[Federal Register Volume 72, Number 166 (Tuesday, August 28, 2007)]
[Rules and Regulations]
[Pages 49352-49566]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 07-3861]



[[Page 49351]]

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Part II





Nuclear Regulatory Commission





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10 CFR Parts 1, 2, 10, et al.



Licenses, Certifications, and Approvals for Nuclear Power Plants; Final 
Rule

Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / 
Rules and Regulations

[[Page 49352]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 1, 2, 10, 19, 20, 21, 25, 26, 50, 51, 52, 54, 55, 72, 
73, 75, 95, 140, 170, and 171

RIN 3150-AG24


Licenses, Certifications, and Approvals for Nuclear Power Plants

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations by revising the provisions applicable to the licensing and 
approval processes for nuclear power plants (i.e., early site permit, 
standard design approval, standard design certification, combined 
license, and manufacturing license). These amendments clarify the 
applicability of various requirements to each of the licensing 
processes by making necessary conforming amendments throughout the 
NRC's regulations to enhance the NRC's regulatory effectiveness and 
efficiency in implementing its licensing and approval processes. The 
NRC has considered and resolved the public comments.

DATES: The effective date is September 27, 2007.

FOR FURTHER INFORMATION CONTACT: Nanette V. Gilles, Office of New 
Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, telephone 301-415-1180, e-mail <a href="/cdn-cgi/l/email-protection#620c1405220c10014c050d14"><span class="__cf_email__" data-cfemail="d6b8a0b196b8a4b5f8b1b9a0">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION: 
I. Background
    A. Development of Proposed Rule
    B. Publication of Revised Proposed Rule
II. Overview of Public Comments
III. Reorganization of Part 52 and Conforming Changes in the NRC's 
Regulations
IV. Responses to Specific Requests for Comments
V. Discussion of Substantive Changes and Responses to Significant 
Comments
    A. Introduction
    B. Testing Requirements for Advanced Reactors
    C. Changes to 10 CFR Part 52
    D. Changes to 10 CFR Part 50
    E. Change to 10 CFR Part 1
    F. Changes to 10 CFR Part 2
    G. Changes to 10 CFR Part 10
    H. Changes to 10 CFR Part 19
    I. Changes to 10 CFR Part 20
    J. Changes to 10 CFR Part 21
    K. Change to 10 CFR Part 25
    L. Changes to 10 CFR Part 26
    M. Changes to 10 CFR Part 51
    N. Changes to 10 CFR Part 54
    O. Changes to 10 CFR Part 55
    P. Changes to 10 CFR Part 72
    Q. Changes to 10 CFR Part 73
    R. Change to 10 CFR Part 75
    S. Changes to 10 CFR Part 95
    T. Changes to 10 CFR Part 140
    U. Changes to 10 CFR Part 170
    V. Changes to 10 CFR Part 171
VI. Section-by-Section Analysis
VII. Availability of Documents
VIII. Agreement State Compatibility
IX. Voluntary Consensus Standards
X. Environmental Impact--Categorical Exclusion
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Certification
XIV. Backfit Analysis
XV. Congressional Review Act

I. Background

A. Development of Proposed Rule

    On July 3, 2003 (68 FR 40026), the NRC published a proposed 
rulemaking that would clarify and/or correct miscellaneous parts of the 
NRC's regulations; update 10 CFR part 52 in its entirety; and 
incorporate stakeholder comments. On March 13, 2006 (71 FR 12781), the 
NRC issued a revised proposed rule that would rewrite part 52, make 
changes throughout the Commission's regulations to ensure that all 
licensing processes in part 52 are addressed, and clarify the 
applicability of various requirements to each of the processes in part 
52 (i.e., early site permit, standard design approval, standard design 
certification, combined license, and manufacturing license). This 
proposed rule superseded the July 3, 2003, proposed rule.
    The NRC issued 10 CFR part 52 on April 18, 1989 (54 FR 15372), to 
reform the NRC's licensing process for future nuclear power plants. 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 allow for 
resolving safety and environmental issues early in licensing 
proceedings and were intended to enhance the safety and reliability of 
nuclear power plants through standardization. Subsequently, the NRC 
certified four nuclear power plant designs under subpart B of 10 CFR 
part 52--the U.S. Advanced Boiling Water Reactor (ABWR) (62 FR 25800; 
May 12, 1997), the System 80+ (62 FR 27840; May 21, 1997), the AP600 
(64 FR 72002; December 23, 1999), and the AP1000 (71 FR 4464; January 
27, 2006). These design certifications are codified in appendices A, B, 
C, and D of 10 CFR part 52, respectively.
    The NRC planned to update 10 CFR part 52 after using the standard 
design certification process. The proposed 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 (SRM) on 
January 14, 1999 (SRM on SECY-98-282), approving the NRC staff's plan 
for revising 10 CFR part 52. Subsequently, the NRC obtained 
considerable stakeholder comment on its planned action, conducted three 
public meetings on the proposed rulemaking, and twice posted draft rule 
language on the NRC's rulemaking Web site before issuance of the July 
2003 proposed rule. \

B. Publication of Revised Proposed Rule

    A number of factors led the NRC to question whether the July 2003 
proposed rule would meet the NRC's objective of improving the 
effectiveness of its processes for licensing future nuclear power 
plants. First, public comments identified several concerns about 
whether the proposed rule adequately addressed the relationship between 
part 50 and part 52, and whether it clearly specified the applicable 
regulatory requirements for each of the licensing and approval 
processes in part 52. In addition, as a result of the NRC staff's 
review of the first three early site permit applications, the staff 
gained additional insights into the early site permit process. The NRC 
also had the benefit of public meetings with external stakeholders on 
NRC staff guidance for the early site permit and combined license 
processes. As a result, the NRC decided that a substantial rewrite and 
expansion of the July 2003 proposed rulemaking was desirable so that 
the agency may more effectively and efficiently implement the licensing 
and approval processes for future nuclear power plants under part 52.
    Accordingly, the Commission decided to revise the July 2003 
proposed rule and published a revised proposed rule for public comment 
on March 13, 2006. This revised proposed rule contained a rewrite of 
part 52, as well as changes throughout the NRC's regulations, to ensure 
that all licensing and approval processes in part 52 are addressed, and 
to clarify the applicability of various requirements to each of the 
processes in part 52. In light of the substantial rewrite of the July 
2003 proposed rule, the expansion of the scope of the rulemaking, and 
the NRC's decision to publish the revised proposed rule for public 
comment, the NRC decided that developing responses to comments received 
on the July 2003 proposed rule would not be an effective use of agency 
resources. The NRC requested that commenters on the July 2003 proposed 
rule who believed that their earlier

[[Page 49353]]

comments were not adequately addressed in the March 2006 proposed rule 
resubmit their comments.

II. Overview of Public Comments

    The public comment period for the March 2006 revised proposed rule 
expired on May 30, 2006. The NRC received 19 comment letters from 
industry stakeholders, other Federal agencies, and individuals during 
the public comment period. The NRC has considered and resolved all of 
the public comments received during the comment period and has made 
modifications to the rule language, as appropriate. The NRC has 
prepared a separate report, entitled Comment Summary Report: 10 CFR 
Part 52, Licenses, Certifications, and Approvals for Nuclear Power 
Plants, in which it summarizes the public comments received and 
discusses the agency's disposition of each comment. This report is 
available to the public as discussed in Section VII of the 
Supplementary Information of this document. The resolution of 
significant public comments is also discussed in Section IV, Responses 
to Specific Requests for Comments and, Section V, Discussion of 
Substantive Changes and Responses to Significant Comments in this 
document.

III. Reorganization of Part 52 and Conforming Changes in the NRC's 
Regulations

    Since the adoption of 10 CFR part 52 in 1989, the NRC and its 
external stakeholders identified a number of interrelated issues and 
concerns with the licensing process. One significant concern was that 
the overall regulatory relationship between part 50 and part 52 was not 
always clear. In the former rules, it was often difficult to tell 
whether general regulatory provisions in part 50 apply to part 52. One 
example is whether the absence of an exemption provision in part 52 
denotes the NRC's determination that exemptions from part 52 
requirements are not available, or that these exemptions are controlled 
by Sec.  50.12. A related problem is the current lack of specific 
delineation of the applicability of NRC requirements throughout 10 CFR 
Chapter I to the licensing and approval processes in part 52. For 
example, the indemnity and insurance provisions in part 140 were not 
revised to address their applicability to applicants for and holders of 
combined licenses under subpart C of part 52. Even where part 52 
provisions referenced specific requirements in part 50, it was not 
always clear from the language of the part 50 requirement how that 
requirement applied to the part 52 processes. For example, Sec.  
52.47(a)(1)(i) provides that a standard design certification 
application must contain the ``technical information which is required 
of applicants for construction permits and operating licenses by 10 
CFR* * *part 50* * *and which is technically relevant to the design and 
not site-specific.''
    The language did not explicitly identify the part 50 requirements 
that are ``technically relevant to the design.'' Even where a specific 
regulation in part 50 is identified as a requirement, the language of 
the referenced regulation itself was not changed to reflect the 
specific requirements as applied to the part 52 processes. For example, 
Sec.  52.79(b) provides that the application must contain the 
``technically relevant information required of applicants for an 
operating license required by 10 CFR 50.34.'' Other than the fact that 
this language shares the problem discussed earlier of what constitutes 
a ``technically relevant'' requirement, Sec.  50.34(b) is based upon 
the two-step licensing process whereby certain important information is 
submitted at the construction permit stage, and then supplemented with 
more detailed information at the operating license stage. Thus, it 
could be asserted that certain information that must be submitted in 
the construction permit application, e.g., the ``principal design 
criteria for the facility'' required by Sec.  50.34(a)(3)(i), may be 
regarded as not required to be submitted for a combined license 
application under the former version of part 52.
    Another potential source of confusion is that the different 
subparts of part 52 and the appendices on standard design approvals and 
manufacturing licenses are not organized using the same format of 
individual sections (e.g., ``Scope of subpart,'' followed by 
``Relationship to other subparts,'' followed by ``Filing of 
application''). Moreover, the organization and textual content of 
identically-titled sections differs among the subparts, and with 
appendices M, N, O, and Q, which establish additional licensing and 
approval processes. While these differences do not constitute an 
insurmountable problem to their use and application, it became apparent 
to the Commission that adoption of a common format, organization, and 
textual content would enhance usability and result in increased 
regulatory effectiveness and efficiency.
    In the 2003 proposed rule, the NRC proposed several changes that 
were intended to address some (but not all) of these issues. However, 
based upon comments received on the 2003 proposed rule, the NRC's 
experience to date with early site permit applications, interactions 
with external stakeholders concerning NRC guidance for combined license 
applications, and NRC's screening of 10 CFR Chapter I requirements 
following the receipt of public comments on the 2003 proposed rule, the 
NRC concluded that the 2003 proposed rule would not adequately address 
and resolve these issues.
    Accordingly, in the March 13, 2006, proposed rule the NRC took a 
more comprehensive approach to addressing these issues by reorganizing 
part 52, implementing a uniform format and content for each of the 
subparts in part 52, using consistent wording and organization of 
sections in each of the subparts, and making conforming changes 
throughout 10 CFR Chapter I to reflect the licensing and approval 
processes in part 52. The NRC also coordinated and reconciled 
differences in wording among provisions in parts 2, 50, 51, and 52 to 
provide consistent terminology throughout all of the regulations 
affecting part 52. Under the NRC's reorganization of part 52, the 
existing appendices O and M on standard design approvals and 
manufacturing licenses, respectively, have been redesignated as new 
subparts in part 52. Redesignating these appendices as subparts in part 
52 has resulted in a consistent format and organization of the 
requirements applicable to each of the licensing and approval 
processes. In addition, the redesignation clarifies that each of the 
licensing and approval processes in these appendices are available to 
potential applicants as an alternative to the processes in part 50 
(construction permit and operating license) and the existing subparts A 
through C of part 52. The Commission does not, by virtue of this 
redesignation, either favor or disfavor the processes in the former 
appendices M and O of part 52. Rather, the Commission is standardizing 
the format and organization of part 52, and clarifying the full range 
of alternatives that are available under part 52 for use by potential 
applicants. Consistent with the broad scope of part 52, the NRC has 
retitled 10 CFR part 52 as ``Licenses, Certifications, and Approvals 
for Nuclear Power Plants.''
    The NRC has also reorganized and expanded the scope of the 
administrative and general regulatory provisions that precede the part 
52 subparts by adding new sections on written communications (analogous 
to Sec.  50.4), employee protection (analogous to Sec.  50.7), 
completeness and accuracy of information (analogous to Sec.  50.9), 
exemptions (analogous to Sec.  50.12), combining licenses (analogous to

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Sec.  50.52), jurisdictional limits (analogous to Sec.  50.53), and 
attacks and destructive acts (analogous to Sec.  50.13). The NRC 
believes that adding the new sections to part 52 rather than revising 
the comparable sections in part 50 is more consistent with the general 
format and content of the Commission's regulations in each of the parts 
of Title 10. The NRC considered whether the numbering of the newly-
added sections to part 52--in particular, the provisions on deliberate 
misconduct, employee protection, and completeness and accuracy of 
information--should match the numbering of the comparable sections in 
part 50. While this may have some benefit, the NRC ultimately decided 
not to adopt such a course for several reasons. First, other parts of 
the NRC's regulations in 10 CFR Chapter I do not maintain the same 
numbering scheme. Rather, it appears that the NRC attempted to maintain 
the order in which these sections are listed in each part. Second, 
there are other provisions in part 50 for which a comparable provision 
needed to be added to the general and administrative provisions in part 
52, but for which it would be impossible to maintain the same numbering 
(for example, Sec.  50.13 (attacks and destructive acts); Sec.  50.32 
(elimination of repetition); Sec.  50.52 (combining licenses)), unless 
the substantive provisions of part 52, beginning with Sec.  52.12, were 
changed.\1\ Maintaining in part 52 the numbering scheme for some, but 
not all, comparable sections from part 50 ultimately would be viewed as 
haphazard and arbitrary. Finally, the NRC does not believe that 
external stakeholders who must constantly refer to part 52 will be 
confused by any difference in numbering of the three sections, given 
that there are other comparable provisions for which the numbering is 
necessarily different between parts 50 and 52. For these reasons, the 
NRC did not attempt to match in the final part 52 rule the numbering of 
the comparable sections in part 50.
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    \1\ The NRC notes, in this regard that nuclear industry 
stakeholders adversely commented on the revised numbering scheme as 
set forth in the 2003 proposed part 52 rule. They suggested that the 
NRC retain, to the greatest extent posible, the numbering of the 
then existing part 52. Inasmuch as Sec.  52.12 is the first 
substantive provision of the former party 52, this placed an upper 
bound on the number of sections available for general provisions--
that is Sec.  52.0 through 52.11.
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    Appendix N, which addresses duplicate design licenses, has been 
retained in both part 52 and part 50 to afford future applicants 
flexibility and to retain the possibility of achieving regulatory 
efficiencies in part 52 combined license proceedings. Since the 
preparation of the March 2006 proposed rule, several industry groups 
have announced their intention to seek combined licenses utilizing the 
same design. In view of this industry development, the NRC believes 
that there is potential utility to keeping the option of appendix N 
open to potential combined license applicants. Accordingly, the NRC is 
retaining in part 52 the procedural alternative provided in appendix N, 
and revising its language to make its provisions applicable to combined 
licenses using identical designs. Appendix Q, which addresses early 
staff review of site suitability issues, is being removed from part 52 
but retained in part 50. Appendix Q provides for NRC staff issuance of 
a staff site report on site suitability issues with respect to a 
specific site for which a potential applicant seeks the NRC staff's 
views. The staff site report is issued after receiving and considering 
the comments of Federal, State, and local agencies and interested 
persons, as well as the views of the Advisory Committee on Reactor 
Safeguards (ACRS), but only if site safety issues are raised. The staff 
site report does not bind the Commission or a presiding officer in any 
hearing under part 2. This process is separate from the early site 
permit process in subpart A of part 52. The NRC recognizes the apparent 
redundancy between the early review of site suitability issues and the 
early site permit process. Accordingly, the NRC is removing appendix Q 
from part 52 and retaining it only in part 50.
    Inasmuch as the NRC may, in the future, adopt other regulatory 
processes for nuclear power plants, the NRC has reserved several 
subparts in part 52 to accommodate additional licensing processes that 
may be adopted by the NRC. The NRC used a standard format and content 
for revising the regulations in the existing subparts and developing 
the new subparts that address the former appendices M and O. The 
standard format and content was modeled on the existing organization 
and content of subparts A and C. Appendix N of part 52, however, has 
not been revised in that fashion because of time constraints in 
developing the final rule.
    Perhaps most importantly, the NRC has reviewed the existing 
regulations in 10 CFR Chapter I to determine if the existing 
regulations must be modified to reflect the licensing and approval 
processes in part 52. First, the NRC determined whether an existing 
regulatory provision must, by virtue of a statutory requirement or 
regulatory necessity, be extended to address a part 52 process, and, if 
so, how the regulatory provision should apply. Second, in situations 
where the NRC has some discretion, the NRC determined whether there 
were policy or regulatory reasons to extend the existing regulations to 
each of the part 52 processes. Most of the conforming changes in this 
final rule occur in 10 CFR part 50. In making conforming changes 
involving 10 CFR part 50 provisions, the NRC has adopted the general 
principle of keeping the technical requirements in 10 CFR part 50 and 
maintaining all applicable procedural requirements in part 52. However, 
due to the complexity of some provisions in 10 CFR part 50 (e.g., Sec.  
50.34), this principle could not be universally followed. A description 
of, and bases for, the substantive conforming changes for each affected 
part is provided in Section V of this document.
    To highlight the relationship between the requirements in part 52 
of this final rule and the requirements in existing part 52, the NRC is 
making two cross-reference tables available to the public. These tables 
can be found on NRC's Agencywide Documents Access and Management System 
(ADAMS) at accession number ML062550U0246. Table 1 matches each part 52 
requirement in this final rule with its counterpart in the existing 
rule. Table 2 is a reverse cross-reference table which identifies the 
section of the existing part 52 requirements from which each part 52 
requirement in this final rule was derived.

IV. Responses to Specific Requests for Comments

    In Section V of the Statements of Consideration for the March 13, 
2006, proposed rule, the NRC posed 15 questions for which it solicited 
stakeholder comments. In the following paragraphs, these questions are 
restated, comments received from stakeholders are summarized, and the 
NRC resolution of the public comments is presented.
    Question 1: General Provisions. Create new subpart for part 50. In 
response to several commenters' concerns about the clarity of the 
applicability of part 50 provisions to part 52, the Commission has 
added provisions to part 52 (Sec. Sec.  52.0 through 52.11) that are 
analogues to comparable provisions in part 50. Another possible way of 
addressing the commenters' concerns would be to transfer all the 
provisions in part 52 to a new subpart (e.g., subpart M) of part 50, 
and retain the existing numbering sequence for the current part 52 with 
the addition of a prefix (e.g., proposed

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50.1001 = current 52.1). The Commission is considering adopting this 
alternative proposal in the final rule and is interested in whether 
stakeholders regard this as a more desirable approach for minimizing 
the ambiguity of the relationship between part 50 and part 52.
    Commenters' Response: Some commenters stated the clarity of the 
regulations would not be enhanced by moving provisions from part 52 to 
a new subpart of part 50. The commenters argued that in addition to not 
eliminating existing confusion, such a content shift would create new 
confusion because current documents referencing part 52 would become 
``obsolete.''
    NRC Response: The NRC has decided not to transfer provisions from 
part 52 to a new subpart in part 50, inasmuch as: (1) no commenter 
favored transferring provisions from part 52 to a new subpart in part 
50, (2) the approaches are legally equivalent, and (3) nearly 17 years 
has passed since the Commission adopted the approach of establishing 
early site permits, standard design certifications, and combined 
licenses in a new part 52, and a reorganization of the regulations at 
this time may engender confusion without any compensating benefits in 
clarity, regulatory stability and predictability, or efficiency.
    Question 2: Currently, Sec. Sec.  52.17(b) of subpart A of 10 CFR 
part 52 requires that an early site permit application identify 
physical characteristics that could pose a significant impediment to 
the development of emergency plans. An early site permit application 
may also propose major features of the emergency plans or propose 
complete and integrated emergency plans in accordance with the 
applicable standards of Sec.  50.47 and the requirements of appendix E 
of 10 CFR part 50. The requirements in Sec.  52.17 do not further 
define major features of emergency plans. Section 52.18 of subpart A 
requires the Commission to determine, after consultation with the 
Federal Emergency Management Agency, whether any major features of 
emergency plans submitted by the applicant under Sec.  52.17(b) are 
acceptable. Section 52.18 does not provide any further explanation of 
the Commission's criteria for judging the acceptability of major 
features of emergency plans.
    The Commission has concluded, after undergoing the review of the 
first three early site permit applications, that Commission review and 
acceptance of major features of emergency plans may not achieve the 
same level of finality for emergency preparedness issues at the early 
site permit stage as that associated with a reasonable assurance 
finding of complete and integrated plans. Therefore, the Commission is 
considering modifying in the final rule the early site permit process 
in proposed subpart A to remove the option for applicants to propose 
major features of emergency plans in early site permit applications and 
requests public comment on this alternative. The NRC believes that, if 
the option for early site permit applicants to include major features 
of emergency plans is to be retained, it would be useful to further 
define in the final rule what a major feature is and establish a 
clearer level of finality associated with the NRC's review and 
acceptance of major features of emergency plans. If the option to 
include major features of emergency plans is retained in the final 
rule, the NRC would define major features of emergency plans as 
follows:

    Major features of the emergency plans means the aspects of those 
plans necessary to: (1) address one or more of the sixteen standards 
in Sec.  50.47(b), and (2) describe the emergency planning zones as 
required in Sec. Sec.  50.33(g), 50.47(c)(2), and appendix E to 10 
CFR part 50.

    In addition, the NRC is considering adopting in the final rule the 
requirement that major features of emergency plans must include the 
proposed inspections, tests, and analyses that the holder of a combined 
license referencing the early site permit shall perform, and the 
acceptance criteria that are necessary and sufficient to provide 
reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met, the facility has been 
constructed and will operate in conformity with the license, the 
provisions of the Atomic Energy Act (AEA), and the NRC's regulations, 
insofar as they relate to the major features under review.
    The NRC believes that, under this alternative, the level of 
finality associated with each major feature that the Commission found 
acceptable would be equivalent, for that individual major feature, to 
the level of finality associated with a reasonable assurance finding by 
the NRC for a complete and integrated plan, including inspections, 
tests, analyses, and acceptance criteria (ITAAC), at the early site 
permit stage.
    Commenters' Response: Several commenters suggested the current 
process for addressing major features of emergency plans (EP) in the 
early site permit (ESP) be retained without modification. Some 
commenters expressed a fear that the loss of this option would result 
in a loss of flexibility to achieve ``finality'' without producing a 
comprehensive EP. Some commenters identified a need to clarify the 
definition of ``major features'' of the EP to make it less restrictive. 
Some commenters believed that the approved major features were 
acceptable elements of a ``complete and integrated emergency plan that 
would be considered later.'' Some commenters believed the information 
should not be reviewed again during the COL process, which would 
instead focus on (1) the integration of these major features with 
information necessary to support the ``reasonable assurance finding,'' 
and (2) the updating of EP information required by Sec.  52.39(b).
    NRC Response: Based on the commenters' feedback, the NRC has 
decided to retain the current process for addressing major features of 
emergency plans in an ESP without modification. The NRC agrees that it 
should clarify the definition of ``major features'' and has done so by 
adding the definition suggested by the commenters to Sec.  52.1 in the 
final rule. For a detailed discussion of the basis for this change, see 
Section V.C.5.b of the Supplementary Information section of this 
document which discusses changes to Sec.  52.1, ``Definitions.''
    Question 3: As indicated in Section IV, Discussion of Substantive 
Changes (in the March 13, 2006, proposed rule), the NRC is proposing to 
remove appendix Q to part 52 entirely from part 52 and retain it in 
part 50. Currently, appendix Q to part 52 provides for NRC staff 
issuance of a staff site report on site suitability issues with respect 
to a specific site, for which a person (most likely a potential 
applicant for a construction permit or combined license) seeks the NRC 
staff's views. The NRC is also considering removing, in the final rule, 
the early site review process in appendix Q to part 52 in its entirety 
from the NRC's regulations and is interested in stakeholder feedback on 
this alternative. One possible reason for removing the early site 
review process in its entirety is that potential nuclear power plant 
applicants would use the early site permit process in subpart A of part 
52, rather than the early site review process as it currently exists in 
appendix Q to parts 50 and 52. Also, in cases where a combined license 
applicant was interested in seeking NRC staff review of selected site 
suitability issues (as appendix Q to part 52 was designed for), the 
applicant could request a pre-application review of these issues. The 
use of pre-application reviews for selected issues has been 
successfully used by applicants for design certification. The NRC is

[[Page 49356]]

especially interested in the views of potential applicants for nuclear 
power plant construction permits and combined licenses as to whether 
there is any value in retaining the early site review process.
    Commenters' Response: Some commenters expressed concern about the 
loss of flexibility to assess site suitability that would result from 
the deletion of appendix Q from parts 50 and 52. These commenters 
believed that appendix Q to parts 50 and 52 (in conjunction with 
subpart F of 10 CFR part 2) was important for allowing ``critical path 
issues'' to be reviewed prior to submission of a combined license (COL) 
application in instances where prior completion of an ESP was not 
feasible. Some commenters argued for the efficiency of appendix Q to 
parts 50 and 52 and subpart F of part 2 because only applicant-selected 
issues would be reviewed during these processes. Some commenters 
recommended changes be made to specifically allow ESP and COL 
applicants to reference an early site review conducted in accordance 
with appendix Q or subpart F. The commenters stated that the NRC should 
not delete the option for a part 52 applicant to reference a review 
performed under appendix Q to 10 CFR part 52.
    NRC Response: After considering these comments the NRC has decided 
to go forward with removal of appendix Q from part 52 in the final 
rule.
    However, the NRC agrees that Sec.  2.101(a-1) and subpart F of part 
2 should be modified to allow applicants for early site permits and 
combined licenses under part 52 to take advantage of those provisions. 
Both Sec.  2.101(a-1) and subpart F of part 2 have been revised in the 
final rule, albeit somewhat differently than the approach recommended 
by the commenter. Inasmuch as the revisions are to the Commission's 
rules of procedure and practice, the Commission may adopt them in final 
form without further notice and comment, under the rulemaking 
provisions of the APA, 5 U.S.C. 553(b)(A). The Commission believes that 
sufficient flexibility will be retained for future combined license 
applicants with the preservation of the provisions in Sec.  2.101(a-1) 
and subpart F of part 2 and that there is little value in also 
retaining the provisions in appendix Q.
    Question 4: Under subpart F of part 52 of the proposed rule, the 
NRC proposes to require approval of, and extend finality to, the final 
design for a reactor to be manufactured under a manufacturing license. 
While the NRC will also review the acceptability of the manufacturing 
license applicant's organization responsible for design and 
manufacturing, as well as the quality assurance (QA) program for design 
and manufacturing, the proposed rule does not provide a regulatory 
structure for further extending the scope of NRC review and issue 
finality to the manufacturing process itself. The NRC is considering 
extending regulatory review approval, and consequently expand issue 
finality, to the manufacturing itself in the final rule. There are two 
models that the Commission is considering adopting if it were to move 
in this direction. The first would be an analogue to the subpart C of 
part 52 combined license process, whereby the NRC would review and 
approve manufacturing ITAAC to be included in the manufacturing 
license. During the manufacturing of each reactor, the NRC would verify 
at the manufacturing location whether the ITAAC have been conducted and 
the acceptance criteria met. A NRC finding of successful completion of 
all the ITAAC would preclude any further inspection of the 
acceptability of the manufacture of the reactor at the site where the 
manufactured reactor is to be permanently sited and operated. The NRC's 
inspections and findings for the combined license or operating license 
would be limited to whether the reactor had been emplaced in undamaged 
condition (or damage had been appropriately repaired) and all interface 
requirements specified in the manufacturing license had been met. The 
NRC believes that it has authority to issue a manufacturing license 
under Section 161.h of the AEA.
    The other model that the NRC could adopt would be a combination of 
the approval processes used by the Federal Communications Commission 
(FCC) and Federal Aviation Administration (FAA) in approving the 
manufacture of electronic devices and airplanes. The NRC's 
manufacturing license would approve: (1) the design of the nuclear 
power reactor to be manufactured; (2) the specific manufacturing and 
quality assurance/quality control processes and procedures to be used 
during manufacture; and (3) tests and acceptance criteria for 
demonstrating that the reactor has been properly manufactured. To be 
completely consistent with the FCC and FAA models, the NRC would issue 
a manufacturing license only after a prototype of the reactor had been 
constructed and tested to demonstrate that all performance requirements 
(i.e., compliance with NRC requirements and manufacturer's 
specifications) can be met by the design to be approved for 
manufacture.
    The NRC requests public comment on whether the manufacturing 
license process in proposed subpart F of part 52 should be further 
extended in the final rule to provide an option for NRC approval of the 
manufacturing, and if so, which model of regulatory oversight, i.e., 
the combined license ITAAC model or the FCC/FAA approval model, should 
be used by the NRC. The NRC also seeks public comment on whether an 
opportunity for hearing is required by the AEA in connection with a NRC 
determination that the manufacturing ITAAC have been successfully 
completed.
    Commenters' Response: Some commenters requested that applicants for 
manufacturing licenses be allowed, but not required, to use ITAAC to 
ensure that an ``as-manufactured plant conforms to the important design 
characteristics specified in the application for the manufacturing 
license.'' Some commenters stated that a manufacturing license for 
evolutionary designs should be subject to proposed Sec.  50.43(e) and 
should not require a prototype. Some commenters stated that 
manufacturing licenses should not be subject to more stringent 
requirements than design certifications.
    NRC Response: The NRC has decided to defer consideration of this 
alternative on ITAAC, for several reasons. First, one commenter's 
proposal to allow ITAAC for assuring that the as-manufactured reactor 
``conforms to the important design characteristics specified in the 
application for the manufacturing license,'' raises questions about 
what those ``important design characteristics'' might be, and why the 
ITAAC would be so narrowly limited. The Commission did not receive any 
in-depth comments presenting arguments one way or the other on the 
feasibility of developing such ITAAC, and the potential legal 
implications of, and technical considerations with respect to, such a 
finding by the manufacturer. Moreover, it is clear that any regulatory 
process that the Commission may adopt in rulemaking would require 
further opportunity for public comment, and therefore could not be 
adopted in a final part 52 rulemaking without substantial delay. In 
light of the lack of any near-term interest by any entity in obtaining 
a manufacturing license, the Commission has decided not to adopt any 
provisions for ITAAC governing approval of manufacturing in the final 
part 52 rule. However, the Commission would address these issues in a 
timely fashion if raised in a rulemaking

[[Page 49357]]

petition which demonstrated near-term interest in an application for a 
manufacturing license.
    The Commission agrees with the commenters'' suggestions that 
manufacturing licenses for evolutionary designs should be subject to 
new Sec.  50.43(e), and that under those provisions a prototype would 
not be prerequisite to issuance of a manufacturing license for an 
evolutionary design. Further discussion is provided below in Testing 
Requirements for Advanced Reactors.
    Question 5: Currently, part 52 allows an applicant for a 
construction permit to reference either an early site permit under 
subpart A of part 52 or a design certification (DC) under subpart B of 
part 52. Specifically, Sec.  52.11 states that subpart A of part 52 
sets out the requirements and procedures applicable to NRC 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 or combined license for such a facility. 
Similarly, Sec.  52.41 states that subpart B of part 52 sets out the 
requirements and procedures applicable to NRC issuance of regulations 
granting standard design certification for nuclear power facilities 
separate from the filing of an application for a construction permit or 
combined license for the facility. However, the current regulations in 
10 CFR part 50 that address the application for and granting of 
construction permits do not make any reference to a construction permit 
applicant's ability to reference either an early site permit or a 
design certification. Also, the NRC has not developed any guidance on 
how the construction permit process would incorporate an early site 
permit or design certification, nor has the nuclear power industry made 
any proposals for the development of industry guidance on this subject. 
The NRC has not received any information from potential applicants 
stating an intention to seek a construction permit for the construction 
of a future nuclear power plant. In addition, the NRC recommends that 
future applicants who want to construct and operate a commercial 
nuclear power facility use the combined license process in subpart C of 
part 52. Therefore, the NRC is considering removing from part 52, in 
the final rule, the provisions allowing a construction permit applicant 
to reference an early site permit or a design certification and is 
interested in stakeholder feedback on this alternative.
    Commenters' Response: Some commenters stated the deletion of 
provisions allowing a construction permit applicant to reference an ESP 
or DC was ill-advised given the untested nature of the COL process and 
the resulting need to retain ``regulatory flexibility'' to deal with 
unexpected issues. As a contingency plan to buffer against difficulties 
with COL process, the commenters proposed the addition of a provision 
in part 50 to specify that a construction permit applicant could 
reference a DC without the inclusion of ITAAC. The commenters suggested 
that in these instances, ``the operating license proceeding would need 
to find under 10 CFR 50.57(a)(1) that construction of the facility has 
been substantially completed, in conformity with the construction 
permit and the application as amended, the provisions of the Act, and 
the rules and regulations of the Commission.'' Commenters stated that 
standard design should be final and not open to review in the 
construction permit and operating licenses proceeding. Commenters 
requested a construction permit applicant be able to reference an ESP 
in the same way as would a COL applicant.
    NRC Response: Based on some of the commenters' responses to this 
question and further consideration of the issue, the NRC has decided 
not to make any changes in the final rule to delete provisions allowing 
a construction permit applicant to reference an early site permit or a 
design certification. The NRC has also decided not to add any 
additional provisions to part 50 or part 52 to address a construction 
permit applicant's ability to reference either a design certification 
or an early site permit. The NRC believes it is unlikely that such a 
construction permit application will be submitted, and the NRC will 
handle any such applications on a case-by-case basis. If such an 
application were submitted, there are many process issues that would 
need to be carefully considered and would need to be discussed with the 
applicant and other stakeholders. In particular, the previously 
certified designs all used design acceptance criteria in lieu of 
detailed design information. A process for completing that design 
information without using ITAAC would have to be developed.
    Question 6: The NRC is considering revising Sec.  52.103(a) in the 
final rule to require the combined license holder to notify the NRC of 
the licensee's scheduled date for loading of fuel into a plant no later 
than 270 days before the scheduled date, and to advise the NRC every 30 
days thereafter if the date has changed and if so, the revised 
scheduled date for loading of fuel. The initial notification would 
facilitate timely NRC publication of the notice required under Sec.  
52.103(a) and NRC staff scheduling of inspection and audit activities 
to support NRC staff determinations of the successful completion of 
ITAAC under Sec.  52.99. The proposed updating would also facilitate 
NRC staff scheduling of those inspection and audit activities, 
Commission completion of hearings within the time frame allotted under 
Sec.  52.103(e), and any Commission determinations on petitions as 
provided under Sec.  52.103(f). The NRC requests public comment on the 
benefits and impacts (including information collection and reporting 
burdens) that would occur if the proposed requirements were adopted.
    Commenters' Response: Some commenters agreed with this concept. 
However, they do not support a rule change because they believe a rule 
change is not necessary. Rather, they believe that the concept should 
be implemented via guidance rather than a rule change. Additionally, 
following the initial notification, a licensee should be required to 
submit a follow-up 30-day notification only if the schedule in the 
prior notification has changed. It would be unnecessarily burdensome to 
require a licensee to submit notifications every 30 days stating that 
the schedule has not changed.
    NRC Response: The NRC has decided to amend Sec.  52.103(a) in the 
final rule to ensure that the combined license holder will notify the 
NRC of its scheduled date for initial loading of fuel into a plant no 
later than 270 days before the scheduled date, and will notify the NRC 
of updates to its schedule every 30 days thereafter. The notification 
will facilitate timely NRC publication of the notice required under 
Sec.  52.103(a), completion of hearings within the time frame allotted 
under Sec.  52.103(e), and completion of any Commission determinations 
on petitions filed under Sec.  52.103(f). The NRC believes that the 
update notifications when the schedule has not changed will not be 
burdensome. Additional discussion on this issue is provided in Section 
V.C.8.b of the supplementary information in this final rule.
    Question 7: As discussed in Section IV.C.6.f of the March 13, 2006, 
proposed rule, the NRC is proposing to modify Sec.  52.79(a) to add 
requirements for descriptions of operational programs that need to be 
included in the final safety analysis report (FSAR) to allow a 
reasonable assurance finding of acceptability. This proposed amendment 
is in support of the Commission's direction to the staff in SRM-SECY-
02-0067 dated September 11, 2002, ``Inspections, Tests, Analyses, and 
Acceptance Criteria for Operational

[[Page 49358]]

Programs (Programmatic ITAAC),'' that a combined license applicant was 
not required to have ITAAC for operational programs if the applicant 
fully described the operational program and its implementation in the 
combined license application. In this SRM, the Commission stated:

    [a]n ITAAC for a program should not be necessary if the program 
and its implementation are fully described in the application and 
found to be acceptable by the NRC at the COL stage. The burden is on 
the applicant to provide the necessary and sufficient programmatic 
information for approval of the COL without ITAAC.

