backfit rule does not apply because the proposed
rule would not impose any modifications on a current holder of an early
site permit, certified design, or combined license.
List of Subjects
10 CFR Part 2
Administrative practice and procedure, Antitrust, Byproduct
material, Classified information, Environmental protection, Nuclear
materials, Nuclear power plants and reactors, Penalties, Sex
discrimination, Source material, Special nuclear material, Waste
treatment and disposal.
10 CFR Part 20
Byproduct material, Criminal penalties, Licensed material, Nuclear
materials, Nuclear power plants and reactors, Occupational safety and
health, Packaging and containers, Radiation protection, Reporting and
record keeping requirements, Source material, Special nuclear material,
Waste treatment and disposal.
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
[[Page 40042]]
Reporting and record keeping requirements.
10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
record keeping requirements.
10 CFR Part 51
Administrative practice and procedure, Environmental impact
statement, Nuclear materials, Nuclear power plants and reactors,
Reporting and record keeping requirements.
10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and record keeping requirements,
Standard design, Standard design certification.
10 CFR Part 72
Administrative practice and procedure, Criminal penalties, Manpower
training programs, Nuclear materials, Occupational safety and health,
Penalties, Radiation protection, Reporting and record keeping
requirements, Security measures, Spent fuel, Whistle blowing.
10 CFR Part 73
Criminal penalties, Export, Hazardous materials transportation,
Import, Nuclear materials, Nuclear power plants and reactors, Reporting
and record keeping requirements, Security measures.
10 CFR Part 140
Criminal penalties, Extraordinary nuclear occurrence, Insurance,
Intergovernmental relations, Nuclear materials, Nuclear power plants
and reactors, Reporting and record keeping requirements.
10 CFR Part 170
Byproduct material, Import and export licenses, Intergovernmental
relations, Non-payment penalties, Nuclear materials, Nuclear power
plants and reactors, Source material, Special nuclear material.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 553, the NRC is proposing to
adopt the following amendments to 10 CFR parts 2, 20, 21, 50, 51, 52,
72, 73, 140, and 170.
PART 2--RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND
ISSUANCE OF ORDERS
1. The authority citation for part 2 continues to read as follows:
Authority: Secs. 161, 181, 68 Stat. 948, 953, as amended (42
U.S.C. 2201, 2231); sec. 191, as amended, Pub. L. 87-615, 76 Stat.
409 (42 U.S.C. 2241); sec. 201, 88 Stat.1242, as amended (42 U.S.C.
5841); 5 U.S.C. 552.
Section 2.101 also issued under secs. 53, 62, 63, 81, 103, 104,
105, 68 Stat. 930, 932, 933, 935, 936, 937, 938, as amended (42
U.S.C. 2073, 2092, 2093, 2111, 2133, 2134, 2135); sec. 114(f), Pub.
L. 97-425, 96 Stat. 2213, as amended (42 U.S.C. 10143(f)); sec. 102,
Pub. L. 91-190, 83 Stat. 853, as amended (42 U.S.C. 4332); sec. 301,
88 Stat. 1248 (42 U.S.C. 5871). Sections 2.102, 2.103, 2.104, 2.105,
2.721 also issued under secs. 102, 103, 104, 105, 183i, 189, 68
Stat. 936, 937, 938, 954, 955, as amended (42 U.S.C. 2132, 2133,
2134, 2135, 2233, 2239). Section 2.105 also issued under Pub. L. 97-
415, 96 Stat. 2073 (42 U.S.C. 2239). Sections 2.200-2.206 also
issued under secs. 161 b, i, o, 182, 186, 234, 68 Stat. 948-951,
955, 83 Stat. 444, as amended (42 U.S.C. 2201 (b), (i), (o), 2236,
2282); sec. 206, 88 Stat 1246 (42 U.S.C. 5846). Section 2.205(j)
also issued under Pub. L. 101-410, 104 Stat. 90, as amended by
section 3100(s), Pub. L. 104-134, 110 Stat. 1321-373 (28 U.S.C. 2461
note). Sections 2.600-2.606 also issued under sec. 102, Pub. L. 91-
190, 83 Stat. 853, as amended (42 U.S.C. 4332). Sections 2.700a,
2.719 also issued under 5 U.S.C. 554. Sections 2.754, 2.760, 2.770,
2.780 also issued under 5 U.S.C. 557. Section 2.764 also issued
under secs. 135, 141, Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C.
10155, 10161). Section 2.790 also issued under sec. 103, 68 Stat.
936, as amended (42 U.S.C. 2133), and 5 U.S.C. 552. Sections 2.800
and 2.808 also issued under 5 U.S.C. 553. Section 2.809 also issued
under 5 U.S.C. 553, and sec. 29, Pub. L. 85-256, 71 Stat. 579, as
amended (42 U.S.C. 2039). Subpart K also issued under sec. 189, 68
Stat. 955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230
(42 U.S.C. 10154). Subpart L also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239). Subpart M also issued under sec. 184 (42
U.S.C. 2234) and sec. 189, 68 stat. 955 (42 U.S.C. 2239). Appendix A
also issued under sec. 6, Pub. L. 91-560, 84 Stat. 1473 (42 U.S.C.
2135).
2. In Sec. 2.110, paragraph (a) is revised to read as follows:
Sec. 2.110 Filing and administrative action on submittals for design
review or early review of site suitability issues.
(a)(1) A submittal under subpart E of part 52 of this chapter must
be subject to Sec. Sec. 2.101(a) and 2.790 to the same extent as if it
were an application for a permit or license.
(2) Except as specifically provided otherwise by the provisions of
subpart B to part 52 of this chapter, a submittal under subpart B must
be subject to Sec. 2.101(a) (2) through (4) to the same extent as if
it were an application for a permit or license.
* * * * *
3. Section 2.400 is revised to read as follows:
Sec. 2.400 Scope of subpart.
This subpart describes procedures applicable to licensing
proceedings that involve the consideration in hearings of a number of
applications, filed by one or more applicants pursuant to subpart I of
part 52 of this chapter, for licenses to construct and operate nuclear
power reactors of essentially the same design to be located at
different sites.
4. Section 2.401 is revised to read as follows:
Sec. 2.401 Notice of hearing on applications under Subpart I of Part
52 for construction permits.
(a) In the case of applications under subpart I of part 52 of this
chapter for construction permits for nuclear power reactors of the type
described in Sec. 50.22 of this chapter, the Secretary will issue
notices of hearing under Sec. 2.104.
(b) The notice of hearing will also state the time and place of the
hearings on any separate phase of the proceeding.
5. In Sec. 2.402, paragraph (a) is revised to read as follows:
Sec. 2.402 Separate hearings on separate issues; consolidation of
proceedings.
(a) In the case of applications under subpart I of part 52 of this
chapter for construction permits for nuclear power reactors of a type
described in Sec. 50.22 of this chapter, the Commission or the
presiding officer may order separate hearings on particular phases of
the proceeding, such as matters related to the acceptability of the
design of the reactor, in the context of the site parameters postulated
for the design; environmental matters; or antitrust aspects of the
application.
* * * * *
6. Section 2.403 is revised to read as follows:
Sec. 2.403 Notice of proposed action on applications for operating
licenses under Subpart I of Part 52.
In the case of applications under subpart I of part 52 of this
chapter for operating licenses for nuclear power reactors, if the
Commission has not found that a hearing is in the public interest, the
Director of Nuclear Reactor Regulation will, prior to acting thereon,
cause to be published in the Federal Register, under Sec. 2.105, a
notice of proposed action with respect to each
[[Page 40043]]
application as soon as practicable after the applications have been
docketed.
7. Section 2.404 is revised to read as follows:
Sec. 2.404 Hearings on applications for operating licenses under
Subpart I of Part 52.
If a request for a hearing and/or petition for leave to intervene
is filed within the time prescribed in the notice of proposed action on
an application for an operating license under subpart I of part 52 of
this chapter with respect to a specific reactor(s) at a specific site
and the Commission or an atomic safety and licensing board designated
by the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel has issued a notice of hearing or other appropriate order,
the Commission or the atomic safety and licensing board may order
separate hearings on particular phases of the proceeding and/or
consolidate for hearing two or more proceedings in the manner described
in Sec. 2.402.
8. Section 2.406 is revised to read as follows:
Sec. 2.406 Finality of decisions on separate issues.
Notwithstanding any other provision of this chapter, in a
proceeding conducted under this subpart and subpart I of part 52 of
this chapter, no matter which has been reserved for consideration in
one phase of the hearing shall be considered at another phase of the
hearing except on the basis of significant new information that
substantially affects the conclusion(s) reached at the other phase or
other good cause.
9. Section 2.500 is revised to read as follows:
Sec. 2.500 Scope of subpart.
This subpart prescribes procedures applicable to licensing
proceedings which involve the consideration in separate hearings of an
application for a license to manufacture nuclear power reactors under
subpart H of part 52 of this chapter, and applications for construction
permits and operating licenses for nuclear power reactors which have
been the subject of such an application for a license to manufacture
such facilities (manufacturing license).
10. In Sec. 2.501, paragraphs (a), (b)(1)(vii) and (b)(3) are
revised to read as follows:
Sec. 2.501 Notice of hearing on application under Subpart H of Part
52 for a license to manufacture nuclear power reactors.
(a) In the case of an application under subpart H of part 52 of
this chapter for a license to manufacture nuclear power reactors of the
type described in Sec. 50.22 of this chapter to be operated at sites
not identified in the license application, the Secretary shall issue a
notice of hearing to be published in the Federal Register at least
thirty (30) days prior to the date set for hearing in the notice. The
notice must be issued as soon as practicable after the application has
been docketed. The notice will state:
(1) The time, place, and nature of the hearing and/or the
prehearing conference;
(2) The authority within which the hearing is to be held;
(3) The matters of fact and law to be considered; and
(4) The time within which answers to the notice shall be filed.
(b) * * *
(1) * * *
(vii) Whether, in accordance with the requirements of subpart A of
part 51 and subpart H of part 52 of this chapter, the license should be
issued as proposed.
* * * * *
(3) That, regardless of whether the proceeding is contested or
uncontested, the presiding officer will, in accordance with subpart A
of part 51 and Sec. 52.245(b) of this chapter,
* * * * *
11. Section 2.502 is revised to read as follows:
Sec. 2.502 Notice of hearing on application for a permit to construct
a nuclear power reactor manufactured under a Commission license issued
under subpart H of part 52 of this chapter at the site at which the
reactor is to be operated.
The issues stated for consideration in the notice of hearing on an
application for a permit to construct a nuclear power reactor(s) which
is the subject of an application for a manufacturing license under
subpart H of part 52 of this chapter, will be those stated in Sec.
2.104(b) and, in addition, whether the site on which the facility is to
be operated falls within the postulated site parameters specified in
the relevant application for a manufacturing license.
PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION
12. The authority citation for part 20 continues to read as
follows:
Authority: Secs. 53, 63, 65, 81, 103, 104, 161, 182, 186, 68
Stat. 930, 933, 935, 936, 937, 948, 953, 955, as amended, sec. 1701,
106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2093, 2095, 2111, 2133,
2134, 2201, 2232, 2236, 2297f), secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
13. Section 20.1002 is revised to read as follows:
Sec. 20.1002 Scope.
The regulations in this part apply to persons licensed by the
Commission to receive, possess, use, transfer, or dispose of byproduct,
source, or special nuclear material or to operate a production or
utilization facility under parts 30 through 36, 39, 40, 50, 52, 60, 61,
70, or 72 of this chapter, and in accordance with 10 CFR 76.60 to
persons required to obtain a certificate of compliance or an approved
compliance plan under part 76 of this chapter. The limits in this part
do not apply to doses due to background radiation, to exposure of
patients to radiation for the purpose of medical diagnosis or therapy,
to exposure from individuals administered radioactive material and
released in accordance with 10 CFR 35.75, or to exposure from voluntary
participation in medical research programs.
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
14. The authority citation for part 21 continues to read as
follows:
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C.
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as
amended, 1246 (42 U.S.C. 5841, 5846).
Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425, 96
Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
15. In Sec. 21.2, paragraphs (a), (b), and (c) are revised to read
as follows:
Sec. 21.2 Scope.
(a) The regulations in this part apply, except as specifically
provided otherwise in Parts 31, 34, 35, 39, 40, 60, 61, 63, 70, or Part
72 of this chapter, to:
(1) Each individual, partnership, corporation, or other entity
licensed pursuant to the regulations in this chapter to possess, use,
or transfer within the United States source material, byproduct
material, special nuclear material, and/or spent fuel and high-level
radioactive waste, or to construct, manufacture, possess, own, operate,
or transfer within the United States, any production or utilization
facility or independent spent fuel storage installation (ISFSI) or
monitored retrievable storage installation (MRS); and each director and
responsible officer of such a licensee; and
(2) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such organization, that holds a permit or
license under part 52 of this chapter or constructs a production or
utilization
[[Page 40044]]
facility licensed for the manufacture, construction, or operation
pursuant to part 50 or part 52 of this chapter, an ISFSI for the
storage of spent fuel licensed pursuant to part 72 of this chapter, an
MRS for the storage of spent fuel or high-level radioactive waste
pursuant to part 72 of this chapter, or a geologic repository for the
disposal of high-level radioactive waste under part 60 or 63 of this
chapter; or supplies basic components for a facility or activity
licensed, other than for export, under parts 30, 40, 50, 52, 60, 61,
63, 70, 71, or part 72 of this chapter.
