[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
------------------------------------------------------------------------

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.
---------------------------------------------------------------------------

    \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.''
---------------------------------------------------------------------------

    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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    \2\ ``System 80+'' is a trademark of Westinghouse Electric 
Company LLC.
---------------------------------------------------------------------------

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