    Accordingly, the NRC is proposing in the final part 52 rulemaking 
to add requirements to Sec.  52.79 that combined license applications 
contain descriptions of operational programs. In doing so, the 
Commission has taken into account NEI's proposal to address SRM-SECY-
04-0032 in its letter dated August 31, 2005 (ML052510037). However, the 
NRC is concerned that there may be operational program requirements 
that it has not captured in its proposed Sec.  52.79. Therefore, the 
NRC is requesting public comment on whether there are additional 
required operational programs that should be described in a combined 
license application that are not identified in proposed Sec.  52.79. If 
additional required operational programs are identified, the Commission 
is considering adding them to Sec.  52.79 in the final rule.
    Commenters' Response: Some commenters believed that requirements 
for operational programs were sufficient as proposed, and that no 
additional operational programs needed to be described in the COL 
application.
    NRC Response: The NRC does not agree that no additional operational 
programs need to be described in a COL application. During the 
preparation of the final rule, the NRC discovered that several of the 
operational programs listed in SECY-05-0197 (October 28, 2005) were not 
addressed in proposed Sec.  52.79. To ensure the list of requirements 
for the contents of applications is complete, the NRC is adding several 
new provisions to address operational programs in the final rule. 
Specifically, the NRC is adding requirements to Sec.  52.79 for COL 
applicants to include a description of: (1) the process and effluent 
monitoring and sampling program required by appendix I to 10 CFR part 
50 [Sec.  52.79(a)(16)(ii)]; (2) a training and qualification plan in 
accordance with the criteria set forth in appendix B to 10 CFR part 73 
[Sec.  52.79(a)(36)(ii)]; (3) a description of the radiation protection 
program required by Sec.  20.1101 [Sec.  52.79(a)(39)]; (4) a 
description of the fire protection program required by Sec.  50.48 
[Sec.  52.79(a)(40)]; and (5) a description of the fitness-for-duty 
program required by 10 CFR part 26 [Sec.  52.79(a)(44)]. During the 
preparation of the final rule, the NRC also noticed that it had not 
completely implemented the Commission's direction regarding the 
treatment of operational programs in a COL application because it had 
failed to add requirements to address program implementation in its 
revisions to Sec.  52.79(a). Therefore, in the final rule, the NRC has 
added requirements to address the implementation of all operational 
programs required to be described in a COL application. This is 
consistent with the Commission's direction to the staff in SRM-SECY-02-
0067 (September 11, 2002, ML022540755) that a combined license 
applicant was not required to have ITAAC for operational programs if 
the applicant fully described the operational program and its 
implementation in the combined license application.
    Question 8: Backfitting--reproduce backfitting requirements in part 
52. The NRC notes that the backfitting provisions applicable to various 
part 52 processes are contained in both part 50 and part 52 and, 
therefore, the proposed language for Sec.  50.109 cross-references to 
applicable provisions of part 52, which may be confusing. The NRC is 
considering adopting in the final rule an alternative which would 
remove from Sec.  50.109 the backfitting provisions applicable to the 
licensing and approval processes in part 52, and place them in part 52. 
There are two possible approaches for doing so: the first would be for 
the NRC to establish a general backfitting provision in part 52 
applicable exclusively to the licensing and approval processes in part 
52. Under this approach, each licensing and approval process in part 52 
would be the subject of a backfitting section in a new subpart of part 
52 (e.g., Sec.  52.201 for standard design approvals, etc.). The 
existing backfitting provisions applicable to early site permits and 
design certification would be transferred to the relevant sections in 
the new subpart. The second approach would be to ensure that each 
subpart of part 52 contains the backfitting provisions applicable to 
the licensing or approval process in that subpart. The NRC is 
considering adopting these alternative approaches in the final rule and 
requests public comment on whether either of these administrative 
approaches is preferable to the approach in the proposed rule.
    Commenters' Response: Some commenters stated that NRC's alternative 
approach to addressing backfitting was unnecessary to clarify the 
application of the backfit rule to part 52 actions. Commenters stated 
that the proposed rule included adequate references to Sec.  50.109 and 
in the various subparts of part 52, making replication of this language 
elsewhere unnecessary. If the NRC deemed the inclusion of such 
information necessary, several commenters suggested each subpart in 
part 52 include its own standards for backfitting to avoid confusion.
    NRC Response: The NRC has decided to revise Sec.  50.109 to include 
the conforming changes necessary to reflect part 52, rather than 
adopting a backfitting provision in part 52, because no commenter 
favored the alternative approach of adopting a backfitting provision in 
part 52, and both approaches are legally equivalent.
    Question 9: The Commission is considering adopting in the final 
part 52 rulemaking an alternative to the re-proposed rule's approach 
for addressing new and significant environmental information with 
respect to matters addressed in the ESP environmental impact statement 
(EIS) which require supplementation.\2\ As a separate matter, the 
Commission is also considering adopting in the final part 52 rulemaking 
an analogous requirement for addressing new information necessary to 
update and correct the emergency plan approved by the ESP, the ITAAC 
associated with EP, or the terms and conditions of the ESP with respect 
to emergency preparedness, or new information materially changing the 
Commission's determinations on emergency preparedness matters 
previously resolved in the ESP. To implement either or both of these 
alternatives, the Commission is also evaluating whether several 
additional concepts should be adopted in the final rulemaking. The two 
alternatives, as well as the additional implementing concepts, are 
described below. The Commission emphasizes that it may, with respect to 
the alternative addressing updating environmental information and 
emergency preparedness information, adopt either or both alternatives 
in the final part 52

[[Page 49359]]

rulemaking, in place of or in addition to the proposed rule's 
alternative of conducting the updating in each combined license 
proceeding. Under the option where multiple alternatives for updating 
environmental and emergency preparedness information would be allowed, 
the Commission proposes that the decision be left to the combined 
license applicant as to which alternative to pursue. Commenters are 
requested to address: (1) the advantages and disadvantages of adopting 
each alternative for updating environmental and emergency preparedness 
information in an ESP proceeding as opposed to the proposed rule's 
alternative of conducting the updating in each combined license 
proceeding; (2) whether the Commission should only allow updating of 
environmental and emergency preparedness information in an ESP 
proceeding or in a COL proceeding, but not both; and (3) if the 
Commission allows updating in either an ESP proceeding or in a COL 
proceeding, whether it should be an option for the COL applicant to 
decide which update process to pursue. The Commission believes it may 
allow COL applicants the option of deciding whether to update 
environmental and emergency preparedness information in either an ESP 
proceeding or in a COL proceeding in order to afford the COL applicant 
the determination which approach best satisfies their business and 
economic interests.
---------------------------------------------------------------------------

    \2\ The scope of environmental information that must be 
supplemented is limited to the matters which were addressed in the 
original EIS for the ESP. Thus, for example, if the ESP applicant 
chose not to address need for power (as is allowed under Sec.  
52.18), the combined license applicant need not address need for 
power in its environmental report (ER) to update the ESP EIS, and 
the NRC need not determine whether there is new and significant 
information with respect to need for power as part of the updating 
of the ESP EIS.
---------------------------------------------------------------------------

Environmental Matters Resolved in ESP

    The Commission is considering requiring a combined license 
applicant planning to reference an ESP to submit a supplemental 
environmental report for the ESP. The supplemental environmental report 
must address whether there is any new and significant environmental 
information with respect to the environmental matters addressed in the 
ESP EIS. Based upon this information, the NRC will prepare a draft 
supplemental environmental assessment (EA) or EIS setting forth the 
agency's proposed determinations with respect to any new and 
significant information. In accordance with existing practice and 
procedure, the draft supplemental EA or EIS will be issued for public 
comment. After considering comments received from the public and 
relevant Federal and State agencies, the NRC will issue a final 
supplemental EA or EIS. Once the final supplemental EA or EIS is 
issued, the ESP finality provisions in proposed Sec.  52.39 would apply 
to the matters addressed in the supplemental EA or EIS, and those 
matters need not be addressed in any combined license proceeding 
referencing the ESP. Thus, for example, if a new and significant 
environmental issue, for example, a newly-designated endangered 
species, is addressed in the supplemental ESP EIS, the matter would be 
resolved for all combined licenses referencing the ESP (unless, of 
course, there is new and significant information identified at the time 
of a subsequent referencing combined license with respect to that 
endangered species). There would be no updating of environmental 
information necessary in the combined license proceeding. The 
Commission considers this approach for updating the ESP as meeting the 
Agency's obligations under the National Environmental Policy Act 
(NEPA), without imposing undue burden on the ESP holder and the NRC 
through continuous or periodic updating, and preserving the distinction 
between the ESP and any referencing combined license proceeding. Since 
an ESP may be referenced more than once, this approach would provide 
for issue finality of the updated information and preclude the need for 
reconsideration of the same environmental issue in successive combined 
license proceedings referencing the ESP. The Commission requests public 
comment on this proposal, which would likely involve changes to 
Sec. Sec.  52.39, 51.50(c), 51.75, and 51.107 (and possibly conforming 
changes in parts 2, 51, and 52).

Emergency Preparedness Information Resolved in ESP

    The Commission is separately considering requiring a combined 
license applicant referencing an ESP to provide to the NRC new EP 
information necessary to correct inaccurate information in the ESP 
emergency plan, EP ITAAC, or the terms and conditions of the ESP with 
respect to EP. Based upon the EP information submitted by the combined 
license applicant, the NRC will, as necessary, approve changes to the 
ESP emergency plan, the EP ITAAC, or the terms and conditions of the 
ESP with respect to EP. Once the Commission has resolved the EP 
updating matters, these matters would be accorded finality under Sec.  
52.39. There would be no separate updating necessary in the combined 
license proceeding. Thus, for example, if an EP ITAAC in an ESP were 
changed by virtue of this updating process, the changed ITAAC for EP 
would be applicable to any combined license referencing the ESP whose 
ITAAC have not yet been satisfied (i.e., the amended EP ITAAC would not 
be applicable to a combined license where the Commission has made the 
Sec.  52.103(g) finding with respect to that EP ITAAC). The NRC's 
consideration of such EP information would be considered to be part of 
the ESP proceeding, and any necessary changes with respect to EP would 
therefore be deemed to be changes within the scope of the ESP. The 
Commission considers this proposal as a means for updating the ESP with 
respect to EP information in a timely fashion, without imposing undue 
burden on the ESP holder and the NRC through continuous or periodic 
updating, while preserving the distinction between the ESP and any 
referencing combined license proceeding.
    Since an ESP may be referenced more than once, this approach would 
provide for issue finality of the updated information and preclude the 
need for reconsideration of the same issue in successive combined 
license proceedings referencing the ESP. The Commission requests 
comment whether this approach should be adopted by the Commission in 
the final rulemaking, which will likely involve changes to Sec.  52.39 
(and possible conforming changes in Sec.  50.47, 50.54, and 10 CFR part 
50, appendix E).

ESP Updating in Advance of Combined License Application Submission

    To minimize the possibility that the ESP updating process may 
adversely affect a combined license proceeding referencing that ESP, 
the Commission proposes to require the combined license applicant 
intending to reference an ESP to submit its application to update the 
ESP with respect to EP and/or environmental information no later than 
18 months before the submission of its combined license application. 
The Commission believes that the 18-month lead time is sufficient to 
complete the NRC's regulatory consideration of the updating, such that 
the combined license applicant will be able to prepare its application 
to reflect the updated ESP. The Commission also recognizes that there 
may be increased regulatory complexity under this approach, as well as 
the possibility that resources may be unnecessarily expended if the 
potential combined license applicant ultimately decides not to proceed 
with its application. The Commission requests public comment on whether 
the 18-month lead time is appropriate, whether the time should be 
decreased or increased, or whether the Commission should simply require 
that the ESP update application be filed no later than simultaneously 
with the filing of the combined license application. Based upon the 
public comments, the Commission will adopt one of these

[[Page 49360]]

alternatives, if it decides that updating of environmental and/or EP 
matters should be accomplished in an ESP proceeding, as opposed to the 
combined license proceeding in which the ESP is referenced.

Expanding the Scope of Resolved Issues After ESP Issuance

    The Commission is also considering whether the final rule should 
include provisions addressing how the ESP holder may request, at any 
time after the issuance of the ESP, that additional issues be resolved 
and given finality under Sec.  52.39. For example, the holder of the 
ESP which does not include an approved emergency plan, may wish to 
submit complete emergency plans for NRC review and approval. Such a 
request is not explicitly addressed in either the current or re-
proposed subpart A to part 52, although it would be reasonable to treat 
that request as an application to amend the ESP.
    The Commission requests public comment on whether the Commission 
should adopt in the final rule new provisions in subpart A to part 52 
that would explicitly address requests by the ESP holder to amend the 
early site permit to expand the scope of issues which are resolved and 
given issue finality under Sec.  52.39. The Commission is also 
considering whether, as part of the ESP updating process discussed 
previously, the ESP holder/combined license applicant should be allowed 
to request an expansion of issues which are resolved and given issue 
finality.
    If the Commission were to allow an ESP holder/combined license 
applicant to expand the scope of resolved issues in the ESP update 
proceeding, the Commission believes that the 18-month time period for 
filing the updating application in the ESP proceeding may be 
insufficient, and is considering adopting in the final rule a 24-month 
(2-year) period for filing the ESP updating application, where the ESP 
holder/combined license applicant seeks to expand the scope of resolved 
issues. The Commission seeks public comment on whether, in such cases, 
the Commission should require in the final rule an 18- or 24-month 
period, or some other period, for submitting its ESP updating 
application.

Approval in ESP of Process and Criteria for Updating ESP After Issuance

    The Commission requests public comment whether the Commission 
should adopt in the final rulemaking provisions affording the ESP 
applicant the option of requesting NRC approval of procedures and 
criteria for identifying and assessing new and significant 
environmental information, and/or new information necessary to update 
and correct the emergency plan approved by the ESP, the ITAAC 
associated with emergency preparedness (EP), or the terms and 
conditions of the ESP with respect to emergency preparedness, or 
otherwise materially changing the Commission's determinations on 
emergency preparedness matters previously resolved in the ESP. These 
procedures and criteria, if approved as part of the ESP issuance, could 
be used by any combined license applicant referencing the ESP to 
identify the need to update the ESP with respect to environmental and/
or emergency preparedness information. There would be no need for the 
NRC to review the adequacy of the ESP holder/combined license 
applicant's process and criteria for determining whether new 
information is of such importance or significance so as to require 
updating; the NRC review could thereby be focused solely on whether the 
ESP holder's updated information, or determination that there is no 
change in either an environmental or emergency preparedness matter, was 
correct and adequate. Under this proposal, Sec.  52.17 and/or Sec.  
51.50(b) would be amended to incorporate such a process for ``pre-
approval'' of ESP updating procedures and criteria.
    While NRC approval of updating procedures and criteria would be 
reflected in the ESP, the Commission does not believe that the ESP 
itself must contain the procedures and criteria in order to be accorded 
finality under Sec.  52.39. An ESP holder/combined license applicant 
need not comply with any or all of the updating process and criteria, 
and would be free to use (and justify) other procedures or criteria in 
the ESP updating proceeding. Naturally, there would be no finality 
associated with such departures from the ESP-approved procedures and 
criteria.
    The Commission does not believe that either subpart A of part 52 or 
an ESP with the contemplated approved updating procedures and criteria 
should contain a ``change process'' akin to Sec.  50.59, allowing the 
ESP holder to make changes to the approved updating procedures and 
criteria without NRC review and approval. Any change (other than 
typographic and administrative corrections) should require an amendment 
to the ESP. However, the Commission seeks public comment on whether a 
different course should be adopted in the final rule.
    The Commission recognizes that any NRC-approved procedures and 
criteria for updating environmental and/or emergency preparedness 
information in an ESP updating process as described previously, would 
be equally valid for updating such information under the updating 
provisions in the re-proposed rule. The Commission requests comments on 
whether, if the Commission adopts in the final rulemaking the re-
proposed rule's concept of updating in the combined license proceeding, 
the Commission should provide the ESP applicant with the option of 
seeking NRC approval of the procedures and criteria for updating 
environmental and/or emergency preparedness information in a combined 
license proceeding which references the ESP.

Public Participation in ESP Updating Process

    The Commission is considering two ways for allowing public 
participation in the updating process, if the updating alternative is 
adopted in the final rule. One approach would be to allow interested 
persons to challenge the proposed updating by submitting a petition, 
analogous to that in proposed Sec.  52.39(c)(2), which would be 
processed in accordance with Sec.  2.206. This approach would be most 
consistent with the existing provisions in Sec.  52.39, inasmuch as 
updating of an ESP is roughly equivalent to a request that the terms 
and conditions of an ESP be modified. A consequence of this approach is 
that the potential scope of matters which may be raised is not limited 
to those ESP matters which the ESP holder/combined license applicant 
and the NRC conclude must be updated.
    The other approach that the Commission may adopt is to treat any 
necessary updating as an amendment to the ESP, for which an opportunity 
to request a hearing is provided. This approach would limit the scope 
of the hearing to those matters for which an amendment is required. 
Where the ESP holder does not request an amendment on the basis that no 
updating is necessary with respect to a matter, an interested person 
could not intervene with respect to that matter. A consequence of this 
approach is that, under the Commission's regulations in 10 CFR part 2 
and its current practice, a hearing granted on any amendment 
necessitated by the updating process would be more formalized than a 
hearing accorded under the Sec.  2.206 petition process. The Commission 
requests public comment on the approach that the Commission should 
adopt, together with the reasons for the commenter's recommendation.
    Commenters' Response: Several commenters believed an ESP holder 
should not be required to update the

[[Page 49361]]

information in the ESP application. These commenters stated that the 
proposal to require updating would add an unnecessary additional level 
of review (and possibly hearings) with little or no additional benefit 
(i.e., the COL applicant would still be under the obligation to update 
the information provided by the ESP holder). Some commenters contended 
that an updating requirement would only serve to erode the finality and 
certainty provided by the ESP, thereby defeating one of the purposes of 
an ESP. These commenters also believed that an updated requirement 
would run counter to NRC regulations. Some commenters stated that while 
the ESP is in effect, the NRC cannot change or impose new requirements, 
including emergency planning requirements, unless it determines that a 
modification is necessary either to bring the permit or the site into 
compliance with the NRC'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. Some commenters argued that the proposed 18-month updating 
requirement may not be feasible. A commenter gave the following 
example, ``under the NRC's current schedule for the existing ESP 
applications for North Anna and Grand Gulf, the ESPs will not be issued 
until 2007, shortly before the planned COL applications for those 
sites. This would result in insufficient time for the updating 
envisioned by the NRC, and it would be unfair to those applicants to 
require them to delay their COL applications to accommodate the 
updating process. Additionally, the proposed updating process would be 
inconsistent with Sec.  52.27(c), which permits a COL application to 
reference an ESP application.''
    Several commenters agreed with NRC's proposal to provide the ESP 
holder with the option of requesting an ESP amendment in order to 
resolve issues that were not addressed at the ESP stage or to achieve 
finality on updated information. These commenters also suggested that a 
COL applicant should be able to reference an application for an ESP 
amendment that is pending approval by the NRC similar to the process 
that already exists in 10 CFR 52.27(c).
    Several commenters expressed the belief that a COL applicant should 
be able to make changes or updates to ESP emergency planning 
information without NRC approval in accordance with the criteria in 10 
CFR 50.54(q) just as the remaining safety information can be revised 
under Sec.  50.59 once it has been reviewed and approved. These 
commenters also stated that this revised information should not be 
considered as an ``amendment'' submitted under Sec.  50.90 for review 
and approval, but rather should be considered to be information 
equivalent to that provided under Sec.  50.71(e) for information.
    NRC Response: Upon consideration of the public comments on this 
subject, the NRC has decided not to require updating of ESP information 
prior to receipt of a COL application referencing the ESP. The NRC is 
retaining the proposed rule structure for dealing with new EP and 
environmental information at the COL stage. The NRC believes this 
structure will provide for the most effective and efficient use of NRC 
and applicant resources. The NRC is, however, making revisions to the 
final rule to allow for voluntary changes to an ESP by the ESP holder 
through the license amendment process. Specifically, the NRC is making 
revisions to Sec. Sec.  50.90 and 50.92 to include ESPs within the 
scope of these requirements. The NRC is also adding a new provision to 
Sec.  52.39 to allow ESP holders to make changes to the ESP, including 
changes to the SSAR, under the license amendment process. These changes 
will provide ESP holders with additional flexibility to resolve issues 
that were not addressed in the original ESP review and to achieve 
finality on new information. The NRC does not believe it is necessary 
to add rule language to address the situation where a COL applicant 
references an ESP for which there is an amendment review pending before 
the NRC. The NRC will address these situations on a case-by-case basis.
    Question 10: The Commission is considering adopting in the final 
part 52 rulemaking a new provision in Sec.  50.71 that would require 
combined license holders to update the PRA [probabilistic risk 
assessment] submitted with the combined license application 
periodically throughout the life of the facility on a schedule similar 
to the schedule for final safety analysis report (FSAR) updates (i.e., 
at least every 24 months) or, alternatively, on a schedule to coincide 
with every other refueling outage. Updates would be required to ensure 
that the information included in the PRA contains the latest 
information developed. The PRA update submittal would be required to 
contain all the changes necessary to reflect information and analyses 
submitted to the Commission by the licensee or prepared by the licensee 
pursuant to Commission requirement since the submittal of the original 
PRA, or as appropriate, the last update to the PRA under this section. 
The submittal would be required to include the effects of all changes 
made in the facility or procedures as reflected in the PRA; all safety 
analyses and evaluations performed by the licensee either in support of 
approved license amendments or in support of conclusions that changes 
did not require a license amendment in accordance with Sec.  
50.59(c)(2) or, in the case of a license that references a certified 
design, in accordance with Sec.  52.98(c); and all analyses of new 
safety issues performed by or on behalf of the licensee at Commission 
request. The Commission requests stakeholder feedback on whether such a 
requirement should be added to the Commission's regulations and, if so, 
what is an appropriate update schedule.
    Commenters' Response: Several commenters noted that the proposed 
rule did not include a frequency for updating the PRA. These commenters 
noted that the Commission stated that PRA scope and methods should be 
addressed in guidance, not in regulations (SRM on SECY-05-0203). These 
commenters stated that they believed that PRA update frequency should 
also be addressed in guidance rather than regulations. These commenters 
indicated a frequency of once every two operating cycles would be 
reasonable and consistent with existing requirements in 10 CFR 
50.69(e).
    Additionally, some commenters stated the plant-specific PRA used to 
support a COL application that references a design certification would 
essentially be the design certification PRA. These commenters expressed 
the belief that the plant-specific PRA would be updated to be 
consistent with the PRA scope and quality standards 6 months before the 
COL was issued as plant-specific design and as-built information was 
developed during construction. Some commenters argued that this would 
allow (1) an updated plant-specific PRA that was representative of the 
as-built plant to be completed, and (2) an updated plant-specific PRA 
that would be available prior to fuel load for NRC audit and to support 
plant operations. These commenters suggested that the update of the 
plant-specific PRA during construction was a matter suitable for 
guidance.
    Some commenters expressed confusion over the NRC proposal to 
require PRA updates to reflect safety analyses and evaluations 
performed by the licensee, and analyses of new safety issues performed 
by or on behalf of the licensee at the NRC's request. These commenters 
stated that new analyses

[[Page 49362]]

and evaluations were often performed using design-basis assumptions 
that may not be appropriate for a PRA. These commenters suggested that 
only new analyses that impact the PRA warrant consideration, and 
requested guidance and examples be developed regarding the information 
that should be considered when updating the plant-specific PRA.
    NRC Response: As discussed in further detail in Section V.D.6.b of 
this document, the Commission is adopting requirements to require 
maintenance of a PRA, and periodic upgrades every 4 years, by a COL 
holder beginning at the time of initial operation. These PRAs and 
upgrades are not required to be submitted to the NRC, but instead 
should be maintained by the licensee for NRC inspection.
    Question 11: In a letter dated July 5, 2005, the Nuclear Energy 
Institute (NEI) submitted comments on the proposed rule for the AP1000 
design certification. Many of those comments have generic applicability 
to the three pre-existing design certification rules (DCRs) in 
appendices A through C of 10 CFR part 52. In the final AP1000 
rulemaking (January 27, 2006; 71 FR 4464), the Commission adopted some 
of the NEI-recommended changes, while rejecting others (71 FR 4465-
4468). For those changes that were adopted in the final AP1000 design 
certification, the Commission indicated that it would consider making 
the same changes to the existing design certifications in appendices A 
through C. For those changes that were not adopted in the final AP1000 
design certification, the Commission stated that it would reconsider 
the issues in the part 52 rulemaking, and if the Commission changes its 
position and the change is adopted, the Commission would make the 
change for all four design certifications, including the AP1000.
    The Commission is considering amending the appropriate sections in 
each DCR based on the comments below. The Commission considers most of 
NEI's proposed changes to be consistent with proposed Sec.  
52.63(a)(1); in particular, the Commission believes that the proposed 
changes would satisfy the ``reduces unnecessary regulatory burden'' 
criterion in proposed Sec.  52.63(a)(1)(iii). The few remaining 
changes, constituting editorial clarifications or corrections 
reflecting the Commission's original intent, are not subject to the 
existing change restrictions in Sec.  52.63(a)(1). Accordingly, the 
Commission believes that it has authority to incorporate some or all of 
the NEI-proposed changes into appendices A through D in the final part 
52 rulemaking.
    The Commission also requests comments on whether some of NEI's 
proposed changes accepted in the AP1000 design certification and 
proposed for inclusion in appendices A through C should not be included 
in those appendices in the final part 52 rulemaking because they are 
unnecessary, or because they would not meet one or more of the change 
criteria in proposed Sec.  52.63(a)(1). The Commission is also 
assessing whether NEI's proposed changes which were not adopted in the 
AP1000 final rulemaking should be adopted in the final part 52 
rulemaking for all four design certifications, including the AP1000. 
The Commission is particularly interested in whether there are reasons, 
other than those presented by NEI, for adopting those changes, as well 
as commenter's views on the Commission's reasons for rejecting the NEI 
proposals as stated in the final AP1000 design certification 
rulemaking.
    a. NEI recommended modification of the generic technical 
specification definition in Section II.B to clarify that bracketed 
information is not part the DCRs for purposes of the change processes 
in Section VIII.C, and an exemption is not required for plant-specific 
departures from bracketed information. The Commission stated in the 
section-by-section analysis for the AP1000 DCR (71 FR 4464) that some 
generic technical specifications and investment protection short-term 
availability controls contain values in brackets. The values in 
brackets are neither part of the DCR nor are they binding. Therefore, 
the replacement of bracketed values with final plant-specific values 
does not require an exemption from the generic technical specifications 
or investment protection short-term availability controls. The 
Commission believes that including this guidance in each DCR is not 
necessary. The Commission requests comment on whether there are 
countervailing considerations that favor inclusion of this provision in 
the DCRs.
    b. NEI recommended modification of the Tier 2 definition in Section 
II.E to clarify that bracketed information in the investment protection 
short-term availability controls is not part of Tier 2 and thus not 
subject to the Section VIII.B change controls. The Commission stated in 
the section-by-section analysis for the AP1000 DCR (71 FR 4464) that 
some generic technical specifications and investment protection short-
term availability controls contain values in brackets. The values in 
brackets are neither part of the DCR nor are they binding. Therefore, 
the replacement of bracketed values with final plant-specific values 
does not require an exemption from the generic technical specifications 
or investment protection short-term availability controls. The 
Commission believes that including this guidance in each DCR is not 
necessary. The Commission requests comment on whether there are 
countervailing considerations that favor inclusion of this provision in 
the DCRs.
    c. NEI recommended modification of the requirement in Section 
VIII.C.2 to delete the phrase ``or licensee'' because that phrase 
conflicted with the requirement in Section VIII.C.6. The Commission 
believes that generic technical specifications should not apply to 
holders of a combined license because the license will include plant-
specific technical specifications. Therefore, the Commission is 
considering amending each of the DCRs to delete the phrase ``or 
licensee'' from Section VIII.C.2 and requests public comment on this 
approach.
    d. NEI recommended modification of the requirement in Section 
VIII.C.6 to delete the last portion, which states ``changes to the 
plant-specific technical specifications will be treated as license 
amendments under 10 CFR 50.90.'' NEI stated that this sentence is not 
necessary because it is redundant with Sec.  50.90. It is not necessary 
to include a provision in each DCR stating that a license amendment is 
necessary to make changes to technical specifications in order to 
render this a legally-binding requirement inasmuch as Section 182.a of 
the AEA requires that technical specifications be part of each license. 
The Commission believes that clarity and understanding by the reader is 
enhanced by repeating this statutory requirement in each DCR. The 
Commission requests comment on whether there are countervailing 
considerations that favor non-inclusion of this provision in the DCRs, 
and may decide to remove this provision in the final part 52 
rulemaking.
    e. NEI recommended modification of the requirement in Section X.A.1 
to require the design certification applicant to include all generic 
changes to the generic technical specifications and other operational 
requirements in the generic DCD. The Commission believes that inclusion 
of changes to the generic technical specifications and other 
operational requirements will enhance the generic DCD and facilitate 
its use by referencing applicants. The Commission is considering 
amending each of the DCRs to include the generic technical 
specifications and other operational requirements in the generic

[[Page 49363]]

DCD and requests public comment on this approach.
    f. NEI recommended modification of the requirement in Sections 
IV.A.2 and IV.A.3 to be consistent with respect to inclusion of 
information in the plant-specific DCD, or explain the difference 
between ``include'' (IV.A.2) and ``physically include'' (IV.A.3). The 
Commission is considering amending each of the DCRs to use the same 
term in both provisions, and requests public comment on this approach.
    g. NEI recommended modification of the definition in Section II.E.1 
to exclude the design-specific probabilistic risk assessment (PRA) and 
the evaluation of the severe accident mitigation design alternatives 
(SAMDA) from Tier 2 information. The Commission believes that the PRA 
and SAMDA evaluations do not need to be included in Tier 2 information 
because they are not part of the design basis information. The 
Commission is considering amending each of the DCRs to modify the 
definition of Tier 2, and requests public comment on this approach.
    h. NEI recommended modification of the requirement in Section III.E 
to use ``site characteristics'' consistently, instead of ``site-
specific design parameters.'' The Commission intends to use the term 
``characteristics'' to refer to actual values and ``parameters'' to 
refer to postulated values. The Commission has proposed amending 
Section III.E of each DCR to use ``site characteristics,'' and requests 
public comment on this approach.
    i. NEI recommended modification of Section IV.A.2 to clarify the 
use of ``same information'' and ``generic DCD'' in that requirement. 
The Commission has proposed amending Section IV.A.2 of each DCR to use 
the phrase ``same type of information'' to avoid confusion, and 
requests public comment on this approach.
    j. NEI recommended modification of the requirement in Section 
VIII.B.6.a to delete the sentence ``The departure will not be 
considered a resolved issue, within the meaning of Section VI of this 
appendix and 10 CFR 52.63(a)(4),'' in order to be consistent with the 
requirement in Section VI.B.5 of the DCRs. The Commission believes that 
departures from Tier 2* information should not receive finality or be 
treated as resolved issues within the meaning of section VI.B of the 
DCRs. The Commission requests comment on whether departures from Tier 
2* information should be considered a resolved issue, and may decide to 
remove this provision from each DCR.
    k. NEI recommended modification of Section VIII.C.3 to require the 
NRC to meet the backfit requirements of 10 CFR 50.109 in addition to 
the special circumstances in 10 CFR 2.758(b) (which has now been 
designated as Sec.  2.335) in order to require plant-specific 
departures from operational requirements. The Commission believes that 
plant-specific departures should not have to meet the backfit 
requirement for generic changes. The Commission will have to 
demonstrate that special circumstances, as defined in Sec.  2.335, are 
present in order to require a plant-specific departure. The Commission 
requests comment on whether there are countervailing considerations 
that would favor modification of this provision in the DCRs.
    l. NEI recommended modification of the requirement in Section 
VIII.C.4 to include a requirement that operational requirements that 
were not completely reviewed and approved by the NRC should not be 
subject to any Tier 2 change controls, e.g., exemptions. However, NEI 
previously proposed that requested departures from Chapter 16 by an 
applicant for a COL require an exemption (62 FR 25808; May 12, 1997). 
The Commission believes that the requirement for an exemption applies 
to technical specifications and operational requirements that were 
completely reviewed and approved in the design certification rulemaking 
(see 62 FR 25825). The Commission requests comment on whether 
departures from technical specifications and operational requirements 
that were not completely reviewed and approved should also require an 
exemption.
    m. NEI recommended modification of the requirement in Section 
VIII.C.4 to delete the sentence ``The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing,'' in order to be consistent with the requirement 
in Section VI.B.5 of the DCRs. The Commission believes that exemptions 
from operational requirements should not receive finality or be treated 
as resolved issues (refer to Section VI.C of the DCRs). The Commission 
requests comment on whether exemptions from operational requirements 
should be considered a resolved issue, and may decide to modify this 
provision in each DCR.
    n. NEI recommended modification of the requirement in Section 
IX.B.1 to better distinguish between NRC staff ITAAC conclusions under 
proposed Sec.  52.99(e) and the Commission's ITAAC finding under 
proposed Sec.  52.103(g). The Commission believes that individual DCRs 
should not address the scope of the NRC staff's activities with respect 
to ITAAC verification. This is a generic matter that, if it is to be 
addressed in a rulemaking, is more appropriate for inclusion in subpart 
C of part 52 dealing with combined licenses. The Commission requests 
comment on whether there are countervailing considerations that favor 
clarification of this provision in the DCRs.
    o. NEI recommended modification of the language in Section IX.B.3 
to make editorial changes for clarity, e.g., ``ITAAC will expire'' vs. 
``their expiration will occur.'' The Commission believes that the 
original rule language is acceptable. The Commission requests comment 
on whether there are countervailing considerations that favor 
clarification of this provision in the DCRs.
    p. NEI recommended modification of the language in Sections X.B.1 
and X.B.3 to clarify references to the design control documents, e.g., 
``plant-specific'' vs. ``generic.'' The Commission agrees that the 
references to plant-specific and generic DCD should be clarified in 
Sections X.B.1 and X.B.3 to ensure that the requirements in these 
sections are properly implemented by applicants referencing the design 
certification rules. The Commission requests public comment on this 
prospective modification.
    Commenters' Response: Several commenters recommended the NRC 
incorporate the NEI recommendations on the AP1000 rule, cited specific 
NEI recommendations (71 FR 12834-12836), and made additional 
suggestions and clarifications.
    Regarding NEI recommendations (a) and (b), several commenters 
suggested it would be sufficient if the statements of considerations 
for the final rule provided the requested clarification, rather than 
the rule itself.
    Regarding NEI recommendation (f), several commenters supported the 
use of the term ``include'' rather than ``physically include'' for 
requirements in Section IV of the design certification rules concerning 
content of COLAs. These commenters also requested clarification on the 
permissible method of incorporating the generic DCD into the plant-
specific DCD portion of the COL application's final safety analysis 
report (FSAR), because the current NRC position has apparently ``led to 
considerable confusion'' among COL preparers. These commenters noted 
that in the statements of consideration accompanying the AP1000 final 
rule, NEI recommended a change to the Definitions (Section III.B of 
that rule, 71 FR 4466). These commenters stated the NRC staff disagreed 
with this

[[Page 49364]]

recommendation, saying that ``the generic DCD should also be part of 
the FSAR, not just incorporated by reference, in order to facilitate 
the NRC staff's review of any departures or exemptions.'' Some 
commenters believed that this NRC position was in conflict with the 
former Sec.  52.79(b), which states that the COL application's FSAR 
``may incorporate by reference the final safety analysis report for a 
certified standard design,'' and with Sec.  50.32, which provides for 
incorporation by reference to eliminate repetitive information. Some 
commenters argued that although the wording had been altered, the 
ability to incorporate by reference was preserved in proposed 
Sec. Sec.  52.79 (b) and (c), respectively. These commenters claimed 
this interpretation of incorporation was validated by NRC staff during 
the Draft Regulatory Guide (DG)-1145 workshops. These commenters stated 
support for this interpretation and requested the NRC explicitly 
describe that either approach is acceptable.
    In discussing NEI recommendation (j), several commenters mentioned 
Section VIII.B.6.a of the design certification rules, which states that 
an applicant who references the design certification rule must obtain 
NRC approval for departures from Tier 2* information in the generic 
DCD. Some commenters believed that this section states the departure is 
not considered to be a resolved issue under Section VI of the design 
certification rules. Some commenters indicated this was inconsistent 
with Section VI.B.5 of the design certification rules, which states 
that license amendments are considered to be resolved. These commenters 
expressed support for the revision of Section VIII.B.6. of the design 
certification rules to make it consistent with Section VIII.B.5 of the 
design certification rules. These commenters stated that departures 
from Tier 2* information that are reviewed and approved by the NRC in 
the combined license proceeding should have finality for the plant in 
question.
    With respect to NEI recommendation (k), several commenters 
expressed concern that Section VIII.C.3 of the design certification 
rules ``inappropriately'' allowed the NRC to make changes to 
operational requirements in the DCD without satisfying the backfit 
requirements in Sec.  50.109. These commenters stated that the 
operational requirements in the design certification proceeding should 
be afforded the protection of the backfit rule. Some commenters 
supported a revision to Section VIII.C.3 of the design certification 
rules to include a reference to Sec.  50.109 for these changes.
    In the discussion of NEI recommendations (l) and (m), several 
commenters mentioned Section VIII.C.4 of the design certification 
rules, which states a COL applicant must request an exemption from the 
NRC if the applicant wants to depart from the generic technical 
specifications or other operational requirements. These commenters 
described this requirement as ``unduly burdensome.'' These commenters 
noted that the operational requirements do not have finality under 
Section VI.C of the design certification rules, and that no basis 
existed for applying such a change control process to a COL applicant 
seeking to change operational requirements. Some commenters cited 
Section VIII.B.5 of the design certification rules, which states a COL 
applicant may depart from final design-related provisions in the design 
certification rule using a ``Sec.  50.59-like'' process, and argued 
that imposing an exemption process with respect to operational 
provisions was not required. Some commenters recommended Section 
VII.C.4 be amended to state that a departure from an operational 
requirement does not require an exemption.
    Several commenters mentioned information from NEI's September 30, 
2003, response to the 2003 part 52 notice of proposed rulemaking. These 
commenters expressed support for the need to add a basic definition of 
``departure'' to the DCRs to be consistent with adding the definition 
of ``departure from a method of evaluation,'' and stated that both 
should be based on Regulatory Guide 1.187. The commenters stated, ``The 
basic definition of `change or departure' should precede the definition 
of departure from a method of evaluation.'' Some commenters recommend 
adding the new definition as paragraph II.G and renaming the final two 
paragraphs as II.H and II.I.
    NRC Response: In response to Question 11.a, the NRC has decided 
that modification of the generic technical specification definition in 
Section II.B of the DCRs is not necessary. As stated in the section-by-
section analysis for the AP1000 DCR (71 FR 4475; January 27, 2006):

    Some generic technical specifications and investment protection 
short-term availability controls contain values in brackets [ ]. The 
brackets are placeholders indicating that the NRC's review is not 
complete, and represent a requirement that the applicant for a 
combined license referencing the AP1000 DCR must replace the values 
in brackets with final plant-specific values. The values in brackets 
are neither part of the design certification rule nor are they 
binding. Therefore, the replacement of bracketed values with final 
plant-specific values does not require an exemption from the generic 
technical specifications or investment protection short-term 
availability controls.