(b) For persons licensed to construct a facility under either a
construction permit issued under Sec. 50.23 of this chapter or a
combined license issued under Sec. 52.227 of this chapter, or approved
to hold a permit for a site or sites for one or more nuclear power
facilities under Sec. 52.24 of this chapter, evaluation of potential
defects and failures to comply and reporting of defects and failures to
comply under Sec. 50.55(e) of this chapter satisfies each person's
evaluation, notification, and reporting obligation to report defects
and failures to comply under this part and the responsibility of
individual directors and responsible officers of such licensees to
report defects under section 206 of the Energy Reorganization Act of
1974.
(c) For persons licensed to operate a nuclear power plant under
part 50 or part 52 of this chapter, evaluation of potential defects and
appropriate reporting of defects under Sec. Sec. 50.72, 50.73 or Sec.
73.71 of this chapter satisfies each person's evaluation, notification,
and reporting obligation to report defects under this part and the
responsibility of individual directors and responsible officers of such
licensees to report defects under section 206 of the Energy
Reorganization Act of 1974.
* * * * *
16. Section 21.3 is revised to read as follows:
Sec. 21.3 Definitions.
As used in this part:
Basic component. (1)(i) When applied to nuclear power plants
licensed pursuant to 10 CFR part 50 or part 52 of this chapter, basic
component means a structure, system, or component, or part thereof that
affects its safety function necessary to assure:
(A) The integrity of the reactor coolant pressure boundary;
(B) The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
(C) 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.
(ii) Basic components are items designed and manufactured under a
quality assurance program complying with 10 CFR part 50, appendix B, or
commercial grade items which have successfully completed the dedication
process.
(2) When applied to other facilities and when applied to other
activities licensed pursuant to 10 CFR parts 30, 40, 50 (other than
nuclear power plants), 60, 61, 63, 70, 71, or 72 of this chapter, basic
component means a structure, system, or component, or part thereof that
affects their safety function, that is directly procured by the
licensee of a facility or activity subject to the regulations in this
part and in which a defect or failure to comply with any applicable
regulation in this chapter, order, or license issued by the Commission
could create a substantial safety hazard.
(3) In all cases, basic component includes safety-related design,
analysis, inspection, testing, fabrication, replacement of parts, or
consulting services that are associated with the component hardware
whether these services are performed by the component supplier or
others.
Commercial grade item. (1) When applied to nuclear power plants
licensed pursuant to 10 CFR part 50 or part 52, commercial grade item
means a structure, system, or component, or part thereof that affects
its safety function, that was not designed and manufactured as a basic
component. Commercial grade items do not include items where the design
and manufacturing process require in-process inspections and
verifications to ensure that defects or failures to comply are
identified and corrected (i.e., one or more critical characteristics of
the item cannot be verified).
(2) When applied to facilities and activities licensed pursuant to
10 CFR parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63,
70, 71, or 72, commercial grade item means an item that is:
(i) Not subject to design or specification requirements that are
unique to those facilities or activities;
(ii) Used in applications other than those facilities or
activities; and
(iii) To be ordered from the manufacturer/supplier on the basis of
specifications set forth in the manufacturer's published product
description (for example, a catalog).
Commission means the Nuclear Regulatory Commission or its duly
authorized representatives.
Constructing or construction means the analysis, design,
manufacture, fabrication, placement, erection, installation,
modification, inspection, or testing of a facility or activity which is
subject to the regulations in this part and consulting services related
to the facility or activity that are safety related.
Critical characteristics. When applied to nuclear power plants
licensed pursuant to 10 CFR part 50 or part 52, critical
characteristics are those important design, material, and performance
characteristics of a commercial grade item that, once verified, will
provide reasonable assurance that the item will perform its intended
safety function.
Dedicating entity. When applied to nuclear power plants licensed
pursuant to 10 CFR part 50 or part 52, dedicating entity means the
organization that performs the dedication process. Dedication may be
performed by the manufacturer of the item, a third-party dedicating
entity, or the licensee. The dedicating entity, pursuant to Sec.
21.21(c) of this part, is responsible for identifying and evaluating
deviations, reporting defects and failures to comply for the dedicated
item, and maintaining auditable records of the dedication process.
Dedication. (1) When applied to nuclear power plants licensed
pursuant to 10 CFR part 50 or part 52, dedication is an acceptance
process undertaken to provide reasonable assurance that a commercial
grade item to be used as a basic component will perform its intended
safety function and, in this respect, is deemed equivalent to an item
designed and manufactured under a 10 CFR part 50, appendix B, quality
assurance program. This assurance is achieved by identifying the
critical characteristics of the item and verifying their acceptability
by inspections, tests, or analyses performed by the purchaser or third-
party dedicating entity after delivery, supplemented as necessary by
one or more of the following: commercial grade surveys; product
inspections or witness at holdpoints at the manufacturer's facility,
and analysis of historical records for acceptable performance. In all
cases, the dedication process must be conducted in accordance with the
applicable provisions of 10 CFR part 50, appendix B. The process is
considered complete when the item is designated for use as a basic
component.
(2) When applied to facilities and activities licensed pursuant to
10 CFR parts 30, 40, 50 (other than nuclear
[[Page 40045]]
power plants), 60, 61, 63, 70, 71, or 72, dedication occurs after
receipt when that item is designated for use as a basic component.
Defect means: (1) A deviation in a basic component delivered to a
purchaser for use in a facility or an activity subject to the
regulations in this part if, on the basis of an evaluation, the
deviation could create a substantial safety hazard; or
(2) The installation, use, or operation of a basic component
containing a defect as defined in this section; or
(3) A deviation in a portion of a facility subject to the
construction permit or manufacturing licensing requirements of part 50
or part 52 of this chapter provided the deviation could, on the basis
of an evaluation, create a substantial safety hazard and the portion of
the facility containing the deviation has been offered to the purchaser
for acceptance; or
(4) A condition or circumstance involving a basic component that
could contribute to the exceeding of a safety limit, as defined in the
technical specifications of a license for operation issued pursuant to
part 50 or part 52 of this chapter.
Deviation means a departure from the technical requirements
included in a procurement document.
Director means an individual, appointed or elected according to
law, who is authorized to manage and direct the affairs of a
corporation, partnership or other entity. In the case of an individual
proprietorship, director means the individual.
Discovery means the completion of the documentation first
identifying the existence of a deviation or failure to comply
potentially associated with a substantial safety hazard within the
evaluation procedures discussed in Sec. 21.21(a).
Evaluation means the process of determining whether a particular
deviation could create a substantial hazard or determining whether a
failure to comply is associated with a substantial safety hazard.
Notification means the telephonic communication to the NRC
Operations Center or written transmittal of information to the NRC
Document Control Desk.
Operating or operation means the operation of a facility or the
conduct of a licensed activity which is subject to the regulations in
this part and consulting services related to operations that are safety
related.
Procurement document means a contract that defines the requirements
which facilities or basic components must meet in order to be
considered acceptable by the purchaser.
Responsible officer means the president, vice-president or other
individual in the organization of a corporation, partnership, or other
entity who is vested with executive authority over activities subject
to this part.
Substantial safety hazard means a loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to public health and safety for any facility or activity
licensed, other than for export, pursuant to parts 30, 40, 50, 52, 60,
61, 63, 70, 71, or 72 of this chapter.
Supplying or supplies means contractually responsible for a basic
component used or to be used in a facility or activity which is subject
to the regulations in this part.
17. Section 21.21 is revised to read as follows:
Sec. 21.21 Notification of failure to comply or existence of a defect
and its evaluation.
(a) Each individual, corporation, partnership, dedicating entity,
or other entity subject to the regulations in this part shall adopt
appropriate procedures to--
(1) Evaluate deviations and failures to comply to identify defects
and failures to comply associated with substantial safety hazards as
soon as practicable, and, except as provided in paragraph (a)(2) of
this section, in all cases within 60 days of discovery, in order to
identify a reportable defect or failure to comply that could create a
substantial safety hazard, were it to remain uncorrected, and
(2) Ensure that if an evaluation of an identified deviation or
failure to comply potentially associated with a substantial safety
hazard cannot be completed within 60 days from discovery of the
deviation or failure to comply, an interim report is prepared and
submitted to the Commission through a director or responsible officer
or designated person as discussed in Sec. 21.21(d)(5). The interim
report should describe the deviation or failure to comply that is being
evaluated and should also state when the evaluation will be completed.
This interim report must be submitted in writing within 60 days of
discovery of the deviation or failure to comply.
(3) Ensure that a director or responsible officer subject to the
regulations of this part is informed as soon as practicable, and, in
all cases, within the 5 working days after completion of the evaluation
described in Sec. 21.21(a)(1) if the construction or operation of a
facility or activity, or a basic component supplied for such facility
or activity--
(i) Fails to comply with the Atomic Energy Act of 1954, as amended,
or any applicable rule, regulation, order, or license of the Commission
relating to a substantial safety hazard, or
(ii) Contains a defect.
(b) If the deviation or failure to comply is discovered by a
supplier of basic components, or services associated with basic
components, and the supplier determines that it does not have the
capability to perform the evaluation to determine if a defect exists,
then the supplier must inform the purchasers or affected licensees
within five working days of this determination so that the purchasers
or affected licensees may evaluate the deviation or failure to comply,
pursuant to Sec. 21.21(a).
(c) A dedicating entity is responsible for--
(1) Identifying and evaluating deviations and reporting defects and
failures to comply associated with substantial safety hazards for
dedicated items; and
(2) Maintaining auditable records for the dedication process.
(d)(1) A director or responsible officer subject to the regulations
of this part or a person designated under Sec. 21.21(d)(5) must notify
the Commission when he or she obtains information reasonably indicating
a failure to comply or a defect affecting--
(i) The construction or operation of a facility or an activity
within the United States that is subject to the licensing requirements
under parts 30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of this chapter
and that is within his or her organization's responsibility; or
(ii) A basic component that is within his or her organization's
responsibility and is supplied for a facility or an activity within the
United States that is subject to the licensing requirements under parts
30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of this chapter.
(2) The notification to NRC of a failure to comply or of a defect
under paragraph (d)(1) of this section and the evaluation of a failure
to comply or a deviation under paragraph (a)(1) of this section, are
not required if the director or responsible officer has actual
knowledge that the Commission has been notified in writing of the
defect or the failure to comply.
(3) Notification required by paragraph (d)(1) of this section must
be made as follows--
(i) Initial notification by facsimile, which is the preferred
method of notification, to the NRC Operations Center at (301) 816-5151
or by
[[Page 40046]]
telephone at (301) 816-5100 within two days following receipt of
information by the director or responsible corporate officer under
paragraph (a)(3) of this section, on the identification of a defect or
a failure to comply. Verification that the facsimile has been received
should be made by calling the NRC Operations Center. This paragraph
does not apply to interim reports described in Sec. 21.21(a)(2).
(ii) Written notification to the NRC at the address specified in
Sec. 21.5 within 30 days following receipt of information by the
director or responsible corporate officer under paragraph (a)(3) of
this section, on the identification of a defect or a failure to comply.
(4) The written report required by this paragraph must include, but
need not be limited to, the following information, to the extent known:
(i) Name and address of the individual or individuals informing the
Commission.
(ii) Identification of the facility, the activity, or the basic
component supplied for such facility or such activity within the United
States which fails to comply or contains a defect.
(iii) Identification of the firm constructing the facility or
supplying the basic component which fails to comply or contains a
defect.
(iv) Nature of the defect or failure to comply and the safety
hazard which is created or could be created by such defect or failure
to comply.
(v) The date on which the information of such defect or failure to
comply was obtained.
(vi) In the case of a basic component which contains a defect or
fails to comply, the number and location of all such components in use
at, supplied for, or being supplied for one or more facilities or
activities subject to the regulations in this part.
(vii) The corrective action which has been, is being, or will be
taken; the name of the individual or organization responsible for the
action; and the length of time that has been or will be taken to
complete the action.
(viii) Any advice related to the defect or failure to comply about
the facility, activity, or basic component that has been, is being, or
will be given to purchasers or licensees.
(5) The director or responsible officer may authorize an individual
to provide the notification required by this paragraph, provided that,
this shall not relieve the director or responsible officer of his or
her responsibility under this paragraph.
(e) Individuals subject to this part may be required by the
Commission to supply additional information related to a defect or
failure to comply. Commission action to obtain additional information
may be based on reports of defects from other reporting entities.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
18. The authority citation for part 50 continues to read as
follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83
Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232,
2233, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242,
as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951, as amended by Pub. L. 102-486, sec. 2902, 106 Stat. 3123 (42
U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68
Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under Pub. L. 97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78
also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections
50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42
U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955
(42 U.S.C. 2237).
19. In Sec. 50.8, paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 50.30, 50.33, 50.33a, 50.34, 50.34a,
50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54,
50.55, 50.55a, 50.59, 50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 50.66,
50.68, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120,
and appendices A, B, E, G, H, I, J, K, R, and S to this part.
* * * * *
20. In Sec. 50.109, paragraph (a)(1) is revised to read as
follows:
Sec. 50.109 Backfitting.
(a)(1) Backfitting is defined as the modification of or addition to
systems, structures, components, or design of a facility; or the design
approval or manufacturing license for a facility; or the procedures or
organization required to design, construct or operate a facility; any
of which may result from a new or amended provision in the Commission
rules or the imposition of a regulatory staff position interpreting the
Commission rules that is either new or different from a previously
applicable staff position after:
(i) The date of issuance of the construction permit for the
facility for facilities having construction permits issued after
October 21, 1985; or
(ii) Six months before the date of docketing of the operating
license application for the facility for facilities having construction
permits issued before October 21, 1985; or
(iii) The date of issuance of the operating license for the
facility for facilities having operating licenses; or
(iv) The date of issuance of the design approval under subpart E of
part 52 of this chapter;
(v) The date of issuance of a manufacturing license under subpart H
of part 52 of this chapter;
(vi) The date of issuance of the first construction permit issued
for a duplicate design under subpart I of part 52 of this chapter; or
(vii) The date of issuance of a combined license under subpart G of
part 52 of this chapter, provided that if the combined license
references an early site permit, the provisions in Sec. 52.39 apply
with respect to the site characteristics, terms, and conditions of the
early site permit. If the combined license references an early site
review, the provisions in Sec. 52.47 apply with respect to the staff
site report. If the combined license references a design certification
rule, the provisions in Sec. 52.127(a) apply with respect to the
design matters resolved in the design certification.
* * * * *
Appendix M to Part 50 [Removed]
21. Appendix M to Part 50 is removed.
Appendix N to Part 50 [Removed]
22. Appendix N to Part 50 is removed.
Appendix O to Part 50 [Removed]
23. Appendix O to Part 50 is removed.
Appendix Q to Part 50 [Removed]
24. Appendix Q to Part 50 is removed.
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
25. The authority citation for Part 51 continues to read as
follows:
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,
[[Page 40047]]
1244 (42 U.S.C. 5841, 5842). 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)).
26. In Sec. 51.20, paragraph (b)(6) is revised to read as follows:
Sec. 51.20 Criteria for and identification of licensing and
regulatory actions requiring environmental impact statements.
* * * * *
(b) * * *
(6) Issuance of a license to manufacture pursuant to Subpart H of
Part 52 of this chapter.
* * * * *
27. Part 52 is revised to read as follows:
PART 52--ADDITIONAL LICENSING PROCESSES FOR NUCLEAR POWER PLANTS
General Provisions
Sec.
52.1 Scope.
52.3 Definitions.
52.5 Applicability of 10 CFR Part 50 provisions.
52.8 Information collection requirements: OMB approval.
Subpart A--Early Site Permits
52.11 Scope of subpart.
52.13 Relationship to Subpart F of 10 CFR Part 2 and Subpart B of
this part.
52.15 Filing of applications.
52.17 Contents of applications.
52.18 Standards for review of applications.
52.19 Applicability of NRC requirements.
52.21 Hearings.
52.23 Referral to the ACRS.
52.24 Issuance of early site permit.
52.25 Extent of activities permitted.
52.27 Duration of 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.
52.37 Reporting of defects and noncompliance; revocation,
suspension, modification of permits for cause.
52.39 Finality of early site permit determinations.
Subpart B--Early Site Reviews
52.41 Scope of subpart.
52.43 Filing and contents of applications.
52.45 Notice of application.
52.46 Referral to the ACRS.
52.47 Issuance of site report.
52.49 Relationship to other subparts.
Subpart C--[Reserved]
Subpart D--Standard Design Certifications
52.101 Scope of subpart.
52.103 Relationship to other subparts.
52.105 Filing of applications.
52.107 Contents of applications.
52.109 Standards for review of applications.
52.111 Applicability of NRC requirements.
52.113 Administrative review of applications.
52.115 Referral to the ACRS.
52.117 Issuance of standard design certification.
52.119 Duration of certification.
52.121 Application for renewal.
52.123 Criteria for renewal.
52.125 Duration of renewal.
52.127 Finality of standard design certifications.
Subpart E--Standard Design Approvals
52.131 Scope of subpart.
52.133 Filing of applications.
52.135 Contents of applications.
52.137 Referral to the ACRS.
52.139 Staff approval of design.
52.141 Finality of the design approval.
52.143 Information requests.
Subpart F--[Reserved]
Subpart G--Combined Licenses
52.201 Scope of subpart.
52.203 Relationship to other subparts.
52.205 Filing of applications.
52.207 Contents of applications; general information.
52.209 Contents of applications; training and qualification of
nuclear power plant personnel.
52.211 Contents of applications; technical information.
52.213 Standards for review of applications.
52.215 Applicability of NRC requirements.
52.217 Administrative review of applications.
52.219 Referral to the ACRS.
52.221 Environmental review.
52.223 Authorization to conduct site activities.
52.225 Exemptions and variances.
52.227 Issuance of combined licenses.
52.229 Inspection during construction.
52.231 Operation under a combined license.
Subpart H--Manufacturing Licenses
52.241 Scope of subpart.
52.243 Relationship to other subparts.
52.245 Filing and contents of applications.
52.247 Standards for review of applications.
52.249 Applicability of NRC requirements.
52.251 Referral to the ACRS.
52.253 Issuance of manufacturing license.
52.255 Duration of design approval.
52.257 Finality of the manufacturing license.
Subpart I--Duplicate Design Licenses
52.261 Scope of subpart.
52.263 Relationship to other subparts.
52.265 Filing and contents of applications.
Subpart J--[Reserved]
Subpart K--[Reserved]
Subpart L--[Reserved]
Subpart M--Enforcement
52.401 Violations.
52.403 Criminal penalties.
APPENDIX A--Design Certification Rule for the U.S. Advanced Boiling
Water Reactor
APPENDIX B--Design Certification Rule for the System 80+ Design
APPENDIX C--Design Certification Rule for the AP600 Design
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).
General Provisions
Sec. 52.1 Scope.
This part governs the issuance of early site permits and staff site
reports, design approvals and certifications, and combined,
manufacturing, and duplicate design licenses for nuclear power
facilities licensed under section 103 or 104b 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 licensee, holder of, or
applicant for an approval, certification, permit, site report, 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 licensee, holder of, or applicant
for an approval, certification, permit, site report, or license,
subject to this part, that they may be individually subject to NRC
enforcement action for violation of the provisions in 10 CFR 50.5.
Sec. 52.3 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 pursuant to
subpart C of this part.
Early site permit means a Commission approval, issued pursuant to
subpart A of this part, for a site or sites for one or more nuclear
power facilities.
Modular design means a nuclear power station that consists of two
or more essentially identical nuclear reactors (modules), where each
module
[[Page 40048]]
is a separate nuclear reactor capable of being safely operated
independent of the state 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 reactor that is used to test design
features, such as the testing required by Sec. 52.107(b)(2). The
prototype plant is similar to the first-of-a-kind or standard plant
design in all features and size, but may include additional safety
features to protect the public, the plant staff, and the plant itself
from the possible consequences of accidents during the testing period.
Standard design means a design which is sufficiently detailed and
complete to support certification in accordance with subpart B 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 certification, design certification, or
certification means a Commission approval, issued pursuant to Subpart B
of this part, of a standard design for a nuclear power facility. A
design so approved 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.
Sec. 52.5 Applicability of 10 CFR part 50 provisions.
Unless otherwise specifically provided for in this part, Sec. Sec.
50.3, 50.4, 50.5, 50.7, 50.9, 50.10, 50.11, 50.12, 50.13, 50.50, 50.51,
50.52, 50.53, 50.54, 50.55, 50.55a, 50.56, 50.57, 50.58, 50.59, 50.70,
50.71, 50.72, 50.73, 50.74, 50.75, 50.78, 50.80, 50.81, 50.82, 50.90,
50.91, 50.92, 50.100, 50.101, 50.102, 50.103 and 50.109 of this chapter
apply to a licensee, holder of, or applicant for an approval,
certification, permit, site report, or license issued under this part.
A licensee, holder of, or applicant for an approval, certification,
permit, site report, or license issued under this part shall comply
with all requirements in these provisions that are otherwise applicable
to applicants or licensees under part 50 of this chapter.
Sec. 52.8 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 3150-0151.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 52.15, 52.17, 52.29, 52.35, 52.39,
52.45, 52.105, 52.107, 52.111, 52.119, 52.121, 52.123, 52.127, 52.205,
52.207, 52.209, 52.211, 52.215, 52.223, 52.225, 52.229, 52.231, 52.243,
and Appendices A, B, and C.
Subpart A--Early Site Permits
Sec. 52.11 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of early site permits for approval of a site or
sites for one or more nuclear power facilities separate from the filing
of an application for a construction permit, combined license, or
duplicate design license for such a facility.
Sec. 52.13 Relationship to Subpart F of 10 CFR Part 2 and Subpart B
of this part.
The procedures of this subpart do not replace those set out in
subpart F of 10 CFR part 2 or subpart B of this part. Subpart F of 10
CFR part 2 applies only when an early partial decision of site
suitability issues is sought in connection with an application for a
permit to construct certain power facilities. Subpart B of this part
applies only when NRC staff review of one or more site suitability
issues is sought separately from and prior to the submittal of an
application for a construction permit, combined license, or duplicate
design license. A Staff Site Report issued under subpart B of this part
in no way affects the authority of the Commission or the presiding
officer in any proceeding under Subparts F or G of 10 CFR part 2. This
subpart A applies when any person who may apply for a construction
permit under 10 CFR part 50 or for a combined license under part 52
seeks an early site permit from the Commission separately from an
application for a construction permit or a combined license for a
facility.
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 of Nuclear
Reactor Regulation. 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 or sites for which a permit is sought.
(b) The application must comply with the filing requirements of 10
CFR 50.30 (a), (b), and (f) as they would apply to an application for a
construction permit. The following portions of 10 CFR 50.4, which is
referenced by 10 CFR 50.30(a)(1), are applicable: Paragraphs (a), (b)
(1) (2) (3), (c), (d), and (e).
(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. 52.17 Contents of applications.
(a)(1) The application must contain the information required by 10
CFR 50.33(a) through (d), the information required by 10 CFR 50.34
(a)(12) and (b)(10), and to the extent approval of emergency plans is
sought under paragraph (b)(2)(ii) of this section, the information
required by Sec. 50.33 (g) and (j), and Sec. 50.34 (b)(6)(v) of this
chapter. The application must also contain 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 the site under the radiological
consequence evaluation factors identified in Sec. 50.34(a)(1) of this
chapter. Site characteristics must comply with part 100 of this
chapter. In addition, the application should describe 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 boundaries of the site;
(iii) The proposed general location of each facility on the site;
(iv) The anticipated maximum levels of radiological and thermal
effluents each facility will produce;
(v) The type of cooling systems, intakes, and outflows that may be
associated with each facility;
(vi) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site;
(vii) The location and description of any nearby industrial,
military, or transportation facilities and routes; and
(viii) The existing and projected future population profile of the
area surrounding the site.
(2) A complete environmental report as required by 10 CFR 51.45 and
51.50 must be included in the application, provided, however, that such
environmental report must focus on the
[[Page 40049]]
environmental effects of construction and operation of a reactor, or
reactors, which have characteristics that fall within the postulated
site parameters, and provided further that the report need not include
an assessment of the benefits (for example, need for power) of the
proposed action or an evaluation of alternative energy sources, but
must include an evaluation of alternative sites to determine whether
there is any obviously superior alternative to the site proposed.
(b)(1) The application must identify physical characteristics
unique to 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.
(2) The application may also either:
(i) Propose major features of the emergency plans, such as the
exact sizes of the emergency planning zones, that can be reviewed and
approved by NRC in consultation with the Federal Emergency Management
Agency (FEMA) 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 FEMA, in accord with the
applicable provisions of 10 CFR 50.47.
(3) Under paragraphs (b)(1) and (b)(2)(i) of this section, the
application must include a description of contacts and arrangements
made with local, state, and Federal governmental agencies with
emergency planning responsibilities.
(i) 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:
(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) The application must contain any certifications that have been
obtained. If these certifications cannot be obtained, the application
must contain information, including a utility plan, sufficient to show
that the proposed plans nonetheless provide reasonable assurance that
adequate protective measures can and will be taken, in the event of a
radiological emergency at the site.
(c) If the applicant wishes to be able to perform, after grant of
the early site permit, the activities at the site allowed by 10 CFR
50.10(e)(1) without first obtaining the separate authorization required
by that section, the applicant shall propose, in the early site permit,
a plan for redress of the site in the event that the activities are
performed and the site permit expires before it is referenced in an
application for a construction permit or a combined license issued
under Subpart G of this part. The application must demonstrate that
there is reasonable assurance that redress carried out under the plan
will achieve an environmentally stable and aesthetically acceptable
site suitable for whatever non-nuclear use may conform with local
zoning laws.
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 as they apply to applications for construction
permits for nuclear power plants. 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, provided, however, that the draft and final environmental
impact statements prepared by the Commission focus on the environmental
effects of construction and operation of a reactor, or reactors, which
have characteristics that fall within the postulated site parameters,
and provided further that the statements need not include an assessment
of the benefits (for example, need for power) of the proposed action or
an evaluation of alternative energy sources, but must include an
evaluation of alternative sites to determine whether there is any
obviously superior alternative to the site proposed. The Commission
shall determine, after consultation with FEMA, 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, whether
any major features of emergency plans submitted by the applicant under
Sec. 52.17(b)(2)(i) are acceptable, 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. 52.19 Applicability of NRC requirements.
(a) An applicant shall comply with all requirements in 10 CFR
Chapter I applicable to applicants for construction permits and limited
work authorizations under 10 CFR 50.10.
(b) A holder of an early site permit shall comply with all
requirements in 10 CFR Chapter I applicable to holders of construction
permits and limited work authorizations under 10 CFR 50.10.
Sec. 52.21 Hearings.