    The NRC believes that the above guidance resolves NEI's concern 
regarding bracketed information in the generic technical 
specifications.
    Regarding Question 11.b, the NRC has decided that modification of 
the Tier 2 definition in Section II.E of the DCRs is not necessary. The 
NRC believes that the previously mentioned guidance resolves NEI's 
concern regarding bracketed information in the investment protection 
short-term availability controls located in the Tier 2 information.
    Regarding Question 11.c, the NRC agrees with NEI's recommendation 
and has decided to delete the phrase ``or licensee'' from Section 
VIII.C.2 of the DCRs because the generic technical specifications will 
not apply to holders of a combined license.
    Regarding Question 11.d, the NRC has decided not to modify the rule 
language in Section VIII.C.6 of the DCRs, which states that ``changes 
to the plant-specific technical specifications will be treated as 
license amendments under 10 CFR 50.90.'' The Commission believes that 
this statement provides clarity to this requirement.
    Regarding Question 11.e, the NRC agrees with NEI's recommendation 
and has decided to modify the requirement in Section X.A.1 of the DCRs. 
The Commission believes that the inclusion of changes to the generic 
technical specifications and other operational requirements in the 
generic design control document (DCD) will enhance the DCD and 
facilitate its use by referencing applicants.
    Regarding Question 11.f, the NRC has decided to modify Section IV 
of the DCRs to consistently use the term ``include'' rather than 
``physically include'' as recommended by NEI.
    Several commenters also requested clarification on the permissible 
method of incorporating the generic DCD in the plant-specific DCD 
portion of the COL application's final safety analysis report (FSAR), 
because the NRC position has apparently ``led to considerable 
confusion'' among COL preparers. The NRC is requiring COL applicants 
that reference the DCRs in appendices A through D of part 52 to include 
the generic DCD in the application's FSAR, in order to facilitate the 
NRC staff's review of any departures or exemptions. Simply 
incorporating the generic DCD by reference into the FSAR is not

[[Page 49365]]

sufficient because of the manner in which these existing DCDs were 
submitted to the NRC. Therefore, Section IV.A.2 of the DCRs overrides 
Sec. Sec.  50.32 and 52.79(d). The NRC is hopeful that future DCRs will 
not have to use this special requirement.
    Regarding Question 11.g, the NRC agrees with NEI's recommendation 
and has decided to modify the definition of Tier 2 in Section II.E.1 of 
the DCRs to exclude the design-specific probabilistic risk assessment 
(PRA) and the evaluation of the severe accident mitigation design 
alternatives (SAMDAs). The NRC believes that the PRA and SAMDA 
evaluations do not need to be included in Tier 2 because they are not 
part of the design basis information. Also, the revised Section II.E.1 
is now consistent with the requirements in the new Sec.  52.80 
regarding PRA and SAMDA evaluations.
    Regarding Question 11.h, the NRC agrees with NEI's recommendation 
to use ``site characteristics'' instead of ``site-specific design 
parameters'' in Section III.E of the DCRs. This modification of the 
rule language in Section III.E was made in the proposed rule and, 
therefore, no change was made to the final rule.
    Regarding Question 11.i, the NRC agrees with NEI's recommendation 
to clarify the rule language in Section IV.A.2.a of the DCRs and adopts 
the phrase ``same type of information'' to avoid confusion. An 
applicant for a combined license must submit, as part of its 
application, a plant-specific DCD that contains the same type of 
information and uses the same organization and numbering as the generic 
DCD. This organization will facilitate the NRC staff's review of the 
plant-specific DCD. The NRC recognizes that the plant-specific DCD will 
not contain the exact, same information as the generic DCD because the 
plant-specific DCD will be modified and supplemented by the applicant's 
exemptions, departures, and COL action items.
    Regarding Question 11.j, the NRC does not agree with NEI's request 
to modify the requirement in Section VIII.B.6.a of the DCRs. The 
Commission decided during the initial design certification rulemakings 
that departures from Tier 2* information (by an applicant) would not 
receive finality or be treated as a resolved issue within the meaning 
of Section VI of the DCR. This provision applies to applicants for a 
combined license and the new information is subject to litigation in 
the same manner as other plant-specific issues in the licensing 
hearing. Also, Tier 2* information has the same safety significance as 
Tier 1 information and would have received the Tier 1 designation, 
except that NRC decided to provide more flexibility for this type of 
information.
    Regarding Question 11.k, the NRC does not agree with NEI's 
recommendation to modify Section VIII.C.3 of the DCRs. NEI requests 
that the NRC meet the backfit requirements in Sec.  50.109 in addition 
to the special circumstances in Sec.  2.335 in order to require plant-
specific departures from operational requirements. In the original 
design certification rulemakings, the Commission decided on different 
standards for changes made under Section VIII.C (see Section VI.C and 
62 FR 25805; May 12, 1997). The Commission has decided that plant-
specific departures should not have to meet the backfit requirements in 
Sec.  50.109.
    Regarding Question 11.l, the NRC does not agree with NEI's 
recommendation to modify Section VIII.C.4 of the DCRs. The requirement 
in Section VIII.C.4 for an applicant to request an exemption applies to 
generic technical specifications and operational requirements that were 
comprehensively reviewed and finalized in the design certification 
rulemaking (see 62 FR 25825; May 12, 1997). Because this guidance is 
already set forth in the section-by-section discussion for the DCRs, 
the NRC has decided that changes to the rule language are not 
necessary.
    Regarding Question 11.m, the NRC does not agree with NEI's 
recommendation to delete the last sentence from Section VIII.C.4 of the 
DCRs. This sentence applies to applicants for a combined license and 
the new information is subject to litigation in the same manner as 
other plant-specific issues in the licensing hearing. The Commission 
believes that exemptions from operational requirements should not 
receive finality or be treated as resolved issues (refer to Section 
VI.C of the DCRs).
    Regarding Question 11.n, the NRC does not agree with NEI's 
recommendation to modify Section IX.B.1 of the DCRs. The NRC has 
decided that individual DCRs should not address the scope of the NRC 
staff's activities with respect to ITAAC verification. This is a 
generic matter that was addressed in Sec.  52.99(e).
    Regarding Question 11.o, the NRC does not agree with NEI's request 
to clarify the phrase ``their expiration will occur'' in Section IX.B.3 
of the DCRs. The NRC has decided that the original rule language is 
acceptable.
    Regarding Question 11.p, the NRC agrees with NEI's recommendation 
to clarify references to the DCDs in Sections X.B.1 and X.B.3 of the 
DCRs. The references to plant-specific and generic DCD were revised in 
Sections X.B.1 and X.B.3 to ensure that the requirements in these 
sections will be properly implemented by applicants and licensees that 
reference the design certification rules.
    Question 12: The Commission is considering adopting in the final 
part 52 rulemaking a new provision that would either require combined 
license applicants to submit a detailed schedule for the licensee's 
completion of ITAAC or require the combined license holder to submit 
the schedule for ITAAC completion. Delaying submission of the schedule 
would allow the combined license holder to develop the schedules based 
on more accurate information regarding construction schedules and would 
allow the schedule to be submitted at a time when it would be most 
useful to the NRC for planning purposes. The Commission could require 
that applicants submit the schedule within a specified time prior to 
scheduled COL issuance--for example, 3 months prior to COL issuance or 
within some time period (e.g., 6 months or 1 year) after COL issuance. 
In addition, the Commission is considering an additional element to 
this provision that would require that the licensee submit an update to 
the ITAAC schedule within 12 months after combined license issuance and 
that the licensee update the schedule every 6 months until 12 months 
before scheduled fuel load, and monthly thereafter until all ITAAC are 
complete. The Commission is considering adopting these requirements to 
support the NRC staff's inspection and oversight with respect to ITAAC 
completion, and to facilitate publication of the Federal Register 
notices of successful completion of ITAAC as required by proposed Sec.  
52.99(e). The Commission requests stakeholder comment on whether such a 
provision, with or without the update element, should be added to the 
Commission's regulations and which time frame for submission of the 
schedule would be most beneficial.
    The Commission is also considering adopting a provision that would 
establish a specific time by which the licensee must complete all ITAAC 
to allow sufficient time for the NRC staff to verify successful 
completion of ITAAC, without adversely affecting the licensee's 
scheduled date for fuel load and operation. The Commission considers 
``60 days prior to the schedule date for initial loading of fuel'' to 
be a

[[Page 49366]]

reasonable time period by which all ITAAC must be completed. However, 
the Commission requests comments on whether this time period would 
provide too much or too little time prior to scheduled fuel load. 
Alternatively, the Commission is considering a 30-day or a 90-day time 
period prior to scheduled fuel load. The 30-day option would allow more 
flexibility for the licensee to complete ITAAC late in construction but 
would require immediate action on the part of the NRC (to determine if 
the final ITAAC were completed successfully and, if so, for the 
Commission to make its finding under Sec.  52.103(g)) so as not to 
delay scheduled fuel load. The 90-day option would reduce licensee 
flexibility to complete ITAAC late in construction but would ensure 
that the NRC had ample time to make its determination on the final 
ITAAC for Commission review of all ITAAC under Sec.  52.103(g). The 
Commission requests stakeholder comment on whether a provision 
requiring completion of ITAAC within a certain time period prior to 
scheduled fuel load should be added to the Commission's regulations.
    Commenters' Response: Several commenters believed it was 
unnecessary to include a requirement for either the COL applicant or 
the COL holder to submit a detailed schedule for ITAAC completion 
because a COL applicant could provide only a progressively less 
accurate estimated completion schedule. Some commenters stated that the 
COL holder would have schedules at the site, and those schedules would 
be available for NRC review. Some commenters believed that COL holders 
would interact and coordinate with the NRC to ensure that NRC had 
sufficient information to schedule its inspection activities for ITAAC, 
making a regulatory requirement for submission of a schedule 
unnecessary. In addition, these commenters noted that a COL applicant/
holder would likely consider detailed schedule information to be 
proprietary information, which would make its submission inappropriate.
    Several commenters also stated it was ``wrong'' to require 
completion of ITAAC in a set time period prior to fuel loading and 
operation. These commenters indicated that a COL holder would likely 
complete several ITAAC within 30 days of fuel loading and argued that 
the NRC should not abrogate responsibility by imposing a mandatory 
delay on licensees. Some commenters stated the importance of the NRC 
providing the appropriate level of inspections and reviews to prevent 
delays in fuel load and emphasized the high cost (stated to be on the 
order of $1,000,000 per day) of such delay. Some commenters suggested 
the NRC should be in a position to make a Sec.  52.103(g) finding 
promptly following the completion of the last ITAAC.
    NRC Response: The NRC has decided to amend Sec.  52.99 to require 
licensees to submit their schedules for completing the inspections, 
tests, or analyses in the ITAAC. The NRC has added a new paragraph (a) 
in Sec.  52.99 that requires a licensee to submit to the NRC, no later 
than 1 year after issuance of the combined license or at the start of 
construction as defined in 10 CFR 50.10, whichever is later, its 
schedule for completing the inspections, tests, or analyses in the 
ITAAC. Licensees are required to submit updates to the ITAAC schedule 
every 6 months thereafter and, within 1 year of its scheduled date for 
initial loading of fuel, licensees must submit updates to the ITAAC 
schedule every 30 days until the final notification is provided to the 
NRC under Sec.  52.99(c)(1). Although commenters did not believe that a 
requirement for submission of a schedule was necessary, the NRC 
believes it is necessary to ensure that the NRC has sufficient 
information to plan all of the activities necessary for the NRC to 
support the Commission's determination as to whether all of the ITAAC 
have been met prior to initial operation. In the event that licensees 
consider their schedule information to be proprietary, they can request 
that the schedule be withheld from public disclosure under Sec.  2.390. 
If an applicant claims that its construction schedule information 
submitted to the NRC is proprietary, and requests the NRC to withhold 
that information under the Freedom of Information Act (FOIA), the NRC 
will consider that request under the existing rules governing FOIA 
disclosure in 10 CFR 2.309(a)(4).
    The NRC has also decided to amend Sec.  52.99(c) which requires the 
licensee to notify the NRC that the prescribed inspections, tests, and 
analyses in the ITAAC have been or will be completed and that the 
acceptance criteria have been met. The NRC is revising Sec.  
52.99(c)(1) in the final rule to more closely follow the language of 
Section 185b. of the AEA and to clarify that the notification must 
contain sufficient information to demonstrate that the prescribed 
inspections, tests, and analyses have been performed and that the 
prescribed acceptance criteria have been met. The NRC is adding this 
clarification to ensure that combined license applicants and holders 
are aware that (1) it is the licensee's burden to demonstrate 
compliance with the ITAAC and (2) the NRC expects the notification of 
ITAAC completion to contain more information than just a simple 
statement that the licensee believes the ITAAC has been completed and 
the acceptance criteria met. The NRC expects the notification to be 
sufficiently complete and detailed for a reasonable person to 
understand the bases for the licensee's representation that the 
inspections, tests, and analyses have been successfully completed and 
the acceptance criteria have been met. The term ``sufficient 
information'' requires, at a minimum, a summary description of the 
bases for the licensee's conclusion that the inspections, tests, or 
analyses have been performed and that the prescribed acceptance 
criteria have been met. The NRC plans to prepare regulatory guidance, 
in consultation with interested stakeholders, to explain how the 
functional requirement to provide ``sufficient information'' with 
regard to ITAAC submittals could be met.
    The NRC is also revising Sec.  52.99(c) by adding a new paragraph 
(c)(2) requiring that, if the licensee has not provided, by the date 
225 days before the scheduled date for initial loading of fuel, the 
notification required by paragraph (c)(1) of this section for all 
ITAAC, then the licensee shall notify the NRC that the prescribed 
inspections, tests, or analyses for all uncompleted ITAAC will be 
performed and that the prescribed acceptance criteria will be met prior 
to operation (consistent with the Section 185.b requirement that the 
Commission, ``prior to operation,'' find that the acceptance criteria 
in the combined license are met). The notification must be provided no 
later than the date 225 days before the scheduled date for initial 
loading of fuel. It is the licensee's burden to demonstrate that it 
will comply with the ITAAC and it must provide sufficient information 
to demonstrate that the prescribed inspections, tests, or analyses will 
be performed and the prescribed acceptance criteria for the uncompleted 
ITAAC will be met. The term ``sufficient information'' requires, at a 
minimum, a summary description of the bases for the licensee's 
conclusion that the inspections, tests, or analyses will be performed 
and that the prescribed acceptance criteria will be met. In addition, 
``sufficient information'' includes, but is not limited to, a 
description of the specific procedures and analytical methods to be 
used for performing the inspections, tests, and analyses and 
determining that the acceptance criteria have been met.
    Paragraph (e) has been revised to require that the NRC make 
available to

[[Page 49367]]

the public the notifications to be submitted under Sec.  52.99(c)(1) 
and (c)(2), no later than the Federal Register notice of intended 
operation and opportunity for hearing on ITAAC under Sec.  52.103(a). A 
conforming change is included in Sec.  2.105(b)(3) to require that the 
Sec.  52.103(a) notice reference the public availability of the Sec.  
52.99(c)(1) and (2) notifications. The NRC is requiring that the 
paragraph (c)(2) notification be made 225 days before the date 
scheduled for initial loading of fuel, in order to ensure that the 
licensee notifications are publicly available through the NRC document 
room and online through the NRC Web site at the same time that the 
Sec.  52.103(a) notice is published in the Federal Register. The NRC's 
goal is to publish that notice 210 days before the date scheduled for 
fuel loading, but in all cases the Sec.  52.103(a) notice would be 
published no later than 180 days before the scheduled fuel load, as 
required by Section 189.a(1)(B) of the AEA.
    Commenters did not support addition of a requirement on completion 
of ITAAC in a set time period prior to fuel load and the NRC has not 
included a provision requiring the completion of all ITAAC by a certain 
time prior to the licensee's scheduled fuel load date. Instead, the NRC 
has decided to modify the concept slightly by requiring the licensee to 
submit, with respect to ITAAC which have not yet been completed 225 
days before the scheduled date for initial loading of fuel, additional 
information addressing whether those inspections, tests, and analyses 
will be successfully completed and the acceptance criteria met before 
initial operation. In the case where the licensee has not completed all 
ITAAC by 225 days prior to its scheduled fuel load date, the NRC 
expects the information that the licensee submits related to 
uncompleted ITAAC to be sufficiently detailed such that the NRC can 
determine what activities it will need to undertake to determine if the 
acceptance criteria for each of the uncompleted ITAAC have been met, 
once the licensee notifies the NRC that those ITAAC have been 
successfully completed and their acceptance criteria met. In addition, 
the NRC is adopting the requirements in paragraphs (c)(1) and (c)(2) to 
ensure that interested persons will have sufficient information to 
address the Atomic Energy Act, Section 189.a(1), threshold for 
requesting a hearing with respect to both completed and as-yet 
uncompleted ITAAC. The NRC plans to prepare regulatory guidance 
providing further explanation of what constitutes ``sufficient 
information'' that must be submitted under paragraphs (c)(1) and (c)(2) 
demonstrating that the inspections, tests, or analyses for ITAAC have 
been or will be completed and the acceptance criteria for the ITAAC 
have been or will be met. The NRC expects that any contentions 
submitted by prospective parties regarding uncompleted ITAAC would 
focus on any inadequacies of the specific procedures and analytical 
methods described by the licensee under paragraph (c)(2), in the 
context of the findings called for by Sec.  52.103(b)(2).\3\
---------------------------------------------------------------------------

    \3\ Inasmuch as the ITAAC themselves have already been approved 
by the NRC and their adequacy may not be challenged except under the 
provisions of Sec.  52.103(f), a contention which alleges the 
deficiency of the ITAAC is not admissible under Sec.  52.103(b).
---------------------------------------------------------------------------

    The NRC notes that, even though it did not include a provision 
requiring the completion of all ITAAC by a certain time prior to the 
licensee's scheduled fuel load date, the NRC will require some period 
of time to perform its review of the last ITAAC once the licensee 
submits its notification that the ITAAC has been successfully completed 
and the acceptance criteria met. In addition, the Commission itself 
will require some period of time to perform its review of the staff's 
conclusions regarding all of the ITAAC and the staff's recommendations 
regarding the Commission finding under Sec.  52.103(g). Therefore, 
licensees should structure their construction schedules to take into 
account these time periods. The NRC staff intends to develop regulatory 
guidance on the licensee's completion and NRC verification of ITAAC and 
will provide estimates of the time it expects to take to verify 
successful completion of various types of ITAAC. The NRC expects that 
such guidance, along with frequent communication with licensees during 
construction, will provide licensees with adequate information to plan 
initial fuel loading and related activities.
    Question 13: ML Hearings. As discussed in Section IV.F.6 of the 
March 13, 2006, proposed rule, the Commission proposes, as a matter of 
policy and discretion, that the Commission hold a ``mandatory'' hearing 
(i.e., a hearing which, under NRC requirements in 10 CFR part 2, is 
held regardless of whether the NRC receives any hearing requests or 
petitions to intervene) in connection with the initial issuance of 
every manufacturing license. The Commission believes that Section 
189.a.(1)(A) of the AEA does not require that a hearing be held in 
connection with the initial issuance of a manufacturing license. 
Nonetheless, there are several reasons for the Commission to require by 
rule, as a matter of discretion, a mandatory hearing. A manufacturing 
license may be viewed as analogous to a construction permit--a 
regulatory approval for which Section 189 of the AEA specifically 
requires that a hearing be held. Even though the Commission's 
regulations did not address the hearing requirements for manufacturing 
licenses, the Commission noticed a ``mandatory'' hearing in connection 
with the only manufacturing license application ever received by the 
Agency. Offshore Power Systems (Floating Nuclear Power Plants), 38 FR 
34008 (December 10, 1973). Accordingly, proposed Sec. Sec.  2.104 and 
52.163 require that a mandatory hearing be held in each proceeding for 
initial issuance of a manufacturing license. However, the Commission 
recognizes that there may be countervailing considerations weighing 
against Commission adoption of a rulemaking provision mandating that a 
hearing be held in connection with the initial issuance of every 
manufacturing license where there has been no stakeholder interest in a 
hearing. If there is no stakeholder interest in a hearing, transparency 
and public confidence would not appear to be relevant considerations in 
favor of holding a mandatory hearing. Considerations of regulatory 
efficiency and effectiveness would be paramount, and would weigh 
against holding of a mandatory hearing. The Commission requests 
comments on whether the Commission should exercise its discretion to 
provide by rule an opportunity for hearing, rather than a mandatory 
hearing, and the reasons in favor of providing an opportunity for 
hearing as opposed to holding a mandatory hearing. Based upon the 
public comments, the Commission may adopt a final rule which deletes 
Sec.  2.104(f), revises Sec.  2.105 (governing the content of a Federal 
Register notice of proposed action where a mandatory hearing is not 
held under Sec.  2.104) to add, as appropriate, references to issuance 
of manufacturing licenses, and revised Sec.  52.163 to provide an 
opportunity for hearing rather than a mandatory hearing in connection 
with the initial issuance of a manufacturing license.
    Commenters' Response: Several commenters stated there was no need 
to require mandatory hearings for manufacturing licenses, or that the 
need for such hearings was unclear. These commenters expressed the 
belief that such hearings were not an appropriate method for reviewing 
and resolving

[[Page 49368]]

technical issues. Some commenters advised that the decision to request 
a hearing be left to either the NRC staff or stakeholders.
    NRC Response: As stated in the statement of considerations for the 
March 13, 2006, proposed rule, the NRC acknowledges that hearings on 
initial issuances of manufacturing licenses are not required by the AEA 
(71 FR 12814). The NRC also agrees with the general premise of the 
commenters that adjudicatory hearings may not be the best approach for 
resolving technical design issues--especially in uncontested 
proceedings. Indeed, the NRC removed the opportunity for adjudicatory-
style hearings for design certifications as part of the 2004 changes to 
10 CFR part 2 (January 14, 2004; 69 FR 2182). The primary 
responsibility for determining the safety of an application is with the 
NRC staff, and not the presiding officer. This is true regardless of 
whether the proceeding is contested or uncontested. Public confidence 
would not seem to be enhanced in any significant manner by the holding 
of a hearing where there is no request that the NRC hold a hearing. 
Accordingly, the NRC has decided not to adopt in the final part 52 rule 
a requirement for a ``mandatory'' hearing in connection with issuance 
of manufacturing licenses.
    Question 14: As discussed in Section IV.C.5.g of the statements of 
consideration of the March 13, 2006, proposed rule, the proposed rule 
would amend the special backfit requirement in 10 CFR 52.63(a)(1) to 
provide the Commission with the ability to make changes to the design 
certification rules (DCRs) or the certification information in the 
generic design control documents that reduce unnecessary regulatory 
burdens. The underlying rationale for this provision also forms the 
basis for amending the Tier 2 change process in the three DCRs 
(appendices A, B, and C of part 52) to incorporate the revised change 
criteria in 10 CFR 50.59.
    The Commission is considering adopting an additional provision 
[Sec.  52.63(a)(1)(iv)] in the final rule that would allow amendments 
of design certification rules to incorporate generic resolutions of 
design acceptance criteria (DAC) or other design information without 
meeting the special backfit requirement in the current Sec.  
52.63(a)(1). The applicants for the current DCRs requested use of DAC 
in lieu of providing detailed design information for certain areas of 
their nuclear plant designs, for example, instrumentation and control 
systems. Under the proposed requirements, a generic change to design 
certification information would have to meet the special backfit 
requirement of Sec.  52.63(a)(1) or reduce an unnecessary regulatory 
burden while maintaining protection to public health and safety and the 
common defense and security. The Commission adopted this special 
backfit requirement to restrict changes and to require that everyone 
meet the same backfit standard for generic changes, thereby ensuring 
that all plants built under a referenced DCR would be standardized. By 
allowing a DCR amendment to include generic resolutions of DAC or other 
design information, the Commission would enhance its goals for design 
certification, for example, early resolution of all design issues and 
finality for those issue resolutions, which would avoid repetitive 
consideration of design issues in individual combined license 
proceedings.
    There are currently three ways of resolving generic design issues: 
(1) the combined license applicant that references a DCR could submit 
plant-specific resolutions in its application, which could result in 
loss of standardization; (2) a vendor could submit generic resolutions 
in topical reports that, if approved, could but would not be required 
to be referenced in a combined license application; or (3) the 
Commission could exempt itself from the special backfit requirement in 
Sec.  52.63(a)(1) and amend the DCR to incorporate a generic 
resolution, which could result in multiple rulemakings to revise each 
DCR to incorporate each generic resolution. The Commission intends that 
any review of a proposed generic resolution would be performed under 
the regulations that are applicable and in effect at the time that the 
approval or amendment is completed.
    Therefore, the NRC is requesting public comments on: (1) whether a 
provision should be added to Sec.  52.63(a)(1) to allow generic 
amendments to design certification information that meet applicable 
regulations in effect at the time that the rulemaking is completed; and 
(2) whether the generic resolutions should be incorporated into a DCR 
without meeting a backfit requirement, which would provide for 
completion of the design certification information and facilitate 
standardization, or whether an application for a generic amendment 
should be required to meet a backfit requirement (e.g., Sec.  50.109).
    Commenters' Response: Some commenters stated that revisions to NRC 
regulations should include the current 10 CFR 52.63, which they 
believed should allow the original design certification applicant (or 
its successor) to obtain amendments to the design certification rule. 
These commenters believed current regulations prevented any amendment 
to a design once the design has been certified by rule (10 CFR 
52.63(a)(1)). Some commenters stated that the design certification 
applicant should be able to petition the NRC for, and obtain, an 
amendment to the design certification rule to incorporate 
``beneficial'' changes to the design certification, including: (1) 
Design changes that would result in significant improvements in safety; 
(2) design changes that would result in significant improvements in 
efficiency, reliability and/or economics; (3) design changes that 
result from continuing engineering or design work or are required 
because of lack of availability of components specified in the original 
design certification; and (4) design changes necessary to correct minor 
errors in the original design certification. Some commenters also 
suggested that where proposed changes involved changes to Tier 2, the 
design certification applicant should be able to make such changes 
using a Sec.  50.59-like change process. One commenter noted that 
changes to allow an amendment to the final design certification could 
potentially simplify COL applications, reduce NRC staff resource 
burden, and help assure standardization across the industry.
    NRC Response: The NRC has decided to include an amendment process 
in the final rule that: (1) Reduces unnecessary regulatory burden and 
maintains protection to public health and safety and common defense and 
security; (2) provides the detailed design information necessary to 
resolve selected design acceptance criteria; (3) corrects material 
errors in the certification information; (4) substantially increases 
overall safety, reliability, or security of a facility and the costs of 
the change are justified; or (5) contributes to increased 
standardization of the certification information, without meeting the 
special backfit requirement in Sec.  52.63(a)(1)(ii). These amendments 
will apply to all plants that have referenced or will reference the 
DCR. The NRC believes that these amendments will enhance 
standardization by further completing or correcting the certification 
information. A detailed discussion of the amendment process is provided 
in Section V.C.7.g of the Supplementary Information of this document.
    Question 15: In Section IV.J of the supplementary information of 
the March 13, 2006, proposed rule, the NRC

[[Page 49369]]

outlines key principles regarding its proposal for reporting 
requirements that implement Section 206 of the Energy Reorganization 
Act, as amended, for part 52 licenses, certifications, and approvals. 
The NRC discusses that the beginning of the ``regulatory life'' of a 
referenced license, standard design approval, or standard design 
certification under part 52 occurs when an application for a license, 
design approval, or design certification is docketed. The NRC also 
cautions, however, that this does not mean that an applicant is without 
Section 206 responsibilities for pre-application activities because 
there are two aspects to the reporting requirements, namely, a 
``backward looking'' or retrospective aspect with respect to existing 
information, and a ``forward looking'' or prospective aspect with 
respect to future information. For an early site permit applicant, the 
retrospective obligation is that the early site permit holder and its 
contractors, upon issuance of the early site permit, must report all 
known defects or failures to comply in ``basic components,'' as defined 
in part 21. Under the proposed part 21 requirements presented in the 
proposed rule, the early site permit holder and its contractors are 
required to meet these requirements upon issuance of the early site 
permit. Accordingly, applicants should procure and control safety-
related design and analysis or consulting services in a manner 
sufficient to allow the early site permit holder and its contractors to 
comply with the above described reporting requirements of Section 206, 
as implemented by part 21. A similar argument applies to design 
certification applicants. Although the Commission has not proposed an 
explicit requirement imposing part 21 on applicants for an early site 
permit or design certification in the proposed rule, it is considering 
adopting such a requirement in the final part 52 rulemaking because, as 
a practical matter, the NRC has to require these applicants to 
implement a part 21 program before approval of the early site permit or 
design certification. Therefore, providing explicit part 21 
requirements for applicants would clarify the Commission's intent. The 
Commission requests stakeholder comment on whether it should, in the 
final rule, impose part 21 reporting requirements on applicants for 
early site permits and design certifications.
    Commenters' Response: Several commenters were opposed to the 
proposed changes to part 21. Some commenters stated part 21 had been in 
existence for almost 30 years, during which it was never applied to 
applicants. They complained that they were not aware, and the NRC had 
not made them aware, of problems that would warrant a change. The 
commenters noted that applicants take measures to ensure that they were 
made aware of any errors and deficiencies identified by contractors and 
suppliers for work performed on commercial nuclear projects, because 
applicants eventually become holders, and licensees and want equipment 
to operate correctly. Several commenters were also concerned that the 
proposal was contrary to the Energy Reorganization Act (ERA), which was 
the basis for part 21. They believed it would be inappropriate and 
contrary to the ERA to apply part 21 to applicants. They stated part 21 
was established to implement Sec.  206 of the ERA, which applies to 
``licensees'' and vendors, suppliers, and contractors of licensees, not 
to ``applicants.'' These commenters cited 10 CFR 21.2, stating that the 
existing regulations of part 21 apply only to entities licensed to 
possess, use, or transfer radioactive material within the United 
States, or to construct, manufacture, possess, own, operate, or 
transfer within the United States, any production or utilization 
facility or fuel storage facility. The commenter believed applicants 
did not fall within the scope of Sec.  206 of the ERA, and it was 
inconsistent with the Act to expand the scope of Sec.  21.2 to include 
applicants.
    Some commenters also noted that it had been the standard practice 
for a construction permit (CP) applicant to specify part 21 
requirements in its procurement contracts for a plant prior to issuance 
of the construction permit. Some commenters agreed with this practice 
because part 21 was applicable to such contracts once the CP was issued 
by the NRC, and expected that this ``good practice'' would be 
implemented by COL applicants as well. From a ``practical 
perspective,'' the commenters believed this negated the need to expand 
part 21 to applicants.
    Some commenters argued that the obligations for applicants to 
provide information to the NRC under proposed Sec.  52.6(a) was broader 
than the obligation in part 21, and would require applicants to update 
and correct their applications to account for the types of defects and 
noncompliances covered by part 21. These commenters stated the industry 
had no objection to proposed Sec.  52.6(a), which should therefore 
eliminate the need to apply part 21 to applicants.
    NRC Response: The Commission proposed part 21 reporting 
requirements on applicants for early site permits, design 
certifications, and standard design approvals in the proposed rule. A 
detailed discussion on the Commission's rationale for imposing these 
requirements in the final rule is provided in Section V.J of the 
supplementary information of this document.

V. Discussion of Substantive Changes and Responses to Significant 
Comments

A. Introduction

    The changes to 10 CFR Chapter I are further discussed by part. 
Changes to parts 52 and 50 are discussed first, followed by changes to 
other parts in numerical order. Within each part, general topics are 
discussed first, followed by discussion of changes to individual 
sections as necessary. In addition to the substantive changes, rule 
language was revised to make conforming administrative changes (e.g., 
identification of regulations containing information collection 
requirements in Sec.  52.11), correct typographic errors, adopt 
consistent terminology (e.g., ``makes the finding under Sec.  
52.103(g)''), correct grammar, and adopt plain English. These changes 
are not discussed further.

B. Testing Requirements for Advanced Reactors

    This rule amends Sec. Sec.  50.43, 52.47, 52.79, and 52.157 to 
achieve clarity and consistency in the testing requirements for 
advanced reactor designs and plants. This amendment requires applicants 
for a combined license, operating license, or manufacturing license 
that use new safety features but do not reference a certified advanced 
reactor design to also perform the design qualification testing 
required of certain applicants for design certification. If a combined 
license application references a certified design, the necessary 
qualification testing will have been performed under Sec.  52.47(c)(2). 
The codification of testing requirements in the original Sec.  52.47 
was a principal issue during the development of 10 CFR part 52 (see 
Section II of 54 FR 15372; April 18, 1989). The requirement to 
demonstrate the performance of new safety features for nuclear power 
plants that differ significantly from evolutionary light-water reactors 
or that use 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 new safety features will perform as 
predicted in the applicant's safety analysis report, to provide 
sufficient data to validate analytical codes, and that the effects of 
systems

[[Page 49370]]

interactions are acceptable. 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 and 
its goal of resolving all safety issues before authorizing 
construction.
    Some commenters stated that it is unnecessary to apply 
qualification testing requirements to combined license applicants. The 
Commission does not agree because, when it reformed the licensing 
process for new nuclear plants with the issuance of part 52, the 
Commission required applicants to demonstrate that new safety features 
will perform as predicted in the final safety analysis report. Although 
the focus of the NRC at that time was on applications for design 
certification, the Commission intended that testing to qualify new 
design features (proof-of-performance testing) would be required for 
all advanced reactors, including custom designs (see Question 6 at 51 
FR 24 646; July 8, 1986). Furthermore, it would make no sense for the 
Commission to require qualification testing for design certification 
applicants (so-called paper designs) and not require testing for 
applications to build and operate an advanced nuclear power plant. 
Therefore, the NRC has implemented its intent in adopting part 52 to 
resolve issues early 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 
reactor designs or plants (including nuclear reactors manufactured 
under a manufacturing license).
    This amendment also includes a requirement in Sec.  50.43(e)(2) for 
licensing a prototype plant, as defined in Sec. Sec.  50.2 and 52.1, if 
the plant is used to meet the testing requirements in Sec.  
50.43(e)(1). The new Sec.  50.43(e) states that, if a prototype plant 
is used to comply with the qualification testing requirements, the NRC 
may impose additional requirements on siting, safety features, or 
operational conditions for the prototype plant to compensate for any 
uncertainties associated with the performance of the new or innovative 
safety features in the prototype plant.
    Some commenters stated that it would be inappropriate to establish 
or impose prototype testing on combined license applicants. Although 
the Commission 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 
Advanced Reactor Policy Statement at 51 FR 24646; July 8, 1986, and the 
statement of consideration for 10 CFR part 52, 54 FR 15372; April 18, 
1989), this rule does not require the use of a prototype plant for 
qualification testing. Rather, this rule provides that if a prototype 
plant is used to qualify an advanced reactor design, then additional 
conditions may be required for the licensed prototype plant to 
compensate for any uncertainties with the unproven safety features. 
Also, the prototype plant could be used for commercial operation.

C. Changes to 10 CFR Part 52

1. Use of Terms: Site Characteristics, Site Parameters, Design 
Characteristics, and Design Parameters in Sec. Sec.  52.1, 52.17, 52.U0 
, 52.39, 52.47, 52.54, 52.79, 52.93, 52.157, 52.158, 52.167, 52.171, 
and Appendices A, B, and C to Part 52
    The NRC is revising 10 CFR part 52 to clarify the use of the terms, 
site characteristics, site parameters, design characteristics, and 
design parameters, in order to ensure that the NRC's requirements 
governing applications for and issuance of early site permits, design 
approvals, design certifications, combined licenses, and manufacturing 
licenses are expressed in clear and unambiguous terms. This final rule 
adds or revises these terms where necessary to reflect this 
clarification. Corresponding changes are made to Sec. Sec.  52.17, 
52.24, 52.39, 52.47, 52.54, 52.79, 52.93, 52.157, 52.158, 52.167, 
52.171, and Section III.E of appendices A, B, and C to part 52.
    The NRC is also adding definitions of the terms design 
characteristics, design parameters, site characteristics, and site 
parameters to Sec.  52.1 to clarify the use of these terms. Design 
characteristics are defined as the actual features of a reactor. Design 
characteristics are specified in a standard design approval, a standard 
design certification, a combined license application, or a 
manufacturing license. Design parameters are defined as the postulated 
features of a reactor or reactors that could be built at a proposed 
site. Design parameters are specified in an early site permit. Site 
characteristics are defined as the actual physical, environmental and 
demographic features of a site. Site characteristics are specified in 
an early site permit or in a final safety analysis report for a 
combined license. Site parameters are defined as the postulated 
physical, environmental and demographic features of an assumed site. 
Site parameters are specified in a standard design approval, standard 
design certification, or a manufacturing license.
    In addition, the NRC is revising Sec.  52.79 to include a 
requirement that a combined license application referencing a certified 
design must contain information sufficient to demonstrate that the 
design of the facility falls within the site characteristics and design 
parameters specified in the early site permit. Former Sec.  52.79 
included a requirement that a combined license application referencing 
an early site permit contain information sufficient to demonstrate that 
the design of the facility falls within the parameters specified in the 
early site permit. The NRC interprets parameters to mean the site 
characteristics and design parameters as defined in Sec.  52.1. The NRC 
is making similar changes to Sec. Sec.  52.39 and 52.93. The need for 
these changes became evident during NRC's review of the pilot early 
site permit applications. Because the NRC is relying on certain design 
parameters specified in the early site permit applications to reach its 
conclusions on site suitability, these design parameters will be 
included in any early site permit issued. The NRC believes that these 
changes, in the aggregate, will provide sufficient clarification on the 
use of the terms in question.
    As the NRC completes its review of the first early site permit 
applications and prepares for the submittal of the first combined 
license application, it is focusing on the interaction among the early 
site permit, design certification, and combined license processes. The 
NRC believes that its review of a combined license application that 
references an early site permit will involve a comparison to ensure 
that the actual characteristics of the design chosen by the combined 
license applicant fall within the design parameters specified in the 
early site permit. NRC review of a combined license application that 
references a design certification will involve a comparison to ensure 
that the actual characteristics of the site chosen by the combined 
license applicant fall within the site parameters in the design 
certification. Similarly, if a combined license applicant references 
both an early site permit and a design certification, the NRC will 
review the application to ensure that the site characteristics in the 
early site permit fall within the site parameters in the referenced 
design certification and that the actual characteristics of the 
certified design fall within the design parameters in the early site 
permit. For these

[[Page 49371]]

reasons, the NRC believes it is important to make the changes described 
above in order to clarify these terms and their use in part 52 
licensing processes.
2. Issuance of Combined and Manufacturing Licenses (Sec. Sec.  52.97 
and 52.167)
    Current Sec.  50.50 sets forth the NRC's authority to include 
conditions and limitations in permits and licenses issued by the NRC 
under part 50. Similar language delineating the NRC's authority in this 
regard is also set forth in Sec.  52.24 for early site permits, but is 
not included in part 52 with respect to either combined licenses or 
manufacturing licenses. There are two possible ways of addressing this 
omission: Sec.  50.50 could be revised to refer to combined licenses 
and manufacturing licenses, or provisions analogous to Sec.  50.50 
could be added to the appropriate sections in part 52 for combined 
licenses and manufacturing licenses. Inasmuch as the NRC's inclusion of 
appropriate conditions in combined licenses is not a technical matter 
per se but rather a matter of regulatory authority, the most 
appropriate location for this provision appears to be in part 52. 
Inclusion of these provisions in appropriate portions of part 52 would 
be consistent with the provision applicable to early site permits in 
Sec.  52.24. Accordingly, the NRC is adding the language in Sec.  
52.97(c) for combined licenses, and Sec.  52.167(b) for manufacturing 
licenses, which are analogous to Sec.  50.50.
3. NRC Staff Information Requests
    Section 52.47(a)(3) of the 1989 part 52 rulemaking provided that 
the NRC staff would advise the design certification applicant on 
whether there was any additional information beyond that required to be 
submitted by that section, that must be submitted. The March 2006 
proposed rule included analogous provisions (Sec. Sec.  52.17(d), 
52.79(a)(42), 52.137(a)(27), and 52.157(p)) for each of the other 
licensing and regulatory approval processes in part 52. Upon further 
consideration in response to a comment on the March 2006 proposed rule, 
the Commission has decided that these provisions are redundant to Sec.  
2.102(a), which provides the NRC staff with overall authority to 
request information to support their review of an application. 
Accordingly, Sec. Sec.  52.17(d), 52.79(a)(42), 52.137(a)(27), and 
52.157(p) of the proposed rule have not been adopted in the final rule, 
and Sec.  52.47(a)(3) is removed from part 52.
4. Changes to a Design Certification, Departures, Variances, Exemptions
    External stakeholders have expressed confusion over the years in 
public meetings and in written comments submitted under various 
circumstances with respect to the meaning of the terms, change to a 
design certification, departures, variances, and exemptions. To clarify 
the meaning of these terms, the Commission provides the following 
explanation of these terms.
a. Change to a Design Certification
    A change to a design certification is a generic change to the 
design certification information which is approved by the Commission in 
a standard design certification rule under subpart B of part 52. In the 
four design certifications currently approved by the Commission, the 
design certification information which is approved by the Commission is 
either ``certified information'' and is designated as ``Tier 1,'' or is 
``approved'' and is designated as ``Tier 2.'' The term ``generic,'' 
means that if the Commission makes a change to the design 
certification, Sec.  52.63(a) requires that the change 
(``modification'' under Sec.  52.63(a)(3)) be applied to each plant 
referencing the design certification rule.
    A change to a design certification may be distinguished from a 
departure or variance by understanding that a change is generic. 
Therefore, a change to a design certification is:
    (1) Requested by the original design certification applicant in 
accordance with 10 CFR 2.811 (see 10 CFR 2.800(c)), or by any other 
member of the public, in a petition for rulemaking under 10 CFR 2.802;
    (2) Applies to all past nuclear power reactors (including 
manufactured reactors) whose applications have referenced the design 
certification, as well as future reactors referencing the design 
certification rule; and
    (3) Requires the Commission provide an exemption to the applicant, 
if the proposed change is inconsistent with the one or more of the 
Commission's regulations.
b. Departure
    A departure as a plant-specific ``deviation'' from design 
information in either a standard design certification or a 
manufacturing license. For a design certification, a departure is a 
deviation from the certification information which is certified by the 
Commission in a standard design certification rule (for the current 
four design certification rules in appendices A through D of part 52, 
the certification information is ``Tier 1'' information). For a 
manufacturing license, a departure is a deviation from any design 
information approved in the manufacturing license, including technical 
specifications, site parameters and design characteristics, and 
interface requirements.\4\ A departure may be distinguished from a 
change to a standard design certification rule (i.e., a change to Tier 
1 or Tier 2 information in a design certification rule) or a change to 
the design approved in a manufacturing license by recalling that a 
departure is plant-specific. Therefore, a departure:
---------------------------------------------------------------------------

    \4\ As discussed in the section-by-section discussion for Sec.  
52.171, a departure requested by a holder of a combined license 
referencing a manufactured reactor must be in the form of a license 
amendment, but the criteria for determining the request will be the 
exemption criteria in Sec.  52.7 even though the departure itself 
may not involve an exemption.
---------------------------------------------------------------------------

    <bullet> Concerns certified design information or manufacturing 
license information.
    <bullet> Is requested by the applicant/licensee referencing a 
design certification or the use of a manufactured reactor.
    <bullet> Applies only to the design of the nuclear power reactor 
referencing the design certification or the manufactured reactor for 
which a departure is sought by the applicant/licensee.
    <bullet> Requires the applicant/licensee to obtain an exemption 
from the referenced design certification if the proposed departure is 
inconsistent with one or more of the Commission's regulations. The 
exemption would be granted under the provisions of Sec.  52.7 (which 
references the same criteria for the granting of exemptions that are 
set forth in Sec.  50.12).
c. Variance
    A variance is a plant-specific ``deviation'' from one or more of 
the site characteristics, design parameters, or terms and conditions of 
an early site permit, or from the site safety analysis report. A 
variance to an early site permit is analogous to a departure to a 
standard design certification, in that it is plant-specific. Therefore, 
a variance:
    (1) Concerns information addressed in an early site permit;
    (2) Is requested by the applicant referencing an early site permit;
    (3) Applies only to the construction permit or combined license 
referencing the early site permit; and
    (4) Requires the applicant to also obtain an exemption from the 
Commission's regulations if the proposed variance is inconsistent with 
one or more of the Commission's regulations.

[[Page 49372]]

d. Exemption
    An exemption is a Commission-granted dispensation from compliance 
with one or more of the Commission's rules and regulations which would 
otherwise apply to an entity, a license, permit or other approval such 
as a standard design certification rule. Exemption from the 
requirements in part 26, or from the requirements in any particular 
design certification rule would be provided under Sec.  52.7. Exemption 
from an underlying technical requirement in part 50 would be provided 
under Sec.  50.12. This would be true even in the course of Commission 
adoption of a design certification rule. For example, if the design 
certification did not, at the time of final rulemaking, comply with a 
technical requirement in part 50, the Commission would provide an 
exemption to that requirement as part of the final design certification 
rulemaking. Moreover, if the nature of the technical requirement is 
such that a subsequent applicant referencing the design certification 
would need an exemption from compliance with the requirement as applied 
to the applicant, then the Commission would include the exemption in 
the design certification rule itself.
5. General Provisions
a. Section 52.0, Scope; Applicability of 10 CFR Chapter I Provisions
    The Commission is redesignating former Sec.  52.1, Scope, as Sec.  
52.0, Scope; applicability of 10 CFR Chapter I provisions, in order to 
add additional sections in the General Provisions portion of part 52. 
As discussed elsewhere, the Commission has decided general provisions, 
common to all substantive parts in 10 CFR Chapter I, should be added to 
part 52. To provide enough section numbers, it is necessary to 
redesignate former Sec.  52.1 as Sec.  52.0.
    Paragraph (a) of Sec.  52.0 is derived from the text of former 
Sec.  52.1, but is revised to include standard design approvals and 
manufacturing licenses within the scope of part 52, and to remove 
references to Section 104.b of Atomic Energy Act of 1954 (AEA), thereby 
providing that licenses issued under part 52 are licenses issued under 
Section 103 of the AEA. After passage of the 1970 amendments to the 
AEA, all licenses for commercial nuclear power plants with construction 
permits issued after the date of the amendments were required to be 
issued as Section 103 licenses. The NRC interprets the 1970 amendment 
as requiring combined licenses under Section 185 to be issued as 
Section 103 licenses.\5\ Accordingly, the NRC is revising the scope of 
part 52 to limit its applicability to licenses issued under Section 103 
of the AEA.
---------------------------------------------------------------------------

    \5\ This may be an academic distinction, in light of the Energy 
Policy Act of 2005, Pub. L. No. 109-58, which removed the need for 
antitrust reviews of new utilization facilities.
---------------------------------------------------------------------------

    Paragraph (b) of Sec.  52.0 is a new provision that makes clear 
that the regulations in 10 CFR Chapter I apply to a holder of, or 
applicant for an approval, certification, permit, or license issued 
under part 52 and that any license, approval, certification, or permit, 
issued under 10 CFR part 52 must comply with these regulations. The 
need for this paragraph was determined as a result of the July 3, 2003 
(68 FR 40026) proposed rule on part 52. In that proposed rule, the 
Commission proposed 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. This proposal responded to a letter dated 
November 13, 2001, from the Nuclear Energy Institute (NEI), which 
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.

    Section 52.5, as proposed in 2003, would have clarified that the 
general provisions in 10 CFR part 50 were also applicable to the new 
licensing processes for early site permits, standard design 
certifications, and combined licenses in part 52 (as well as the 
licensing and approval processes in appendices M, N, O, and Q which 
were added to part 52 by the 1989 part 52 rulemaking). Although the 
general provisions in part 50 did not specifically refer to the 
additional licensing processes in 10 CFR part 52 (and no changes to the 
language of those general provisions was proposed), the Commission 
believed that proposed Sec.  52.5 would make clear that a holder of, or 
applicant for an approval, certification, permit, or license issued 
under part 52 must also comply with those general provisions.
    However, few commenters on the July 2003 proposed rule believed 
that the proposed Sec.  52.5 would provide greater clarity. On the 
contrary, some commenters indicated that Sec.  52.5 was overly broad 
and would impose burdensome and seemingly inappropriate new 
requirements on applicants for design certifications that were 
unwarranted.
    Accordingly, in the March 2006 proposed rule, the Commission 
proposed a different approach, viz., making conforming changes to all 
of the regulations in 10 CFR Chapter I to specify their applicability 
to the relevant part 52 regulatory processes, and to add proposed Sec.  
52.0(b) to make clear that the regulations in 10 CFR Chapter I apply to 
the relevant part 52 regulatory processes, and holders and applicants 
under part 52. The Commission did not receive any comments calling into 
question the legality of this approach, or otherwise questioning the 
clarity of the proposed regulatory language. Accordingly, the 
Commission is adopting this approach in the final part 52, including 
Sec.  52.0(b).
    As discussed elsewhere in this document, the NRC is retaining 
appendix N in part 52, and revising this appendix to apply to part 52 
combined licenses. The provisions of appendix N to part 52 concern 
applicants for combined licenses under part 52. Therefore, the 
applicability language in Sec.  52.0, by referring to ``licenses'' 
under part 52, need not specifically refer to appendix N to part 52.
b. Section 52.1, Definitions
    Section 52.1 (formerly, Sec.  52.3) is revised by adding 
definitions for decommission, license, licensee, major feature of the 
emergency plans, manufacturing license, modular design, prototype 
plant, and standard design approval. A definition of decommission, 
which is identical to that in 10 CFR part 50, is added to part 52 
because the final part 52 rulemaking addresses decommissioning of 
nuclear power reactors with combined licenses under part 52. 
Definitions of license and licensee are added to facilitate the use of 
these terms throughout part 52. These definitions were derived from the 
definitions in Sec.  2.4, but were modified to reflect the regulatory 
processes in part 52. The definitions of these terms in part 2 are 
modified to be consistent with the definitions in part 52, and the 
definitions of these terms are added in part 50, to ensure consistency 
among parts 2, 50, and 52. Definitions of manufacturing license and 
standard design approval are added to part 52 so that each of these 
part 52 license types are defined.
    A definition of modular design is added to explain the type of 
modular reactor design which is the subject of the second sentence of 
Sec.  52.103(g). That provision is added to 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

[[Page 49373]]

(PRISM) designs, consisting of three or four nuclear reactors in a 
single power block with a shared power conversion system. During the 
period that the power block is under construction, the NRC could 
separately authorize operation for each nuclear reactor when each 
reactor and all of its necessary support systems were completed. In 
view of the several definitions of ``modular reactor'' which are used 
within the nuclear industry, the Commission intends to avoid future 
disputes regarding the intended applicability of Sec.  52.103(g) by 
defining the term, modular design, for purposes of part 52.
    The definition of major feature of the emergency plans is being 
added in the final rule, based on commenters' responses to Question 2 
in Section V of the Supplementary Information of the 2006 proposed 
rule, to clarify what is meant by this term as it is used in Sec. Sec.  
52.17, 52.18, 52.39, and 52.79. The definition states that a major 
feature of the emergency plans means an aspect of those plans necessary 
to: (1) address in whole or part, one or more of the sixteen standards 
in Sec.  50.47(b), or (2) describe the emergency planning zones as 
required in Sec.  50.33(g). The goal of the ``major features'' option 
in Sec.  52.17(b) is an NRC finding that the proposed major features 
are acceptable as elements of a complete and integrated emergency plan 
that would be considered later, when the early site permit is 
referenced in a license application. This is not the same level of 
finality as the ``reasonable assurance'' finding that would be made in 
connection with the approval of a completed and integrated plan. 
However, the NRC would not re-review, at the COL stage, information 
that provided the basis for the NRC approval of major features in an 
ESP but would address integration of approved major features with the 
balance of emergency planning information provided in the COL 
applications necessary to support the NRC's reasonable assurance 
finding; and updated emergency planning information required by Sec.  
52.39(b).
    A definition of prototype plant is added to explain the type of 
nuclear power plant that the NRC is addressing in Sec. Sec.  52.43, 
52.47(b), 52.79, and 52.157. A prototype plant is a licensed nuclear 
reactor test facility that is similar to and representative of either 
the first-of-a-kind or standard 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 safety 
features for the first-of-a-kind nuclear plant design, as well as being 
used as a commercial nuclear power facility.
c. Section 52.2, Interpretations; and Sec.  52.4, Deliberate Misconduct
    The former section on interpretations in Sec.  52.5 is retained and 
redesignated without change as Sec.  52.2. The former section on 
deliberate misconduct in Sec.  52.9 is retained and redesignated 
without change as Sec.  52.4.
d. Section 52.3, Written Communications; Sec.  52.5, Employee 
Protection; Sec.  52.6, Completeness and Accuracy of Information; Sec.  
52.7, Specific Exemptions; Sec.  52.8, Combining Licenses; Sec.  52.9, 
Jurisdictional Limits; and Sec.  52.10, Attacks and Destructive Acts
    Section 52.3, Written communications, which is essentially 
identical with the current Sec.  50.4, is added to address the 
requirements for correspondence, reports, applications, and other 
written communications from applicants, licensees, or holders of a 
standard design approval to the NRC concerning the regulations in part 
52.
    Section 52.5, which is largely identical with the current Sec.  
50.7, is added to make clear that discrimination against an employee 
for engaging in certain protected activities concerning the regulations 
in part 52 is prohibited. This section differs from its part 50 
counterpart, in that the Commission has added a provision on 
coordination with the requirements in 10 CFR part 19.
    Section 52.6, which is identical with the current Sec.  50.9, is 
added to require that information provided to the Commission by a 
licensee, a holder of a standard design approval, and an applicant 
under part 52, and information required by statute or by the NRC's 
regulations, orders, or license conditions to be maintained by a 
licensee, holder of a standard design approval, and applicant under 
part 52 (including the applicant for a standard design certification 
under part 52 following Commission adoption of a final design 
certification rule) be complete and accurate in all material respects. 
The Commission has corrected an error in the proposed rule version of 
paragraph (a) of Sec.  52.6. In the proposed rule, the first sentence 
began, ``Information provided to the Commission by a licensee 
(including a construction permit holder, and a combined license holder) 
* * *.'' In the final rule, this phrase has been corrected to read, 
``Information provided to the Commission by a licensee (including an 
early site permit holder, a combined license holder, and a 
manufacturing license holder) * * *.'' This provision applies to 
licenses issued under part 52 and not to licenses issued under part 50.
    Section 52.7, which is essentially identical with current Sec.  
50.12, is added to address the procedure and criteria for obtaining an 
exemption from the requirements of part 52. Although part 50 contains a 
provision (Sec.  50.12) for obtaining specific exemptions, Sec.  50.12 
by its terms applies only to exemptions from part 50. Although it would 
be possible to revise Sec.  50.12 so that its provisions apply to 
exemptions from part 52, this is inconsistent with the general 
regulatory structure of 10 CFR, wherein each part is treated as a 
separate and independent regulatory unit. The NRC notes that the 
exemption provisions in Sec.  52.7 are generally applicable to part 52, 
and do not supercede or otherwise diminish more specific exemption 
provisions that are in part 52.
    Section 52.8, which combines into a single section regulatory 
provisions which are addressed in separate regulations in part 50, is 
added to clarify that these regulatory provisions also apply to part 52 
licenses.
    Paragraph (a) of Sec.  52.8, which is analogous to Sec.  50.31, is 
added to make clear that an applicant for a license under part 52 may 
combine in one application, several applications for different kinds of 
licenses under various regulations in 10 CFR Chapter I. Section 50.31 
currently provides that an applicant may combine in one application, 
several applications for different kinds of licenses under various 
regulations in 10 CFR Chapter I. The plain reading of this language, 
given that this provision is located in part 50, is that a part 50 
application may contain in one application other applications for 
different licenses in other parts of 10 CFR Chapter I. Thus, Sec.  
50.31 would not appear to allow a part 52 application (as for a 
combined license) to combine in one application other applications for 
different license in other parts of 10 CFR Chapter I. Accordingly, 
paragraph (a) of Sec.  52.8 of the final rule makes clear that a part 
52 application may be combined with applications for different licenses 
in other parts of 10 CFR Chapter I. This provision was not included in 
the March 2006 proposed rule, inasmuch as the NRC determined the 
desirability of including in part 52 a provision analogous to Sec.  
50.31 only after the publication of the March 2006 proposed rule.
    Paragraph (b) of Sec.  52.8, which is analogous to Sec.  50.32, is 
added to make clear that an applicant for a license, standard design 
certification, or design approval under part 52 may incorporate by 
reference in its application information contained in other documents 
provided to the Commission,

[[Page 49374]]

but must clearly specify the information to be incorporated. This 
provision was also not included in the March 2006 proposed rule, 
inasmuch as the NRC determined the desirability of including in part 52 
a provision analogous to Sec.  50.32 only after the publication of the 
March 2006 proposed rule.
    Paragraph (c) of Sec.  52.8, which is analogous to Sec.  50.52, is 
added to clarify the Commission's authority under Section 161.h of the 
AEA to combine NRC licenses, such as a special nuclear materials 
license under part 70 for the reactor fuel, with a combined license 
under part 52. Analogous to the situation with respect to Sec.  50.31, 
the language in Sec.  50.52 would not appear to allow the Commission to 
combine into a single part 52 license, other non-part 52 licenses. 
Inasmuch as these changes to Sec.  52.8 constitute revisions to the 
Commission's rules of procedure and practice, the Commission may adopt 
them in final form without further notice and comment, under the 
rulemaking provisions of the APA, 5 U.S.C. 553(b)(A).
    Section 52.9, which is identical with Sec.  50.53, is added to 
clarify that NRC licenses issued under part 52 do not authorize 
activities which are not under or within the jurisdiction of the United 
States; an example would be the construction of a nuclear power reactor 
outside the territorial jurisdiction of the United States which uses a 
design identical to that approved in a standard design certification 
rule in part 52.
    Section 52.10 is added because there is no specific provision in 
part 52 specifying that the Commission's longstanding determination 
with respect to the lack of need for design features and other measures 
for protection of nuclear power plants against attacks by enemies of 
the United States, or the use of weapons deployed by United States 
defense activities, applies to part 52 applicants. The Commission's 
determination, which was upheld by the U.S. Court of Appeals for the 
D.C. Circuit, see Siegel v. Atomic Energy Commission, 400 F.2d 778 
(D.C. Cir 1968), is currently codified for part 50 applicants in Sec.  
50.13. Although it would be possible to revise Sec.  50.13 so that its 
provisions apply to applications under part 52, this would be 
inconsistent with the overall regulatory pattern of 10 CFR Chapter I, 
whereby each part is treated as a separate and independent regulatory 
unit. Moreover, any changes to Sec.  50.13 might erroneously be viewed 
as changes to the Commission's substantive determination on this 
matter. For these reasons, the Commission is adding new Sec.  52.10 to 
part 52, which is essentially identical with Sec.  50.13. Inclusion of 
this provision in part 52 makes clear that applications for combined 
licenses, manufacturing licenses, design certification rulemakings, 
standard design approvals, and amendments to these licenses, 
rulemakings, and approvals under part 52 need not provide design 
features or other measures for protection of nuclear power plants 
against attacks by enemies of the United States, or the use of weapons 
deployed by U.S. defense activities. In adding Sec.  52.10, the 
Commission emphasizes that it is not changing in any way, nor is it 
intending to revisit in this rulemaking, the Commission's determination 
with respect to the lack of need for design features or other measures 
for protection of nuclear power plants against attacks by enemies of 
the United States, or the use of weapons deployed by U.S. defense 
activities. The Commission is simply making it clear that its 
longstanding determination applies to applications under part 52 just 
as it applies to applications under part 50.
6. Subpart A, Early Site Permits
a. Emergency Preparedness Requirements for Early Site Permit Applicants
    The NRC is amending Sec. Sec.  52.17(b), 52.18, and 52.39 to 
address changes to emergency preparedness requirements for early site 
permit applicants. The NRC is amending Sec.  52.17(b)(1), which 
requires that an early site permit application identify physical 
characteristics unique to the proposed site that could pose a 
significant impediment to the development of emergency plans. The NRC 
is adding a sentence to require that, if physical characteristics that 
could pose a significant impediment to the development of emergency 
plans are identified, the application must identify measures that 
would, when implemented, mitigate or eliminate the significant 
impediment. The NRC believes this addition is necessary to clarify the 
NRC's expectations in cases where a physical characteristic exists that 
could pose a significant impediment to the development of emergency 
plans. Simply identifying these physical characteristics alone does not 
provide the NRC with enough information to determine if these 
characteristics are likely to pose a significant impediment to the 
development of emergency plans. Similarly, the Commission is amending 
Sec.  52.18 to require that the Commission determine whether the 
information required of the applicant by Sec.  52.17(b)(1) shows that 
there is no significant impediment to the development of emergency 
plans that cannot be mitigated or eliminated by measures proposed by 
the applicant [emphasis added].
    The NRC is amending Sec. Sec.  52.17(b)(2)(i), 52.17(b)(2)(ii), and 
52.18 to clarify that any emergency plans or major features of 
emergency plans proposed by early site permit applicants must be in 
accordance with the applicable standards of 10 CFR 50.47 and the 
requirements of appendix E to part 50. These changes clarify the 
standards applicable to emergency preparedness information supplied 
with an early site permit application. The NRC is also amending 
Sec. Sec.  52.17(b)(1), (b)(2), and (b)(4) to indicate that the 
emergency preparedness information supplied in the early site permit 
application must be included in the site safety analysis report. This 
change is necessary for consistency with past practice and with the 
requirements for combined license applicants in Sec.  52.79(a) that 
require emergency preparedness information to be included in the final 
safety analysis report. Note that the proposed rule only included these 
changes in Sec.  52.17(b)(2). In the final rule, the NRC is making the 
additional conforming changes in Sec. Sec.  52.17(b)(1) and (b)(4).
    The NRC is adding new Sec.  52.17(b)(3) to require that any 
complete and integrated emergency plans submitted for review in an 
early site permit application must include the proposed inspections, 
tests, and analyses that the holder of a combined license referencing 
the early site permit shall perform, and the acceptance criteria that 
are necessary and sufficient to provide reasonable assurance that, if 
the inspections, tests, and analyses are performed and the acceptance 
criteria met, the facility has been constructed and would operate in 
conformity with the license, the provisions of the AEA, and the NRC's 
regulations. The NRC is making these amendments for consistency with 
the requirements in subpart C of part 52 regarding the review of 
emergency plans and to provide additional finality to ESP holders. The 
NRC believes that its review of complete and integrated plans included 
in an early site permit application should be no different than its 
review of emergency plans submitted in a combined license application, 
given that the NRC must make the same findings in both cases, namely, 
that the plans submitted by the applicant provide reasonable assurance 
that adequate protective measures can and will be taken in the event of 
a radiological emergency. The NRC will

[[Page 49375]]

not be able to make the required finding without the inclusion of 
proposed ITAAC in an early site permit application that includes 
complete and integrated emergency plans. In the final rule, the NRC has 
added an allowance that major features of an emergency plan submitted 
under paragraph (b)(2)(i) of Sec.  52.17 may include proposed ITAAC. 
This will give an applicant that has proposed major features additional 
opportunities to achieve finality on major features in cases where 
ITAAC can be included to address implementation aspects of the major 
feature.
b. Section 52.13, Relationship to Other Subparts
    The title of Sec.  52.13 is revised from ``Relationship to subpart 
F of 10 CFR part 2 and appendix Q of this part,'' to ``Relationship to 
other subparts,'' to reflect the revised scope of this section, which 
has been refocused on part 52.
c. Section 52.16, Contents of Applications; General Information and 
Sec.  52.17, Contents of Applications; Technical Information
    The NRC is adding Sec.  52.16 to include the general content 
requirements from Sec.  52.17(a)(1).
    The title of Sec.  52.17 is revised to read, ``Contents of 
applications; technical information.'' In response to several comments 
on the proposed rule, the NRC is including a general grandfathering 
provision in Sec.  52.17(a) that states, ``For applications submitted 
before September 27, 2007, the rule provisions in effect at the date of 
docketing apply unless otherwise requested by the applicant in 
writing.'' This revision reflects the Commission's belief that ESPs 
currently under review or issued prior to the effective date of the 
final part 52 rule should not be required to be modified by this rule. 
Section 52.17(a)(1) is amended to state that the early site permit 
application must specify the range of facilities for which the 
applicant is requesting site approval (e.g., one, two, or three 
pressurized-water reactors). This new language provides a clearer and 
more complete statement of the applicant's proposal with respect to the 
facilities which may be located under the early site permit. This 
facilitates NRC review, as well as providing adequate notice to 
potentially-affected members of the public and State and local 
governmental entities. The NRC assumes that an applicant for an early 
site permit may not know what type of nuclear plant may be built 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 approval of the site will resolve 
issues with respect to the actual plant or plants that the combined 
license or construction permit applicant decides to build. In a letter 
dated November 13, 2001 (comment 27 on draft proposed rule text), 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-vii (sic.), 
industry recommends new Sec.  52.17a.2.'' The NRC 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 NRC expects that some applicants for an early site permit may 
not have decided on a particular type of nuclear power plant, 
therefore, Sec.  52.17(a)(1) was revised to address this situation.
    The NRC is amending Sec.  52.17(a)(1) to eliminate all references 
to Sec.  50.34. The references to Sec.  50.34(a)(12) and (b)(10) are 
removed because these provisions require compliance with the earthquake 
engineering criteria in appendix S to part 50 and are not requirements 
for the content of an application. The reference to Sec.  
50.34(b)(6)(v), which requires plans for coping with emergencies, is 
also being removed. All requirements related to emergency planning for 
early site permits are addressed in Sec.  52.17(b) and other plans for 
coping with emergencies will be addressed in a combined license 
application. Finally, the reference to the radiological consequence 
evaluation factors identified in Sec.  50.34(a)(1) is being removed and 
the requirements are included in Sec.  52.17(a)(1). The NRC is 
modifying the existing requirement for early site permit applications 
to describe the seismic, meteorological, hydrologic, and geologic 
characteristics of the proposed site to add that these descriptions 
must reflect appropriate consideration of the most severe of the 
natural phenomena that have been historically reported for the site and 
surrounding area and with sufficient margin for the limited accuracy, 
quantity, and time in which the historical data have been accumulated. 
This addition is to ensure that future plants built at the site would 
be in compliance with general design criterion 2 from appendix A to 
part 50 which requires that structures, systems, and components 
important to safety be designed to withstand the effects of natural 
phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, 
and seiches without loss of capability to perform their safety 
functions. The design bases for these structures, systems, and 
components are required to reflect appropriate consideration of the 
most severe of the natural phenomena that have been historically 
reported for the site and surrounding area, with sufficient margin for 
the limited accuracy, quantity, and time in which the historical data 
have been accumulated.
    The NRC is adding several requirements to Sec.  52.17(a)(1). A 
requirement is added to Sec.  52.17(a)(1)(x) that applications for 
early site permits include information to demonstrate that adequate 
security plans and measures can be developed. This requirement is 
inherent in current Sec.  52.17(a)(1) which states that site 
characteristics must comply with 10 CFR part 100. Section 100.21(f) 
states that site characteristics must be such that adequate security 
plans and measures can be developed. A new Sec.  52.17(a)(1)(xi) is 
added to require early site permit applications to include a 
description of the quality assurance program applied to site activities 
related to the future design, fabrication, construction, and testing of 
the structures, systems, and components of a facility or facilities 
that may be constructed on the site. This change was made for 
consistency with changes to Sec.  50.55 and appendix B to part 50. A 
discussion of these changes can be found in this section under the 
heading ``Appendix B to Part 50.''
    An additional requirement is added to Sec.  52.17(a)(1) that is 
taken from Sec.  50.34(h), and that the NRC believes should be 
applicable to early site permits. Section 52.17(a)(1)(xii) requires 
that early site permit applications include an evaluation of the site 
against the applicable sections of the standard review plan (SRP) 
revision in effect 6 months before the docket date of the application. 
The SRP requirement currently exists for applicants for construction 
permits, operating licenses, and combined licenses. The NRC also 
believes it should be applicable to applicants for early site permits 
because they are partial construction permits that can be referenced in 
applications for construction permits or combined licenses and because 
it will facilitate the NRC's review of the early site permit 
application.
    The NRC is not requiring applicants to evaluate their site against 
the applicable sections of Regulatory Guide (RG) 1.206, ``Combined 
License Applications for Nuclear Power Plants.'' However, the NRC 
believes that the applicable portions of RG 1.206 can provide useful 
guidance to ESP applicants in preparing their

[[Page 49376]]

applications and that use of this guidance will facilitate the NRC's 
review.
    The NRC is making a change to Sec.  52.17(a)(1) based on several 
comments on the proposed rule. The NRC is deleting the requirement in 
proposed Sec.  52.17(a)(1)(x) that required ESP applicants to address 
impacts on operating units of constructing new units on existing sites, 
as well as include a description of the managerial and administrative 
controls to be used to assure that the limiting conditions of operation 
for existing units will not be exceeded. The NRC is deleting this 
requirement because it was contrary to the industry-NRC understanding 
documented in correspondence in 2003 regarding ESP Topic ESP-19 [see 
NEI letter dated May 14, 2003 (ML031920U0 6), and NRC letter dated 
August 11, 2003 (ML031490478)] and because the COL applicant is in the 
best position to provide such information, since it will have final 
information regarding the facility design and construction plans. The 
NRC may include a condition in early site permits that would require 
the permit holder to notify the operating plant licensee prior to 
conducting any activities authorized under Sec.  52.25. These controls 
should be sufficient to evaluate construction activities at a site with 
an existing operating unit. The NRC has deleted this provision from 
subpart A in the final rule. COL applicants will, however, continue to 
be required to meet this provision under Sec.  52.79(a)(31).
    The NRC is moving the environmental provisions in former Sec.  
52.17(a)(2) to Sec.  51.50(b). Revised Sec.  52.17(a)(2) simply states 
that an early site permit application must contain a complete 
environmental report as required by 10 CFR 51.50(b). A discussion of 
the final rule provisions related to the NRC's environmental review at 
the ESP stage can be found in the Supplementary Information section 
that discusses changes to 10 CFR part 51.
    The NRC is amending Sec.  52.21 to reflect clarifications provided 
in part 51 that an early site permit applicant has the flexibility of 
either addressing the matter of alternative energy sources in the 
environmental report supporting its early site permit application, or 
deferring consideration of alternative energy sources to the time that 
the early site permit is referenced in a licensing application. These 
changes to Sec.  52.21 clarify that the NRC's EIS need not address the 
need for power or alternative energy sources (and therefore these 
matters may not be litigated) if the early site permit applicant 
chooses not to address these matters in its environmental report.
    The NRC is amending Sec.  52.17(c) to clarify that if the applicant 
wants to request authorization to perform limited work activities at 
the site after receipt of the early site permit, the application must 
contain an identification and description of the specific activities 
that the applicant seeks authorization to perform. This request by the 
early site permit applicant would be separate from, but not in addition 
to, a request to perform activities under 10 CFR 50.10(e)(1). The 
submittal of this descriptive information will enable the NRC staff to 
perform its review of the request, consistent with past practice, to 
determine if the requested activities are acceptable under Sec.  
50.10(e)(1). If an applicant for a construction permit or combined 
license references an early site permit with authorization to perform 
limited work activities at the site and subsequently decides to request 
authorization to perform activities beyond those authorized under Sec.  
52.U0 (c), those additional activities will have to be requested 
separately under Sec.  50.10(e)(1). Some minor changes were made to the 
rule language in Sec.  52.17(c) in the final rule to remove references 
to information being included in either the site safety analysis report 
or the environmental report. The NRC concluded that it is preferable to 
include both the list of proposed activities and the redress plan as a 
separate document in the application, outside of both the site safety 
analysis report and the environmental report. The NRC's conclusion is 
based on the fact that the requirements in Sec.  50.10(e) address both 
safety and environmental issues. Additional changes were made to 
Sec. Sec.  51.50, 52.79(a), and 52.80 to implement this concept.
d. Section 52.24, Issuance of Early Site Permit
    The NRC is revising Sec.  52.24 to clarify the information that the 
NRC must include in the early site permit when it is issued. Section 
52.24 is also being amended to be more consistent with the parallel 
provision in Sec.  50.50, Issuance of licenses and construction 
permits, by requiring the NRC to ensure that there is reasonable 
assurance that the site is in conformity with the provisions of the 
AEA, and the NRC's regulations; that the applicant is technically 
qualified to engage in any activities authorized; and that issuance of 
the permit will not be inimical to the common defense and security or 
to the health and safety of the public.
    Section 52.24 is being amended to provide that the early site 
permit must state the site characteristics and design parameters, as 
well as the ``terms and conditions,'' of the early site permit, rather 
than the ``conditions and limitations'' as was formerly provided. The 
change provides consistency with Sec.  52.39(a)(2), and in particular 
Sec.  52.39(a)(2)(iii) of the former regulations, which also refers to 
``site parameters'' (corrected to ``site characteristics'' in the final 
rule) and ``terms and conditions.'' Section 52.24(c) is being added to 
require that the early site permit state the activities that the permit 
holder is authorized to perform at the site. This change is consistent 
with the revision to Sec.  52.17(c) where the applicant must specify 
the activities that it is requesting authorization to perform at the 
site under Sec.  50.10(e)(1).
    The NRC is revising paragraph (b) of this section based on public 
comments. Paragraph (b) states that the early site permit shall specify 
the site characteristics, design parameters, and terms and conditions 
of the early site permit the NRC deems appropriate. Paragraph (b) 
further states that, before issuance of either a construction permit or 
combined license referencing an early site permit, the Commission shall 
find that any relevant terms and conditions of the early site permit 
have been met. The NRC is revising this paragraph to add a provision 
that any terms or conditions of the early site permit that could not be 
met by the time of issuance of the construction permit or combined 
license, must be set forth as terms or conditions of the construction 
permit or combined license. This provision is needed to address terms 
or conditions of the early site permit that are related to activities 
that will not take place until after issuance of the construction 
permit or combined license, such as construction activities. A similar 
change is being made to Sec.  52.79(b)(3).
e. Section 52.27, Duration of Permit
    Section 52.27 provides for the duration of an early site permit. 
The NRC did not propose any changes to this section in the proposed 
rule. However, in the final rule, the NRC is making several revisions. 
First, the NRC is revising former Sec.  52.27(b)(1) [final Sec.  
52.27(b)]. This paragraph states that an early site permit continues to 
be valid beyond the date of expiration in any proceeding on a 
construction permit application or a combined license application that 
references the early site permit and is docketed before the date of 
expiration of the early site permit, or, if a timely application for 
renewal of the permit has been filed, before the Commission has 
determined whether to

[[Page 49377]]

renew the permit, consistent with the ``Timely Renewal'' doctrine of 
the Administrative Procedure Act. This section is changed in the final 
rule by deleting the term, ``filing,'' and substituting the term, 
``docketing.'' The NRC believes that timely renewal protection should 
only be provided to those applications which are of sufficient quality 
to be docketed. This is consistent with the requirement in Sec.  
2.109(b) requiring filing of a ``sufficient'' application for renewal 
of operating licenses as a prerequisite for the applicability of the 
timely renewal protection. Inasmuch as the changes to former Sec.  
52.72(b)(1) constitute revisions to the NRC's rules of procedure and 
practice, the NRC may adopt them in final form without further notice 
and comment, under the rulemaking provisions of the APA, 5 U.S.C. 
553(b)(A).
    The NRC is also making revisions to Sec.  52.27 based on public 
comments. The NRC is deleting proposed Sec.  52.27(b)(2) because it was 
inconsistent with proposed Sec.  52.39(d) and the NRC's intention that 
the early site permit be subsumed into the construction permit or 
combined license once the construction permit or combined license is 
issued. To make this intention clear, the NRC is also adding new Sec.  
52.27(d) in the final rule. This provision states that upon issuance of 
a construction permit or combined license, a referenced early site 
permit is subsumed, to the extent referenced, into the construction 
permit or combined license. By ``subsumed'' the NRC means that the 
information that was contained in the early site permit site safety 
analysis report (SSAR) becomes part of the referencing combined license 
final safety analysis report upon issuance of the combined license in 
the same manner as if the combined license applicant had not referenced 
an early site permit. The NRC is including the phrase ``to the extent 
referenced,'' to indicate that it is not all of the information 
submitted in the early site permit application that is subsumed into 
the combined license, but, only that information that is contained in 
the SSAR and identified by the applicant as being referenced in the 
combined license application. This subsumption of the early site permit 
into the referencing license affects the way changes to the early site 
permit information will be handled because it breaks the tie to the 
finality provisions in Sec.  52.39. After issuance of the construction 
permit or combined license, Sec.  52.39 no longer applies to the early 
site permit information and such information will be covered by the 
same finality provisions as the rest of the information in the FSAR 
(with the exception of any referenced design certification 
information), as outlined in Sec.  52.98 (e.g., in accordance with 
Sec. Sec.  50.54, 50.59, etc.).
f. Section 52.28, Transfer of Early Site Permit
    Section 52.28 is being added to state that transfer of an early 
site permit from its existing holder to a new applicant would be 
processed under Sec.  50.80, which contains provisions for transfer of 
licenses. In a letter dated November 13, 2001 (comment 19 on draft 
proposed rule text), the NEI recommended that a new section be added to 
part 52 to clarify the process for transfer of an early site permit. 
The NRC 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 NRC believes that the 
procedures and criteria for transfer of utilization facility licenses 
in 10 CFR 50.80 (and the procedures in subpart M of part 2 for the 
conduct of any hearing) should apply to the transfer of an early site 
permit. Changes that the NRC has made to Sec.  50.80 in the final rule 
to address comments made regarding requirements for transfer of an 
early site permit can be found in Section V.D.8.a of the supplementary 
information of this document.
g. Section 52.33, Duration of Renewal
    Section 52.33 has been revised in the final rule to clarify that 
the renewal period for an early site permit includes any remaining 
years on the early site permit then in effect before renewal. This 
change was made to be consistent with the NRC's regulations concerning 
renewal of nuclear power plant operating licenses as specified in Sec.  
54.31 of this chapter.
h. Section 52.37, Reporting of Defects and Noncompliance; Revocation, 
Suspension, Modification of Permits for Cause
    Section 52.37 is removed because this provision only contains a 
cross-reference to 10 CFR part 21 and Sec.  50.100, and the NRC is 
making conforming changes to those requirements to account for 
requirements for early site permits.
i. Section 52.39, Finality of Early Site Permit Determinations
    The NRC is revising Sec.  52.39 to address the finality of an early 
site permit. While some of the changes are conforming or clarifying, 
others represent a change from the finality provisions in the former 
Sec.  52.39. Paragraph (a)(2) of the former rule distinguishes among 
issues alleging that: (1) a ``reactor does not fit within one or more 
of the site parameters,'' which are to be treated as valid contentions 
(paragraph (a)(2)(i)); (2) a ``site is not in compliance with the terms 
of an early site permit,'' which are to be subject to hearings under 
the provisions of the Administrative Procedure Act (paragraph 
(a)(2)(ii)); and (3) the ``terms and conditions of an early site permit 
should be modified,'' which are to be processed in accordance with 10 
CFR 2.206(a)(2)(iii). With the benefit of hindsight and experience 
gained in reviewing the first three early site permit applications, the 
NRC believes that all issues concerning a referenced early site permit 
may be characterized as:
    (1) Questions regarding whether the site characteristics, design 
parameters, or terms and conditions specified in the early site permit 
have been met;
    (2) Questions regarding whether the early site permit should be 
modified, suspended, or revoked; or
    (3) Significant new emergency preparedness or environmental 
information not considered on the early site permit.
    Questions about the referencing application demonstrating 
compliance with the early site permit are fundamentally questions of 
compliance with the early site permit. They do not attack the 
underlying validity of the permit. For example, if a person questions 
whether the design characteristics of the nuclear power facility that 
the referencing applicant proposes to construct on the site falls 
within the design parameters specified in the early site permit, it is 
a matter of compliance with the early site permit. These compliance 
matters are specific to the proceeding for the referencing application, 
and the NRC concludes that a question about whether the referencing 
application complies with the early site permit may be viewed as 
question/material to the proceeding and appropriate for consideration 
in the referencing application proceeding (assuming that all relevant 
Commission requirements in 10 CFR part 2, such as standing and 
admissibility, are met).
    The NRC also regards new emergency preparedness information 
submitted in the referencing application that substantially alters the 
bases for a previous NRC conclusion or constitutes a sufficient basis 
for the Commission to modify or impose new terms and conditions related 
to emergency preparedness as an issue material to the