An early site permit is a partial construction permit and is
therefore subject to all procedural requirements in 10 CFR Part 2 which
are applicable to construction permits, including the requirements for
docketing in 10 CFR 2.101(a)(1)-(4), and the requirements for issuance
of a notice of hearing in 10 CFR 2.104(a), (b)(1)(iv) and (v), (b)(2)
to the extent it runs parallel to Sec. 2.104(b)(1)(iv) and (v), and
(b)(3). However, 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 the proposed
action or an evaluation of alternative energy sources. In the hearing,
the presiding officer shall also determine whether, taking into
consideration the site criteria contained in 10 CFR Part 100, a
reactor, or reactors, having characteristics that fall within the
parameters for the site can be constructed and operated without undue
risk to the health and safety of the public. All hearings conducted on
applications for early site permits filed under this part are governed
by the procedures contained in subpart G of 10 CFR part 2.
Sec. 52.23 Referral to the ACRS.
The Commission shall refer a copy of the application to the
Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report
on those portions of the application which concern safety.
Sec. 52.24 Issuance of early site permit.
After conducting a hearing under Sec. 52.21 of this subpart and
receiving the report to be submitted by the Advisory Committee on
Reactor Safeguards under Sec. 52.23 of this subpart, and upon
determining that an application for an early site permit meets the
applicable standards and requirements of the Atomic Energy Act and the
Commission's regulations, and that notifications, if any, to other
agencies or bodies have been duly made, the Commission shall issue an
early site permit, in the form the Commission deems appropriate and
necessary. The early site permit shall specify the site parameters and
the terms and conditions of the early site permit.
[[Page 40050]]
Sec. 52.25 Extent of activities permitted.
(a) If an early site permit contains a site redress plan, the
holder of the permit, or the applicant for a construction permit or a
combined license who references the permit, may perform the activities
at the site allowed by 10 CFR 50.10(e)(1) without first obtaining the
separate authorization required by that section, if the final
environmental impact statement prepared for the permit has concluded
that the activities will not result in any significant adverse
environmental impact which cannot be redressed.
(b) If the activities permitted by paragraph (a) of this section
are performed at any site for which an early site permit has been
granted, and the site is not referenced in an application for a
construction permit or a combined license issued under subpart G of
this part while the permit remains valid, then the early site permit
must remain 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 10 CFR 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. 52.27 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
ten nor more than twenty years from the date of issuance.
(b)(1) 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 either before the date of expiration of the early site
permit, or, if a timely application for renewal of the permit has been
filed, before the Commission has determined whether to renew the
permit.
(2) An early site permit also continues to be valid beyond the date
of expiration in any proceeding on an operating license application
which is based on a construction permit that references the early site
permit, and in any hearing held under 10 CFR 52.231 before operation
begins under a combined license which references the early site 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.
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 twelve nor more than thirty-six months prior to
the expiration date, 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 2.714. If a hearing is
granted, notice of the hearing will be published in accordance with 10
CFR 2.703.
(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 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 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. 52.31.
Sec. 52.31 Criteria for renewal.
(a) The Commission shall grant the renewal if the Commission
determines that the site complies with:
(1) The Atomic Energy Act and the Commission's regulations and
orders applicable and in effect at the time the site permit was
originally issued;
(2) Any new requirements the Commission may wish to impose after a
determination that 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
(3) The direct and indirect costs of implementation of those
requirements are justified in view of this increased protection.
(b) A denial of renewal on this basis 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 ten
nor more than twenty years.
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 of Nuclear Reactor Regulation
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 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 shall terminate the permit.
Sec. 52.37 Reporting of defects and noncompliance; revocation,
suspension, modification of permits for cause.
For purposes of 10 CFR part 21 and 10 CFR 50.100, an early site
permit is a construction permit.
Sec. 52.39 Finality of early site permit determinations.
(a)(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, terms or
conditions of the early site permit, including emergency planning
requirements, on the early site permit or the site for which it was
issued, unless the Commission determines that a modification is
necessary either 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, or to assure adequate protection of the
public health and safety or the common defense and security.
(2) In making the findings required for issuance of a construction
permit, operating license, combined license, or duplicate design
license, or the findings required by Sec. 52.231 of this part, if the
application for the construction permit, operating license, combined
license, or duplicate design license references an early site permit,
the Commission shall treat as resolved those matters resolved in the
proceeding on the application for issuance or renewal of the early site
permit (with the exception of the
[[Page 40051]]
matters in paragraph (b) of this section), unless a contention is
admitted that a nuclear reactor does not fit within one or more of the
site parameters in the early site permit, or a petition is filed which
alleges either that the site does not conform to the site
characteristics in the early site permit, or that the terms and
conditions of the early site permit should be modified.
(i) A contention that a nuclear reactor does not fit within one or
more of the site parameters included in the site permit may be
litigated in the same manner as other issues material to the
proceeding.
(ii) A petition which alleges that the site does not conform to the
site characteristics in the early site permit must include, or clearly
reference, official NRC documents, documents prepared by or for the
permit holder, or evidence admissible in a proceeding under subpart G
of part 2 of this chapter, which show, prima facie, that the site does
not conform to the site characteristics. The permit holder and NRC
staff may file answers to the petition within the time specified in 10
CFR 2.730 for answers to motions by parties and staff. If the
Commission, in its judgment, decides, on the basis of the petitions and
any answers thereto, that the petition meets the requirements of this
paragraph, that the issues are not exempt from adjudication under 5
U.S.C. 554(a)(3), that genuine issues of material fact are raised, and
that settlement or other informal resolution of the issues is not
possible, then the genuine issues of material fact raised by the
petition must be resolved in accordance with the provisions in 5 U.S.C.
554, 556, and 557 which are applicable to determining application for
initial licenses.
(iii) A petition which alleges that the terms and conditions of the
early site permit should be modified will be processed in accordance
with 10 CFR 2.206. Before construction commences, the Commission shall
consider the petition and determine whether any immediate action is
required. If the petition is granted, then 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.
(iv) Prior to construction, the Commission shall find that the
terms and conditions of the early site permit have been met.
(b) An applicant for a construction permit, operating license,
duplicate design license, or combined license who has filed an
application referencing an early site permit issued under this subpart
shall update and correct the information that was provided under Sec.
52.17(b), and discuss whether the new information materially changes
the bases for compliance with the applicable requirements. New
information which materially changes the bases for the Commission's
determination on the matters in Sec. 52.17(b) must be subject to
litigation during the construction permit, operating license, duplicate
design license, or combined license proceeding in the same manner as
other issues material to those proceedings.
(c) An applicant for a construction permit, operating license,
duplicate design license, or combined license who has filed an
application referencing an early site permit issued under this subpart
may include in the application a request for a variance from one or
more elements of the permit. 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. Issuance of the variance must be subject to
litigation during the construction permit, operating license, duplicate
design license, or combined license proceeding in the same manner as
other issues material to those proceedings.
Subpart B--Early Site Reviews
Sec. 52.41 Scope of subpart.
This subpart sets out procedures for the filing, staff review, and
referral to the Advisory Committee on Reactor Safeguards (ACRS) of
requests for early review of one or more site suitability issues
relating to the construction and operation of certain utilization
facilities separately from and prior to the submittal of applications
for construction permits, combined licenses, or duplicate design
licenses for the facilities. The subpart also sets out procedures for
the preparation and issuance of Staff Site Reports and for their
incorporation by reference in applications for the construction and
operation of certain utilization facilities. The utilization facilities
are those which are subject to Sec. 51.20(b) of this chapter and are
of the type specified in Sec. 50.21(b)(2) or (3) or Sec. 50.22 of
this chapter or are testing facilities. This subpart does not apply to
proceedings conducted pursuant to subpart F of part 2 of this chapter.
Sec. 52.43 Filing and contents of applications.
(a) Any person may submit information regarding one or more site
suitability issues to the Commission's Staff for its review separately
from and prior to an application for a construction permit, a combined
license, or a duplicate design license for a facility. The submittal
must consist of the portion of the information required of applicants
for construction permits by Sec. Sec. 50.33(a) through (c) and (e) of
this chapter, and, insofar as it relates to the issue(s) of site
suitability for which early review is sought, by Sec. Sec. 50.34(a)(1)
and 50.30(f) of this chapter. Information with respect to operation of
the facility at the projected initial power level need not be supplied.
(b) The submittal for early review of site suitability issue(s)
must be made in the same manner and in the same number of copies as
provided in Sec. Sec. 50.4 and 50.30 of this chapter for license
applications. The submittal must include sufficient information
concerning the range of postulated facility design and operation
parameters to enable the NRC staff to perform the requested review of
site suitability issues. The submittal must contain suggested
conclusions on the issues of site suitability submitted for review and
must be accompanied by a statement of the bases or the reasons for
those conclusions. The submittal must also list, to the extent
possible, any long-range objectives for ultimate development of the
site, state whether any site selection process was used in preparing
the submittal, describe any site selection process used, and explain
what consideration, if any, was given to alternative sites.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.45 Notice of application.
The NRC staff shall publish a notice of docketing of the submittal
in the Federal Register, and shall send a copy of the notice of
docketing to the Governor of the State, local government bodies
(county, municipality, or other political subdivision), and affected,
Federally-recognized Indian Tribes. This notice must identify the
location of the site, briefly describe the site suitability issue(s)
under review, and invite comments from Federal, State, Tribal, and
local agencies and interested persons within 120 days of publication or
such other time as may be specified, for consideration by the staff in
connection with the initiation or outcome of the review and, if
appropriate, by the ACRS in connection with the outcome of their
review. The person requesting the review shall serve a copy of the
submittal on the Governor or other appropriate official of the State in
which the site is located, and on the chief executive of the
municipality in
[[Page 40052]]
which the site is located or, if the site is not located in a
municipality, on the chief executive of the county.
Sec. 52.46 Referral to the ACRS.
The portion of the submittal containing information requested of
applicants for construction permits by Sec. Sec. 50.33 (a) through (c)
and (e) and 50.34(a)(1) of this chapter will be referred to the ACRS
for a review and report. There will be no referral to the ACRS unless
early review of the site safety issues under Sec. 50.34(a)(1) is
requested.
Sec. 52.47 Issuance of site report.
(a) Upon completion of review by the NRC staff and, if appropriate,
by the ACRS of a submittal under this subpart, the NRC staff shall
prepare a Staff Site Report which identifies the location of the site,
states the site suitability issues reviewed, explains the nature and
scope of the review, states the conclusions of the staff regarding the
issues reviewed and, states the reasons for those conclusions. Upon
issuance of an NRC Staff Site Report, the NRC staff shall publish a
notice of the availability of the report in the Federal Register and
shall make available a copy of the report at the NRC Web site, http://www.nrc.gov. The NRC staff shall also send a copy of the report to the
Governor of the State, local government bodies (county, municipality,
or other political subdivision), and affected, Federally-recognized
Indian Tribes.
(b) Any Staff Site Report prepared and issued in accordance with
this subpart may be incorporated by reference, as appropriate, in an
application for a construction permit, a combined license, or a
duplicate design license for a utilization facility which is subject to
Sec. 51.20(b) of this chapter and is of the type specific in Sec.
50.21(b)(2) or (3) or Sec. 50.22 of this chapter or is a testing
facility. The conclusions of the Staff Site Report will be reexamined
by the staff where five years or more have elapsed between the issuance
of the Staff Site Report and its incorporation by reference in an
application.
(c) Issuance of a Staff Site Report does not constitute a
commitment to issue a permit or license, to permit on-site work under
Sec. 50.10(e) of this chapter, or in any way affect the authority of
the Commission, Atomic Safety and Licensing Board Panel, and other
presiding officers in any proceeding under 10 CFR part 2 of this
chapter.
Sec. 52.49 Relationship to other subparts.
The NRC staff will not conduct more than one review of site
suitability issues with regard to a particular site prior to the full
construction permit, combined license, or duplicate design license
review required by subpart A of part 51 of this chapter. The NRC staff
may decline to prepare and issue a Staff Site Report in response to a
submittal under this subpart where it appears that--
(a) In cases where no review of the relative merits of the
submitted site and alternative sites under subpart A of part 51 of this
chapter is requested, there is a reasonable likelihood that further
staff review would identify one or more preferable alternative sites
and the staff review of one or more site suitability issues would lead
to an irreversible and irretrievable commitment of resources prior to
the submittal of the analysis of alternative sites in the Environmental
Report that would prejudice the later review and decision on
alternative sites under subpart F and/or G of part 2 and subpart A of
part 51 of this chapter; or
(b) In cases where, in the judgment of the staff, early review of
any site suitability issue or issues would not be in the public
interest, considering:
(1) The degree of likelihood that any early findings on those
issues would retain their validity in later reviews;
(2) The objections, if any, of cognizant state or local government
agencies to the conduct of an early review on those issues; and
(3) The possible effect on the public interest of having an early,
if not necessarily conclusive, resolution of those issues.
Subpart C--[Reserved]
Subpart D--Standard Design Certifications
Sec. 52.101 Scope of subpart.
This subpart sets forth the requirements and procedures applicable
to Commission issuance of rules granting standard design certification
for nuclear power facilities separate from the filing of an application
for a construction permit, duplicate design license, or combined
license for such a facility.
Sec. 52.103 Relationship to other subparts.
(a) Subpart H 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
I of this part governs licenses to construct and operate nuclear power
reactors of duplicate design at multiple sites. These subparts may be
used independently of the provisions in this subpart unless the
applicant also wishes to use a certified standard design approved under
this subpart.
(b) Subpart E of this part governs the NRC staff review and
approval of preliminary and final standard designs. An NRC staff
approval under subpart E of this part in no way affects the authority
of the Commission or the presiding officer in any proceeding under
subpart G of 10 CFR part 2.
Sec. 52.105 Filing of applications.
(a)(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 (a)(1) of this section or
utilizes simplified, inherent, passive, or other innovative means to
accomplish its safety functions.