[[Page 49378]]

proceeding and appropriate for consideration as a contention in the 
referencing application proceeding (assuming that all relevant 
Commission requirements in 10 CFR part 2, such as standing and 
admissibility, are met). This is a change to the standard that was 
provided in the proposed rule for new emergency preparedness 
information and is based on public comments. The proposed rule standard 
for litigation of emergency preparedness matters was ``new or 
additional information * * * which materially affects the Commission's 
earlier determination on emergency preparedness, or is needed to 
correct inaccuracies in the emergency preparedness information approved 
in the early site permit.'' Because the final rule language suggested 
by the commenters is the definition that the NRC gave for information 
that could ``materially affect'' the Commission's earlier decision, as 
indicated in the supplementary information section of the 2006 proposed 
rule, the NRC believes it appropriate to use this language in the final 
rule itself. The NRC has decided to drop the language that referred to 
information ``needed to correct inaccuracies'' because the language, by 
itself, could have allowed litigation of issues not significant to 
safety. The NRC believes that the final rule language encompasses all 
significant emergency preparedness matters that should be subject to 
litigation.
    Any significant environmental issue that was not resolved in the 
early site permit proceeding, or any issue involving the impacts of 
construction and operation of the facility that was resolved in the 
early site permit proceeding for which significant new information has 
been identified may also be the subject of a contention during the 
proceeding on the referencing application. The NRC is also making a 
change to this standard in the final rule based on public comment. The 
standard in the final rule more closely reflects the NRC's obligation 
under NEPA to address new and significant information in a COL that 
references an early site permit. Additional discussion of this subject 
can be found in the discussion of changes in 10 CFR part 51, in the 
supplementary information section of this document.
    Because new emergency planning or environmental information, if 
any, will be identified only at the time a license application 
referencing the early site permit is submitted to the NRC, the NRC 
believes it is appropriate to address these issues in the proceeding on 
the referencing application. Other questions regarding whether the 
permit should be modified, suspended, or revoked will be challenges to 
the validity of the early site permit. These challenges may be framed 
in many different ways, e.g., a Commission error at the time of 
issuance; or actual changes to the site have occurred since issuance of 
the permit that render some aspect of the permit irrelevant or 
inadequate to protect public health and safety or common defense and 
security. The Commission's process for challenges to the validity of a 
license is contained in 10 CFR 2.206. Accordingly, the Commission 
concludes that challenges to the validity of an early site permit 
should be processed in accordance with Sec.  2.206. In the Commission's 
view, a variance is not fundamentally a challenge to the validity of 
the early site permit, because it requests dispensation from compliance 
with some aspect of the permit whose validity remains undisputed. 
Therefore, the Commission concludes that variances should be treated as 
proceeding-specific issues of compliance that are potentially valid 
subjects of a contention in a proceeding for a referencing application.
    The revisions to Sec.  52.39 are in agreement with these Commission 
conclusions. Section 52.39 is being divided into five paragraphs 
addressing different aspects of early site permit finality. Each 
paragraph is provided with a subtitle characterizing the subject matter 
addressed in that paragraph. Section 52.39(a) focuses on how the NRC 
accords finality to an early site permit, with Sec.  52.39(a)(1) 
setting forth the circumstances under which the NRC may modify an early 
site permit. The rule language is based upon the existing regulation, 
but adds additional circumstances. Section 52.39(a)(1)(iii) provides 
that the NRC may modify the early site permit if it determines a 
modification is necessary based on an update to the emergency 
preparedness information under Sec.  52.39(b). Section 52.39(a)(1)(iv) 
provides that the NRC may modify the early site permit if a variance is 
issued under proposed Sec.  52.39(d) (paragraph (b) in the former 
regulations); the NRC considers this a conforming change inasmuch as 
the former regulation provided for issuance of variances.
    The NRC is clarifying what aspects of the early site permit are 
subject to the change restrictions in Sec.  52.39(a)(1) by substituting 
the phrase, ``terms and conditions'' of an early site permit for the 
former term, ``requirements.'' Under the new language, the NRC may not 
change or impose new site characteristics, design parameters, or terms 
and 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 language would specify more clearly the broad scope 
of matters in an early site permit which the NRC intended to finalize. 
The phrase, ``site characteristics, or terms, or conditions, including 
emergency planning requirements,'' is used consistently throughout 
Sec.  52.39 and corresponding provisions in the revisions to Sec.  
52.79.
    Section 52.39(a)(2) describes how the NRC treats matters resolved 
in the early site permit proceeding in subsequent proceedings on 
applications referencing the early site permit, and is drawn from the 
former language of Sec.  52.39(a)(2). In the final rule, the NRC has 
included a provision extending this finality to enforcement hearings 
other than those proceedings initiated by the Commission under 
paragraph (a)(1) of this section. This will ensure that finality of an 
early site permit extends to NRC-initiated enforcement proceedings and 
petitions for enforcement action filed under Sec.  2.206. In addition, 
under Sec. Sec.  52.39(a)(2)(i) and (ii), the NRC grants finality to 
changes to an early site permit's emergency plan (or major features of 
it, under Sec.  52.17(b)(2)) that are made after the issuance of the 
early site permit (1) if the early site permit approved an emergency 
plan (or major features thereof) that is in use by a licensee of a 
nuclear power plant and the changes to the emergency plan (or major 
features thereof) are identical to changes made to the licensee's 
emergency plans in compliance with Sec.  50.54(q); or (2) if the early 
site permit approved an emergency plan (or major features thereof) that 
is not in use by a licensee of a nuclear power plant, and the changes 
are equivalent to those that could be made under Sec.  50.54(q) without 
prior NRC approval had the emergency plan been in use by a licensee. 
This change is premised on the view that changes to emergency plans 
which are properly implemented under Sec.  50.54(q) do not require NRC 
review and approval before implementation. Therefore, by analogy, 
similar changes to an early site permit's emergency preparedness plan 
made with similar controls, or changes which are equivalent to those 
that could be made under Sec.  50.54(q) without prior NRC approval, 
should not require NRC review and approval as part of the licensing 
process. Any issues related to compliance with Sec.  50.54(q) should be 
treated as an enforcement matter. Note that the NRC is making some 
adjustments to this position in the final

[[Page 49379]]

rule based on public comments. The proposed rule would not have 
excepted changes to early site permit emergency plans not in use by a 
current licensee that could be made under Sec.  50.54(q) without prior 
NRC approval had the emergency plans been in use by a licensee. The NRC 
is making this change in the final rule because the Sec.  50.54(q) 
standard ensures adequate protection of safety, and has been accepted 
and used by the industry and NRC and it is appropriate to apply this 
same standard to changes in all emergency plans approved by the NRC in 
the ESP proceeding. The NRC is making similar changes to Sec.  
52.79(b)(4) in the final rule to require that all COL applicants 
referencing early site permits with complete and integrated emergency 
plans or major features of emergency plans identify changes that have 
been incorporated into the proposed facility emergency plans and that 
constitute or would constitute a decrease in effectiveness under Sec.  
50.54(q) of this chapter.
    Section 52.39(b) is discussed separately under Section V.C.6.a of 
this document, which discusses emergency preparedness requirements for 
a combined license applicant referencing an early site permit.
    Section 52.39(c) replaces the former criteria in Sec. Sec.  
52.39(a)(2)(i) through (iii), governing how the NRC will treat various 
issues with respect to the early site permit and its referencing in a 
combined license application. Matters regarding compliance with the 
early site permit which would be potentially valid subjects of a 
contention are listed in Sec. Sec.  52.39(c)(1)(i) through (iii), e.g., 
whether the reactor proposed to be built under the referencing 
application fits within the site characteristics and design parameters 
specified in the early site permit; whether one or more of the terms 
and conditions of the early site permit have been met; and whether a 
variance requested by the referencing applicant is unwarranted or 
should be modified. The NRC notes that all contentions at the early 
site permit stage, including a contention pertaining to a variance, 
must meet the requirements for contentions in Sec.  2.309(f). Matters 
regarding significant new emergency preparedness or environmental 
information material to the combined license proceeding, which would be 
potentially valid subjects of contention under the proposed rule, are 
listed in Sec. Sec.  52.39(c)(1)(iv) and (v).
    Other matters, including changes to the site characteristics, 
design parameters, or terms and conditions of the early site permit, 
are treated under Sec.  52.39(c)(2) as challenges to the permit and 
processed in accordance with Sec.  2.206. The NRC is retaining the 
former provision in Sec.  52.39(a)(2)(iii) requiring that the 
Commission consider a petition filed under Sec.  2.206, and determine 
whether immediate action is required before construction commences, as 
well as the former provision indicating that if a petition is granted, 
the Commission will issue an appropriate order which does not affect 
construction unless the Commission makes its order immediately 
effective.
    The final rule redesignates the former provision in Sec.  52.39(b) 
allowing an applicant for a license referencing an early site permit to 
request a variance from one or more ``elements'' of the early site 
permit as Sec.  52.39(d). The rule clarifies ``elements'' for which a 
variance may be sought by substituting the phrase, ``site 
characteristics, design parameters, or terms and conditions of the 
early site permit.'' In addition, the NRC is revising this provision 
further to include an allowance for applicants to request a variance 
from the site safety analysis report (SSAR). The allowance for 
requesting variances to the SSAR was inadvertently omitted in the 
proposed rule. Because the majority of the early site permit 
information that a combined license applicant will be referencing will 
be the information in the SSAR, it is logical that the allowance to 
request variances be extended to the information in the SSAR given that 
the NRC is allowing variances to the permit itself. The NRC notes that 
the admission of a contention on a proposed variance, which was 
formerly addressed in Sec.  52.39(b), is addressed in Sec.  
52.39(c)(iii). The NRC is also adding a provision that precludes the 
Commission from issuing a variance once a construction permit or 
combined license referencing the early site permit is issued. Any 
changes that would otherwise require a variance should instead be 
treated as an amendment to the construction permit or combined license.
    Finally, the NRC is adding a new paragraph to the ``finality'' 
section in each subpart of part 52, in this instance Sec.  52.39(f), 
entitled ``Information requests,'' which delineates the restrictions on 
the NRC for information requests to the holder of the early site 
permit. This provision is analogous to the former provision on 
information requests in paragraph 8 of appendix O to parts 50 and 52, 
and is based upon the language of Sec.  50.54(f). For early site 
permits, this provision is contained in Sec.  52.39(d), and requires 
the NRC to evaluate each information request on the holder of an early 
site permit to determine that the burden imposed by the information 
request is justified in light of the potential safety significance of 
the issue to be addressed in the information request. The only 
exceptions would be for information requests seeking to verify 
compliance with the current licensing basis of the early site permit. 
If the request is from the NRC staff, the request would first have to 
be approved by the Executive Director for Operations (EDO) or his or 
her designee.
7. Subpart B, Standard Design Certifications
a. Section 52.41, Scope of Subpart
    This section defines the scope of subpart B of part 52. The 
requirements on scope and type of nuclear power plants that are 
eligible for design certification were moved from former Sec.  52.45(a) 
to this section, to ensure a consistent format and presentation among 
all the subparts of part 52.
b. Section 52.43, Relationship to Other Subparts
    This section defines the relationship of subpart B to other 
subparts in 10 CFR part 52. Conforming changes were made to make clear 
that an application for a manufacturing license may, but is not 
required to, reference a design certification rule (DCR). The 
requirements formerly located in Sec. Sec.  52.43(c), 52.45(c), and 
52.47(b)(2)(ii) were removed because the Commission decided not to 
require a final design approval (FDA) under subpart E as a prerequisite 
for certification of a standard plant design. This requirement was 
included in part 52, at the time of the original rulemaking, because 
the NRC had no experience with design certifications. By requiring an 
FDA as a prerequisite to design certification, the NRC indicated that 
the licensing processes for design certifications and FDAs were 
similar, even though the requirements for and finality of a design 
certification differ from that of an FDA. The NRC now has considerable 
experience with design certification reviews, and the former 
requirement to apply for an FDA as part of an application for design 
certification is no longer needed. Future applicants have the option to 
apply for either an FDA, a design certification, or both.
c. Section 52.45, Filing of Applications
    This section presents the requirements for filing design 
certification applications. This section was reformatted for 
consistency with the other subparts in part 52 and the references to 
specific paragraphs within Sec. Sec.  50.4 and 50.30 were replaced with 
references to subpart H of part 2. A new

[[Page 49380]]

Sec.  52.45(c) on design certification review fees, was moved from 
Sec.  52.49.
d. Section 52.46, Contents of Applications; General Information
    This section was added to set forth general content requirements 
from 10 CFR 50.33.
e. Section 52.47, Contents of Applications; Technical Information
    This section presents the requirements for contents of a design 
certification application and is organized into three sections. The 
requirements for the final safety analysis report (FSAR) are set forth 
in Sec. Sec.  52.47(a) and 52.47(c), and the technical requirements for 
the remainder of the design certification application are in Sec.  
52.47(b). The former Sec.  52.47(a)(1)(i) required the submittal of 
information required for construction permits and operating licenses by 
parts 20, 50 (including the applicable requirements from 10 CFR 50.34), 
73, and 100, which were technically relevant to the design and not 
site-specific. That general requirement was removed and replaced with 
specific requirements that describe what must be included in an FSAR. 
In addition, the NRC included technical positions that were developed 
after part 52 was originally codified in 1989, e.g., Sec.  52.47(a)(22) 
which requires a description of how relevant operating experience was 
incorporated into the standard design (see SRM on SECY-90-377, dated 
February 15, 1991, ML003707892). Also, the relevant requirements were 
revised to clarify their applicability to design certifications and 
renumbered. This effort resulted in a comprehensive list of 
requirements for a design certification application.
    Some commenters recommended that the requirement to demonstrate 
technical qualifications [now Sec.  52.47(a)(7)] be deleted because the 
AEA only imposes that requirement on applicants for a license. Although 
the NRC agrees that the AEA imposes the technical qualification finding 
specifically for license applicants, it does not preclude the NRC from 
a determination that such a finding is also necessary in other 
contexts. The applicant creates information that may become the bases 
for a future license and, therefore, must be qualified to perform 
design, analyses, and safety determinations. Accordingly, the NRC has 
concluded that a technical qualification finding should also be made 
for design certification applicants.
    Some commenters recommended that the requirement to address the 
standard review plan (SRP) be revised to apply to light-water reactors. 
The NRC agrees with this comment and has revised this requirement [now 
Sec.  52.47(a)(9)] to be applicable to light-water-cooled nuclear power 
plants, but notes that much of the SRP review guidance and criteria are 
general and would also apply to reviews of gas-cooled reactor designs.
    Some commenters recommended that the requirement to provide 
information required by Sec.  50.49(d) [now Sec.  52.47(a)(13)] be 
deleted because the applicant will not be able to establish 
qualification files for all applicable components. The NRC agrees that 
applicants may not be able to establish qualification files, but 
applicants can provide the electric equipment list required by Sec.  
50.49(d). Therefore, the NRC revised the wording in Sec.  52.47(a)(13) 
to be consistent with the wording for the same provision in Sec.  
52.79(a), which requires that applicants provide the list of electrical 
equipment important to safety required by Sec.  50.49(d).
    Some commenters recommended that the requirement in Sec.  
52.47(a)(22) to demonstrate how operating experience insights have been 
incorporated into the plant design be deleted. The NRC disagrees with 
this comment. The NRC developed this requirement for future plants (see 
SRM on SECY-90-377) and it was implemented in past design certification 
applications by addressing NRC's generic letters and bulletins. The NRC 
agrees that insights from generic letters and bulletins should be 
incorporated into the latest revision of the standard review plan 
(SRP). Therefore, for plant designs that are based on or are evolutions 
of nuclear plants that have operated in the United States, the 
applicant should use NRC's generic letters and bulletins issued after 
the most recent revision of the applicable SRP and 6 months before the 
docket date of the application. If the application is for a nuclear 
plant design that is not based on or is not an evolution of a nuclear 
plant that operated in the United States, the applicant should address 
how insights from any relevant international operating experience has 
been incorporated into that plant design.
    Some commenters recommended that the requirement to describe severe 
accident design features in the FSAR [now Sec.  52.47(a)(23)] be 
deleted. The NRC disagrees with this comment because the Commission has 
determined that this requirement is necessary for future light-water 
reactor designs (see SRM on SECY-93-087) and was applied to previous 
applications. The commenters confused the meaning of design bases 
information (see Sec.  50.2) with the requirements for design-basis 
accidents (DBAs). Postulated severe accidents are not design-basis 
accidents and the severe accident design features do not have to meet 
the requirements for DBAs (see SECY-93-087). However, the severe 
accident design features are part of a plant's design bases 
information.
    A new Sec.  52.47(b) was created to set forth the required 
technical contents of a design certification application that are not 
required to be located in the FSAR. In response to public comments on 
the proposed rule, the NRC has deleted proposed Sec.  52.47(b)(1) which 
required design certification applicants to submit a design-specific 
probabilistic risk assessment (PRA). In its place, the NRC has added 
new Sec.  52.47(a)(27) which requires that design certification 
applicants submit a description of the design-specific PRA and its 
results in the FSAR. The NRC agrees with some commenters that 
applicants should not be required to submit their complete design-
specific PRA and that, instead, applicants should only be required to 
provide a summary description of the PRA and its results in their FSAR 
with the understanding that the complete PRA (e.g., codes) would be 
available for NRC inspection at the applicant's offices, if needed. The 
NRC expects that, generally, the information that it needs to perform 
its review of the design certification application from a PRA 
perspective is that information that will be contained in applicants' 
FSAR Chapter 19.
    The rule language for ITAAC [now Sec.  52.47(b)(1)] was conformed 
with the statutory language in the AEA. This clarification of the 
language in the former Sec.  52.47(a)(1)(vi), which was a condensed 
version of the language in the former Sec.  52.97(b)(1), was intended 
to avoid any misunderstandings regarding the statutory requirement. 
Some commenters recommended that the rule language in Sec.  52.47(b)(1) 
be modified to maintain the language in the former Sec.  
52.47(a)(1)(vi) claiming the proposed language could be misconstrued as 
expanding the scope of ITAAC needed for design certification. The NRC 
disagrees with this comment and notes that it is well understood that 
the requirements that are applicable to design certification are 
limited to the scope of the certified design.
    Some commenters recommended that the requirement in proposed Sec.  
52.47(b)(3) (now in 10 CFR 51.55) to evaluate severe accident 
mitigation design alternatives (SAMDAs) be deleted and that the NRC 
should initiate a rulemaking or policy statement to disposition SAMDA 
generically. The NRC disagrees with this comment. The

[[Page 49381]]

NRC has required SAMDA evaluations for previous applications in order 
to achieve greater finality for the design features that are resolved 
in design certification rulemakings. Further, the initiation of a 
rulemaking or policy statement for SAMDAs is outside the scope of the 
part 52 update rulemaking. As for the perspective that SAMDA 
evaluations need not be performed for current reactor designs because 
the severe accident risk for such designs is too remote and 
speculative, the NRC has already addressed this issue in other 
contexts. The NRC has considered petitions to eliminate the 
consideration of SAMDAs previously. The NRC position, both then and now 
is that it is not prepared to reach the conclusion that the risks of 
all severe accidents are so unlikely as to warrant their elimination 
from consideration in our NEPA reviews. As the NRC has stated in 
response to other requests to confine or eliminate such issues from 
consideration, if new information in the future provides a firm basis 
for concluding that severe accidents are remote and speculative, then 
the NRC may revisit the issue.
    Former Sec.  52.47(b) was reorganized by separating the 
requirements on scope of design and modular configuration [now located 
in Sec.  52.47(c)] from the testing requirements. This action is part 
of the NRC's goal to put the procedural requirements for the licensing 
processes in part 52 and maintain the reactor safety requirements in 
part 50 (or other parts of 10 CFR Chapter I. As a result, the testing 
requirements were relocated to Sec.  50.43(e). Also, see the discussion 
on testing for advanced nuclear reactors in Section V.B of this 
document.
f. Section 52.54, Issuance of Standard Design Certification
    This section was amended to be consistent with the parallel 
provisions in Sec. Sec.  50.50 and 50.57 by including requirements 
that, after conducting a rulemaking proceeding and receiving the report 
submitted by the ACRS, the NRC will determine whether there is 
reasonable assurance that the design conforms with the provisions of 
the AEA, and the NRC's regulations; that the applicant is technically 
qualified; and that issuance of the design certification will not be 
inimical to the common defense and security or to the health and safety 
of the public. In addition, a new Sec.  52.54(a)(8) was added to state 
that the NRC will not issue a design certification unless it finds that 
the design certification applicant has implemented the quality 
assurance program described in the safety analysis report. This 
requirement was added to indicate the NRC's expectation that design 
certification applicants will implement the QA program that is required 
to be included in their application under Sec.  52.47(a)(19), which is 
consistent with the requirement for licensees.
    A new Sec.  52.54(b) was added to require that a design 
certification specify the site parameters and design characteristics 
and any additional requirements and restrictions of the rule, as the 
Commission deems necessary and appropriate. Some commenters recommended 
that the requirement in Sec.  52.54(b) to list ``design 
characteristics'' be removed and noted that the design control document 
will contain this information. The NRC disagrees with this comment. The 
NRC wants to specifically identify this information to facilitate 
future comparisons with ``design parameters'' specified in an early 
site permit. The NRC staff will use its experience with current early 
site permit reviews to determine what an appropriate list will be for 
future design certification reviews.
    The NRC also modified Sec.  52.54 to require that applicants for a 
design certification agree to withhold access to National Security 
Information from individuals until the requirements of 10 CFR parts 25 
and/or 95, as applicable, are met. Section 52.54 was amended to include 
a new paragraph (c) which requires that every DCR contain a provision 
stating that, after the Commission has adopted the final design 
certification rule, the applicant for that design certification will 
not permit any individual to have access to, or any facility to 
possess, Restricted Data or classified National Security Information 
until the individual and/or facility has been approved for access under 
the provisions of 10 CFR parts 25 and/or 95. The NRC believes that this 
amendment, along with the changes to parts 25, 95, and 10 CFR 50.37, 
are necessary to ensure that access to classified information is 
adequately controlled by all entities applying for NRC certifications.
g. Section 52.63, Finality of Standard Design Certifications
    The final rule revises the finality provisions in Sec.  52.63(a) to 
provide processes for amending design certification information without 
meeting the special backfit requirement in Sec.  52.63(a)(1)(ii). The 
special backfit requirement restricted changes to certification 
information, thereby ensuring that all plants built under a referenced 
certified design would be standardized. Section 52.63(a)(1) was also 
revised to replace ``a modification'' with ``the change,'' to clarify 
that the criteria for changes apply to modifications, rescissions, or 
imposition of new requirements. In addition, Sec.  52.63 was revised to 
use the phrase ``certification information'' in order to distinguish 
the rule language in the DCRs from the design certification information 
(e.g., Tier 1 and Tier 2 information) that is incorporated by reference 
in the DCRs.
    Section 52.63(a)(1)(iii) was added to provide the NRC with the 
ability to make generic changes to the design certification rule 
language that reduce unnecessary regulatory burdens. The former Sec.  
52.63(a)(1) stated 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. This requirement did not appear to permit changes 
to the rule language 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 could not be made under the former Sec.  52.63(a)(1) was 
a change to the rule language in appendices A, B, and C of 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, to reduce 
unnecessary regulatory burden (64 FR 53582, October 4, 1999).
    In Section V of the 2006 proposed rule, Question 14, the NRC stated 
that it was considering adopting an additional provision in Sec.  
52.63(a)(1) that would allow amendments of DCRs to incorporate generic 
resolutions of design acceptance criteria (DAC) or other design 
information without meeting the special backfit requirement in the 
former Sec.  52.63(a)(1). By allowing for an amendment to generically 
resolve DAC, the NRC would achieve resolution of additional design 
issues, would achieve finality for those issue resolutions, and would 
avoid repetitive consideration of those design issues in individual 
combined license proceedings. The final rule includes an amendment 
process in Sec.  52.63(a)(1)(iv) that allows for generic resolutions of 
DAC without meeting the special backfit requirement. These amendments 
will

[[Page 49382]]

apply to all plants that have or will reference the DCR under Sec.  
52.63(a)(2). The NRC believes that these amendments will enhance 
standardization by further completing the certification information. 
The NRC will review the amendment application to ensure that the design 
acceptance criteria are met and that the new design information 
conforms with the applicable regulations.
    Some commenters proposed that the amendment process should allow 
for generic resolutions of errors in the certification information. The 
NRC is aware that design certification applicants have discovered 
errors in their design information after the NRC has completed its 
review and even after the NRC has certified their design. The final 
rule includes a new provision in Sec.  52.63(a)(1)(v) to correct 
material errors in the certification information. This provision is 
only to be used to correct a material error, which is an error that 
significantly and adversely affects a design function or analysis 
conclusion described in the design control document (certification 
information). The NRC wants to correct material errors by amendment so 
that these errors will not have to be addressed in individual licensing 
proceedings.
    Many commenters encouraged the NRC to adopt an amendment process 
that would allow for ``beneficial'' changes to certification 
information, would apply the amendment to all plants referencing the 
certified design, and would only allow amendments prior to issuance of 
the first combined license that referenced the DCR. The NRC agreed with 
these comments and included paragraph (a)(1)(vi) to allow for 
amendments of certification information that will substantially 
increase the overall safety, reliability, or security of facility 
design, construction, or operation provided that the direct and 
indirect costs of implementation of the amendment are justified in view 
of this increased safety, reliability, or security. However, the NRC 
does not agree with precluding amendments after issuance of the first 
combined license. If licensees who referenced a DCR want to adopt a 
proposed amendment in order to achieve enhanced standardization and the 
beneficial changes that the amendment would bring, then the NRC may 
amend the DCR and apply the amendment to all plants referencing the 
DCR.
    Also, some commenters requested that the amendment process allow 
for changes to the certification information for a wide variety of 
other reasons. These commenters claimed that the need for a design 
change may be discovered during detailed design work performed after 
the original design information was approved by the NRC (so-called 
first-of-a-kind-engineering) or that certain components in the original 
design may no longer be available for purchase due to the long duration 
of a DCR. The NRC's deliberations on this proposal considered the 
Commission's goal for design certification, which is to achieve and 
maintain the benefits of standardization. The NRC is still determined 
to maintain standardization, but has decided to allow amendments for 
other design changes [see paragraph (a)(1)(vii)] provided that the 
amendment will be applied to all plants that reference the DCR, thereby 
increasing standardization. In determining whether to codify a proposed 
amendment, the NRC will give special consideration to comments from 
applicants or licensees who reference the DCR regarding whether they 
want to backfit their plants with these additional design changes.
    The final rule includes a new Sec.  52.63(a)(2), which sets forth 
procedures for rulemakings conducted under Sec.  52.63(a)(1). Paragraph 
(a)(2)(i) requires that for rulemakings under Sec.  52.63(a)(1), except 
for rulemakings under Sec.  52.63(a)(1)(ii) necessary to provide 
adequate protection, the NRC will give consideration to whether the 
benefits justify the costs for plants that are already licensed or for 
which an application for a license is under consideration.
    The final rule also revised the former Sec.  52.63(a)(2) [now Sec.  
52.63(a)(3)] to delete the reference to the former Sec.  52.63(a)(4) 
[now Sec.  52.63(a)(5)]. The reference to the former 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 the new Sec.  52.63(a)(3) deals with modifications 
that the NRC may impose on a DCR under Sec. Sec.  52.63(a)(4) or 
52.63(b)(1). No substantive change is intended by this revision, which 
merely clarifies the intent of the rule.
    Finally, the NRC restates its previous decision regarding the 
ability of any person to request an amendment to a DCR. In Section 
II.1.h of the 1989 SOC for part 52 (54 FR 15372), the Commission stated 
that Sec.  52.63(a)(1) places a designer on the same footing as the NRC 
or any other interested member of the public. Therefore, anyone may 
submit a petition for rulemaking to the NRC to correct an error or 
otherwise amend the certification information. All amendments to the 
certification information must be accomplished through rulemaking, with 
an opportunity for public comment under Sec.  52.63(a)(2). Once a 
certified design is amended by rulemaking, the new rule would apply to 
all applications referencing the DCR as well as all plants referencing 
the DCR, unless the change has been rendered ``technically irrelevant'' 
through other action taken under Sec. Sec.  52.63(a)(4) or (b)(1). 
Also, the NRC will decide whether to codify the proposed amendment 
based on comments from the referencing applicants and licensees. Thus, 
standardization is maintained by ensuring that any generic change to 
the certification information is imposed upon all nuclear power plants 
referencing the DCR. The duration of the amended DCR will be for the 
same period of time as the original DCR and have the same expiration 
date.
8. Subpart C, Combined Licenses
a. Emergency Preparedness Requirements for a Combined License Applicant 
Referencing an Early Site Permit
    The NRC is revising former Sec. Sec.  52.39 and 52.79 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 
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 or in 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 could be deferred until a ``lessons learned'' 
rulemaking, updating 10 CFR part 52, was conducted after the first 
design certification rulemakings were 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, NEI submitted a letter 
dated September 8, 1998 (comment 2.d), which expressed opposition to a 
requirement for updating emergency preparedness information throughout

[[Page 49383]]

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. 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 NRC agrees in part with the Illinois Department of Nuclear 
Safety. Emergency plans and/or offsite certificates in support of 
emergency plans, approved as part of an early site permit review, 
should be updated. However, emergency plans do not need to be kept up-
to-date throughout the duration of an early site permit. There is no 
need to update the emergency plans approved in an early site permit 
until the time the permit is referenced in a combined license 
application. At that time, the emergency plans would have to be 
reviewed to confirm that they are up-to-date and to provide any new 
information that may materially affect the NRC's earlier determination 
on emergency preparedness, or correct inaccuracies in the emergency 
preparedness information approved in the early site permit in support 
of a reasonable assurance determination, in accordance with Sec.  50.47 
and appendix E to part 50. In addition, the NRC 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 determined that Sec. Sec.  52.39 and 
52.79 should contain an updating requirement to be imposed upon the 
applicant referencing an early site permit.
    A new Sec.  52.39(b) is added to require an applicant for a 
construction permit, operating 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). In 
addition, the applicant must discuss whether the new information could 
materially change the bases for compliance with the applicable NRC 
requirements. A parallel requirement is included in Sec.  52.79 to 
ensure that applicants for combined licenses referencing an early site 
permit will submit the updated emergency preparedness information. 
Section 52.39(a)(1)(iii) is also added stating that the Commission may 
modify an early site permit if it determines that a modification is 
necessary based on updated emergency preparedness information provided 
in a referencing license application. New information that materially 
changes the bases for compliance includes information that 
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, and information that 
would constitute a basis for the Commission to modify or impose new 
terms and conditions on the early site permit related to emergency 
preparedness in accordance with Sec.  52.39(a)(1). New information that 
materially changes the NRC'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) will be subject to litigation during the 
construction permit, operating license, or combined license proceedings 
in accordance with Sec.  52.39(c).
    Not all new information on emergency preparedness will be subject 
to challenge in a hearing under Sec.  52.39(c). For example, an 
emergency plan may have to be updated to reflect current telephone 
numbers, 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 current names of hospital 
facilities. These corrections do not materially change the NRC's 
previously-stated bases for accepting the early site permit emergency 
plan, and a hearing contention will not be admitted under Sec.  
52.39(c) in a proceeding for a license referencing the early site 
permit. In 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. This 
type of information might 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(c), assuming that the requirements in 10 CFR part 2 for 
admission of a contention are met.
b. Resolution of ITAAC
    Sections 52.99 and 52.103 are revised to incorporate rule language 
from the design certification regulations 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 NRC codified these ITAAC 
requirements into Section IX of the regulations. The purpose of the 
requirement in Sec.  52.99(b) is to clarify 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 met.
    Section 52.99(c) requires the licensee to notify the NRC that the 
prescribed inspections, tests, and analyses in the ITAAC have been or 
will be completed and that the acceptance criteria have been met. The 
NRC is revising Sec.  52.99(c)(1) in the final rule to more closely 
follow the language of Section 185b. of the AEA (in response to a late-
filed comment) and to clarify that the notification must contain 
sufficient information to demonstrate that the prescribed inspections, 
tests, and analyses have been performed and that the prescribed 
acceptance criteria have been met. The NRC is adding this clarification 
to ensure that combined license applicants and holders are aware that 
(1) it is the licensees' burden to demonstrate compliance with the 
ITAAC and (2) the NRC expects the notification of ITAAC completion to 
contain more information than just a simple statement that the licensee 
believes the ITAAC has been completed and the acceptance criteria met. 
The NRC expects the notification to be sufficiently complete and 
detailed for a reasonable person to understand the bases for the 
licensee's representation that the inspections, tests, and analyses 
have been successfully completed and the acceptance criteria have been 
met. The term ``sufficient information'' requires, at a minimum, a 
summary description of the bases for the licensee's conclusion that the 
inspections, tests, or analyses have been performed and that the 
prescribed acceptance criteria have been met. The

[[Page 49384]]

NRC plans to prepare regulatory guidance, in consultation with 
interested stakeholders, to explain how the functional requirement to 
provide ``sufficient information'' with regard to ITAAC submittals 
could be met.
    The NRC is also revising Sec.  52.99(c) in the final rule by adding 
a new paragraph (c)(2) requiring that, if the licensee has not 
provided, by the date 225 days before the scheduled date for initial 
loading of fuel, the notification required by paragraph (c)(1) of this 
section for all ITAAC, then the licensee shall notify the NRC that the 
prescribed inspections, tests, or analyses for all uncompleted ITAAC 
will be performed and that the prescribed acceptance criteria will be 
met prior to operation (consistent with the Section 189.a(1)(B) 
requirement governing a request for hearing on acceptance criteria, and 
the Section 185.b. requirement that the Commission find that the 
acceptance criteria in the combined license are met). The notification 
must be provided no later than the date 225 days before the scheduled 
date for initial loading of fuel. It is the licensee's burden to 
demonstrate that it will comply with the ITAAC and it must provide 
sufficient information to demonstrate that the prescribed inspections, 
tests, or analyses will be performed and the prescribed acceptance 
criteria for the uncompleted ITAAC will be met. The term ``sufficient 
information'' requires, at a minimum, a summary description of the 
bases for the licensee's conclusion that the inspections, tests, or 
analyses will be performed and that the prescribed acceptance criteria 
will be met. In addition, ``sufficient information'' includes, but is 
not limited to, a description of the specific procedures and analytical 
methods to be used for performing the inspections, tests, and analyses 
and determining that the acceptance criteria have been met.
    Paragraph (e) has been revised to require that the NRC make 
available to the public the notifications to be submitted under Sec.  
52.99(c)(1) and (c)(2), no later than the Federal Register notice of 
intended operation and opportunity for hearing on ITAAC under Sec.  
52.103(a). A conforming change is included in Sec.  2.105(b)(3) to 
require that the Sec.  52.103(a) notice reference the public 
availability of the Sec.  52.99(c)(1) and (2) notifications. The NRC is 
requiring that the paragraph (c)(2) notification be made 225 days 
before the date scheduled for initial loading of fuel, in order to 
ensure that the licensee notifications are publicly available through 
the NRC document room and online through the NRC Web site at the same 
time that the Sec.  52.103(a) notice is published in the Federal 
Register. The NRC's goal is to publish that notice 210 days before the 
date scheduled for fuel loading, but in all cases the Sec.  52.103(a) 
notice would be published no later than 180 days before the scheduled 
fuel load, as required by Section 189.a(1)(B) of the AEA.
    In Section V of the Supplementary Information of the proposed rule, 
the NRC requested stakeholder feedback on whether a provision on 
completion of ITAAC in a set time period prior to fuel load should be 
added to the final rule. Commenters did not support addition of a 
requirement on completion of ITAAC in a set time period prior to fuel 
load and the NRC has not included a provision requiring the completion 
of all ITAAC by a certain time prior to the licensee's scheduled fuel 
load date. Instead, the NRC has decided to modify the concept slightly 
by requiring the licensee to submit, with respect to ITAAC which have 
not yet been completed 225 days before the scheduled date for initial 
loading of fuel, additional information addressing whether those 
inspections, tests, and analyses will be successfully completed and the 
acceptance criteria met before initial operation. In the case where the 
licensee has not completed all ITAAC by 225 days prior to its scheduled 
fuel load date, the NRC expects the information that the licensee 
submits related to uncompleted ITAAC to be sufficiently detailed such 
that the NRC can determine what activities it will need to undertake to 
determine if the acceptance criteria for each of the uncompleted ITAAC 
have been met, once the licensee notifies the NRC that those ITAAC have 
been successfully completed and their acceptance criteria met. In 
addition, the NRC is adopting the requirements in paragraphs (c)(1) and 
(c)(2) to ensure that interested persons will be able to meet the 
Atomic Energy Act, Section 189.a(1), threshold for requesting a hearing 
with respect to both completed and as-yet uncompleted ITAAC. The NRC 
therefore expects that the information submitted by licensees in the 
Sec.  52.99(c)(2) notification will be sufficiently complete and 
detailed. Furthermore, the NRC expects that any contentions submitted 
by prospective intervenors regarding uncompleted ITAAC would focus on 
the inadequacies of the procedures and analytical methods described by 
the licensee for completing those ITAAC in the context of the 
reasonable assurance finding under Sec.  52.103(b)(2). Therefore, the 
level of detail provided by the licensee should be sufficient to allow 
a prospective intervenor to form such judgments by reference to that 
information. The NRC plans to prepare regulatory guidance providing 
further explanation of what constitutes ``sufficient information'' to 
demonstrate that the inspections, tests, or analyses for uncompleted 
ITAAC will be successfully completed and the acceptance criteria for 
the uncompleted ITAAC will be met.
    The NRC notes that, even though it did not include a provision 
requiring the completion of all ITAAC by a certain time prior to the 
licensee's scheduled fuel load date, the NRC will require some period 
of time to perform its review of the last ITAAC once the licensee 
submits its notification that the ITAAC has been successfully completed 
and the acceptance criteria met. In addition, the Commission will 
require some period of time to perform its review of the staff's 
conclusions regarding all of the ITAAC and the staff's recommendations 
regarding the Commission finding under Sec.  52.103(g). Therefore, 
licensees should structure their construction schedules to take into 
account these time periods. The NRC intends to develop regulatory 
guidance on the licensee's completion and NRC verification of ITAAC and 
will provide estimates of the time it expects to take to verify 
successful completion of various types of ITAAC. The NRC expects that 
such guidance, along with frequent communication with licensees during 
construction, will provide licensees with adequate information to plan 
initial fuel loading and related activities.
    Section 52.99(d) 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. The NRC is revising Sec.  52.99(d) in the 
final rule as a result of comments made on the proposed rule. Proposed 
Sec.  52.99(d) stated that, in the event that an activity is subject to 
an ITAAC derived from a referenced early site permit or standard design 
certification and the licensee has not demonstrated that the ITAAC has 
been met, the licensee may take corrective actions to successfully 
complete that ITAAC, request a variance from the early site permit 
ITAAC, or request an exemption from the standard design certification 
ITAAC, as applicable. The language in proposed Sec.  52.99(d) that 
referred to requesting variances to ESP ITAAC after the COL is issued 
is inconsistent with rule language in other sections of proposed part 
52 (e.g., Sec.  52.39(d)). Therefore, the NRC has adopted the 
commenters' suggestion to delete references to ESP ITAAC and ESP 
variances from Sec.  52.99(d).

[[Page 49385]]

    Paragraph (e)(1) requires the NRC to publish, at appropriate 
intervals until the last date for submission of requests for hearing 
under Sec.  52.103(a), notices in the Federal Register of the NRC 
staff's determination of the successful completion of inspections, 
tests, and analyses. Paragraph (e)(2) provides that the NRC shall make 
publicly available the licensee notifications under paragraphs (c)(1) 
and (c)(2). In general, the NRC expects to make the paragraph (c)(1) 
notifications availability shortly after the NRC has received the 
notifications and concluded that they are complete and detailed. 
Furthermore, by the date of the Federal Register notice of intended 
operation and opportunity to request a hearing on whether acceptance 
criteria have been or will be met (under Sec.  52.103(a)), the NRC will 
make available the notifications under paragraph (c)(2), and the 
notifications under paragraph (c)(2) for all ITAAC for which paragraph 
(c)(1) notifications have not been provided by the licensee.
    Finally, Sec.  52.103(h) states that ITAAC do not, by virtue of 
their inclusion in the combined license, 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 Sec.  52.97 (proposed Sec.  
52.98) and Section VIII of the design certification rules have been 
complied with.
    In a letter dated April 3, 2001 (comment 23), NEI requested that 
the NRC ``consider incorporating DCR [Design Certification Rule] 
general provisions into Subpart C as appropriate.'' The NRC has added 
these ITAAC requirements to Sec.  52.99, consistent with NEI's 
proposal, because it believes that these provisions embody general 
principles that are applicable to all holders of combined licenses.
    The NRC revised Sec.  52.99 in the final rule to delete the 
requirements in proposed Sec.  52.99(a). Proposed Sec.  52.99(a) 
required holders of COLs to comply with the provisions of Sec. Sec.  
50.70 and 50.71. Because the language in proposed Sec. Sec.  50.70 and 
50.71 requires COL holders to comply with their provisions, and because 
of the applicability provisions in Sec.  52.0(b), this duplicate 
requirement in Sec.  52.99 is unnecessary.
    The NRC has added a new paragraph (a) in Sec.  52.99 that requires 
a licensee to submit to the NRC, no later than 1 year after issuance of 
the combined license or at the start of construction as defined in 10 
CFR 50.10, whichever is later, its schedule for completing the 
inspections, tests, or analyses in the ITAAC. Licensees are required to 
submit updates to the ITAAC schedule every 6 months thereafter and, 
within 1 year of its scheduled date for initial loading of fuel, 
licensees must submit updates to the ITAAC schedule every 30 days until 
the final notification is provided to the NRC under Sec.  52.99(c). In 
Section V of the Supplementary Information of the 2006 proposed rule, 
the NRC requested stakeholder feedback on whether such a provision 
should be added to the final rule. Although some commenters did not 
believe that a regulatory requirement for submission of a schedule was 
necessary, the NRC believes it is necessary to ensure the NRC has 
sufficient information to plan all of the activities necessary for the 
NRC to support the Commission's finding whether all of the ITAAC have 
been met prior to the licensee's scheduled date for fuel load.
c. Section 52.73, Relationship to Other Subparts
    Section 52.73 clarifies that a design approval issued under subpart 
E of part 52 or a manufacturing license under subpart F of part 52 may 
also be referenced in an application for a combined license filed under 
10 CFR part 52. The former Sec.  52.73 only stated that a combined 
license may reference a standard design certification or an early site 
permit. The final rule incorporates into new Sec.  52.73(b) the 
requirement in the current Sec.  52.63(c) in order to clarify that this 
requirement applies to applicants for a combined license. This 
provision requires that, before 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 the 
information is necessary for the NRC 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 former Sec.  52.79 rather than the former Sec.  52.73. The 
NRC agrees with NEI that 10 CFR part 52 should be modified to clarify 
that the requirement in former Sec.  52.63(c) applied to applicants for 
a combined license, and that the last sentence be deleted. However, the 
Commission is adding the remaining provision to the original Sec.  
52.73(b), and not to Sec.  52.79, as recommended by NEI.
d. Section 52.75, Filing of Applications
    Section 52.75 provides requirements for the filing of combined 
license applications. The NRC has reformatted this section for 
consistency with the other subparts in 10 CFR part 52 and to replace 
the references to specific paragraphs within Sec. Sec.  50.4 and 50.30 
with general references to those sections. The specific references are 
no longer needed because the NRC is adopting conforming changes to 
Sec. Sec.  50.4 and 50.30 in this final rule which clarify which 
provisions are applicable to combined license applications.
e. Section 52.78, Content of Applications; Training and Qualification 
of Nuclear Power Plant Personnel
    Section 52.78 has been removed, and the requirements applicable to 
an applicant for, and holder of, a combined license with respect to the 
training program are moved to Sec.  50.120, where the requirements 
currently exist for holders of operating licenses.
f. Section 52.79, Contents of Applications; Technical Information in 
Final Safety Analysis Report; and Sec.  52.80, Contents of Application; 
Additional Technical Information
    Section 52.79 is reformatted to divide the requirements for the 
technical contents of a combined license application into two separate 
provisions. Section 52.79 covers requirements for the contents of the 
FSAR, and Sec.  52.80 covers requirements for the remainder of the 
technical content of a combined license application.
    Former Sec.  52.79 states that a combined license application must 
contain the technically relevant information required of applicants for 
an operating license by 10 CFR 50.34. The reference to 10 CFR 50.34 is 
removed and replaced with Sec.  52.79(a), which contains all of the 
relevant requirements from 10 CFR 50.34 that describe what must be 
included in the FSAR for a combined license application, including 
requirements that are currently applicable to both construction permit 
and operating license applications. In addition, requirements from 
other sections of 10 CFR part 50 (e.g., Sec. Sec.  50.48 and 50.63) are 
included. These requirements were issued after the current fleet of 
operating reactors were licensed and, therefore, were not required 
contents for these earlier FSARs. In making these modifications,

[[Page 49386]]

the NRC has attempted to capture all relevant requirements regarding 
contents of the FSAR for a combined license application.
    In addition, Sec.  52.79(a) contains requirements for descriptions 
of operational programs that need to be included in the FSAR to allow a 
reasonable assurance finding of acceptability. This amendment is in 
support of the Commission's direction to the staff in SRM-SECY-02-0067 
dated September 11, 2002, ``Inspections, Tests, Analyses, and 
Acceptance Criteria for Operational Programs (Programmatic ITAAC),'' 
that a combined license applicant was not required to have ITAAC for 
operational programs if the applicant fully described the operational 
program and its implementation in the combined license application. In 
this SRM, the Commission stated:

    [a]n ITAAC for a program should not be necessary if the program 
and its implementation are fully described in the application and 
found to be acceptable by the NRC at the COL stage. The burden is on 
the applicant to provide the necessary and sufficient programmatic 
information for approval of the COL without ITAAC.

    The Commission clarified its definition of fully described in SRM-
SECY-04-0032, ``Programmatic Information Needed for Approval of a 
Combined License Application Without Inspections, Tests, Analyses, and 
Acceptance Criteria,'' dated May 14, 2004, as follows:

    In this context, fully described should be understood to mean 
that the program is clearly and sufficiently described in terms of 
the scope and level of detail to allow a reasonable assurance 
finding of acceptability. Required programs should always be 
described at a functional level and at an increased level of detail 
where implementation choices could materially and negatively affect 
the program effectiveness and acceptability.

    Accordingly, the NRC is adding requirements for descriptions of 
operational programs. In doing so, the NRC has taken into account NEI's 
proposal to address SRM-SECY-04-0032 in its letter dated August 31, 
2005 (ML052510037). That proposal was reflected in SECY-05-0197 
(October 28, 2005, ML052770225), Attachment 1, and approved by the 
Commission in SRM-SECY-05-0197 dated February 22, 2006 (ML060530316). 
During the preparation of the final rule, the NRC discovered that 
several of the operational programs listed in SECY-05-0197 were not 
addressed in proposed Sec.  52.79. To ensure the list of requirements 
for the contents of applications is complete, the NRC is adding several 
new provisions to address operational programs in the final rule. 
Specifically, the NRC is adding requirements to Sec.  52.79 for COL 
applicants to include a description of: (1) The process and effluent 
monitoring and sampling program required by appendix I to 10 CFR part 
50 [Sec.  52.79(a)(16)(ii)]; (2) a training and qualification plan in 
accordance with the criteria set forth in appendix B to 10 CFR part 73 
[Sec.  52.79(a)(36)(ii)]; (3) a description of the radiation protection 
program required by Sec.  20.1101 [Sec.  52.79(a)(39)]; (4) a 
description of the fire protection program required by Sec.  50.48 
[Sec.  52.79(a)(40)]; and (5) a description of the fitness-for-duty 
program required by 10 CFR part 26 [Sec.  52.79(a)(44)]. During the 
preparation of the final rule, the NRC also noticed that the proposed 
rule had not completely implemented the Commission's direction 
regarding the treatment of operational programs in a COL application 
inasmuch as requirements to address operational program implementation 
were not included in proposed Sec.  52.79(a). Therefore, in the final 
rule, the NRC has added requirements to address the implementation of 
all operational programs required to be described in a COL application. 
This is consistent with the Commission's position in SRM-SECY-02-0067 
that a combined license applicant is not required to have ITAAC for 
operational programs if the applicant ``fully describes the operational 
program and its implementation'' in the combined license application 
[emphasis added].
    In addition, the NRC added a new provision to Sec.  52.79(a) in the 
final rule to address the application requirements in current Sec.  
20.1406. Section 20.1406 requires applicants for a license to describe 
in their application how facility design and procedures for operation 
will minimize, to the extent practicable, contamination of the facility 
and the environment, facilitate eventual decommissioning, and minimize, 
to the extent practicable, the generation of radioactive waste. To 
ensure that Sec.  52.79 contains a complete list of the requirements 
for the contents of a COL application, the NRC added paragraph (a)(45) 
to Sec.  52.79 to require COL applications to include the information 
required by Sec.  20.1406. This is not a new requirement but merely a 
pointer to an existing requirement to include this information.
    Section 52.79(a) requires that emergency plans submitted with a 
combined license application be included in the FSAR. This modification 
from the former rule is being made for consistency with Sec.  50.34 
which requires that emergency plans be included in the FSAR for 
operating license applications.
    The NRC is adding a new provision in Sec.  52.79(a)(29)(ii) that 
the applicant submit plans for coping with emergencies, other than the 
plans required by Sec.  52.79(a)(21). Paragraph 52.79(a)(21) requires 
the applicant to submit emergency plans complying with the requirements 
of Sec.  50.47 and 10 CFR part 50, appendix E. This requirement was 
drawn from the existing requirement in Sec.  50.34(b)(6)(v) which 
requires applicants to submit ``Plans for coping with emergencies, 
which shall include the items specified in appendix E.'' When this 
requirement was translated into the associated requirement for combined 
license applicants, the NRC inadvertently only included a portion of 
the requirements in Sec.  50.34(b)(6)(v), namely, the requirement in 
proposed Sec.  52.79(a)(21) to submit emergency plans. The NRC has 
corrected this omission in the final rule by including the new 
provision in Sec.  52.79(a)(29)(ii) to include other plans for coping 
with emergencies. This requirement is meant to capture, for example, 
emergency operating procedures as discussed in SRP Section 13.5.2.1, 
``Operating and Emergency Operating Procedures.''
    The NRC has moved the requirements contained in proposed Sec.  
52.79(a)(23) that addressed a request to conduct activities under Sec.  
50.10(e) and added them in a new Sec.  52.80(c). The NRC concluded that 
it is preferable to include both the list of proposed Sec.  50.10(e) 
activities and the redress plan as separate documents in the 
application, outside of both the site safety analysis report and the 
environmental report. The NRC's conclusion is based on the fact that 
the requirements in Sec.  50.10(e) address both safety and 
environmental issues. Additional changes were made to Sec. Sec.  51.50 
and 52.17 to implement this concept.
    Some commenters recommended that the requirement in Sec.  
52.79(a)(37) to demonstrate how operating experience insights have been 
incorporated into the plant design be deleted. The NRC disagrees with 
this comment. The NRC developed this requirement for future plants (see 
SRM on SECY-90-377) and it was implemented in past design certification 
applications by addressing NRC's generic letters and bulletins. The NRC 
agrees that insights from generic letters and bulletins should be 
incorporated into the latest revision of the standard review plan 
(SRP). Therefore, for plant designs that are

[[Page 49387]]

based on or are evolutions of nuclear plants that have operated in the 
United States, the applicant should use NRC's generic letters and 
bulletins issued after the most recent revision of the applicable SRP 
and 6 months before the docket date of the application. If the 
application is for a nuclear plant design that is not based on or is 
not an evolution of a nuclear plant that operated in the United States, 
the applicant should address how insights from any relevant 
international operating experience has been incorporated into that 
plant.
    Section 52.79(a)(41) requires that the applicant evaluate the 
facility against the standard review plan (SRP). For COL applicants 
that reference the same design certification rule and adopt a design-
centered approach in preparing their COL applications, the NRC expects 
that the ``reference application'' will fully conform with this 
requirement and then any follow-on applications will not need to 
provide the evaluations for the application information that is 
identical to the reference application. The NRC did not require 
applicants to evaluate their facility against RG 1.206, ``Combined 
License Applications for Nuclear Power Plants.'' However, the NRC 
believes that RG 1.206 can provide useful guidance to COL applicants in 
preparing their applications and that use of this guidance will 
facilitate the NRC's review.
    The NRC has moved the requirement that COL applicants submit a 
plant-specific PRA that was in proposed Sec.  52.80(a) to a new Sec.  
52.79(a)(46) in the final rule based on public comments. In addition, 
the NRC has revised the provision to require the applicants submit a 
description of their PRA and its results in their COL FSAR. The NRC 
agrees with some commenters who believed that applicants should not be 
required to submit their complete plant-specific PRA and that, instead, 
applicants should only be required to provide a summary description of 
the PRA and its results in their FSAR with the understanding that the 
complete PRA (e.g., codes) would be available for NRC inspection at the 
applicant's offices, if needed. The NRC expects that, generally, the 
information that it needs to perform its review of the COL application 
from a PRA perspective is that information that will be contained in 
applicants' FSAR Chapter 19. The NRC believes that COL application 
guidance that the NRC is developing is consistent with the industry 
comment in that the staff does not expect the complete PRA to be 
included in the COL applicant's FSAR. The guidance focuses on 
qualitative description of insights and uses, but also acknowledges 
that some quantitative PRA results should be submitted.
    Section 52.79(b) describes the variant on the requirements in Sec.  
52.79(a) for a combined license application that references an early 
site permit. Former Sec.  52.79(a) did not explicitly require the 
application to address whether the terms and conditions specified in 
the early site permit under Sec.  52.24 have been or will be 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, Sec.  52.79(b)(3) requires that the FSAR 
demonstrate that all terms and conditions that have been included in 
the early site permit will be satisfied by the date of issuance of the 
combined license. The NRC is revising Sec.  52.79(b)(3) in the final 
rule based on public comments to add an exclusion for terms and 
conditions imposed under Sec.  50.36(b) because such environmental 
conditions should be addressed in the environmental report and not in 
the final safety analysis report. In addition, the Commission is 
revising this paragraph to add a provision that any terms or conditions 
of the early site permit that could not be met by the time of issuance 
of the combined license must be set forth as terms or conditions of the 
combined license. This provision is needed to address terms or 
conditions of the early site permit that are related to activities that 
will not take place until after issuance of the combined license, such 
as construction activities. A similar change is being made to 
Sec. Sec.  52.79(d)(3) and (e)(3) for referenced design certifications 
and manufacturing licenses.
    The NRC is making a revision to the language in proposed Sec.  
52.79(b)(1) in the final rule. Proposed Sec.  52.79(b)(1) stated that 
the FSAR for a combined license application referencing an early site 
permit need not contain information or analyses submitted to the NRC in 
connection with the early site permit. This rule language led to a 
great deal of discussion both within the NRC and in public meetings on 
combined license application guidance as to what the NRC expected to 
see in a combined license application that referenced an early site 
permit. The NRC has concluded that the FSARs in these combined licenses 
applications must either include or incorporate by reference the SSAR 
for the early site permit. The SSAR must be included or incorporated 
into the COL FSAR to ensure that matters addressed in the SSAR legally 
become part of the FSAR upon issuance of the COL. This will also ensure 
that the information in the SSAR is subject to control under Sec.  
50.59 after issuance of the COL. For these reasons, the NRC is 
modifying the language in Sec.  52.79(b)(1) to state that the final 
safety analysis report need not contain information or analyses 
submitted to the NRC in connection with the early site permit. However, 
the final safety analysis report must either include or incorporate by 
reference the early site permit site safety analysis report. With this 
modification, the NRC intends to convey that the combined license 
applicant referencing the early site permit does not need to resubmit, 
for NRC review, information or analyses that were already reviewed and 
resolved in the early site permit proceeding (such as information 
provided in responses to NRC requests for additional information). At 
the same time, the NRC's goal is to provide COL applicants clear 
guidance as to what the combined license application must contain to be 
considered complete. For similar reasons, the NRC is also modifying the 
language in proposed Sec. Sec.  52.79(c)(1), (d)(1), and (e)(1) to 
include the provision that the FSAR in the COL application must either 
include or incorporate by reference the FSAR for the design approval, 
design certification, or manufacturing license that it is referencing. 
Note that each of the existing design certification rules covered in 
appendices A through D of part 52 prohibit the use of incorporation by 
reference in COL FSARs that reference them. At the time those rules 
were issued, the NRC was concerned that the staff would not have easy 
access to the final version of the design certification FSAR (i.e., 
DCD) if it were not included in the COL application. The NRC will 
continue to put restrictions in individual design certification rules 
(and possibly in early site permits, design approvals, or manufacturing 
licenses) if it does not have confidence that the safety analysis 
reports can be easily accessed by the staff if they are incorporated by 
reference in COL applications.
    Section 52.79(c) describes the requirements for combined license 
applications that reference a standard design approval. Previously, no 
guidance was provided regarding a combined license application that 
referenced a standard design approval. The requirements in Sec.  
52.79(c) are essentially the same as those for a combined license 
application that references a standard design certification in Sec.  
52.79(d).
    Section 52.79(d) describes the requirements for combined license 
applications that reference a standard

[[Page 49388]]

design certification. Section 52.79(d) states that the FSAR for a 
combined license application referencing a standard design 
certification need not contain information or analyses submitted to the 
NRC in connection with the design certification. However, the final 
safety analysis report must either include or incorporate by reference 
the standard design certification final safety analysis report (see 
discussion above) and must contain, in addition to the information and 
analyses otherwise required, information sufficient to demonstrate that 
the characteristics of the site fall within the site parameters 
specified in the design certification. In addition, paragraph (d) 
requires that the plant-specific PRA information must use the PRA 
information for the design certification and must be updated to account 
for site-specific design information and any design changes or 
departures. In the case where a COL application is referencing a design 
certification, the NRC only expects the design changes and differences 
in the modeling (or its uses) pertinent to the PRA information to be 
addressed to meet the submittal requirement of Sec.  52.79(d)(1). 
Section 52.79(d) also requires that the FSAR demonstrate that the 
interface requirements established for the design under Sec.  52.47 
have been met and that all requirements and restrictions that may have 
been set forth in the referenced design certification rule be satisfied 
by the date of issuance of the combined license.
    Section 52.79(e) describes the requirements for a combined license 
application that references a manufactured reactor. Previously, no 
guidance was provided regarding a combined license application that 
referenced a manufactured reactor. These requirements are similar to 
those for the content of an FSAR for a combined license referencing a 
design certification. Specifically, Sec.  52.79(e) states that the FSAR 
need not contain information or analyses submitted to the NRC in 
connection with the manufacturing license. However, the final safety 
analysis report must either include or incorporate by reference the 
manufacturing license final safety analysis report and must contain, in 
addition to the information and analyses otherwise required, 
information sufficient to demonstrate that the site characteristics 
fall within the site parameters specified in the manufacturing license. 
This language was slightly different in the proposed rule and has been 
corrected in the final rule to be consistent with Sec.  52.79(d). In 
addition, Sec.  52.79(e) requires that the plant-specific PRA 
information must use the PRA information for the manufactured reactor 
and must be updated to account for site-specific design information and 
any design changes or departures. Section 52.79(e) also requires that 
the FSAR demonstrate that the interface requirements established for 
the design have been met and that all terms and conditions that have 
been included in the manufacturing license be satisfied by the date of 
issuance of the combined license.
    Section 52.80 is added to cover the required technical contents of 
a combined license application that are not contained in the FSAR. 
These application contents include the ITAAC, the environmental report, 
and the request to perform activities under Sec.  50.10(e) with the 
associated redress plan. This last item was moved to Sec.  52.80(c) in 
the final rule from its location in Sec.  52.79(a)(23) in the proposed 
rule. The NRC concluded that it is preferable to include both the list 
of proposed activities and the redress plan as separate documents in 
the application, outside of both the site safety analysis report and 
the environmental report. The NRC's conclusion is based on the fact 
that the requirements in Sec.  50.10(e) address both safety and 
environmental issues. Additional changes were made to Sec. Sec.  51.50 
and 52.17 to implement this concept.
g. Section 52.81, Standards for Review of Applications
    10 CFR parts 54 and 140 are added to the list of standards that the 
NRC will use to review combined license applications. Part 54 addresses 
applications for renewal of combined licenses and part 140 includes the 
requirements applicable to nuclear reactor licensees with respect to 
financial protection and Indemnity Agreements to implement Section 170 
of the AEA, commonly referred to as the Price-Anderson Act.
h. Section 52.83, Finality of Referenced NRC Approvals; Partial Initial 
Decision of Site Suitability
    The former Sec.  52.83, Applicability of part 50 provisions, is 
removed and replaced by a new section addressing the finality of NRC 
approvals which are referenced in a combined license application. 
Former Sec.  52.83 provides that, unless otherwise specifically 
provided for in subpart C to part 52, all provisions of 10 CFR part 50 
and its appendices applicable to holders of construction permits for 
nuclear power reactors also apply to holders of combined licenses. 
Similarly, Sec.  52.83 provides that all provisions of 10 CFR part 50 
and its appendices applicable to holders of operating licenses also 
apply to holders of combined licenses issued under this subpart, once 
the Commission has made the findings required under Sec.  52.99. The 
NRC believes that the former Sec.  52.83 is not necessary because this 
proposed rulemaking will provide conforming changes throughout 10 CFR 
part 50 (as well as all other parts in Title 10 Chapter I) to identify 
which requirements are applicable to combined license applicants and 
holders. Former Sec.  52.83 also provides provisions that address the 
duration of a combined license and these provisions would be moved to 
proposed Sec.  52.104, Duration of combined license.
    The new Sec.  52.83 states that, if an application for a combined 
license references an early site permit, design certification rule, 
standard design approval, or manufacturing license, the scope and 
nature of matters resolved for the application and any combined license 
issued are governed by the relevant provisions addressing finality, 
including Sec. Sec.  52.39, 52.63, 52.98, 52.145, and 52.171. This 
provision clarifies the relationship between a combined license 
application and any other license or regulatory approval that an 
applicant may reference in the combined license application as far as 
issue resolution is concerned.
i. Section 52.89, Environmental Review
    Section 52.89 is removed and reserved for future use. Former Sec.  
52.89 required that, if a combined license application references an 
early site permit or a certified standard design, the environmental 
review must focus on whether the design of the facility falls within 
the 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. Former Sec.  52.89 further stated 
that, if the application does not reference an early site permit or a 
certified standard design, the environmental review procedures set out 
in 10 CFR part 51 must be followed, including the issuance of a final 
environmental impact statement, but excluding the issuance of a 
supplement under Sec.  51.95(a). This provision is removed because the 
requirements for compliance with NEPA are now captured in Sec.  
52.79(a) and in the revisions to part 51.

[[Page 49389]]

j. Section 52.91, Authorization To Conduct Site Activities
    Section 52.91(a)(2) formerly provided requirements for a combined 
license application that does not reference an early site permit, but 
that contains a site redress plan and states that 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 accordance with Sec.  
52.79(a)(3), and obtaining the separate authorization required by 10 
CFR 50.10(e)(1). This provision further states that authorization must 
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). This provision is amended to state that 
authorization may [emphasis added] 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). This 
amendment is consistent with Sec.  52.91(a)(3), which states that 
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). The NRC believes that may is the proper term to 
use in both of these provisions, to reflect the NRC's residual 
authority to decline to authorize the ESP holder to conduct Sec.  
50.10(e)(3)(i) activities, even if the NRC's regulations are met.
k. Section 52.93, Exemptions and Variances
    Paragraph (a) of Sec.  52.93, which includes a discussion of the 
requirements regarding requests for an exemption from any part of a 
referenced design certification, is revised to state that the 
Commission may grant the request if it determines that the exemption 
complies with any exemption provisions of the referenced design 
certification rule, or with Sec.  52.63 if there are no applicable 
exemption provisions in the referenced design certification rule. This 
provision formerly referred to compliance with Sec.  50.12(a). The NRC 
is revising paragraph (b) of this section in the final rule to include 
an allowance for applicants to request a variance from the early site 
permit SSAR. The allowance for requesting variances to the SSAR was 
inadvertently omitted in the proposed rule. Because the majority of the 
early site permit information that a combined license applicant will be 
referencing will be the information in the SSAR, it is logical that the 
allowance to request variances be extended to the information in the 
SSAR given that the NRC is allowing variances to the permit itself. In 
the final rule, the NRC is also adding a provision to paragraph (b) of 
this section that precludes the NRC from issuing a variance once a 
construction permit, operating license, or combined license referencing 
the early site permit is issued; any changes that would otherwise 
require a variance should instead be treated as an amendment to the 
construction permit or combined license.
    Section 52.93 is also revised in the final rule to add a discussion 
of requests for departures from a referenced nuclear power reactor 
manufactured under a manufacturing license in new paragraph (c) of this 
section. This provision was inadvertently omitted in the proposed rule, 
although similar provisions were addressed in the proposed rule in 
Sec. Sec.  52.98 and 52.171. However, the proposed rule incorrectly 
used the term ``variance'' to describe an application-specific change 
to a reactor manufactured under a manufacturing license. The NRC has 
corrected these provisions in the final rule to use the term 
``departure'' for such changes, consistent with the terminology used 
for changes to a referenced design certification. New paragraph (c) of 
this section is consistent with these other sections and states that an 
applicant for a combined license who has filed an application 
referencing a nuclear power reactor manufactured under a manufacturing 
license may include in the application a request for a departure from 
one or more design characteristics, site parameters, terms and 
conditions, or approved design of the manufactured reactor. The NRC may 
grant a request only if it determines that the departure will comply 
with the requirements of 10 CFR 52.7, and that the special 
circumstances outweigh any decrease in safety that may result from the 
reduction in standardization caused by the departure. The criteria for 
granting the departure is the exemption criterion in Sec.  52.7; 
however, the departure itself is not considered an exemption (unless, 
of course, the departure also involves a non-compliance with an 
underlying Commission regulatory requirement in 10 CFR Chapter I). 
Thus, the Commission will not approve a departure unless the Commission 
finds, in addition to the routine exemption criteria in Sec.  52.7, 
that special circumstances outweigh any decrease in safety that may 
result from the reduction in standardization caused by the departure. 
These limitations are intended to maintain the standardization of 
manufactured reactors in operation to the extent practicable. The 
licensee may not depart from the design characteristics, site 
parameters, terms and conditions, or approved design of the 
manufactured reactor through the provisions of Sec.  50.59.
    Finally, the provision contained in paragraph (c) of this section 
in the 2006 proposed rule (and in paragraph (b) in the former rule) has 
been moved to paragraph (d) of this section in the final rule. This 
provision states that issuance of a variance under paragraph (b) or a 
departure under paragraph (c) is subject to litigation during the 
combined license proceeding in the same manner as other issues material 
to that proceeding.
l. Section 52.97, Issuance of Combined Licenses
    The NRC has modified Sec.  52.97 to be more consistent with the 
parallel provision in Sec.  50.50, Issuance of licenses and 
construction permits, by including requirements that, after conducting 
a hearing and receiving the report submitted by the ACRS, the NRC finds 
that there is reasonable assurance that the applicant is technically 
and financially qualified to engage in activities authorized; and that 
issuance of the license will not be inimical to the common defense and 
security or to the health and safety of the public. Section 52.97(c) is 
added, consistent with Sec.  50.50, which states that a combined 
license shall contain conditions and limitations, including technical 
specifications, as the NRC deems necessary and appropriate. Former 
Sec.  52.97(b)(2) is moved to new Sec.  52.98 because the issues 
addressed in this section are issues associated with finality of 
combined license provisions.
m. Section 52.98, Finality of Combined Licenses; Information Requests
    Section 52.98, which addresses the finality associated with the 
issuance of combined licenses, is added to subpart C of part 52, 
consistent with the other subparts in 10 CFR part 52. Section 52.98(a) 
states that, after issuance of a combined license, the Commission may 
not modify, add, or delete any term or condition of the combined 
license, the design of the facility, the inspections, tests, analyses, 
and acceptance criteria contained in the license which are not derived 
from a referenced standard design certification or manufacturing

[[Page 49390]]

license, except in accordance with the provisions of Sec. Sec.  52.103 
or 50.109, as applicable.
    Section 52.98 includes provisions 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. Section 
52.98(b) states that the change processes in 10 CFR part 50 apply to a 
combined license that does not reference a design certification rule or 
a reactor manufactured under a manufacturing license. Section 52.98(c) 
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.
    Section 52.98(d) is added to address changes to a combined license 
that references a reactor manufactured under a manufacturing license. 
Section 52.98(d)(1) states that, if the combined license references a 
reactor manufactured under a subpart F manufacturing license, then 
changes to or departures from information within the scope of the 
manufactured reactor's design are subject to the change processes in 
Sec.  52.171. Note that the proposed rule incorrectly used the term 
``variance'' to describe an application-specific change to a reactor 
manufactured under a manufacturing license. The NRC has corrected this 
provision in the final rule to use the term ``departure'' for such 
changes, consistent with the terminology used for changes to a 
referenced design certification. Section 52.98(d)(2) states that 
changes that are not within the scope of the manufactured reactor's 
design are subject to the applicable change processes in 10 CFR part 50 
(e.g., Sec. Sec.  50.54, 50.59, and 50.90). The NRC made all of these 
requirements to clarify, in one location, the finality provisions 
applicable to all portions of a combined license.
    Finally, the NRC has added a new paragraph (g) to the ``finality'' 
section in each subpart of part 52, including Sec.  52.98, entitled 
``Information requests,'' which delineates the restrictions on the NRC 
for information requests to the holder of the combined license. This 
provision is analogous to the former provision on information requests 
in paragraph 8 of appendix O to parts 50 and 52, and is based upon the 
language of Sec.  50.54(f). For combined licenses, this proposed 
provision is in Sec.  52.98(g), and requires the NRC to evaluate each 
information request of the holder of a combined license to determine 
that the burden imposed by the information request is justified in 
light of the potential safety significance of the issue to be addressed 
in the information request. The only exception is for information 
requests seeking to verify compliance with the current licensing basis 
of the facility. If the request is from the NRC staff, the request will 
first have to be approved by the EDO or his or her designee.
n. Section 52.103, Operation Under a Combined License
    Section 52.103(g) formerly required the NRC to find that the 
acceptance criteria in the combined license are met before operation of 
the facility, but did not refer to loading of fuel. However, Sec.  
52.103(f) stated that fuel loading and operation under the combined 
license will not be affected by the granting of a petition to modify 
the terms and conditions of the combined license unless a Commission 
order is made immediately effective. In the proposed rule, this section 
was amended to require the NRC to find that the acceptance criteria in 
the combined license are met before fuel load and operation of the 
facility. The NRC has decided not to adopt the proposed rule language 
which would have precluded loading of fuel into the reactor until 
acceptance criteria have been met. The NRC believes that the rule 
should reflect, as closely as possible, the statutory requirement in 
Section 185.b of the AEA. The NRC has historically viewed ``operation'' 
as including loading of fuel into the reactor, however it is not 
necessary to change the language of Sec.  52.103(g) to continue the 
historical practice. The NRC believes that this is the common 
interpretation of Sec.  52.103(g).
o. Section 52.104, Duration of Combined License; Sec.  52.105, Transfer 
of Combined License; Sec.  52.107, Application for Renewal; Sec.  
52.109, Continuation of Combined License; and Sec.  52.110, Termination 
of License
    Five new provisions are added to subpart C of part 52 for 
consistency with the other subparts in 10 CFR part 52 and to parallel 
requirements in 10 CFR part 50 for operating licenses. Section 52.104, 
addresses the duration of a combined license and contains requirements 
that formerly existed in Sec.  52.83. In addition, the Commission has 
amended these requirements to indicate that, where the Commission has 
allowed operation under a combined license during an interim period 
under Sec.  52.103(c), the period of operation is not to exceed 40 
years from the date allowing operation during the interim period.
    Section 52.105 provides requirements for the transfer of a combined 
license that refer the applicant to Sec.  50.80. Section 52.107 
provides a reference to 10 CFR part 54 for the renewal of a combined 
license.
    Section 52.109 provides provisions for the continuation of a 
combined license and Sec.  52.110 would provide requirements for the 
termination of a combined license. Formerly, part 52 did not address 
decommissioning of combined licenses (reactors that are manufactured 
under a part 52 manufacturing license do not raise decommissioning 
concerns until they are emplaced at a site, inasmuch as a manufacturing 
license does not permit loading of fuel or operation) and the 
termination of the combined license. By contrast, Sec. Sec.  50.51 and 
50.82 address the permanent shutdown of a nuclear power plant, its 
decommissioning, and the termination of the part 50 operating license. 
There are two possible ways of addressing this omission: Sec. Sec.  
50.51 and 50.82 could be modified to reference combined licenses under 
part 52, or the provisions analogous to these sections could be added 
to part 52. The NRC believes that the second alternative is the best 
approach. The combined license holder's responsibilities upon 
expiration of its license is more a matter of regulatory authority and 
therefore is best placed in part 52. While the question is closer with 
respect to decommissioning, the NRC believes that most users would 
likely turn to part 52 rather than part 50 to determine the 
requirements for decommissioning, inasmuch as decommissioning involves 
questions of both procedure and technical requirements.
9. Subpart D, Reserved
10. Subpart E, Standard Design Approvals (Sec. Sec.  52.131 Through 
52.147)
    The former appendix O to part 52 set forth the requirements for NRC 
staff approval of a standard design for a nuclear plant or a major 
portion of a nuclear plant. This licensing process was first adopted by 
the NRC in 1975 and has been used many times, including issuance of 
four final design approvals (FDAs) under appendix O to part 52 from 
1994 through 2004. These FDAs were issued during previous design 
certification reviews when FDAs were a prerequisite to certification of 
a standard plant design (see SOC

[[Page 49391]]

discussion on 10 CFR 52.43 in this document).
    When the NRC adopted part 52 in 1989, the Commission did not re-
examine the regulatory scheme for standard design approvals to 
determine if the bases for adopting part 52 and the licensing processes 
codified in part 52 would also be an impetus for reorganizing the 
design approval process. However, the Commission did undertake a re-
examination of appendix O to part 52 in the 2003 proposed rule and 
proposed certain changes. In view of the substantial reorganization and 
rewriting of part 52 in this rulemaking, the Commission gave further 
consideration to the licensing process in appendix O to part 52 and has 
made additional changes to enhance the regulatory effectiveness and 
efficiency of that licensing process.
    The Commission continues to believe that the best approach for 
obtaining early resolution of design issues is through the design 
certification process in subpart B of part 52. Design certification 
will provide greater finality and standardization than the design 
approval process. Consequently, the Commission favors use of the design 
certification process, which suggests that the design approval process 
could be eliminated. However, given the frequent use of appendix O to 
part 52 in the past, the Commission has decided to retain this process 
and to reorganize and reformat the design approval process to be 
consistent with other subparts.
    The design approval process, formerly located in appendix O to part 
52, has been moved to subpart E of part 52 and reformatted to be 
consistent with other subparts. A new Sec.  52.133 was created to 
describe the relationship of the design approval process with other 
subparts. An FDA may be referenced in an application for a construction 
permit or operating license under part 50 or a design certification, 
combined license, or manufacturing license under part 52.
    The filing requirements for design approvals are consistent with 
other subparts of part 52. The applicants may still request approval of 
either the entire facility or major portions thereof, but the 
applications are limited to final design information. There are several 
reasons for this change. First, the Commission's recent experience with 
FDAs and design certifications demonstrates that nuclear plant 
designers are technically capable of developing essentially complete 
and final design information for NRC review and approval. Furthermore, 
the economic incentives with respect to design certification also apply 
to final design approvals. In addition, approval of final design 
information removes the unpredictability of issuing a construction 
permit that references only preliminary design information and 
initiating construction while the final design information is being 
developed. Approval of a final design ensures early consideration and 
resolution of technical matters before there is any substantial 
commitment of resources associated with the construction of the plant, 
which will greatly enhance regulatory stability and predictability.
    The Commission has decided that the contents of applications for 
design approvals should contain essentially the same technical 
information that is required of design certification applications 
(e.g., demonstration of compliance with technically relevant Three Mile 
Island requirements, proposed technical resolutions of unresolved 
safety issues and medium- and high-priority generic safety issues, and 
design-specific probabilistic risk assessment information).
    Regarding applications for a major portion of the standard plant 
design, such as the nuclear steam supply system, the application only 
needs to contain the information required for the contents of 
applications that are applicable to the major portion of the plant for 
which NRC staff approval is requested.
    The requirements for contents of applications for design approvals 
(Sec.  52.137) were renumbered to be consistent with the numbering of 
requirements in Sec.  52.47. Also, many of the public comments on 
contents of applications for design certification apply to the 
requirements for design approvals (see the SOC of this document for the 
discussion for Sec.  52.47). Some commenters recommended that the 
requirement for coping with emergencies [Sec.  52.137(a)(11)] be 
deleted because applicants for design approvals will not be responsible 
for certain emergency planning design features. The Commission 
disagrees with this comment. This requirement was taken from the 
original appendix O of part 52, paragraph 3, and it applies to design 
features for coping with emergencies in the operation of the reactor, 
not for emergency planning.
    A new Sec.  52.139, which specifies the standards that will be used 
to review applications for design approvals and new Sec. Sec.  52.145 
and 52.147, which specify the finality and duration of design approvals 
was added to be consistent with other subparts. In a letter dated 
November 13, 2001, NEI commented that ``Industry recommends FDAs be 
valid for 15 years.'' The Commission agrees with NEI's recommendation 
and has decided that the duration of standard design approvals should 
correspond to the duration of design certifications, inasmuch as both 
design approvals and design certifications constitute approvals of 
nuclear power plant designs, and the period of effectiveness of the 
approval from a technical standpoint is not a function of whether the 
approval is granted by the NRC staff or the Commission. Some commenters 
recommended that Sec.  52.147 be rewritten to provide for renewals of 
standard design approvals. The Commission disagrees with this comment. 
The original appendix O to part 52 did not contain a process for 
renewing design approvals and most of the design approvals issued under 
appendix O to part 52 were for a 5-year duration. In this rulemaking, 
the Commission has tripled the duration for a design approval and 
believes that renewals will not be necessary. Also, as stated before, 
the Commission favors the use of the design certification process, 
which includes a process for renewals.
11. Subpart F, Manufacturing Licenses
    The following discussion explains the requirements in subpart F of 
part 52 generically, and covers Sec. Sec.  52.151, 52.153, 52.155, 
52.156, 52.157, 52.159, 52.161, 52.163, 52.165, 52.167, 52.169, 52.171, 
52.173, 52.175, 52.177, 52.179, and 52.181.
    Former appendix M of parts 50 and 52 set forth the NRC's 
requirements governing manufacturing licenses. Appendix M, which was 
first adopted by the NRC in 1973 as an appendix to part 50, provided 
for issuance of a license authorizing the manufacture of a nuclear 
power reactor to be incorporated into a nuclear power plant under a 
construction permit and operated under an operating license at a 
different location from the place of manufacture. Under the licensing 
regime in former appendix M, the NRC did not approve a final reactor 
design to be manufactured as part of the issuance of the manufacturing 
license. Rather, analogous to the two-step construction permit/
operating license process, the NRC would issue a manufacturing license 
based upon the review and approval of a preliminary design equivalent 
to that provided in a construction permit application. Upon issuance of 
the manufacturing license, manufacturing of the reactor can commence, 
although the NRC must approve the final design of the manufactured 
reactor by license amendment before the manufactured reactor may be 
transported from the