(b) An application for certification may be filed notwithstanding
the fact that an application for a construction permit, a duplicate
design license, or a combined license for such a facility has not been
filed.
(c) The applicant must comply with the filing requirements of 10
CFR 50.30(a) and 50.30(b) as these requirements would apply to an
application for a nuclear power plant construction permit.
(d) 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. 52.107 Contents of applications.
(a) The requirements of this paragraph apply to all applications
for design certification.
(1) An application for design certification must contain:
(i) The technical information required of applicants for
construction permits and operating licenses by 10 CFR parts 20, 50 and
its appendices, and 10 CFR parts 73 and 100, and that is technically
relevant to the design and not site-specific;
(ii) Demonstration of compliance with any technically relevant
portions of the Three Mile Island requirements set forth in 10 CFR
50.34(f);
(iii) The site parameters postulated for the design, and an
analysis and evaluation of the design in terms of those site
parameters;
(iv) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority Generic
[[Page 40053]]
Safety Issues that are identified in the version of NUREG-0933 current
on the date six months prior to application and that are technically
relevant to the design;
(v) A design-specific probabilistic risk assessment;
(vi) Proposed inspections, tests, analyses, and acceptance criteria
(ITAAC) that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met, a plant that references the design is
built and will operate in accordance with the design certification, the
provisions of the Act, and the applicable Commission's rules and
regulations.
(vii) 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
final safety analysis and design-specific probabilistic risk assessment
required by paragraph (a)(1)(v) of this section;
(viii) Justification that compliance with the interface
requirements of paragraph (a)(1)(vii) of this section is verifiable
through inspection, testing (either in the plant or elsewhere), or
analysis. The method to be used for verification of interface
requirements must be included as part of the proposed inspections,
tests, analyses, and acceptance criteria required by paragraph
(a)(1)(vi) of this section; and
(ix) A representative conceptual design for those portions of the
plant for which the application does not seek certification, to aid the
NRC staff in its review of the final safety analysis and probabilistic
risk assessment required by paragraph (a)(1)(v) of this section, and to
permit assessment of the adequacy of the interface requirements in
paragraph (a)(1)(vii) of this section.
(2) 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, and procurement
specifications and construction and installation specifications by an
applicant. The Commission will require, prior to 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.
(3) The NRC staff shall advise the applicant on whether any
technical information beyond that required by this section must be
submitted.
(b) This paragraph applies, according to its provisions, to
particular applications:
(1) The application for certification of a 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, 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) Certification of a standard design that 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 will be granted only if--
(i)(A) The performance of each safety feature of the design has
been demonstrated through either analysis, appropriate test programs,
experience, or a combination thereof;
(B) Interdependent effects among the safety features of the design
have been found acceptable by analysis, appropriate test programs,
experience, or a combination thereof;
(C) Sufficient data exist on the safety features of the design to
assess the analytical tools used for safety analyses over a sufficient
range of normal operating conditions, transient conditions, and
specified accident sequences, including equilibrium core conditions;
and
(D) The scope of the design is complete except for site-specific
elements such as the service water intake structure and the ultimate
heat sink; or
(ii) There has been acceptable testing of a prototype plant over a
sufficient range of normal operating conditions, transient conditions,
and specified accident sequences, including equilibrium core
conditions. If the criterion in paragraph (b)(2)(i)(D) of this section
is not met, the testing of the prototype plant must demonstrate that
the non-certified portion of the plant cannot significantly affect the
safe operation of the plant.
(3) An application seeking certification of a modular design must
describe the various options for the configuration of the plant and
site, including variations in, or sharing of, common systems, interface
requirements, and system interactions. The final safety analysis and
the probabilistic risk assessment should also account for differences
among the various options, including any restrictions which will be
necessary during the construction and startup of a given module to
ensure the safe operation of any module already operating.
Sec. 52.109 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 as they apply to applications
for construction permits and operating licenses for nuclear power
plants that are technically relevant to the design proposed for the
facility.
Sec. 52.111 Applicability of NRC requirements.
An applicant shall comply with all requirements in 10 CFR Chapter I
applicable to applicants for construction permits and operating
licenses under 10 CFR Chapter I.
Sec. 52.113 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 this
subpart and shall specify the procedures to be used for the rulemaking.
(b) The rulemaking procedures must provide for notice and comment
and an opportunity for an informal hearing before an Atomic Safety and
Licensing Board. The procedures for the informal hearing must include
the opportunity for written presentations made under oath or
affirmation and for oral presentations and questioning if the Board
finds them either necessary for the creation of an adequate record or
the most expeditious way to resolve controversies. Ordinarily, the
questioning in the informal hearing will be done by members of the
Board, using either the Board's questions or questions submitted to the
Board by the parties. The Board may also request authority from the
Commission to use additional procedures, such as direct and cross
examination by the parties, or may request that the Commission convene
a formal hearing under subpart G of 10
[[Page 40054]]
CFR part 2 on specific and substantial disputes of fact, necessary for
the Commission's decision, that cannot be resolved with sufficient
accuracy except in a formal hearing. The NRC staff will be a party in
the hearing.
(c) The decision in such a hearing will be based only on
information on which all parties have had an opportunity to comment,
either in response to the notice of proposed rulemaking or in the
informal hearing.
(d) Proprietary information will be protected in the same manner
and to the same extent as proprietary information submitted in
connection with applications for construction permits and operating
licenses under 10 CFR part 50. However, the design certification is
published in 10 CFR Chapter I. The provisions of 10 CFR 2.790 do not
limit the protection provided under this paragraph.
Sec. 52.115 Referral to the ACRS.
The Commission shall refer a copy of the application to the
Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report
on those portions of the application which concern safety.
Sec. 52.117 Issuance of standard design certification.
After conducting a rulemaking proceeding under Sec. 52.113 on an
application for a standard design certification and receiving the
report to be submitted by the Advisory Committee on Reactor Safeguards
under Sec. 52.115, and upon determining that the application meets the
applicable standards and requirements of the Atomic Energy Act and the
Commission's regulations, the Commission shall issue a standard design
certification in the form of a rule for the design which is the subject
of the application.
Sec. 52.119 Duration of certification.
(a) Except as provided in paragraph (b) of this section, a standard
design certification issued under this subpart is valid for fifteen
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. 52.231 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. 52.121 Application for renewal.
(a) Not less than twelve nor more than thirty-six months before the
expiration of the initial fifteen-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, prior to 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 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. 52.119.
(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. 52.123.
Sec. 52.123 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. The Commission may impose other requirements
after it determines that 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 that the direct
and indirect costs of implementing those requirements are justified in
view of this increased protection. 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.
(b) 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.
Sec. 52.125 Duration of renewal.
Each renewal of certification for a standard design will be for not
less than ten nor more than fifteen years.
Sec. 52.127 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. 52.119 or
52.125, 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; or
(iii) Reduces unnecessary regulatory burden and maintains
protection to public health and safety and the common defense and
security.
(2) The rulemaking procedures must provide for notice and comment
and an opportunity for the party which applied for the certification to
request an informal hearing which uses the procedures described in
Sec. 52.113 of this subpart.
(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
[[Page 40055]]
the certified design, except those to which the modification has been
rendered technically irrelevant by action taken under paragraphs (a)(3)
or (b)(1) of this section.
(4) While a design certification rule is in effect under Sec.
52.119 or Sec. 52.125, 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 50.12(a) are
present, 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. In addition to the factors listed in 10 CFR
50.12(a), the Commission shall consider whether the special
circumstances which 10 CFR 50.12(a)(2) 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 2.758, in making the findings
required for issuance of a combined license or operating license, or
for any hearing under Sec. 52.231, 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 standard design
certification rule may request an exemption from one or more elements
of the design certification 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). In addition to the factors listed in
Sec. 50.12(a), the Commission shall consider whether the special
circumstances that Sec. 50.12(a)(2) 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 must be subject to litigation in the same
manner as other issues in the operating license or combined license
hearing.
(2) Subject to Sec. 50.59, a licensee who references a standard
design certification rule may make changes to the design of the nuclear
power facility, without prior Commission approval, unless the proposed
change involves a change to the design as described in the rule
certifying the design. The licensee shall maintain records of all
changes to 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, prior to granting a construction
permit, combined license, or operating license which references a
standard design certification rule, that information normally contained
in certain procurement specifications and construction and installation
specifications be completed and available for audit if such information
is necessary for the Commission to make its safety 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 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 Filing of applications.
(a) Any person may submit a proposed preliminary or final 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 preliminary or final design for the entire reactor facility
or the preliminary or final design of major portions thereof.
(b) The submittal for review of the standard design must be made in
the same manner and in the same number of copies as provided in
Sec. Sec. 50.4 and 50.30 of this chapter for license applications.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.135 Contents of applications.
The submittal for review of the standard design must include the
information described in Sec. Sec. 50.33 (a) through (d) of this
chapter and the applicable technical information required by Sec.
50.34 of this chapter, as appropriate (other than that required by 10
CFR 50.34(a)(6) and (10), 50.34(b)(1), (6)(i), (ii), (iv), and (v) and
50.34(b)(7) and (8)), 10 CFR 50.34a, and 52.107(a)(1)(i) through (v),
and (vii). The submittal must also include a description, analysis, and
evaluation of the interfaces between the submitted design and the
balance of the nuclear power plant. With respect to the requirements of
Sec. 50.34(a)(1) of this chapter, the submittal for review of a
standard design must include the site parameters postulated for the
design, and an analysis and evaluation of the design in terms of the
postulated site parameters. The information submitted under Sec.
50.34(a)(7) of this chapter, must be limited to the quality assurance
program to be applied to the design, procurement, and fabrication of
the structures, systems, and components for which design review has
been requested. The information submitted under Sec. 50.34(a)(9) of
this chapter must be limited to the qualifications of the person
submitting the standard design to design the reactor or major portion
thereof. The submittal must also include information pertaining to
design features that affect plans for coping with emergencies in the
operation of the reactor or a major portion thereof.
Sec. 52.137 Referral to the ACRS.
Once the NRC staff has initiated a technical review of a submittal
under this subpart, the submittal will be referred to the Advisory
Committee on Reactor Safeguards (ACRS) for a review and report.
Sec. 52.139 Staff approval of design.
(a) Upon completion of their review of a submittal under this
subpart, the NRC staff shall publish a determination in the Federal
Register as to whether or not the preliminary or final design is
acceptable, subject to appropriate conditions, and make an analysis of
the design in the form of a report available at the NRC Web site,
http://www.nrc.gov.
(b) A standard design approval issued under this subpart is valid
for 15 years from the date of issuance. 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 which
references the design approval and is docketed before the date of
expiration of the design approval.
Sec. 52.141 Finality of the design approval.
(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 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
[[Page 40056]]
license, or in any way affect the authority of the Commission, Atomic
Safety and Licensing Board Panel, and other presiding officers in any
proceeding under part 2 of this chapter.
Sec. 52.143 Information requests.
Information requests to the approval holder regarding an approved
design must be evaluated prior to 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 such 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 prior to issuance of the request.
Subpart F--[Reserved]
Subpart G--Combined Licenses
Sec. 52.201 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of combined licenses for nuclear power facilities.
Sec. 52.203 Relationship to other subparts.
(a) An application for a combined license under this subpart may,
but need not, reference a standard design certification or standard
design approval issued under Subparts D or E of this part, or an early
site permit or site report issued under subparts A or B 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 such design, the Commission will entertain an application for
a combined license that references a standard design certification
issued under subpart D of this part only if the entity that sponsored
and obtained the certification supplies the certified design for the
applicant's use.
(b) The Commission will require, prior to 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 such information is necessary for the Commission
to make its safety determinations, including the determination that the
application is consistent with the certification information.
Sec. 52.205 Filing of applications.
(a) Any person except one excluded by 10 CFR 50.38 may file an
application for a combined license for a nuclear power facility with
the Director of Nuclear Reactor Regulation. The applicant shall comply
with the filing requirements of 10 CFR 50.30 (a) and (b), as they would
apply to an application for a nuclear power plant construction permit.
(b) The fees associated with the filing and review of the
application are set forth in 10 CFR Part 170.
Sec. 52.207 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33, as that section would apply to applicants for construction
permits and operating licenses, and 10 CFR 50.33a, as that section
would apply to an applicant for a nuclear power plant construction
permit. In particular, the applicant shall comply with the requirement
of 10 CFR 50.33a(b) regarding the submission of antitrust information.
Sec. 52.209 Contents of applications; training and qualification of
nuclear power plant personnel.
The application must describe the training program required by
Sec. 50.120 of this chapter. The training program described in the
application must be established, implemented and maintained no later
than eighteen (18) months prior to the scheduled date for initial
loading of fuel, as provided for in Sec. 52.231(a).
Sec. 52.211 Contents of applications; technical information.
(a) Early site permit.
(1) If the application references an early site permit, the
application need not contain information or analyses submitted to the
Commission in connection with the early site permit, but must contain,
in addition to the information and analyses otherwise required:
(i) Information sufficient to demonstrate that the design of the
facility falls within the site parameters specified in the early site
permit;
(ii) Information necessary to resolve any other significant
environmental issue with respect to the site not considered in any
previous proceeding on the site or the design; and
(iii) A demonstration that all terms and conditions of the early
site permit have been satisfied.
(2) If the application does not reference an early site permit, the
applicant must comply with the requirements of 10 CFR 50.30(f) by
including with the application an environmental report prepared in
accordance with the provisions of Subpart A of 10 CFR part 51.