[[Page 49392]]

place of manufacture to the site where it is to be operated.
    When the NRC adopted part 52 in 1989, it added appendix M to part 
52. However, the NRC did not re-examine the regulatory scheme for 
manufacturing licenses in order to determine if the bases for adopting 
part 52 would also be an impetus for changing the regulatory scheme for 
manufacturing licenses. Nor did the NRC undertake such a re-examination 
as part of the process leading to the 2003 proposed rule. However, the 
NRC has reconsidered the efficacy of the manufacturing license process 
in former appendix M to part 52, and has decided to adopt substantial 
changes to those requirements in order to enhance regulatory 
effectiveness and efficiency. These new requirements are contained in a 
new subpart F to part 52.
    The most important shift in the manufacturing license concept in 
subpart F is that a final reactor design, equivalent to that required 
for a standard design certification under part 52 or an operating 
license under part 50, must be submitted and approved before issuance 
of a manufacturing license. There are several reasons for this shift. 
First, the Commission's experience with standard design certifications 
demonstrates that nuclear power plant designers are technically capable 
of developing a complete reactor design for Commission review. 
Furthermore, the economic incentives and limitations with respect to 
approval of a standard reactor design certification also apply to the 
approval of a design of a manufactured reactor. Indeed, one could argue 
that the holder of a manufacturing license may structure the commercial 
transaction to reduce the economic risk associated with the application 
for a manufacturing license for a final reactor design, as compared to 
the economic risk associated with a standard design certification. 
Second, approval of a final reactor design removes the former awkward 
regulatory process of issuing a manufacturing license, and subsequently 
amending the license when a final design is submitted. Approval of a 
final design ensures early consideration and resolution of technical 
matters before there is any substantial commitment of resources 
associated with the actual manufacture of the reactor, which will 
greatly enhance regulatory stability and predictability. Finally, 
Commission approval of standardized manufacturing processes, coupled 
together with the potential for a stable workforce and the application 
of manufacturing process feedback, has great opportunities for 
maintaining and even improving the quality and consistency of 
manufacture, as compared to the traditional method of constructing 
reactors onsite by a variety of contractors and subcontractors.
    The technical information required to be included in an application 
for a manufacturing license, as set forth in Sec. Sec.  52.157 and 
52.158, reflects both the expansion of the scope of approval to include 
the final design of the reactor to be manufactured, as well as lessons 
learned with respect to the NRC's review of early site permits. Section 
52.157, which sets forth the technical information to be submitted in 
support of the design of a reactor, is derived from the existing 
requirements in current part 52, subparts B and C, governing the 
technical information to be submitted in support of an application for 
a standard design certification and combined license. In addition, 
Sec.  52.157 requires that the application address the provisions with 
respect to the demonstration by test, analysis, experience, or a 
combination thereof, of simplified, inherent, passive, or other 
innovative means to accomplish safety functions, or the results of 
testing of a prototype plant, as set forth in revisions to Sec.  50.43. 
As discussed separately with respect to Sec.  50.43, these testing and 
prototype requirements incorporated into Sec.  50.43 were derived from 
the former requirements in Sec.  52.47(b).
    Information which must be submitted as part of an application, but 
is not typically considered part of a final safety analysis report, is 
identified in Sec.  52.158. This includes proposed ITAAC to be used by 
the licensee who will construct and operate a nuclear power plant at 
its site using the manufactured reactor and an environmental report for 
the manufactured reactor. Note that, in the final rule, the NRC has 
moved proposed Sec.  52.158(a) to a new Sec.  52.157(f)(31) which 
requires that manufacturing license applicants submit a description of 
the design-specific PRA and its results in the FSAR. The NRC agrees 
with some commenters that applicants should not be required to submit 
their complete design-specific PRA and that, instead, applicants should 
only be required to provide a summary description of the PRA and its 
results in their FSAR with the understanding that the complete PRA 
(e.g., codes) would be available for NRC inspection at the applicant's 
offices, if needed. The NRC expects that, generally, the information 
that it needs to perform its review of the manufacturing license 
application from a PRA perspective is that information that will be 
contained in applicants' FSAR Chapter 19.
    The environmental report must address SAMDAs, similar to standard 
design certifications, because the design approval stage is usually the 
most cost-effective opportunity for incorporating design features for 
addressing severe accidents. The NRC notes that the environmental 
report need not address environmental impacts associated with the 
actual manufacture of the reactor at any manufacturing location, 
inasmuch as a manufacturing license does not represent NRC approval of 
any specific location, facility, or appurtenance for manufacturing. 
Rather, the NRC is approving a reactor design for manufacture and the 
ITAAC for verifying that it has been acceptably manufactured and 
integrated into a nuclear power facility so that it can be safely 
operated in accordance with the approved manufactured reactor design, 
the NRC's regulations, and the requirements of the AEA. These 
determinations were reflected in proposed Sec. Sec.  52.158(c)(1), 
51.54, and 51.75(c)(3). However, in the final rule, the Commission has 
removed from proposed Sec. Sec.  52.158(c)(1) and (2) (final Sec. Sec.  
52.158(b)(1) and (2)) the rule language addressing the content of the 
environmental report, and integrated that language into Sec. Sec.  
51.54 and 51.75(c)(3). Proposed Sec.  52.158(c)(2) (final Sec.  
52.158(b)(2)) has been revised in the final rule to address the scope 
of the environmental report if the manufacturing license application 
has referenced a standard design certification.
    Section 52.163 of the March 2006 proposed rule would have required 
that the NRC conduct a ``mandatory'' hearing in connection with the 
initial issuance of a manufacturing license, even though the AEA does 
not require a mandatory hearing for issuance of manufacturing licenses. 
For the reasons set forth in the NRC's response to Commission Question 
2, and the discussion on Sec. Sec.  2.104 and 2.105, the NRC has 
decided not to require a ``mandatory'' hearing for initial issuance of 
a manufacturing license, and Sec.  52.163 is revised in the final rule 
to refer to a publication of a notice of proposed action under Sec.  
2.105, rather than a notice of hearing under Sec.  2.104.
    In light of the NRC's review and approval of a final design as part 
of issuance of a manufacturing license, the final rule provides a 
greater degree of finality to a manufacturing license as compared with 
a standard design certification. Under Sec.  52.171(a)(1), the same 
degree of issue finality accorded to the ``certified design'' applies 
throughout the term of the manufacturing license. Under this

[[Page 49393]]

provision, the NRC may not impose any change or modification to the 
approved design (including site parameters, or design characteristics) 
for the manufacturing license unless the NRC determines that the change 
or modification is necessary either for adequate protection or for 
compliance with requirements applicable and in effect at the time the 
manufacturing license was issued. Similarly, the manufacturing license 
holder may not make changes to the design under the provisions of 10 
CFR 50.59. Any change to the design will require a license amendment. 
The Commission regards this as similar to the level of change control 
imposed on designs which are the subject of a standard design 
certification. The Commission is imposing this stringent level of 
change control because one of the key reasons for licensing 
manufactured reactors is to enhance standardization--one of the 
original objectives of the 1989 part 52 rulemaking. Unlike design 
certification, which is an approval of a ``paper design,'' the NRC's 
proposed concept of a manufacturing license is pre-approval of the 
procurement, manufacturing, and quality assurance processes that 
translates the approved reactor design into a manufactured assembly in 
a controlled environment, with the capability to optimize techniques 
and procedures based upon feedback. Some of these advantages may be 
lost if each ``manufactured'' reactor were treated as a ``one-off'' 
custom product. Imposing the discipline of a license amendment process 
should ensure that a profusion of changes are not made to the approved 
design at random intervals. The Commission disagrees with commenters on 
the proposed rule that the design of a manufactured reactor should be 
subject to less-stringent change provisions than a standard design 
certification. The commenters have not demonstrated that there are 
special or unique aspects of manufacturing, as compared with the 
construction of a nuclear power plant based upon a referenced standard 
design certification, that would weigh against maintaining the high 
degree of design standardization achieved by design certification. One 
commenter correctly noted that changes in such manufacturing matters as 
procurement, manufacturing processes, or quality assurance are not 
subject to the proposed Sec.  52.171(b)(1) change restriction, because 
these matters do not constitute changes to the approved design of the 
reactor to be manufactured. These changes would be governed by the 
applicable change process and restrictions already established in the 
Commission's regulations such as Sec.  50.59, and Sec.  50.54(a), and 
may not require license amendments.
    The only relevant rationale provided by the commenters is that 
obsolescence of components and component manufacturers' changes would 
necessitate minor changes to the reactor design over a 15-year period. 
Although the Commission acknowledges the likelihood of these factors, 
the NRC staff does not see any reason why these factors are more likely 
to affect the design of a manufactured reactor as compared with the 
design approved in a design certification. It is not clear why a change 
in component sourcing would necessarily result in a ``design change'' 
requiring an amendment to the manufacturing license. Finally, the 
Commission notes that the proposed rule does not mandate ``zero changes 
in a reactor design.'' As specifically stated in the SOC of the March 
13, 2006 (71 FR 12801), proposed rule (second column), proposed Sec.  
52.171(b)(1) would allow the manufacturer to make changes to the 
approved design to be manufactured, albeit by license amendment.
    The final rule provides that the term of a manufacturing license to 
be for no less than 5, or more than 15 years from the date of issuance. 
The Commission established the 15-year maximum term to be consistent 
with the maximum term for a standard design certification. The 5-year 
minimum term was established by the Commission to encourage the use of 
a manufacturing license for the manufacture of more than one nuclear 
power reactor. The language of Sec.  52.171 has been corrected in the 
final rule by replacing the reference in paragraph (b)(1) to Sec.  
50.12 with a reference to Sec.  52.7, and replacing the term, 
``exemption,'' in paragraph (b)(2) with ``departure.''
    In proposed Sec.  52.167(b)(3), the Commission included a provision 
which would have required the manufacturing license to specify the 
number of reactors authorized to be manufactured under the 
manufacturing license. Upon further consideration in response to a 
comment on the proposed rule, the Commission has decided that there is 
no valid regulatory basis for including this provision, and it may in 
fact serve as a disincentive for the manufacturer to improve the 
efficiency and productivity of the manufacturing process. Accordingly, 
this provision is not included in the final rule.
    Under Sec.  52.177(c), the holder of a manufacturing license may 
not commence manufacturing of a reactor less than 3 years before the 
expiration date, but may continue the manufacturing of a reactor whose 
manufacture commenced before the 3-year deadline up to license 
expiration. If, however, an application for renewal is timely-filed 
with the NRC, manufacturing of a reactor whose manufacture commenced 
before the 3-year deadline may continue until the time that the NRC 
completes action on the renewal application in accordance with the 
Timely Renewal Doctrine of the Administrative Procedure Act (APA). The 
Commission believes that the timely renewal period should be based upon 
the time reasonably needed by the agency to complete action on a 
renewal application, so that an applicant's reliance upon timely 
renewal is the rare exception rather than the rule. The NRC selected 
the 3-year deadline as a reasonable period for completing the 
manufacture of a nuclear power reactor, based in large part upon public 
statements by various reactor vendors that they have set goals for 
constructing complete nuclear power plants onsite within 3 years. It 
seems reasonable, therefore, that a manufactured reactor, built in a 
controlled environment using industrial manufacturing processes, would 
be able to be manufactured in the same 3-year period as the 
construction of an entire facility onsite. Paragraph (b) is corrected 
in the final rule by removing the phrase, ``that the Commission may 
impose,'' in order to avoid the possible misinterpretation that the 
Commission could choose not to impose new adequate protection 
requirements identified by the Commission. In addition, paragraph 
(b)(2) is corrected by removing the reference to ``site permit'' and 
substituting the term, ``manufacturing license.''
    The final rule does not require that the manufacturing license 
specify an earliest and latest date for completion of manufacture of 
any individual reactor. Section 185 of the AEA directs that ``[t]he 
construction permit shall state the earliest and latest date for 
completion of the construction or modification.'' Inasmuch as a 
manufacturing license is not a construction permit, there does not 
appear to be any legal need for the manufacturing license to specify 
the earliest and latest date of completion of manufacture. The language 
of this section has been corrected in the final rule to make clear that 
the duration of the renewed manufacturing license consists of the 
renewed term plus any period remaining on the superseded license 
(analogous to the determination

[[Page 49394]]

of the duration of a renewed operating license under part 54).
12. Subpart G of Part 52 [Reserved]
13. Subpart H of Part 52--Enforcement
    This subpart contains two provisions, Sec.  52.301 and Sec.  
52.303, which are comparable to former Sec.  52.111 and Sec.  52.113, 
and are analogous to provisions contained in other parts of 10 CFR 
Chapter I imposing requirements on regulated entities. Section 52.301 
reiterates, and provides notice to licensees and applicants under part 
52 of the Commission's authority to obtain injunctions or other court 
orders for the enumerated violations. Section 52.113 provides notice to 
all persons and entities subject to part 52 that they are subject to 
criminal sanctions for willful violations, attempted violations, or 
conspiracy to violate certain regulations under part 52. The 
regulations listed in paragraph (b), for which criminal sanctions do 
not apply, have been updated to reflect the final part 52 rulemaking. 
Section 52.99 was erroneously listed in paragraph (b) in the proposed 
rule. Because that regulation contains substantive requirements which 
are promulgated under Section 161.b., i, and o of the AEA, it has been 
removed from the list of regulations in paragraph (b).
14. Appendices A, B, C, and D to Part 52--Design Certifications for 
ABWR, System 80+, AP600, and AP1000
    The NRC amended paragraphs VI.B.4, 5, and 6 of the design 
certification rules (DCRs) in appendices A, B, and C to part 52 for the 
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 DCR are 
resolved when an exemption has been granted for a plant referencing the 
DCR. Paragraph VI.B.5 describes how issues are resolved when a plant 
referencing the DCR 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 obtain 
NRC approval for such departures. Thus, once a matter (e.g., an 
exemption in the case of paragraph VI.B.4) is addressed for a specific 
plant referencing a DCR, 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.
    Each of the DCRs includes a Section VIII on processes for changes 
and departures. These processes apply to changes and departures 
depending upon the category of certification information affected. For 
plant-specific Tier 2 information, the departure process established in 
the rule mirrors, in large part, that in the former 10 CFR 50.59. The 
final rule amends paragraph VIII.B.5 of the DCRs in appendices A, B, 
and C to conform the terminology in the Sec.  50.59-like process to 
that used in the current Sec.  50.59. This amendment deleted references 
to unreviewed safety questions and safety evaluations, and conformed 
the evaluation criteria concerning when prior NRC approval is needed. 
Also, a definition was added to the DCRs (paragraph II.G) for 
``departure from a method of evaluation'' to support the evaluation 
criterion in paragraph VIII.B.5.b(8) of appendices A, B, and C to part 
52.
    In an earlier rulemaking (see 64 FR 53582; October 4, 1999), the 
NRC revised Sec.  50.59 to incorporate new thresholds for permitting 
departures from a plant design as described in the FSAR without NRC 
approval. For consistency and clarity, similar changes were adopted for 
part 52 applicants or licensees. Because of some differences in how the 
requirements are structured in the DCRs, certain criteria contained in 
Sec.  50.59 are not necessary for or applicable to part 52 and are not 
being included in this rule. One criterion definition that the NRC did 
include was from 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 paragraph VIII.B.5.b of appendices A, B, 
and C to part 52 will be implemented as intended.
    Each of the DCRs includes a special process in Section VIII for 
departures from selected severe accident issues. The Commission 
believes that the resolution of severe accident issues should be 
preserved and maintained in the same fashion as all other safety issues 
that were resolved during the design certification review (refer to SRM 
on SECY-90-377). However, because of the increased uncertainty in 
severe accident issue resolutions, the Commission codified separate 
criteria in paragraph B.5.c of Section VIII for determining if a 
departure from design information that resolves these severe accident 
issues would require a license amendment. The final rule amends 
paragraph B.5.c to clarify that the special process applies to ex-
vessel severe accident design features that are described in the plant-
specific design control document (DCD).
    For purposes of applying the special criteria in paragraph B.5.c of 
Section VIII, severe accident resolutions are limited to those design 
features where the intended function of the design feature is relied 
upon to resolve postulated accidents when the reactor core has melted 
and exited the reactor vessel (ex-vessel severe accidents) and the 
containment is challenged. The location of the ex-vessel severe 
accident design information in the DCD is not important to the 
application of this special departure process in paragraph B.5.c. Some 
design features may have intended functions to meet both ``design 
basis'' requirements and to resolve ex-vessel severe accidents. If 
these design features are reviewed under paragraph VIII.B.5, then the 
appropriate criteria from either paragraph B.5.b or B.5.c are selected 
depending upon which function the departure is being taken from.
    Each of the DCRs in appendices A, B, and C to part 52 includes a 
section on records and reporting. The NRC revised paragraph X.B.3.b in 
appendices A, B, and C to part 52 to change the reporting frequency 
from quarterly to semi-annually, and to extend the period of increased 
reporting frequency, relative to the frequency of 10 CFR 50.59(d) and 
50.71(e)(4), from the date of a license application that references a 
DCR to the date that the Commission makes the finding under 10 CFR 
52.103(g). The requirement to report plant-specific departures from, 
and updates to, the design control document during the interval from 
the application for a combined license until the Commission makes the 
finding under Sec.  52.103(g) is to facilitate NRC's monitoring of 
changes to the nuclear power plant, to achieve a common understanding 
of how the as-built facility conforms to the design information, and to 
adjust the inspection program to reflect the design changes.
    The amendment to paragraph X.B.3.b of appendices A, B, and C to 
part 52 reduced the frequency of reporting during the period of 
construction and increased the frequency of reporting during the 
application review period. The NRC believes that these changes in the 
reporting burden balance each other and provide the information needed 
by the NRC to fulfill its responsibilities in the licensing of future 
nuclear power plants. In order to make the finding

[[Page 49395]]

under Sec.  52.103(g), the NRC must monitor the design changes made 
under Section VIII of the DCRs. Frequent reporting of design changes 
will be particularly important in times when the number of design 
changes could be significant, such as during the procurement of 
components and equipment, the detailed design of the plant before and 
during construction, and during pre-operational testing. After the 
facility begins operation, the frequency of reporting would revert to 
the requirement in paragraph X.B.3.c, which is consistent with 
operating plant requirements.
    Additional editorial changes to the design certification rule 
language in appendices A, B, C, and D to part 52 are discussed in the 
NRC's responses to public comments on Question 11 (see Section IV of 
this document).
15. Appendix N to Part 52--Combined Licenses for Nuclear Power Reactors 
of Identical Design
    Prior to this final rulemaking, appendix N in parts 50 and 52 
contained the NRC's procedures governing the review and issuance of 
licenses for nuclear power plants of ``duplicate design.'' Hearings for 
applications filed under appendix N in both parts 50 and 52 are 
governed by subpart D of part 2. In the March 2006 proposed rule, the 
NRC proposed deleting appendix N in part 52, and retaining these 
provisions only in part 50. Although no comment was received on this 
proposal, the NRC has decided to withdraw its proposal to delete 
appendix N in part 52. Since the preparation of the March 2006 proposed 
rule, several industry groups have announced their intention to seek 
combined licenses utilizing the same design. In view of this industry 
development, the NRC believes that there is potential utility to 
keeping the option of appendix N in part 52 open to potential combined 
license applicants. Accordingly, the NRC is retaining in part 52 the 
procedural alternative provided in appendix N to part 52, and to revise 
its language to make its provisions applicable to combined licenses 
using identical designs. As part of this revision, the NRC set forth 
more explicit direction on the information to be submitted, the NRC 
docketing review, notice, and the content of the EIS under appendix N 
of part 52. However, the NRC decided against a wholesale revision of 
appendix N to part 52, together with conforming changes in part 51, 
inasmuch as these changes were not the subject of public comment, and 
because such a course of action would have delayed the overall part 52 
rulemaking. Inasmuch as the changes to appendix N of part 52 
constitute, in essence, revisions to the NRC's rules of procedure and 
practice (albeit located within part 52), the NRC may adopt them in 
final form without further notice and comment, under the rulemaking 
provisions of the APA, 5 U.S.C. 553(b)(A).
    The overall concept of the revised appendix N to part 52 is that 
each application is to be treated as a separate application, with the 
exception of the common design. Hence, appendix N to part 52 requires 
separate applications, separate determinations of sufficiency for 
docketing, separate notices of docketing, and so forth. Sections 
requiring further explanation are discussed below.
    Paragraph 2 of appendix N to part 52 requires that each application 
state that the applicant wishes to have the application considered 
under appendix N to part 52, and to list all of the applications that 
are to be treated together. This requirement ensures that the NRC is 
clearly informed of the intentions of all applicants, and to ensure 
that any individual reviewing the application can easily determine all 
of the applications using the identical (``common'') design.
    Paragraph 3 of appendix N to part 52 requires that each application 
identify the common design, and that the FSAR either incorporate by 
reference or include the common design. This ensures that there will be 
a single physical FSAR document that may be utilized by the NRC, and 
viewed by members of the public.
    Paragraph 5 of appendix N to part 52 provides that, upon an NRC 
determination that each application is acceptable for docketing under 
10 CFR 2.101, each application will be separately docketed (i.e., each 
application will be given a separate docket number, but that docket 
number may include a special designator signifying that it is part of a 
group of applications filed under appendix N to part 52). Ordinarily, 
the NRC will publish in the Federal Register a separate notice of 
docketing for each application, so that delays in the docketing of one 
application will not delay the docketing and subsequent technical 
review of other applications filed in accordance with appendix N to 
part 52. However, if circumstances allow (e.g., sufficiency review for 
multiple applications are completed simultaneously), the NRC may 
publish a single notice of docketing for multiple applications. The 
notice of docketing must state that the application will be processed 
under the provisions of 10 CFR part 52, appendix N and subpart D of 
part 2. As discussed under subpart D of part 2, the NRC also has 
discretion to either publish a notice of hearing for each application 
(possibly with the period for the filing of petitions to intervene 
running from the notice of hearing for the last application of the 
group), or to publish a joint notice of hearing for multiple 
applications.
    Paragraph 6 of appendix N to part 52 sets forth the procedures by 
which the NRC will fulfill its obligations under NEPA. The NRC staff 
will prepare a separate draft EIS for each application, but the NRC may 
conduct joint scoping on environmental issues related to the common 
design. If the applications reference a standard design certification 
or the use of a manufactured reactor, then the EIS must incorporate by 
reference the EA prepared for either the design certification or the 
manufacturing license, as applicable. The NRC has decided that the EA 
need not be included in the EIS. The Commission has required other 
documents to be incorporated into the FSAR in order to maximize the 
utility and ease of use of the FSAR, which is used repeatedly by the 
NRC staff over the lifetime of the licensed reactor. By contrast, the 
EIS is not typically utilized by the staff in such a manner; hence, the 
NRC deemed it unnecessary to require physical incorporation of the 
referenced design certification or manufacturing license EA into the 
referencing combined license EIS.
    Paragraph 7 of appendix N to part 52 requires the ACRS to report on 
each of the combined license applications, as required by Sec.  52.87. 
Each ACRS report is to be limited to the safety matters which are not 
relevant to the common design. In addition, the ACRS must issue a 
report on the safety of the common design--except for those matters 
relevant to the safety of a referenced design certification or 
manufactured reactor. Issuance of separate reports for each application 
will facilitate NRC staff internal review, consideration, and response 
to the ACRS report. It will also ensure that issues relevant to one 
application (e.g., siting) are not addressed in the proceeding and 
hearing for another application. Issuance of a single report on the 
common design will also facilitate the issuance of the presiding 
officer's partial initial decision on the common design, as required by 
paragraph 8 of appendix N to part 52, and 10 CFR 2.405 of subpart D of 
part 2. The NRC notes that there may be circumstances where the common 
design extends beyond the design matters covered in a referenced design

[[Page 49396]]

certification or manufactured reactor. For example, a common design 
could reference the use of a specific design certification and a common 
ultimate heat sink. In such circumstances, the ACRS would issue a 
common report limited to the safety matters for the ultimate heat 
sink.\6\
---------------------------------------------------------------------------

    \6\ The site-specific environmental impacts of the heat sink 
would ordinarily be addressed in each of the separate EISs prepared 
for each application, inasmuch as the environmental impacts would 
differ depending upon factors and characteristics at each site. 
Section 7 does not govern the scope of EISs prepared for common 
design elements.
---------------------------------------------------------------------------

    Paragraph 8 of appendix N to part 52 provides that the NRC will 
designate a presiding officer to conduct the portion of the hearing on 
matters related to the common design, and that the presiding officer 
must issue a partial initial decision on the common design. As 
discussed previously, hearing procedures for appendix N to part 52 
proceedings are set forth in subpart D to part 2. To avoid duplication 
and possible (future) conflicts with subpart D to part 2, the NRC did 
not include in appendix N to part 52 further provisions addressing the 
conduct of hearings.

D. Changes to 10 CFR Part 50

1. General Provisions, Sec.  50.2, Definitions
    New definitions are added as conforming changes to Sec.  50.2. A 
definition of an applicant is added to clarify that a person or entity 
applying for Commission ``permission or approval'' is an applicant. 
This will ensure that part 50 requirements for applicants apply to a 
person or entity seeking an NRC approval not constituting a license, 
such as a standard design approval under part 52.
    Definitions for license and licensee are added to clarify that 
early site permits and combined licenses under part 52 are licenses, 
and that holders of these types of licenses are licensees for purposes 
of part 50.
    A definition for prototype plant is added to describe the type of 
nuclear reactor that is the subject of Sec.  50.43(e). A prototype 
plant is a licensed nuclear reactor test facility that is similar to 
and representative of the first-of-a-kind nuclear plant 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 nuclear plant design, as well as being 
used as a commercial nuclear power facility.
2. Requirement of License, Exceptions, Sec.  50.10, License Required
    Section 50.10 addresses the circumstances under which a license for 
a production or utilization facility is required, and describes 
activities which do not constitute ``construction'' for purposes of 
obtaining a license for a nuclear power plant. Section 50.10(b) 
formerly prohibited a person from beginning construction of a 
production or utilization facility unless a construction permit has 
been issued. Inasmuch as activities constituting construction (as 
defined in Sec.  50.10(b)) are authorized under a combined license, 
Sec.  50.10(b) is revised to refer to combined licenses.
    Formerly Sec.  52.17(c) authorized an early site permit applicant 
to request authority to perform the activities allowed under Sec.  
50.10(e)(1). The NRC notes that the regulation did not provide for the 
holder of an early site permit to request authority to conduct Sec.  
50.10(e)(1) activities after the early site permit has been issued, and 
the NRC does not plan to change the current restriction. It will 
conserve the NRC's resources to consider the safety and environmental 
issues associated with Sec.  50.10(e)(1) activities during the agency's 
consideration of the early site permit application. Late consideration 
of these requests after completion of the NRC's consideration of the 
application could entail substantial diversion of resources from other 
application reviews. For these reasons, the NRC does not allow an early 
site permit holder to request authority to perform activities allowed 
under Sec.  50.10(e)(1) after issuance of the early site permit (the 
Commission notes that under former part 52, early site permit holders 
may not seek authority to perform activities allowed under Sec.  
50.10(e)(3) after issuance of the early site permit).
3. Classification and Description of Licenses
a. Section 50.23, Construction Permits
    Section 50.23 formerly provided that a construction permit for the 
construction of a production or utilization facility must be issued 
before issuance of a license for the facility, and then only upon ``due 
completion'' of the facility. Section 50.23 is revised to clarify that 
if the NRC issues a combined license for a nuclear power plant under 
part 52, the construction permit and operating license are issued 
simultaneously (i.e., are merged into a ``combined license'' under 
subpart C of part 52). This is consistent with Section 185.b of the 
AEA, which provides the NRC with explicit statutory authority to 
combine a construction permit and an operating license for a nuclear 
power plant into a single combined license. The Commission notes that 
Sec.  50.23 is not limited to nuclear power plants; it also allows the 
NRC to combine, under Section 161.h of the AEA, a construction permit 
and operating license for production facilities or utilization 
facilities other than nuclear power plants.
4. Applications for Licenses, Certifications, and Regulatory Approvals; 
Form; Contents; Ineligibility of Certain Applicants
a. Section 50.30, Filing of Application; Oath or Affirmation
    Section 50.30 establishes the NRC's general procedural requirements 
on filing of applications for licenses (including construction permits) 
for production and utilization facilities. The NRC is making conforming 
changes throughout Sec.  50.30 to include necessary references to part 
52 processes other than design certification (subpart H of part 2 
governs the filing of standard design certification applications), 
viz., early site permits, combined licenses, standard design approvals, 
and manufacturing licenses. In addition, Sec.  50.30(a) is revised to 
ensure that the submission requirements governing applications (and 
amendments to these applications) in Sec.  52.3 apply to part 52 
processes other than design certification.
b. Section 50.33, Contents of Applications; General Information
    Section 50.33 identifies the general information that must be 
included in applications for licenses (including construction permits) 
for production and utilization facilities. Section 50.33(f) requires 
certain applicants for nuclear power plant licenses to submit 
information sufficient to determine whether the applicant has the 
financial qualifications to carry out, in accordance with the NRC's 
regulations, the activities for which a license or permit is sought. 
Section 50.33 is revised to require applicants for combined licenses to 
submit financial qualifications information. Financial qualifications 
information need not be submitted by applicants for early site permits, 
standard design certifications, standard design approvals, and 
manufacturing licenses. An NRC review to determine whether an applicant 
has adequate financial qualifications to conduct the activities 
authorized by an early site permit would contribute little, if 
anything, to providing reasonable assurance of adequate protection with 
respect to early site permit activities. Ordinarily, an early site 
permit authorizes no activities, unless the early site permit 
application requested

[[Page 49397]]

authority to conduct the activities permitted under Sec.  50.10(e)(1). 
The NRC has determined that no safety finding per se is necessary to 
authorize the licensee to conduct these activities. The NRC's review of 
a Sec.  50.10(e)(1) application is focused on siting and environmental 
matters.
    With respect to a standard design approval, the argument applies 
with even more force, inasmuch as a design approval authorizes no 
activities of any kind, and the finality associated with a design 
approval is significantly less than for an early site permit. The NRC 
concludes that no regulatory purpose appears to be served by a 
financial qualifications review for early site permits and standard 
design approvals. The NRC believes that there is little additional 
regulatory value in requiring a financial qualifications review for a 
manufacturing license. While it is true that a lack of sufficient 
financial resources could result in inadequate manufacture of a 
reactor, under the NRC's proposed concept of a manufacturing license 
under subpart F of part 52, each manufactured reactor cannot be 
operated until ITAAC specified in the manufacturing license are 
successfully completed by the licensee authorized to construct the 
nuclear power facility using the manufactured reactor. Successful 
completion of the manufactured reactor's ITAAC should ensure that any 
problems with manufacture attributable to lack of financial resources 
of the manufacturing license holder can be identified before operation. 
Moreover, the licensee authorized to construct the facility (either 
under a construction permit or a combined license) using a manufactured 
reactor would have been subject to a financial qualifications review. 
This review should be sufficient to determine if the applicant has 
sufficient financial resources to carry out facility construction and 
the completion of the manufactured reactor's inspections, tests, and 
acceptance criteria. Finally, the NRC notes that it does not require 
the fabricators of safety-related and important to safety structures, 
systems, and components (SSCs) to be licensed and subject to a 
financial qualifications review. The NRC believes that a holder of a 
manufacturing license conducts activities which appear to be, in large 
part, analogous to these current non-licensed fabricators. Accordingly, 
the NRC concludes that a financial qualifications review of the 
applicant for a manufacturing license will not add significant 
regulatory value to justify the cost of such a review.
    Section 50.33(g) addresses radiological emergency response plans 
for State and local government entities that must be submitted in 
applications for operating licenses. The final rule makes a conforming 
change to ensure that applicants for combined licenses must also submit 
this information, as well as applicants for early site permits who 
decide under Sec.  52.17(b)(2)(ii) to seek NRC review and approval of 
complete emergency plans. In addition, Sec.  50.33(g) provides 
requirements for the plume exposure pathway emergency planning zone 
(EPZ) and the ingestion pathway EPZ. The NRC has made a conforming 
change to Sec.  50.33(g) in the final rule to address early site permit 
applications that propose major features of emergency plans describing 
the EPZs under 10 CFR 52.17(b)(2)(i). Such provisions were 
inadvertently left out of the proposed rule. For an application for an 
early site permit that proposes major features of the emergency plans 
describing the EPZs, the change requires the descriptions of the EPZs, 
to meet the requirements of Sec.  50.33(g). This is necessary for the 
NRC to be able to find that major features describing the EPZs are 
acceptable under Sec.  52.18.
    Section 50.33(h) formerly required applicants that propose to 
construct or alter a production or utilization facility to state in 
their application the earliest and latest dates for completion of the 
construction or alteration. This section is being revised in the final 
rule, based on public comments, to exclude combined license applicants. 
The NRC believes that combined license applications need not specify 
the earliest and latest date for completion of construction, in light 
of the amendment to Section 185 of the AEA that was made by the Energy 
Policy Act of 1992. By adding a new Section 185.b. of the AEA, the 
Commission believes that Congress intended that Section 185.b supersede 
Section 185.a of the AEA, so that the Section 185.a requirements for 
``stand-alone'' construction permits, such as the need to specify the 
earliest and latest date for completion of construction, do not apply 
to the construction permit portion of a combined license under Section 
185.b of the AEA. Accordingly, the final rule removes the requirements 
from Sec. Sec.  50.33(h), 52.77, and 52.79(a)(39) that the combined 
license application specify the earliest and latest date for completion 
of construction.
    Section 50.33(k) currently requires applicants for operating 
licenses to provide a report, as described in Sec.  50.75, indicating 
how reasonable assurance that funds will be available for the 
decommissioning process is provided. The final rule makes a conforming 
change to add a reference to combined licenses. The content of this 
report, reflecting the unique considerations of a combined license, is 
addressed separately in the revision to Sec.  50.75.
c. Section 50.34, Contents of Construction Permit and Operating License 
Applications; Technical Information
    The NRC is changing the heading of Sec.  50.34 from Contents of 
applications; technical information to read, Contents of construction 
permit and operating license applications; technical information. 
Section 50.34(a) currently provides the requirements for the technical 
contents of an application for a stationary power reactor construction 
permit, design certification or combined license, and Sec.  50.34(b) 
provides the requirements for the technical contents of an application 
for a stationary power reactor operating license application. However, 
the former version of 10 CFR part 52 provides requirements for design 
certification and combined license applications that are not consistent 
with the current version of Sec.  50.34. For example, former Sec.  
52.47 stated that an application for design certification must contain 
the technical information which is required of applicants for 
construction permits and operating licenses by part 50 which is 
technically relevant to the design and not site-specific. This would 
encompass requirements in both Sec. Sec.  50.34(a) and (b). Also, 
former Sec.  52.79 stated that applications for combined licenses must 
contain the technically relevant information required of applicants for 
an operating license by 10 CFR 50.34, which are found in Sec.  
50.34(b). In addition to the requirements for technical information in 
Sec. Sec.  50.34(a) and (b), Sec. Sec.  50.34(c) through (h) provide 
requirements for the contents of licensing applications related to 
security plans, compliance with Three Mile Island (TMI) related 
requirements, combustible gas control, and conformance with the 
standard review plan. Finally, the NRC notes that the subject of 
contents of an application is an administrative matter, rather than a 
strictly technical matter. Therefore, these administrative requirements 
for part 52 processes are more properly located in part 52, rather than 
in Sec.  50.34. To provide maximum clarity in the requirements for the 
content of each of the different types of licensing applications, the 
NRC is revising Sec.  50.34 to make it applicable to construction 
permit and operating license applications only and to provide separate 
sections for the technical

[[Page 49398]]

contents of applications for the other types of licenses or regulatory 
approvals in 10 CFR part 52 (early site permits in Sec.  52.17, design 
certifications in Sec.  52.47, combined licenses in Sec.  52.79, design 
approvals in Sec.  52.137, and manufacturing licenses in Sec.  52.157). 
In its revisions to 10 CFR part 52, the NRC has brought forward the 
requirements from Sec.  50.34 that are applicable to each of the 
licensing and approval processes in 10 CFR part 52. One exception to 
this structure is the provisions in Sec.  50.34(f) related to 
compliance with TMI related requirements. Due to the length and 
complexity of the requirements in this paragraph, Sec.  50.34(f) is 
being amended to indicate that each applicant for a design 
certification, design approval, combined license, or manufacturing 
license under part 52 of this chapter must demonstrate compliance with 
any technically relevant portions of the requirements in Sec.  
50.34(f)(1) through (3), except for paragraphs (f)(1)(xii), (f)(2)(ix), 
and (f)(3)(v). The NRC chose this approach rather than repeat the 
requirements in each of the relevant sections in part 52. The NRC is 
adding the phrase ``except for paragraphs (f)(1)(xii), (f)(2)(ix), and 
(f)(3)(v)'' in the last sentence of Sec.  50.34(f) based on public 
comments. The commenters pointed out that proposed Sec.  50.34(f) was 
inconsistent with proposed Sec. Sec.  52.47(a)(17), 52.79(a)(17), 
52.137(a)(17), and 52.157(e)(12), which included the exceptions that 
are being added to Sec.  50.34(f) in the final rule.
d. Section 50.34a, Design Objectives for Equipment To Control Releases 
of Radioactive Material in Effluents--Nuclear Power Reactors; and Sec.  
50.36a, Technical Specifications on Effluents From Nuclear Power 
Reactors
    Section 50.34a requires that construction permit and operating 
license applications include a description of the equipment and 
procedures for the control of gaseous and liquid effluents and for the 
maintenance and use of equipment installed in radioactive waste 
systems. Section 50.34a also requires these applications to include an 
estimate of (1) the quantity of each of the principal radionuclides 
expected to be released annually to unrestricted areas in liquid 
effluents produced during normal reactor operations; and (2) the 
quantity of each of the principal radionuclides of the gases, halides, 
and particulates expected to be released annually to unrestricted areas 
in gaseous effluents produced during normal reactor operations. In 
addition, Sec.  50.34a requires a general description of the provisions 
for packaging, storage, and shipment offsite of solid waste containing 
radioactive materials resulting from treatment of gaseous and liquid 
effluents and from other sources. Section 50.34a is revised to clarify 
its applicability to the 10 CFR part 52 licensing and approval 
processes. Section 50.34a applies to combined licenses by virtue of the 
provision in former Sec.  52.83, Applicability of Part 50 Provisions, 
which states that all provisions of 10 CFR part 50 and its appendices 
applicable to holders of construction permits and operating licenses 
also apply to holders of combined licenses. Applicants for design 
certification are also required to include the information required by 
Sec.  50.34a in their applications by virtue of the provision in former 
Sec.  52.47(a)(1)(i), which states that an application for design 
certification must contain the technical information which is required 
of applicants for construction permits and operating licenses by 10 CFR 
part 50 which is technically relevant to the design and not site-
specific. Former appendix O to 10 CFR part 52, Section O.3, explicitly 
required applicants for design approvals to include the applicable 
technical information required by Sec.  50.34a. Finally, former 
appendix M to 10 CFR part 52, Section M.1, states that the provisions 
in part 50 applicable to construction permits apply in context, with 
respect to matters of radiological health and safety, environmental 
protection, and the common defense and security, to manufacturing 
licenses. Therefore, new provisions in Sec.  50.34a(d) are adopted to 
address the applicable requirements for combined license applications 
that parallel the requirements for an operating license application. 
New provisions in Sec.  50.34a(e) are adopted to address the applicable 
requirements for applications for design approvals, design 
certifications, and manufacturing licenses to include: (1) A 
description of the equipment for the control of gaseous and liquid 
effluents and for the maintenance and use of equipment installed in 
radioactive waste systems; and (2) an estimate of the quantity of each 
of the principal radionuclides expected to be released annually to 
unrestricted areas in liquid effluents produced during normal reactor 
operations, and the quantity of each of the principal radionuclides of 
the gases, halides, and particulates expected to be released annually 
to unrestricted areas in gaseous effluents produced during normal 
reactor operations.
e. Section 50.36, Technical Specifications
    Section 50.36(a) currently requires that each applicant for a 
license authorizing operation of a production or utilization facility 
include in its application proposed technical specifications in 
accordance with the requirements of Sec.  50.36. The existing language 
in Sec.  50.36(a) encompasses combined license applicants. However, 
applicants for design certification are also required to include 
proposed technical specifications in their applications by virtue of 
the provision in former Sec.  52.47(a)(1)(i) stating that an 
application for design certification must contain the technical 
information required of applicants for construction permits and 
operating licenses by 10 CFR part 50 that is technically relevant to 
the design and not site-specific. Similarly, applicants for design 
approvals are also required to include proposed technical 
specifications in their applications by virtue of the provision in 
former appendix O to part 52, Section O.3, which states that the 
submittal for review of a standard design shall include the applicable 
technical information under Sec.  50.34 (a) and (b), as appropriate.
    Section 50.36 is revised to clarify that design certification and 
manufacturing license applications must also include proposed technical 
specifications. The new provisions in Sec.  50.36(c) require each 
applicant for a design certification or a manufacturing license to 
include proposed generic technical specifications in its application 
for the portion of the plant that is within the scope of the design 
certification or manufacturing license application.
f. Section 50.36a, Technical Specifications on Effluents From Nuclear 
Power Reactors
    Section 50.36a(a) requires each licensee of a nuclear power reactor 
to include technical specifications to keep releases of radioactive 
materials to unrestricted areas during normal conditions, including 
expected occurrences, as low as is reasonably achievable. The former 
language in Sec.  50.36a(a) encompassed combined license holders. 
However, applicants for design certification are also required to 
include proposed technical specifications on effluents in their 
applications by virtue of the provision in current Sec.  52.47(a)(1)(i) 
which states that an application for design certification must contain 
the technical information which is required of applicants for 
construction permits and operating licenses by 10 CFR part 50