(3) If the application does not reference an early site permit
which contains a site redress plan as described in Sec. 52.17(c), and
if the applicant wishes to be able to perform the activities at the
site allowed by 10 CFR 50.10(e)(1), then the application must contain
the information required by Sec. 52.17(c).
(b) The application must contain the technically relevant
information required of applicants for an operating license by 10 CFR
50.34 in a final safety analysis report.
(1) If the application does not reference a certified design, the
application must comply with the requirements of Sec. 52.107(a)(2) for
level of design information, and must contain the technical information
required by Sec. Sec. 52.107(a)(1) (i), (ii), (iv), and (3); Sec.
52.107(b)(2); and, if the design is modular, Sec. 52.107(b)(3).
(2) If the application does not reference a certified design, the
application must contain a plant-specific probabilistic risk assessment
(PRA).
(3) If a prototype plant is used to comply with the requirements of
Sec. 52.107(b)(2), then the NRC may impose additional licensing
requirements on siting, safety features, or operational conditions for
the prototype plant to protect the public, the plant staff, and the
plant itself from the possible consequences of failures during the
testing period.
(4) An application referencing a certified design must include in
the final safety analysis report the information approved for
incorporation by reference in a design certification rule; describe
those portions of the design that are not described in the certified
design, such as the service water intake structure and the ultimate
heat sink; demonstrate compliance with the interface requirements
established for the design under Sec. 52.107(a)(1); and have available
for audit procurement specifications and construction and installation
specifications in accordance with Sec. Sec. 52.107(a)(2) and
52.203(b).
(5) An application referencing a certified design must include a
plant-specific PRA that uses the design-specific PRA and is updated to
account for site-specific design information and any design changes.
(c) The application must include the proposed inspections, tests
and analyses, including those applicable to emergency planning, which
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
[[Page 40057]]
performed and the acceptance criteria met, the facility has been
constructed and will operate in conformity with the combined license,
the provisions of the Atomic Energy Act, and the NRC's regulations.
(1) If the application references a certified standard design, the
inspections, tests, analyses, and acceptance criteria contained in the
certified design must apply to those portions of the facility design
that are covered by the design certification.
(2) 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. 52.217 must
indicate that the application includes this notification.
(d) The application must contain emergency plans that provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency at the site.
(1) If the application references an early site permit, the
application may incorporate by reference emergency plans, or major
features of emergency plans, approved in connection with the issuance
of the permit. If the application incorporates by reference an
emergency plan or major features of such a plan, the application must
include information that updates and corrects the information
previously provided under Sec. 52.17(b), and discuss whether the new
information materially changes the bases for compliance with the
applicable requirements. New information that materially changes the
bases for the Commission's determination on the matters in Sec.
52.17(b) must be subject to litigation during the combined license
proceeding in the same manner as other issues material to those
proceedings.
(2)(i) If the application does not reference an early site permit,
or if no emergency plans were approved in connection with the issuance
of the permit, the applicant shall make good faith efforts to obtain
certifications from the local and State governmental agencies with
emergency planning responsibilities 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) The application must contain any certifications that have been
obtained. If these certifications cannot be obtained, the application
must contain information, including a utility plan, sufficient to show
that the proposed plans nonetheless provide reasonable assurance that
adequate protective measures can and will be taken in the event of a
radiological emergency at the site.
Sec. 52.213 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, 55, 73, and 100 as
they apply to applications for construction permits and operating
licenses for nuclear power plants, and as those standards are
technically relevant to the design proposed for the facility.
Sec. 52.215 Applicability of NRC requirements.
(a) An applicant shall comply with all requirements in 10 CFR
Chapter I applicable to applicants for construction permits and limited
work authorizations under 10 CFR 50.10.
(b) After a combined license is issued but before the Commission
has authorized operation under Sec. 52.231, the licensee shall comply
with all requirements in this chapter of Title 10 applicable to holders
of construction permits for nuclear power reactors.
(c) After the Commission has authorized operation under Sec.
52.231, the licensee shall comply with all requirements in 10 CFR
Chapter I applicable to holders of operating licenses for nuclear power
reactors. Any limitations contained in 10 CFR part 50 regarding
applicability of the provisions to certain classes of facilities
continue to apply. Provisions of 10 CFR part 50 that do not apply to
holders of combined licenses issued under this subpart include
Sec. Sec. 50.55(a), (b) and (d), and 50.58(a).
Sec. 52.217 Administrative review of applications.
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) and issuance of a notice of
hearing (Sec. 2.104). 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. 52.219 Referral to the ACRS.
The Commission shall refer a copy of the application to the
Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report
on those portions of the application that concern safety and shall
apply the criteria set forth in Sec. 52.213, in accordance with the
finality provisions of this part.
Sec. 52.221 Environmental review.
If the application references an early site permit and/or a design
certification rule, the environmental review must focus on whether the
design of the facility falls within the site parameters specified in
the early site permit and any other significant environmental issue not
considered in any previous proceeding on the site or the design. If the
application does not reference an early site permit, the environmental
review procedures set out in 10 CFR part 51 with respect to a
construction permit must be followed, including the issuance of a final
environmental impact statement, but excluding the issuance of a
supplement under 10 CFR 51.95(a).
Sec. 52.223 Authorization to conduct site activities.
(a)(1) If the application references an early site permit that
contains a site redress plan as described in Sec. 52.17(c) the
applicant is authorized by Sec. 52.25 to perform the site preparation
activities described in 10 CFR 50.10(e)(1).
(2) If the application does not reference an early site permit
which contains a redress plan, the applicant may not perform the site
preparation activities allowed by 10 CFR 50.10(e)(1) without first
submitting a site redress plan in accord with Sec. 52.211(a)(3) and
obtaining the separate authorization required by 10 CFR 50.10(e)(1).
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)(2) and has determined that the site redress
plan meets the criteria in Sec. 52.17(c).
(3) Authorization to conduct the activities described in 10 CFR
50.10(e)(3)(i) may be granted only after the presiding officer in the
combined license proceeding makes the additional finding required by 10
CFR 50.10(e)(3)(ii).
(b) If, after an applicant for a combined license has performed the
activities permitted by paragraph (a) of this section, the application
for the license is withdrawn or denied, and the early site permit
referenced by the application expires, then the applicant shall redress
the site in accord with the
[[Page 40058]]
terms of the site redress plan. If a use not envisaged in the redress
plan is found for the site or parts thereof before redress is complete,
the applicant shall carry out the redress plan to the greatest extent
possible consistent with the alternate use.
Sec. 52.225 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, under 10 CFR 50.12, for an exemption from one or more of the
Commission's regulations, including any part of a design certification
rule. The Commission may grant such a request if it determines that the
exemption will comply with the requirements of 10 CFR 50.12(a) or
52.127(b)(1) if the exemption includes any part of the design
certification rule.
(b) An applicant for a combined license, or any amendment to a
combined license, who has filed an application referencing an early
site permit issued under this subpart may include in the application a
request for a variance from one or more elements of the permit. 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. Issuance of the
variance is subject to litigation during the combined license
proceeding in the same manner as other issues material to that
proceeding.
Sec. 52.227 Issuance of combined licenses.
(a)(1) The Commission shall issue a combined license for a nuclear
power facility upon finding that the applicable requirements of 10 CFR
50.40, 50.42, 50.43, 50.47, and 50.50 have been met, and that 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 rules and regulations.
(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
the combined license have been met. Such a finding will preclude any
required finding under Sec. 52.231(g) with respect to that ITAAC.
(b)(1) 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.
(2) Any modification to, addition to, or deletion from the terms of
a combined license, including 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 these
amendments.
(3) 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.
(c) If the combined license does not reference a certified design,
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. 50.54, Sec. 50.59, or Sec.
50.90).
(d) 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 involve changes to or non-compliance with
information within the scope of the referenced design certification
rule, in which case the applicable provisions of this section and/or
the design certification rule apply.
(e) A combined license is issued for a specified period not to
exceed 40 years from the date on which the Commission makes the finding
required under Sec. 52.231(g).
Sec. 52.229 Inspection during construction.
(a) Holders of combined licenses shall comply with the provisions
of 10 CFR 50.70 and 50.71.
(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 particular ITAAC has been
satisfied.
(c) The licensee shall notify the NRC that the inspections, tests,
or analyses in the ITAAC have been successfully completed and that the
corresponding acceptance criteria have been met.
(d) In the event that an activity is subject to an ITAAC and the
licensee has not demonstrated that the ITAAC has been satisfied, the
licensee may take corrective actions to successfully complete that
ITAAC, request an exemption from the ITAAC in accordance with the
applicable change process in the referenced design certification rule,
or request a license amendment under Sec. 52.227(b), as applicable.
(e) The NRC staff shall ensure that the required inspections,
tests, and analyses in the ITAAC are performed. At appropriate
intervals during construction, the NRC shall publish notices in the
Federal Register of the successful completion of inspections, tests,
and analyses.
Sec. 52.231 Operation under a combined license.
(a) Not less than one hundred and eighty days before the date
scheduled for initial loading of fuel into a plant by a licensee that
has been issued a combined license under Subpart G of this part, the
Commission shall publish notice of intended operation in the Federal
Register. That document 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 of
the ITAAC in the combined license, except for those ITAAC that the
Commission found were met under Sec. 52.227(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
[[Page 40059]]
protection of the public health and safety.
(c) After receiving a request for a hearing, the Commission
expeditiously shall either deny or grant the request. If the request is
granted, the Commission shall determine, after considering petitioners'
prima facie showing and any answers thereto, whether during a period of
interim operation, there will be reasonable assurance of adequate
protection of the public health and safety. If the Commission
determines that 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 therefor.
(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
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 accord 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) Prior to operation of the facility, the Commission shall find
that the acceptance criteria of the ITAAC in the combined license are
met, except for those ITAAC that the Commission found were met under
Sec. 52.227(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
design certification rule or combined license, constitute regulatory
requirements either for licensees or for renewal of the license; except
for specific ITAAC, which are the subject of a hearing under paragraph
(a) of this section, their expiration will occur upon final Commission
action in such 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. 52.227(c) or (d),
as applicable.
Subpart H--Manufacturing Licenses
Sec. 52.241 Scope of subpart.
(a) Section 101 of the Atomic Energy Act of 1954, as amended, and
Sec. 50.10 of this chapter require a Commission license to transfer or
receive in interstate commerce, manufacture, produce, transfer,
acquire, possess, use, import or export any production or utilization
facility. The regulations in 10 CFR part 50 require the issuance of a
construction permit by the Commission before commencement of
construction of a production or utilization facility, and the issuance
of an operating license before operation of the facility. The
provisions of 10 CFR part 50 relating to the facility licensing process
are, in general, predicated on the assumption that the facility will be
assembled and constructed on the site at which it is to be operated. In
those circumstances, both facility design and site-related issues can
be considered in the initial, construction permit stage of the
licensing process.
(b) Under the Atomic Energy Act, a license may be sought and issued
authorizing the manufacture of facilities but not their construction
and installation at the sites on which the facilities are to be
operated. Prior to the ``commencement of construction,'' as defined in
Sec. 50.10(c) of this chapter, of a facility (manufactured under such
a Commission license) on the site at which it is to operate--that is
preparation of the site and installation of the facility--a
construction permit, combined license, or duplicate plant license that,
among other things, reflects approval of the site on which the facility
is to be operated, must be issued by the Commission. This subpart sets
out the particular requirements and provisions applicable to situations
where nuclear power reactors to be manufactured under a Commission
license and subsequently installed at the site under a Commission
construction permit, combined license, or duplicate plant license, are
of the type described in Sec. 50.22 of this chapter.
Sec. 52.243 Relationship to other subparts.
(a) Referencing a manufacturing license. An application for a
construction permit, operating license or combined license to construct
a nuclear power plant which is to be manufactured under a manufacturing
license issued under this subpart need not contain the information or
analyses that have been previously approved by the Commission in
connection with the issuance of the manufacturing license. The
application must reference the manufacturing license, and provide
sufficient information to demonstrate that the site on which the
reactor(s) is to be located and operated fits within the postulated
site parameters specified in the manufacturing license.
(b) Amendment of manufacturing license to reflect final reactor
design. The holder of a manufacturing license issued under this subpart
shall submit to the Commission the final design of the nuclear power
reactor(s) covered by the license as soon as such design has been
completed. The submittal must be in the form of an application for
amendment of the manufacturing license.
(c) Application for construction permit or combined license
referencing a manufacturing license. An application for a permit to
construct a nuclear power reactor(s) or a combined license that is the
subject of an application for a manufacturing license pursuant to this
subpart need not contain information or analyses that have previously
been submitted to the Commission in connection with the application for
a manufacturing license. However, the application must comply with
Sec. Sec. 50.34(a) and 50.34a of this chapter, and provide sufficient
information to demonstrate that the site on which the reactor(s) is to
be operated falls within the postulated site parameters specified in
the relevant manufacturing license application.
(d) Approval of construction permit or combined license referencing
a manufacturing license. The Commission may issue a permit to construct
a nuclear power reactor(s) or a combined license that is the subject of
an application for a manufacturing license pursuant to this subpart if
the Commission--
(1) Finds that the site on which the reactor is to be operated
falls within the postulated site parameters specified in the relevant
application for a manufacturing license; and
(2) Makes the findings otherwise required by 10 CFR part 50. A
construction permit or combined license may not be issued until the
relevant manufacturing license has been issued.
(e) Approval of operating license referencing a manufacturing
license. An operating license for a nuclear power reactor(s) that has
been manufactured under a Commission license issued under this subpart
may be issued by the Commission under 10 CFR 50.57 and
[[Page 40060]]
subpart A of part 51 of this chapter except that the Commission shall
find, under 10 CFR 50.57(a)(1), that construction of the reactor(s) has
been substantially completed in conformity with both the manufacturing
license and the construction permit and the applications therefor, as
amended, and the provisions of the Act, and the rules and regulations
of the Commission. Notwithstanding the other provisions of this
paragraph, no application for an operating license for a nuclear power
reactor(s) that has been manufactured under a Commission license issued
under this subpart will be docketed until the application for an
amendment to the relevant manufacturing license required by Sec.
52.249 has been docketed.
(f) Prohibition against transport of nuclear power reactor
manufactured under this subpart. The prohibition in Sec. 50.10(c) of
this chapter against commencement of construction of a production or
utilization facility prior to issuance of a construction permit applies
to the transport of a nuclear power reactor(s) manufactured pursuant to
this subpart from the manufacturing facility to the site at which the
reactor(s) will be installed and operated. In addition, such nuclear
power reactor(s) may not be removed from the manufacturing site until
the final design of the reactor(s) has been approved by the Commission
in accordance with Sec. 52.249.
Sec. 52.245 Filing and contents of applications.
(a) An application for a manufacturing license under this subpart
must be submitted, as specified in Sec. 50.4 of this chapter and meet
all the requirements of Sec. Sec. 50.34(a)(1)-(9) and 50.34a(a) and
(b) of this chapter except that the preliminary safety analysis report
must be designated as a ``design report'' and any required information
or analyses relating to site matters must be predicated on postulated
site parameters which must be specified in the application. The
application must also include information pertaining to design features
of the proposed reactor(s) that affect plans for coping with
emergencies in the operation of the reactor(s).
(b) An applicant for a manufacturing license under this subpart
shall submit with the application an environmental report as required
of applicants for construction permits in accordance with subpart A of
part 51 of this chapter. However, the report must be directed at the
manufacture of the reactor(s) at the manufacturing site; and, in
general terms, at the construction and operation of the reactor(s) at a
hypothetical site or sites having characteristics that fall within the
postulated site parameters. The related draft and final environmental
impact statement prepared by the NRC staff will be similarly directed.
(c) The financial information submitted under Sec. 50.33(f) of
this chapter and Appendix C of part 50 must be directed at a
demonstration of the financial qualifications of the applicant for the
manufacturing license to carry out the manufacturing activity for which
the license is sought.
(d) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.247 Standards for review of application.
Applications filed under this subpart will be reviewed for
compliance with the standards set out in 10 CFR part 20, part 50 and
its appendices, and parts 73 and 100 as they apply to applications for
construction permits and operating licenses for nuclear power plants,
except as otherwise specified in this subpart or as the context
otherwise indicates. The requirement in Sec. 50.58 of this chapter for
review of the application by the Advisory Committee on Reactor
Safeguards and the holding of a public hearing, apply in context, with
respect to matters of radiological health and safety, environmental
protection, and the common defense and security, to licenses under this
subpart to manufacture nuclear power reactors (manufacturing licenses)
to be operated at sites not identified in the license application.
Sec. 52.249 Applicability of NRC requirements.
An applicant shall comply with all requirements in this chapter of
Title 10 applicable to applicants for construction permits and
operating licenses under this chapter of Title 10, except Sec. Sec.
50.10(b) and (c), 50.12(b), 50.23, 50.30(d), 50.34(a)(10), 50.34a(c),
50.35(a) and (c), 50.40(a), 50.45, 50.55(d), 50.56 of this chapter and
Appendix J of 10 CFR part 50 do not apply to manufacturing licenses.
Appendices E and H of 10 CFR part 50 apply to manufacturing licenses
only to the extent that the requirements of these appendices involve
facility design features.
Sec. 52.251 Referral to the ACRS.
The Commission shall refer a copy of the application to the
Advisory Committee on Reactor Safeguards (ACRS). The ACRS shall report
on those portions of the application which concern safety.
Sec. 52.253 Issuance of manufacturing license.
(a) The Commission may issue a license to manufacture one or more
nuclear power reactors to be operated at sites not identified in the
license application if the Commission finds that:
(1) The applicant has described the proposed design of and the site
parameters postulated for the reactor(s), including, but not limited
to, the principal architectural and engineering criteria for the
design, and has identified the major features of components
incorporated therein for the protection of the health and safety of the
public.
(2) Further technical or design information that may be required to
complete the design report and which can reasonably be left for later
consideration, will be supplied in a supplement to the design report.
(3) Safety features or components, if any, that require research
and development have been described by the applicant and the applicant
has identified, and there will be conducted a research and development
program reasonably designed to resolve any safety questions associated
with the features of components; and
(4) On the basis of the foregoing, there is reasonable assurance
that:
(i) Such safety questions will be satisfactorily resolved before
any of the proposed nuclear power reactor(s) are removed from the
manufacturing site; and
(ii) Taking into consideration the site criteria contained in part
100 of this chapter, the proposed reactor(s) can be constructed and
operated at sites having characteristics that fall within the site
parameters postulated for the design of the reactor(s) without undue
risk to the health and safety of the public.
(5) The applicant is technically and financially qualified to
design and manufacture the proposed nuclear power reactor(s).
(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.
(7) On the basis of the evaluations and analyses of the
environmental effects of the proposed action required by subpart A of
part 51 of this chapter and Sec. 52.245(b), the action called for is
the issuance of the license.
(b) When an applicant has supplied initially all of the technical
information required to complete the application, including the final
design of the reactor(s), the findings required for the issuance of the
license will be
[[Page 40061]]
appropriately modified to reflect that fact.
(c) Each manufacturing license issued under this subpart will
specify the number of nuclear power reactors authorized to be
manufactured and the latest date of the completion of the manufacture
of all such reactors. Upon good cause shown, the Commission will extend
the completion date for a reasonable period of time.
Sec. 52.255 Duration of design approval.
A nuclear plant design that is approved as part of the issuance of
a manufacturing license is valid for five years from the date of
issuance of the manufacturing license.
Sec. 52.257 Finality of the manufacturing license.
In making the findings required by this part for the issuance of a
construction permit or an operating license for a nuclear power
reactor(s) that has been manufactured under a Commission license issued
under this subpart, or an amendment to such a manufacturing license,
construction permit, or operating license, the Commission will treat as
resolved those matters which have been resolved at an earlier stage of
the licensing process, unless there exists significant new information
that substantially affects the conclusion(s) reached at the earlier
stage or other good cause.
Subpart I--Duplicate Design Licenses
Sec. 52.261 Scope of subpart.
(a) Section 101 of the Atomic Energy Act of 1954, as amended, and
Sec. 50.10 of this chapter require a Commission license to transfer or
receive in interstate commerce, manufacture, produce, transfer,
acquire, possess, use, import or export any production or utilization
facility. The regulations in 10 CFR part 50 require the issuance of a
construction permit by the Commission before commencement of
construction of a production or utilization facility, except as
provided in Sec. 50.10(e) of this chapter, and the issuance of an
operating license before the operation of the facility.
(b) The Commission's regulations in 10 CFR part 2 specifically
provide for the holding of hearings on particular issues separately
from other issues involved in hearings in licensing proceedings (10 CFR
2.761a and 10 CFR part 2, appendix A, section I(c)), and for the
consolidation of adjudicatory proceedings and of the presentations of
parties in adjudicatory proceedings such as licensing proceedings (10
CFR 2.715a and 2.716).
(c) This subpart sets out the particular requirements and
provisions applicable to situations in which applications are filed by
one or more applicants for licenses to construct and operate nuclear
power reactors of essentially the same design to be located at
different sites.
(d) 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 subpart and subpart D of part 2 of this chapter.
Sec. 52.263 Relationship to other subparts.
Except as otherwise specified in this subpart or as the context
otherwise indicates, the provisions of 10 CFR part 50, applicable to
construction permits and operating licenses, including the requirement
in Sec. 50.58 of this chapter for review of the application by the
Advisory Committee on Reactor Safeguards and the holding of public
hearings, apply to construction permits and operating license subject
to this subpart.
Sec. 52.265 Filing and contents of applications.
(a) Applications for construction permits submitted under this
subpart must include the information required by Sec. Sec. 50.33,
50.33a, 50.34(a) and 50.34a (a) and (b) of this chapter, and be
submitted as specified in Sec. 50.4 of this chapter. The applicant
shall also submit the information required by Sec. 51.50 of this
chapter.
(b) For the technical information required by Sec. Sec.
50.34(a)(1) through (5) and (8) and 50.34a (a) and (b) of this chapter,
reference may be made to a single preliminary safety analysis of the
design \1\ which, for the purposes of 10 CFR 50.34(a)(1) includes one
set of site parameters postulated for the design of the reactors, and
an analysis and evaluation of the reactors in terms of such postulated
site parameters. This single preliminary safety analysis must also
include information pertaining to design features of the proposed
reactors that affect plans for coping with emergencies in the operation
of the reactors, and must describe the quality assurance program with
respect to aspects of design, fabrication, procurement and construction
that are common to all of the reactors.
---------------------------------------------------------------------------
\1\ As used in this subpart, 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.
---------------------------------------------------------------------------
(c) Applications for operating licenses submitted pursuant to this
subpart must include the information required by Sec. Sec. 50.33,
50.34(b) and (c), and 50.34a(c) of this chapter. The applicant shall
also submit the information required by Sec. 51.53 of this chapter.
For the technical information required by Sec. Sec. 50.34(b)(2)
through (5) and 50.34a(c), reference may be made to a single final
safety analysis of the design.
(d) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Subpart J--[Reserved]
Subpart K--[Reserved]
Subpart L--[Reserved]
Subpart M--Enforcement
Sec. 52.401 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) Section 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 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. 52.403 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 issued under
sections 161b, 161i, or 161o of the Act. For purposes of section 223,
all the regulations in this 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 this part 52 that are not issued under
sections 161b,
[[Page 40062]]
161i, or 161o for the purposes of section 223 are as follows:
Sec. Sec. 52.1, 52.3, 52.5, 52.8, 52.11, 52.13, 52.15, 52.17, 52.18,
52.19, 52.21, 52.23, 52.24, 52.27, 52.29, 52.31, 52.33, 52.37, 52.39,
52.101, 52.103, 52.105, 52.107, 52.109, 52.111, 52.113, 52.115, 52.117,
52.119, 52.121, 52.123, 52.125, 52.201, 52.203, 52.205, 52.207, 52.209,
52.211, 52.213, 52.215, 52.217, 52.219, 52.221, 52.225, 52.227, 52.231,
52.401, 52.403.
Appendix A--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 (hereinafter 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 10 CFR 52.107, with the exception of
generic technical specifications and conceptual design information;
2. Information required for a final safety analysis report under
10 CFR 50.34;
3. 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
4. Combined license (COL) action items (COL license
information), which identify certain matters that shall 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: (i) 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 (ii) 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, 10 CFR 52.3, 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 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, VA 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-specific design parameters, 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 license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10
CFR 52.207, 52.209, and 52.211, 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 information and
utilizing 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 Section 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 52.107(a) that is not within
the scope of this appendix.
3. Physically 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;
[[Page 40063]]
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 52.127(a)(4) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held pursuant to 10
CFR 52.231, 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 pursuant to 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 pursuant to 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 Section VIII.B.5.f of this appendix,
all departures from Tier 2 pursuant to and in compliance with the
change processes in Section 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 52.127(a)(4). 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 52.217 or the hearing provided under 10 CFR
52.231, 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 52.217 or 10 CFR 52.231. If GE Nuclear
Energy declines to provide the information sought, GE Nuclear Energy
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 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 52.119(b) and
52.121(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 52.127(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 52.127(a)(3).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.127(b)(1) and 52.227(b). 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 52.127(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. 52.119 or 52.125, 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 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.
[[Page 40064]]
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 pursuant to 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 a
severe accident issue identified in the plant-specific DCD, requires
a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular 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 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.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.231(a), who believes that an applicant or licensee who
references this appendix has not complied with Section 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
in compliance with the general requirements of 10 CFR 2.714(b)(2),
the petition must demonstrate that the departure does not comply
with Section VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears an asserted noncompliance with an
ITAAC acceptance criterion in the case of a 10 CFR 52.231
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 Section 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
52.127(a)(4).
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 52.231(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 N-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 technical specifications and other
operational requirements 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 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 2.758(b) 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 50.12(a). 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 52.231(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. The petition must
comply with the general requirements of 10 CFR 2.714(b)(2) and must
demonstrate why special circumstances as defined in 10 CFR 2.758(b)
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
[[Page 40065]]
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
satisfied.
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 satisfied, 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 52.227(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.227(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 52.231(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
52.231(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. 52.231(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 52.227 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 and Tier 2.
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
pursuant to 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
departures from the plant-specific 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 50.4.
2. An applicant or licensee who references this appendix shall
submit updates to its plant-specific DCD, which reflect the generic
changes to the generic DCD and the plant-specific departures made
pursuant to Section VIII of this appendix. These updates must be
filed in accordance with the filing requirements applicable to final
safety analysis report updates in 10 CFR 50.4 and 50.71(e).
3. The reports and updates required by paragraphs B.1 and B.2 of
this section 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 plant-specific DCD.
b. During the interval from the date of application to the date
of issuance of a license, the report and any updates to the plant-
specific DCD must be submitted annually and may be submitted along
with amendments to the application.
c. During the interval from the date of issuance of a license to
the date the Commission makes its findings under 10 CFR 52.231(g),
the report must be submitted quarterly. Updates to the plant-
specific DCD must be submitted annually.
d. After the Commission has made its finding under 10 CFR
52.231(g), reports and updates to the plant-specific DCD may be
submitted annually or along with updates to the site-specific
portion of the final safety analysis report for the facility at the
intervals required by 10 CFR 50.71(e), or at shorter intervals as
specified in the license.
Appendix B--Design Certification Rule for the System 80+ Design
I. Introduction
Appendix B constitutes design certification for the System
80+\2\ 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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\2\ ``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 (hereinafter 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 10 CFR 52.107, with the exception of
generic technical specifications and conceptual design information;
2. Information required for a final safety analysis report under
10 CFR 50.34;
3. 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
4. Combined license (COL) action items (COL license
information), which identify
[[Page 40066]]
certain matters that shall 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:
(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, 10 CFR 52.3, 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, VA 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-specific design parameters, 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 license that wishes to reference this
appendix shall, in addition to complying with the requirements of 10
CFR 52.207, 52.209, and 52.211, 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 information and
utilizing 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 Section 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 52.107(a) that is not within
the scope of this appendix.
3. Physically 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
part 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 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 52.127(a)(4) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held pursuant to 10
CFR 52.231, 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 pursuant to 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 pursuant to 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 Section VIII.B.5.f of this appendix,
all departures from Tier 2 pursuant to and in compliance with the
change processes in Section 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 52.127(a)(4). 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:
[[Page 40067]]
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 52.217
or the hearing provided under 10 CFR 52.231, 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 52.217 or 10 CFR 52.231. 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 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 52.119(b) and
52.121(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 52.127(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 52.127(a)(3).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.127(b)(1) and Sec. 52.227(b). 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 52.127(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. 52.119 or 52.125, 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 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 technical specifications,
or requires a license amendment pursuant to 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 a
severe accident issue identified in the plant-specific DCD, requires
a license amendment if--
(1) There is a substantial increase in the probability of a
severe accident such that a particular 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 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.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.231(a), who believes that an applicant or licensee who
references this appendix has not complied with Section 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 2.714(b)(2), the
petition must demonstrate that the departure does not comply with
Section VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC
[[Page 40068]]
acceptance criterion in the case of a 10 CFR 52.231 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 Section 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
52.127(a)(4).
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 52.231(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 N-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 & controls setpoint methodology.
(8) Instrumentation & controls hardware and software changes.
(9) Instrumentation & 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 technical specifications and other
operational requirements 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 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 2.758(b) 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 50.12(a). 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 52.231(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 2.714(b)(2) and must
demonstrate why special circumstances as defined in 10 CFR 2.758(b)
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
satisfied.
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 satisfied, 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 52.227(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.227(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 52.231(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
52.231(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. 52.231(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 52.227 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 and Tier 2.
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
pursuant to Section VIII of this appendix throughout the period of
application and for the term of the license (including any period of
renewal).
[[Page 40069]]
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
departures from the plant-specific 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 50.4.
2. An applicant or licensee who references this appendix shall
submit updates to its plant-specific DCD, which reflect the generic
changes to the generic DCD and the plant-specific departures made
pursuant to Section VIII of this appendix. These updates must be
filed in accordance with the filing requirements applicable to final
safety analysis report updates in 10 CFR 50.4 and 50.71(e).
3. The reports and updates required by paragraphs B.1 and B.2 of
this section 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 plant-specific DCD.
b. During the interval from the date of application to the date
of issuance of a license, the report and any updates to the plant-
specific DCD must be submitted annually and may be submitted along
with amendments to the application.
c. During the interval from the date of issuance of a license to
the date the Commission makes its findings under 10 CFR 52.231(g),
the report must be submitted quarterly. Updates to the plant-
specific DCD must be submitted annually.
d. After the Commission has made its finding under 10 CFR
52.231(g), reports and updates to the plant-specific DCD may be
submitted annually or along with updates to the site-specific
portion of the final safety analysis report for the facility at the
intervals required by 10 CFR 50.71(e), or at shorter intervals as
specified in the license.
Appendix C--Design Certification Rule for the AP600 Design
I. Introduction
Appendix C constitutes the standard design certification for the
AP600\3\ 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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\3\ 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 (hereinafter 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 10 CFR 52.107, with the exception of
generic technical specifications and conceptual design information;
2. Information required for a final safety analysis report under
10 CFR 50.34;
3. 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
4. Combined license (COL) action items (combined 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.
5. 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 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, 10 CFR 52.3, 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 Mr. Michael Corletti, Westinghouse Electric
Company, P.O. Box 355, Pittsburgh, PA 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-specific design parameters, 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 license that wishes to reference this
appendix shall, in addition
[[Page 40070]]
to complying with the requirements of 10 CFR 52.207, 52.209, and
52.211, 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 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 Section 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 52.107(a) that is not within
the scope of this appendix.
3. Physically 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;
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 52.127(a)(4) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held pursuant to 10
CFR 52.231, 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 pursuant to 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 pursuant to 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 Section VIII.B.5.f of this appendix,
all departures from Tier 2 pursuant to and in compliance with the
change processes in Section 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 (SAMDAs) 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 SAMDA 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 52.127(a)(4). 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 52.217
or the hearing provided under 10 CFR 52.231, 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 52.217 or 10 CFR 52.231. 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 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 52.119(b) and
52.121(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 52.127(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
[[Page 40071]]
plant-specific orders are governed by the requirements in 10 CFR
52.127(a)(3).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.127(b)(1) and 52.227(b). 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 52.127(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. 52.119 or 52.125, 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 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 technical specifications,
or requires a license amendment pursuant to 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 a
severe accident issue identified in the plant-specific DCD, requires
a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular 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 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.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.231(a), who believes that an applicant or licensee who
references this appendix has not complied with Section 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
comply with the general requirements of 10 CFR 2.714(b)(2), the
petition must demonstrate that the departure does not comply with
Section 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 52.231
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 Section 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
52.127(a)(4).
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 52.231(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 N-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 technical specifications and other
operational requirements 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 C.3 or C.4 of this section.
3. The Commission may require plant-specific departures on
generic technical
[[Page 40072]]
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 2.758(b) 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 50.12(a). 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 52.231(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 2.714(b)(2) and must
demonstrate why special circumstances as defined in 10 CFR 2.758(b)
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
satisfied.
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 satisfied, 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 52.227(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.227(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 52.231(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
52.231(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. 52.231(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 52.227 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 and Tier 2.
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
pursuant to 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
departures from the plant-specific 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 50.4.
2. An applicant or licensee who references this appendix shall
submit updates to its plant-specific DCD, which reflect the generic
changes to the generic DCD and the plant-specific departures made
pursuant to Section VIII of this appendix. These updates must be
filed in accordance with the filing requirements applicable to final
safety analysis report updates in 10 CFR 50.4 and 50.71(e).
3. The reports and updates required by paragraphs B.1 and B.2 of
this section 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 plant-specific DCD.
b. During the interval from the date of application to the date
of issuance of a license, the report and any updates to the plant-
specific DCD must be submitted annually and may be submitted along
with amendments to the application.
c. During the interval from the date of issuance of a license to
the date the Commission makes its findings under 10 CFR 52.231(g),
the report must be submitted quarterly. Updates to the plant-
specific DCD must be submitted annually.
d. After the Commission has made its finding under 10 CFR
52.231(g), reports and updates to the plant-specific DCD may be
submitted annually or along with updates to the site-specific
portion of the final safety analysis report for the facility at the
intervals required by 10 CFR 50.71(e), or at shorter intervals as
specified in the license.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE
28. The authority citation for Part 72 continues to read as
follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat.
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-
486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135,
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153,
10155, 10157, 10161, 10168).
[[Page 40073]]
Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),
10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2224 (42 U.S.C. 10101,
10137(a), 10161(h)). Subparts K and L are also issued under sec.
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252
(42 U.S.C. 10198).
29. Section 72.210 is revised to read as follows:
Sec. 72.210 General license issued.
A general license is hereby issued for the storage of spent fuel in
an independent spent fuel storage installation at power reactor sites
to persons authorized to possess or operate nuclear power reactors
under 10 CFR part 50 or under a combined license or duplicate design
license under 10 CFR part 52.
30. In Sec. 72.218, paragraph (b) is revised to read as follows:
Sec. 72.218 Termination of licenses.
* * * * *
(b) An application for termination of the reactor operating,
combined, or duplicate design license submitted under Sec. 50.82 of
this chapter must contain a description of how the spent fuel stored
under this general license will be removed from the reactor site.
* * * * *
PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS
31. The authority citation for Part 73 continues to read as
follows:
Authority: Secs. 53, 161, 68 Stat. 930, 948, as amended, sec.
147, 94 Stat. 780 (402 U.S.C. 2073, 2167, 2201); sec. 201, as
amended, 204, 88 Stat. 1242, as amended, 1245, sec. 1701, 106 Stat.
2951, 2952, 2953 (42 U.S.C. 5841, 5844, 2297f).
Section 73.1 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 73.37(f) also
issued under sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841
note). Section 73.57 is issued under sec. 606, Pub. L. 99-399, 100
Stat. 876 (42 U.S.C. 2169).
32. In Sec. 73.1, paragraph (b)(1)(i) is revised to read as
follows:
Sec. 73.1 Purpose and scope.
* * * * *
(b) * * *
(1) * * *
(i) The physical protection of production and utilization
facilities licensed pursuant to 10 CFR parts 50 or 52.
* * * * *
PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY
REQUIREMENTS
33. The authority citation for Part 140 continues to read as
follows:
Authority: Secs. 161, 170, 68 Stat. 948, 71 Stat. 576, as
amended (42 U.S.C. 2201, 2210); secs. 201, as amended, 202, 88 Stat.
1242, as amended, 1244 (42 U.S.C. 5841, 5842).
34. In Sec. 140.2, paragraph (a)(1) is revised to read as follows:
Sec. 140.2 Scope.
(a) * * *
(1) To each person who is an applicant for or holder of a license
issued pursuant to 10 CFR parts 50, 52, or 54 to operate a nuclear
reactor, and
* * * * *
35. Section 140.10 is revised to read as follows:
Sec. 140.10 Scope.
This subpart applies to applicants for and holders of licenses
issued pursuant to 10 CFR parts 50, 52, or 54 authorizing operation of
nuclear reactors, except licenses for the conduct of educational
activities issued to, or applied for, by persons found by the
Commission to be nonprofit educational institutions and except persons
found by the Commission to be Federal agencies. This subpart also
applies to persons licensed to possess and use plutonium in a plutonium
processing and fuel fabrication plant.
36. Section 140.11 is amended by revising paragraph (b) and adding
paragraph (c) to read as follows:
Sec. 140.11 Amounts of financial protection for certain reactors.
* * * * *
(b) In any case where a person is authorized pursuant to parts 50
or 52 of this chapter to operate two or more nuclear reactors at the
same location, the total primary financial protection required of the
licensee for all such reactors is the highest amount which would
otherwise be required for any one of those reactors: Provided, That
such primary financial protection covers all reactors at the location.
(c) A holder of a combined license issued under part 52 of this
chapter must comply with paragraphs (a) and (b) of this section when
the Commission authorizes operation under Sec. 52.231(g).
37. Section 140.13 is revised to read as follows:
Sec. 140.13 Amount of financial protection required of certain
holders of construction permits and combined licenses.
(a) Each holder of a construction permit under part 50 of this
chapter authorizing construction of a nuclear reactor who is also the
holder of a license under part 70 of this chapter authorizing
ownership, possession, and storage only of special nuclear material at
the site of the nuclear reactor for use as fuel in operation of the
nuclear reactor after issuance of an operating license under part 50 of
this chapter, shall (during the period prior to issuance of the license
authorizing operation of the reactor) have and maintain financial
protection in the amount of $1,000,000. Proof of financial protection
shall be filed with the Commission in the manner specified in Sec.
140.15 prior to issuance of the license under part 70 of this chapter.
(b) Each holder of a combined license for a nuclear power reactor
under part 52 of this chapter, who is also the holder of a license
under part 70 of this chapter authorizing ownership, possession, and
storage only of special nuclear material at the site of the nuclear
reactor for use as fuel in operation of the nuclear reactor after
authorization to operate under part 52 of this chapter, shall (during
the period prior to Commission authorization to operate the reactor
under Sec. 52.231 of this chapter) have and maintain financial
protection in the amount of $1,000,000. Proof of financial protection
shall be filed with the Commission in the manner specified in Sec.
140.15 prior to issuance of the license under part 70 of this chapter.
PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT
OF 1954, AS AMENDED
38. The authority citation for part 170 continues to read as
follows:
Authority: Sec. 9701, Pub. L. 97-258, 96 Stat. 1051 (31 U.S.C.
9701); sec. 301, Pub. L. 92-314, 86 Stat. 227 (42 U.S.C. 2201w);
sec. 201, Pub. L. 93-438, 88 Stat. 1242, as amended (42 U.S.C.
5841); sec. 205a, Pub. L. 101-576, 104 Stat. 2842, as amended (31
U.S.C. 901, 902).
39. In Sec. 170.2, paragraphs (g) and (k) are revised to read as
follows:
Sec. 170.2 Scope.
* * * * *
(g) An applicant for or holder of a production or utilization
facility construction permit or operating license issued under 10 CFR
part 50, or an
[[Page 40074]]
approval, certification, permit, or license issued under 10 CFR part
52;
* * * * *
(k) Applying for or already has applied for review, under 10 CFR
part 52, of a facility site prior to the submission of an application
for a construction permit;
* * * * *
Dated at Rockville, Maryland, this 24th day of June, 2003.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 03-16413 Filed 7-2-03; 8:45 am]
BILLING CODE 7590-01-P