[[Page 49399]]

which is technically relevant to the design and not site-specific. In 
addition, former appendix M to 10 CFR part 50, Section M.1, states that 
the provisions in part 50 applicable to construction permits apply in 
context, with respect to matters of radiological health and safety to 
manufacturing licenses. Therefore, Section 50.36a(a) is revised to 
state that each licensee of a nuclear power reactor and each applicant 
for a design certification or a manufacturing license will include 
technical specifications to keep releases of radioactive materials to 
unrestricted areas during normal conditions, including expected 
occurrences, as low as is reasonably achievable. The proposed rule did 
not include the provisions for manufacturing licenses. However, 
proposed Sec.  52.157(e)(18) did require manufacturing license 
applicants to include proposed technical specifications in accordance 
with Sec.  50.36a. Therefore, it was clearly the NRC's intent that the 
provisions of Sec.  50.36a be applicable to manufacturing license 
applications and the NRC has corrected this omission in the final rule.
    Some commenters on the 2006 proposed rule identified an additional 
conforming change needed in Sec.  50.36a that the NRC did not make in 
the proposed rule. Section 50.36(a)(2) currently requires that each 
licensee submit a report to the Commission annually that specifies the 
quantity of each of the principal radionuclides released to 
unrestricted areas in liquid and in gaseous effluents during the 
previous 12 months, including any other information as may be required 
by the Commission to estimate maximum potential annual radiation doses 
to the public resulting from effluent releases. The NRC has modified 
this provision to state that each holder of a combined license is only 
required to begin submitting reports after the Commission has made the 
finding under Sec.  52.103(g) that allows fuel load and operation. This 
would apply the requirements in Sec.  50.36a consistently for part 50 
and part 52 licensees, because for a part 50 licensee, the annual 
reporting requirement is effective only after an operating license is 
issued.
    The NRC is also making conforming changes to appendix I to 10 CFR 
part 50. These changes parallel the changes to Sec. Sec.  50.34a and 
50.36a.
g. Section 50.36b, Environmental Conditions
    Section 50.36b authorizes the Commission to include conditions to 
protect the environment in each license authorizing operation of a 
production or utilization facility and each license for a nuclear power 
reactor facility for which the certification of permanent cessation of 
operations required under Sec.  50.82(a)(1) has been submitted. These 
conditions are to be derived from information contained in the 
environmental report and the supplement to the environmental report as 
analyzed and evaluated in the NRC record of decision. The conditions 
must identify the obligations of the licensee in the environmental 
area, including, as appropriate, requirements for reporting and keeping 
records of environmental data, and any conditions and monitoring 
requirement for the protection of the nonaquatic environment.
    The NRC has made conforming changes to Sec.  50.36b in the final 
rule to address all applicable part 52 licenses. The changes were made 
in response to public comments that highlighted the need for 
clarification in Sec.  50.36b. The NRC provided proposed requirements 
for identifying environmental conditions on early site permits and 
combined licenses in the proposed rule in Sec. Sec.  51.50(b) and (c). 
Requirements for identifying environmental conditions for construction 
permits were contained in former Sec.  51.50 and proposed Sec.  
51.50(a). The proposed rule stated that, in an application for a 
construction permit, an early site permit, or a combined license, the 
applicant shall identify ``any conditions and monitoring requirements 
for protecting the non-aquatic environment, proposed for possible 
inclusion in the license as environmental conditions in accordance with 
Sec.  50.36b of this chapter.'' However, the NRC neglected to make the 
additional conforming changes to Sec.  50.36b in the proposed rule. To 
correct this oversight, the NRC has modified Sec.  50.36b in the final 
rule to make the requirements in this section consistent with the 
requirements in Sec.  51.50. In doing so, the NRC has provided separate 
paragraphs for imposing conditions during construction and for imposing 
conditions during operation and decommissioning. Paragraph 50.36b(a) 
addresses requirements for imposing conditions on construction permits, 
early site permits, and combined licenses to protect the environment 
during construction. Paragraph 50.36b(b) addresses requirements for 
imposing conditions on licenses authorizing operation and licenses for 
a facility in decommissioning to protect the environment during 
operation and decommissioning. These changes provide consistency in 
requirements for environmental conditions across parts 50 and 51.
h. Section 50.37, Agreement Limiting Access to Classified Information
    Section 50.37 requires that a license or construction permit 
applicant agree in writing that it will not permit any individual to 
have access to or any facility to possess Restricted Data or classified 
National Security Information until the individual and/or facility has 
been approved for access under the provisions of 10 CFR parts 25 and/or 
95. Section 50.37 also requires that this agreement be part of the 
application for a license or construction permit and that the agreement 
of the applicant shall be deemed part of the license or construction 
permit, whether stated or not. The former language of Sec.  50.37 
encompassed early site permit, combined license, and manufacturing 
license applicants under 10 CFR part 52 because these products are all 
licenses. However, the NRC is revising Sec.  50.37 to encompass 
applicants for design certification and for standard design approvals 
under 10 CFR part 52 for consistency with the changes to 10 CFR part 
25. Part 25 sets forth the NRC's requirements governing the granting of 
access authorization to classified information to certain individuals, 
and the Commission is making modifications to part 25 to reflect the 
licensing and regulatory approval processes in part 52. Accordingly, 
the Commission is revising Sec.  50.37. Section 50.37 is revised to 
require that an applicant for a license, construction permit, design 
certification, or design approval under part 52 agree in writing that 
it will not permit any individual to have access to or any facility to 
possess Restricted Data or classified National Security Information 
until the individual and/or facility has been approved for access under 
the provisions of 10 CFR parts 25 and/or 95. Section 50.37 also 
requires that this agreement be part of the application and be deemed 
part of the license, or construction permit, or NRC standard design 
approval whether stated or not. Section 52.54 is revised to include a 
new provision which requires that every standard design certification 
rule issued contain a provision that states that, after the Commission 
has adopted the final standard design certification rule, the applicant 
will not permit any individual to have access to or any facility to 
possess Restricted Data or classified National Security Information 
until the individual and/or facility has been approved for access under 
the provisions of 10 CFR parts 25 and/or 95. The NRC believes that 
these revisions, along with the complementary changes to parts 25 and 
95, are necessary to

[[Page 49400]]

ensure that access to classified information is adequately controlled 
by all entities applying for NRC licenses, design certifications, or 
design approvals.
5. Standards for Licenses, Certifications, and Approvals
a. Section 50.40, Common Standards
    This section sets forth standards for issuance of a license. 
Sections 50.40(a), (b), and (c) are revised to add conforming 
references to the additional licensing processes issued under 10 CFR 
part 52 that are applicable to these standards.
b. Section 50.43, Additional Standards and Provisions Affecting Class 
103 Licenses and Certifications for Commercial Power
    The text and heading of this section are revised to clarify that 
certain additional standards and provisions for class 103 licenses 
apply to applications for combined licenses, design certifications, and 
manufacturing licenses issued under part 52, in addition to 
applications for construction permits and operating licenses issued 
under part 50. Section 50.43(e) is added to clarify that the 
requirements to demonstrate new safety features by testing, which were 
previously set forth in part 52, apply to applicants for operating 
licenses issued under part 50 and applicants for combined licenses, 
design certifications, and manufacturing licenses issued under part 52. 
This amendment conforms to the goal of having reactor safety 
requirements in part 50 and procedural requirements in part 52. Only 
the requirements in Sec.  50.43(e) apply to applications for design 
certification. Refer to the generic discussion on testing requirements 
for advanced reactors in Section V.B of this document.
c. Section 50.45, Standards for Construction Permits, Operating 
Licenses, and Combined Licenses
    This section is revised to include the standards for review of an 
application to alter a facility that was constructed under a combined 
license, after the findings under Sec.  52.103(g) of this chapter are 
made by the Commission. Some commenters recommended that the proposed 
rule be revised to reference the applicable requirements in part 52 
rather than the requirements in 10 CFR 50.31 through 50.43 and claimed 
that most of those requirements were moved to part 52 in the proposed 
rule. The Commission does not agree with that claim but does 
acknowledge that most of Sec.  50.34 was moved to the contents of 
application section for each of the licensing processes in part 52. 
Therefore, Sec.  50.45 was revised to set forth the standards for 
review of an application to alter a facility after the Commission makes 
the finding under Sec.  52.103(g) of this chapter. The standards for 
issuance of a combined license are set forth in Sec.  52.97.
d. Section 50.46, Acceptance Criteria for Emergency Core Cooling 
Systems for Light-Water Nuclear Power Reactors
    Section 50.46(a)(3) contains reporting requirements for changes to 
or errors in emergency core cooling system (ECCS) evaluation models. 
Conforming references to design approvals, design certifications, and 
licenses issued under part 52 were made to Sec.  50.46, so that the NRC 
will be notified of changes to or errors in acceptable evaluation 
models, or the application of such models, that were used in licenses, 
certifications, and approvals issued under part 52.
e. Section 50.47, Emergency Plans, Sec.  50.54(gg), and Appendix E to 
Part 50, Emergency Planning and Preparedness for Production and 
Utilization Facilities
    Section 50.47 and appendix E to 10 CFR part 50 contain emergency 
planning requirements for nuclear power plants. Prior to this 
rulemaking, these regulations did not clearly address early site permit 
or combined license applicants or holders. Accordingly, the NRC is 
making a number of changes in these regulations. Section 50.47(a)(1) 
states that no initial operating license for a nuclear power reactor 
will be issued unless a finding is made by the NRC that there is 
reasonable assurance that adequate protective measures can and will be 
taken in the event of a radiological emergency, and that no finding 
under Sec.  50.47 is necessary for issuance of a renewed nuclear power 
reactor operating license. The NRC is revising Sec.  50.47(a)(1) to 
include provisions to address combined licenses and early site permits 
which include either complete and integrated plans or major features of 
the emergency plans. The NRC inadvertently left out provisions to 
address early site permits that include major features of the emergency 
plans in the proposed rule and a new provision has been added to 
address applicants in the final rule.
    The NRC is making some additional changes to Sec.  50.47(a)(1) in 
the final rule. Proposed Sec.  50.47(a)(1)(ii) stated that ``Except as 
provided in paragraph (e) of this section, no initial combined license 
under part 52 of this chapter will be issued unless a finding is made 
by the NRC that there is reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological 
emergency.'' In the final rule, the NRC is removing the phrase ``except 
as provided in paragraph (e)'' because paragraph (e) does not address 
issuance of the combined license, but, rather, addresses the Commission 
finding under Sec.  52.103(g). Likewise, the NRC is making a change to 
paragraph (e) of this section in the final rule to remove the reference 
to paragraph (a) of this section.
    Finally, the NRC is removing the statement in proposed Sec.  
50.47(a)(1)(iii) that ``No finding under this section is necessary for 
issuance of a renewed early site permit.'' The NRC included this 
provision in the proposed rule to be consistent with the existing 
requirement for operating licenses. However, upon further 
consideration, the NRC concludes that the basis for this exclusion for 
an operating license and for a combined license does not apply to an 
early site permit. The original license renewal rule, which limited the 
scope of matters to be addressed in the renewal proceeding, was based 
upon a determination that the regulatory process maintains and updates 
the licensing basis for operating licenses, that matters like the state 
of the emergency preparedness plans need not be addressed in license 
renewal. The bases for the license renewal rule described the process, 
in each substantive regulatory area, for maintaining and updating the 
current licensing basis. This logic does not directly apply to 
emergency preparedness information submitted in an early site permit 
application, because there is no maintenance or update requirement for 
the early site permit. Therefore, the NRC cannot exclude the need to 
address emergency preparedness in an early site permit renewal 
proceeding.
    Section 50.47(c)(1) provides a process for operating license 
applicants that fail to meet the applicable standards of Sec.  
50.47(b). The NRC is revising Sec.  50.47(c)(1) to clarify that this 
process is applicable to combined license applicants as well.
    Section 50.47(d) formerly provided that no NRC or Department of 
Homeland Security (DHS) review, findings, or determinations concerning 
the state of offsite emergency preparedness or the adequacy of and 
capability to implement State and local or utility offsite emergency 
plans are required before issuance of an operating license authorizing 
only fuel loading or low-power testing and training (up to 5 percent of 
the rated power). Section 50.47(d) further stated that a license 
authorizing fuel loading and/or low-power testing and training may be

[[Page 49401]]

issued after a finding is made by the NRC that the state of onsite 
emergency preparedness provides reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency and provides the standards by which the NRC will 
base such a finding. The NRC is adding a new Sec.  50.47(e) to provide 
essentially parallel provisions for a combined license holder by 
stating that a combined license holder may not load fuel or operate 
except as provided in accordance with appendix E to part 50 and, 
because of the nature of the combined license process, the NRC is 
adding new Sec.  50.54(gg) that would add a condition to all combined 
licenses. This is necessary to account for the fact that the combined 
license will already be issued at the time of the first full or partial 
participation exercise.
    The NRC's findings regarding the state of emergency preparedness 
for a combined license holder will be taken into account in the NRC's 
review under Sec.  52.103(g). The NRC will make its determination by 
judging whether the licensee has met the acceptance criteria in the 
combined license for the inspections, tests, and analyses related to 
the conduct of the first full or partial participation exercise under 
paragraph IV.F.2.a of appendix E to part 50. Paragraph 50.54(gg) states 
that if, following the conduct of the exercise required by paragraph 
IV.F.2.a of appendix E to part 50, DHS identifies one or more 
deficiencies in the state of offsite emergency preparedness, the holder 
of a combined license may operate at up to 5 percent of rated thermal 
power only if the Commission finds that the state of onsite emergency 
preparedness provides reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological 
emergency. Paragraph 50.54(gg) also provides the standards by which the 
NRC will base such a finding.
    The NRC is revising appendix E to part 50 to conform to the changes 
proposed for Sec. Sec.  50.47 and 50.54. The introduction to appendix E 
to part 50 states that each applicant for an operating license is 
required by Sec.  50.34(b) to include in the final safety analysis 
report plans for coping with emergencies. The NRC is adding a parallel 
statement for combined license applicants, and a statement that an 
early site permit applicant may submit emergency plans. The final rule 
also makes additional conforming changes to the second paragraph of the 
introduction that were inadvertently overlooked in the proposed rule. 
Similar modifications are proposed in Section III of appendix E to part 
50 regarding the content of final safety analysis reports and site 
safety analysis reports for an early site permit. The NRC is making a 
correction to Section III in the final rule to replace references to 
the early site permit application with references to the site safety 
analysis report. The NRC is also adding a statement that the site 
safety analysis report for an early site permit which proposes major 
features must address the relevant provisions of 10 CFR 50.47 and 10 
CFR part 50, appendix E, within the scope of emergency preparedness 
matters addressed in the major features. This is consistent with the 
requirements in Sec.  52.17(b).
    In Section IV of appendix E to part 50, the NRC is modifying 
paragraph F.2.a, to address combined licenses in addition to operating 
licenses. Paragraph F.2.a currently provides requirements regarding the 
conduct of full participation exercises and states that a full 
participation exercise shall be conducted within 2 years before the 
issuance of the first operating license for full power of the first 
reactor. Paragraph F.2.a also requires that, if the full participation 
exercise is conducted more than 1 year before issuance of an operating 
licensee for full power, an exercise which tests the licensee's onsite 
emergency plans shall be conducted within 1 year before issuance of an 
operating license for full power. The NRC is designating the 
requirements for operating licenses as paragraph F.2.a.i, and adding a 
new paragraph F.2.a.ii that contains the requirements for combined 
licenses. Paragraph F.2.a.ii states that, for a combined license, the 
first full participation exercise must be conducted within 2 years of 
the scheduled date for initial loading of fuel and operation under 
Sec.  52.103. Paragraph F.2.a.ii also requires that, if the first full 
participation exercise is conducted more than 1 year before the 
scheduled date for initial loading of fuel and operation under Sec.  
52.103, an exercise which tests the licensee's onsite emergency plans 
must be conducted within 1 year before the scheduled date for initial 
loading of fuel and operation under Sec.  52.103. The modifications 
further state that, if DHS identifies one or more deficiencies in the 
state of offsite emergency preparedness as the result of the first full 
participation exercise, or if the NRC finds that the state of emergency 
preparedness does not provide reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency, the provisions of Sec.  50.54(gg) will apply, 
as previously discussed.
    The NRC is adding a new paragraph IV.F.2.a.iii to appendix E to 
part 50 to require that, if the applicant has an operating reactor at 
the site, an exercise, either full or partial participation, be 
conducted for each subsequent reactor constructed on the site. This 
exercise may be incorporated in the exercise requirements of paragraphs 
(2)(b) and (2)(c) of Section IV.F. If DHS identifies one or more 
deficiencies in the state of offsite emergency preparedness as the 
result of this exercise for the new reactor, or if the NRC finds that 
the state of emergency preparedness does not provide reasonable 
assurance that adequate protective measures can and will be taken in 
the event of a radiological emergency, the provisions of Sec.  
50.54(gg) apply just as they do for the first reactor at a site. This 
new provision is desirable because of the nature of ITAAC for emergency 
preparedness requirements. The emergency preparedness ITAAC, 
specifically ITAAC that will be demonstrated through an exercise, 
provide the necessary reasonable assurance for programs and facilities 
associated with the yet-unbuilt reactor. Recent agreements between the 
NRC and external stakeholders on emergency preparedness ITAAC are based 
on the understanding that ITAAC on the emergency preparedness exercise 
would serve to demonstrate various aspects of emergency preparedness 
(e.g., programs and facilities) that did not warrant their own 
specific/detailed ITAAC. For example, there is no ITAAC for determining 
whether an adequate staffing roster exists for the technical support 
center or emergency offsite facility, but its existence and adequacy 
could be demonstrated during an exercise. Therefore, appendix E to part 
50 requirements for emergency preparedness exercises must be included 
for the current concepts regarding emergency preparedness ITAAC to be 
viable. With regard to subsequent reactors, those aspects of an 
exercise which address currently untested (i.e., unexercised) aspects 
of emergency preparedness for the proposed new reactor must be 
addressed in new emergency preparedness ITAAC for the subsequent 
reactor. If various generic exercise-related aspects of emergency 
preparedness for the site have been previously addressed and satisfied, 
then there would be no ITAAC for those emergency preparedness aspects 
for subsequent reactors.
    The NRC is also modifying Section V of appendix E to part 50, which 
states

[[Page 49402]]

that no less than 180 days before the scheduled issuance of an 
operating license for a nuclear power reactor or a license to possess 
nuclear material, the applicant's detailed implementing procedures for 
its emergency plan shall be submitted to the Commission. Paragraph V 
also requires that licensees submit any changes to the emergency plan 
or procedures to the NRC within 30 days of these changes. The NRC is 
clarifying that paragraph V is also applicable to COL holders by 
stating that they must submit their detailed implementing procedures 
for their emergency plans to the NRC no less than 180 days before the 
scheduled date for initial loading of fuel. The wording of this 
requirement has been changed slightly in the final rule. In the 
proposed rule, this provision required that COL holders submit their 
detailed implementing procedures for their emergency plans to the NRC 
no less than 180 days before the date that the Commission authorizes 
fuel load and operation under Sec.  52.103. The NRC has modified the 
provision to make the target date 180 days before scheduled date for 
initial loading of fuel because this will be a known date whereas the 
licensee would not know the date that the Commission will make the 
Sec.  52.103(g) finding. This change is also consistent with other 
requirements in appendix E that are tied to the scheduled date for 
initial fuel load.
f. Section 50.48, Fire Protection
    Section 50.48(a)(1) is revised to clarify that holders of an 
operating license issued under part 50 and a combined license issued 
under part 52 must have a fire protection plan. Section 50.48(a)(4) is 
added to clarify that applications for design approvals, design 
certifications, and manufacturing licenses issued under part 52 must 
meet the fire protection design requirements set forth in general 
design criterion 3 of appendix A to part 50.
g. Section 50.49, Environmental Qualification of Electric Equipment 
Important to Safety for Nuclear Power Plants
    Section 50.49(a) is revised to clarify that these programmatic 
requirements apply to applicants for and holders of operating licenses 
issued under part 50 and combined licenses and manufacturing licenses 
under part 52.
h. Section 50.54, Conditions of Licenses; and Sec.  50.55, Conditions 
of Construction Permits, Early Site Permits, Combined Licenses, and 
Manufacturing Licenses
    Section 50.54 sets forth various provisions that are deemed to be 
conditions ``in every license issued,'' while Sec.  50.55 sets forth 
the provisions deemed to be conditions of every construction permit. In 
making the conforming changes to these regulations to reflect part 52, 
the NRC has decided to maintain this dichotomy. Conditions applicable 
to part 52 processes which are either licenses or prerequisites to 
licenses, and do not address activities analogous to construction for 
which a construction permit license is required under the AEA, are 
addressed in Sec.  50.54. By contrast, conditions applicable to part 52 
processes which address construction activities, or activities 
analogous to construction for which a construction permit license is 
required under the AEA, are covered in Sec.  50.55. Combined licenses 
represent a special case, inasmuch as they address both construction 
and operation. The NRC addresses combined licenses by placing the 
conditions applicable only to construction in Sec.  50.55, which 
indicates that these conditions are applicable until the date that the 
Commission makes the finding under Sec.  52.103(g). Conditions which 
are applicable during construction and operation or only during 
operation are set forth in Sec.  50.54. The NRC is revising the 
introductory paragraph of Sec.  50.54 to refer to combined licenses, 
and to exclude manufacturing licenses from its provisions. The NRC is 
making revisions to Sec.  50.54 in the final rule based on public 
comments. In the proposed rule, the NRC did not distinguish which 
provisions in Sec.  50.54 are applicable only during operation from 
those that are applicable during both construction and operation. In 
the final rule, the NRC has revised the introductory paragraph to 
indicate which provisions are applicable only after the Commission 
makes the finding under Sec.  52.103(g). In making these revisions, the 
NRC determined that the provisions that need to be applied during both 
construction and operation are paragraphs (a) through (h), (o), (p), 
(q), (t), (v), and (aa) through (ee). All of these provisions have some 
requirements that will be implemented prior to the Commission finding 
under Sec.  52.103(g).
    In addition, the NRC is adding paragraphs (r) and (u) to the list 
of provisions in the introduction that are not applicable to combined 
licenses. This is because paragraph (r) only applies to research and 
test reactor facilities and paragraph (u) was only applicable for 60 
days after the amendment to Sec.  50.54 that added paragraph (u). 
Finally, the NRC is also revising the first sentence of the 
introduction to indicate that paragraphs (r) and (gg) do not apply to 
nuclear power reactor operating licenses. In the proposed rule, the 
introduction stated that they did not apply to operating licenses, 
which would have included research and test reactor operating licenses.
    The NRC is revising Sec.  50.54(a)(1) to indicate that the quality 
assurance (QA) requirements applicable to operation, as described in a 
combined license holder's SAR, become effective 30 days before the 
scheduled date for the initial loading of fuel.
    The NRC is revising Sec.  50.54(i-1) to indicate its applicability 
to combined licenses. Specifically, Sec.  50.54(i-1) requires that 
within 3 months after the date that the Commission makes the finding 
under Sec.  52.103(g) for a combined license, the licensee shall have 
in effect an operator requalification program that must, as a minimum, 
meet the requirements of Sec.  55.59(c) of this chapter.
    The NRC has added changes to Sec.  50.54(p) and (q) in the final 
rule. The changes to paragraph (p) are being made to include references 
to appropriate part 52 sections in addition to the existing references 
to part 50 sections. The change to paragraph (q) is being added to 
include a statement that, for combined licenses, the requirement to 
follow and maintain in effect emergency plans which meet the standards 
in Sec.  50.47(b) and the requirements in appendix E of part 50 is only 
applicable after the Commission makes the finding under Sec.  
52.103(g). However, the remainder of the requirements in paragraph (p) 
apply from the time the combined license is issued (e.g., requirements 
to retain records of emergency plan changes). This is consistent with 
the change made to the introductory paragraph of Sec.  50.54 discussed 
earlier in this section.
    The NRC is adding a new Sec.  50.54(gg). These revisions are 
discussed with related requirements in Section IV.D.4.f of this 
document, ``Section 50.47, Emergency plans, Sec.  50.54(gg), and 
appendix E to part 50.''
    Although the NRC generally views Sec.  50.55 as the appropriate 
section in part 50 for specifying the conditions applicable to 
construction permits and part 52 processes analogous to construction 
permits, the NRC does not believe that all of the conditions in Sec.  
50.55 should apply equally to all of the part 52 processes. 
Accordingly, the introductory text to Sec.  50.55 is revised to specify 
which paragraphs apply to a construction permit, early site permit, 
combined license, and manufacturing license.

[[Page 49403]]

    Sections 50.55(a) and (b) of the March 2006 proposed rule would 
have required a combined license to state the earliest and latest dates 
for completion of construction or modification, and to provide for 
forfeiture of the combined license if the construction or modification 
is not completed by the stated date. The Commission has reconsidered 
this position and has decided to remove this requirement from the final 
rule. The statutory requirement for a construction permit to state the 
earliest and latest date for completion of construction is now 
contained in Section 185.a of the AEA. The combined license, by 
contrast, is address in Section 185.b. The Commission believes that in 
the absence of specific language regarding the restriction in paragraph 
a. applicable to combined licenses in paragraph b., the combined 
license is not subject to any of the statutory restrictions in 
paragraph a. The NRC believes that the provisions of Section 185 of the 
AEA do not apply to a manufacturing license, inasmuch as a 
manufacturing license is not, per se, a construction permit. 
Accordingly, no earliest and latest date for completion of manufacture 
would be required to be stated in a manufacturing license.
    Section 50.55(c) makes the license conditions in Sec.  50.54 also 
apply to construction permits, unless otherwise modified. In the 
proposed rule, the NRC revised this paragraph to add a reference to 
combined licenses. However, upon further consideration, the NRC has 
determined that no change to Sec.  50.55(c) is necessary because the 
introduction to Sec.  50.54 outlines which provision in that section 
apply to combined licenses.
    Section 50.55(e) addresses the obligation of holders of 
construction permits and their contractors and subcontractors, to 
report defects constituting a substantial safety hazard. These 
requirements, which implement Section 206 of the ERA, as amended, are 
comparable to the requirements in 10 CFR part 21. As discussed with 
respect to the NRC's changes to part 21, the NRC is retaining the 
current regulatory structure, whereby persons and entities engaged in 
activities constituting construction (and their contractors and 
subcontractors) are subject to Sec.  50.55(e), and persons and 
licensees who are authorized to operate a nuclear power plant (and 
their contractors and subcontractors) are subject to part 21. Inasmuch 
as a combined license under part 52 authorizes both construction and 
operation, a combined license holder would be subject to the reporting 
requirements in Sec.  50.55(e) from the date of issuance of the 
combined license until the Commission makes the finding under Sec.  
52.103. Thereafter, the combined license holder would be governed by 
the reporting requirements in part 21. The manufacture of a nuclear 
power reactor under a manufacturing license is the functional 
equivalent of construction. Accordingly, the NRC's view is that the 
holder of a manufacturing license should be subject to reporting under 
Sec.  50.55(e). Standard design approvals under subpart E to part 50 
(former appendix M to part 52) and design certifications under subpart 
B of part 52 are not directly associated with construction, and the NRC 
believes that their reporting should be addressed under part 21. 
Accordingly, the NRC is revising Sec.  50.55(e)(1) to provide that the 
reporting requirements in Sec.  50.55(e) apply to a holder for a 
combined license (until the NRC makes the finding under Sec.  
52.103(g)), and a manufacturing license under part 52. As discussed 
further in Section J on part 21 of this document, early site permits do 
not authorize ``construction'' or its functional equivalent. Therefore, 
early site permits are subject to the requirements of part 21 rather 
than Sec.  50.55(e) under the final rule.
    Section 50.55(f) sets forth the NRC's requirements with respect to 
compliance with the QA requirements in 10 CFR part 50, appendix B, and 
implementation of the construction permit holder's QA program as 
described in its SAR. Comparable provisions applicable to holders of 
operating licenses are contained in Sec.  50.54(a); requirements 
governing the SAR's description of the QA program are contained in 
Sec.  50.34. A detailed discussion of all changes related to QA 
requirements can be found in Section IV.D.13.b of this document.
i. Section 50.55a, Codes and Standards
    Section 50.55a provides requirements relating to codes and 
standards for construction permits and operating licenses for boiling 
or pressurized water-cooled nuclear power facilities. The NRC is 
revising Sec.  50.55a to clarify how the regulations in Sec.  50.55a 
apply to approvals, certifications, and licenses issued under 10 CFR 
part 52. Section 50.55a formerly applied to combined licenses by virtue 
of the provision in current Sec.  52.83, which stated that all 
provisions of 10 CFR part 50 and its appendices applicable to holders 
of construction permits and operating licenses also apply to holders of 
combined licenses. Also, Sec.  50.55a formerly applied to design 
certifications by virtue of the provision in former Sec.  52.48, which 
states that design certification applications will be reviewed for 
compliance with the standards set out in 10 CFR part 50 as it applies 
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. Although former appendix O to 
part 52 does not explicitly require applicants for design approvals to 
comply with the requirements of Sec.  50.55a, the NRC is requiring 
design approval holders to comply with Sec.  50.55a because the NRC 
believes that the requirements for a design approval should be the same 
as the requirements for design certification, given that the reviews 
performed by the NRC staff for the two products are essentially 
identical. Finally, appendix M to part 52, Section M.1, states that the 
provisions in part 50 applicable to construction permits apply in 
context, with respect to matters of radiological health and safety, 
environmental protection, and the common defense and security, to 
manufacturing licenses. Therefore, the NRC is modifying Sec.  50.55a to 
state that each combined license for a utilization facility is subject 
to the conditions in Sec.  50.55a, but is only subject to the 
conditions in Sec. Sec.  50.55a(f) and (g) after the NRC makes the 
finding under Sec.  52.103. The modifications to Sec.  50.55a also 
state that each manufacturing license, design approval, and design 
certification application is subject to the conditions in Sec. Sec.  
50.55a(a), (b)(1), (b)(4), (c), (d), (e), (f)(3), and (g)(3), which are 
the provisions related to nuclear power facility design.
j. Section 50.59, Changes, Tests, and Experiments
    This section presents a change process for information contained in 
the FSAR. Section 50.59(b) is revised to clarify that this change 
process is applicable to holders of operating licenses issued under 
part 50 and combined licenses issued under part 52. If the combined 
license references a design certification rule, then the information in 
the design control document is controlled by the change process in the 
applicable design certification rule. Section 50.59(d)(2) is revised to 
conform the frequency that summary reports are submitted for holders of 
combined licenses with the frequency set forth in the design 
certification rules. Section 50.59(d)(3) is revised to clarify that the 
requirement for maintaining records applies to holders of operating 
licenses issued under part 50 and combined licenses issued under part 
52.

[[Page 49404]]

k. Section 50.61, Fracture Toughness Requirements for Protection 
Against Pressurized Thermal Shock Events
    This section is revised to clarify that the fracture toughness 
requirements apply to an operating license for a pressurized water 
reactor issued under part 50 or a combined license for a pressurized 
water reactor issued under 10 CFR part 52.
l. Section 50.62, Requirements for Reduction of Risk From Anticipated 
Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear 
Power Plants
    Paragraph (d) of Sec.  50.62 provides implementation requirements 
for the requirements of the section. This paragraph is revised to 
indicate that these implementation requirements only apply to light-
water-cooled nuclear power plant operating licenses issued before the 
effective date of this final rule. Section 50.62 is revised to require 
each light-water-cooled nuclear power plant operating license 
application submitted after the effective date of this final rule to 
submit information in its final safety analysis report demonstrating 
how it will comply with paragraphs (c)(1) through (c)(5) of Sec.  
50.62. Similarly, the NRC is adding provisions to Sec. Sec.  52.47, 
52.79, 52.137, and 52.157 requiring that applicants for standard design 
certifications, combined licenses, standard design approvals, and 
manufacturing licenses include the information required by this section 
in their final safety analysis reports.
m. Section 50.63, Loss of All Alternating Current Power
    Conforming changes are made to this section to clarify that the 
requirements for station blackout apply to applications for 
construction permits, combined licenses, design approvals, design 
certifications, manufacturing licenses, and operating licenses.
n. Section 50.65, Requirements for Monitoring the Effectiveness of 
Maintenance at Nuclear Power Plants
    This section presents the requirements for monitoring the 
effectiveness of maintenance at nuclear power plants. Paragraph 
50.65(a) is revised to clarify that holders of operating licenses 
issued under part 50 and combined licenses issued under part 52 must 
comply with the requirements in this section. In the proposed rule, 
Sec.  50.65(c) was revised to specify that, for new licenses issued 
after the effective date of this regulation, the requirements of this 
section must be implemented 30 days before the initial fuel loading of 
the reactor. Commenters recommended that NRC should not require 
implementation prior to fuel load when not all systems will have been 
placed in service. The NRC agrees with this comment and has deleted the 
proposed revision to Sec.  50.65(c). Under the final rule, licensees 
are required to implement the requirements of this section by the time 
that initial fuel loading has been authorized.
6. Inspections, Records, Reports, Notifications
a. Section 50.70, Inspections
    Section 50.70(a) requires that each licensee and each holder of a 
construction permit allow inspection, by duly authorized 
representatives of the Commission, of its records, premises, 
activities, and of licensed materials in possession or use, related to 
the license or construction permit as may be necessary to effectuate 
the purposes of the AEA. The language in Sec.  50.70(a) encompasses 
combined license holders and manufacturing license holders because they 
are licensees. In addition, the provision in former Sec.  52.83, states 
that all provisions of 10 CFR part 50 and its appendices applicable to 
holders of construction permits and operating licenses also apply to 
holders of combined licenses. Also, former Section M.1 of appendix M to 
part 52, states that the provisions in part 50 applicable to 
construction permits apply in context, with respect to matters of 
radiological health and safety, environmental protection, and the 
common defense and security, to manufacturing licenses. Section 
50.70(a) is revised to clarify that these inspection requirements also 
apply to holders of early site permits under 10 CFR part 52. An early 
site permit is a partial construction permit and therefore should be 
subject to the same inspection requirements as a construction permit. 
In addition, the NRC is clarifying that the inspection requirements 
also apply to applicants for licenses, construction permits, and early 
site permits. It is common for applicants to perform activities related 
to NRC regulations before issuance of the license or permit for which 
they are applying and it has been the NRC's practice to inspect these 
activities whenever they are performed. Therefore, the modification to 
require that the inspection requirements in Sec.  50.70(a) apply to 
applicants is simply a codification of the NRC's current practices.
    Section 50.70(b)(1) requires that each licensee and each holder of 
a construction permit provide rent-free office space for the exclusive 
use of NRC inspection personnel. The existing language in this 
provision encompasses combined license holders and manufacturing 
license holders. Section 50.70(b)(2) provides requirements regarding 
the space to be provided for a site with a single power reactor 
facility licensed under 10 CFR part 50 and for sites containing 
multiple power reactor units. The NRC is revising Sec.  50.70(b)(2) to 
clarify that these requirements also apply to sites for combined 
license holders under 10 CFR part 52 and to facilities issued 
manufacturing licenses under 10 CFR part 52.
b. Section 50.71, Maintenance of Records, Making of Reports
    Section 50.71 establishes the NRC's requirements for maintenance 
and retention of records and reports, and updating of FSARs. Section 
50.71(a) requires each licensee and each holder of a construction 
permit to maintain all records and make all reports as may be required 
by license, or by the NRC's regulations. The former language does not 
apply to non-licensees, such as holders of standard design approvals 
and applicants for standard design certifications, even though it would 
appear that these requirements should Accordingly, the NRC is revising 
Sec.  50.71(a) to make its provisions applicable to holders of standard 
design approvals and all applicants for design certification during the 
period of NRC consideration of the application for design 
certification, and those applicants for design certification whose 
designs are certified via rulemaking in accordance with subpart B of 10 
CFR part 52.
    Section 50.71(c) specifies that the default record retention period 
(i.e., the period that applies if a record retention period is not 
specified by the regulation requiring the record) ends when the NRC 
``terminates the facility license.'' A manufacturing license is not a 
``facility'' license, inasmuch as subpart F of part 52 is limited to 
the manufacture of reactors, not a ``facility.'' Finally, some licenses 
(e.g., early site permits and manufacturing licenses) may either be 
terminated by the NRC, or ``expire'' as a matter of law at the end of 
their term. Accordingly, the NRC is revising Sec.  50.71(c) to 
establish the records retention period and to properly refer to 
manufacturing licenses, early site permits, and construction permits.
    Section 50.71(e) establishes the updating requirements for the 
FSAR, including the information that must be included in each update. 
The former regulation, however was deficient in two respects. First, it 
did not address the updating requirements for combined license 
applicants and holders. Second,

[[Page 49405]]

the regulation, if applied to manufacturing licenses under subpart F of 
part 52, imposed unnecessary regulatory burden with respect to periodic 
updating.
    Accordingly, the NRC is revising Sec.  50.71(e) to specify the FSAR 
updating requirements for combined license applicants and holders. In 
addition, current Sec.  50.71(f) is redesignated as Sec.  50.71(g), and 
a new Sec.  50.71(f) is added.
    Section 50.71(e)(3)(iii) is added to contain the provisions 
applicable to combined license holders during the period of time from 
docketing of the application to the Commission finding under Sec.  
52.103(g). The update frequency during this period is established as 
annually, which is consistent with requirements in Section X.B.3.b of 
the design certification rules in appendices A through D of part 52 for 
combined license holders that reference those rules. After the 
Commission finding under Sec.  52.103(g), the frequency would be 
governed by Sec.  50.71(e)(4), as for other operating reactors.
    Section 50.71(f) is revised to require the holder of the 
manufacturing license to update the FSAR to reflect any modifications 
to the design of the reactor authorized to be manufactured which have 
been approved by the NRC under Sec.  52.171, or any new analyses 
requested to be performed by the NRC. Periodic updating of an FSAR for 
a manufacturing license is not required by Sec.  50.71(f), inasmuch as 
the NRC's concept for a manufacturing license is for the design of the 
reactor authorized to be manufactured to be stable with no changes 
except as specifically approved by the NRC as necessary for adequate 
protection to public health and safety or common defense and security, 
or to ensure compliance with the NRC's requirements in effect at the 
time of issuance of the manufacturing license. The provision in Sec.  
50.71(f) requiring the FSAR for a manufacturing license to be updated 
to reflect new safety analyses required by the NRC is analogous to the 
existing updating requirement in Sec.  50.71(e). This assures that new 
analyses performed to demonstrate the continuing adequacy of the 
unchanged manufactured reactor design are appropriately reflected in 
the FSAR.
    Paragraph (g), formerly (f), is being revised to add reference to 
Sec.  52.110(a)(1) for permanent cessation of operation for plants 
licensed under part 52.
    Finally, paragraph (h) is being added to 50.71. This paragraph 
contains requirements for licensees to maintain and upgrade the PRA 
periodically throughout the plant life. These provisions apply only to 
COLs under part 52, but are included in part 50 in this section 
covering maintenance of records and making of reports, consistent with 
the Commission's practice elsewhere in development of the requirements 
for the part 52 processes.
    These new requirements are a culmination of the Commission's 
interest in use of risk-informed processes as articulated in its 1995 
Policy Statement (``Use of Probabilistic Risk Assessment Methods in 
Nuclear Activities: Final Policy Statement,'' (60 FR 42622; August 16, 
1995)).In the original part 52 rule, each design certification holder 
was required to include as part of the application a design-specific 
PRA. The Commission has been engaged in an effort to improve PRA 
quality through support and endorsement of consensus standards on PRA 
methods.
    In the proposed rule published in March 2006, the Commission 
included a specific request for comment (Question 10, ``New 
Requirements for Periodic Updates to the PRA''--see section IV of this 
document) about part 52 licensees periodically updating the PRA 
throughout the life of the facility, on a schedule similar to that for 
FSAR updates. Several commenters noted that the proposed rule did not 
include a frequency for updating the PRA. These commenters stated that 
they believed that PRA update frequency should be addressed in guidance 
rather than regulations. These commenters indicated a frequency of once 
every two operating cycles would be reasonable and consistent with 
existing requirements in 10 CFR 50.69(e). After considering the 
comments received, the Commission has decided to require combined 
license holders to maintain and upgrade a PRA to meets endorsed 
standards over the lifetime of the facility. To implement this 
decision, new requirements are being placed in Sec.  50.71(h).
    Paragraph (h)(1) requires each holder of a combined license, by the 
time of the scheduled fuel load date for the facility, to develop a 
plant-specific PRA. The PRA is to be both level 1 and level 2 and must 
cover those modes of operation and initiating events for which NRC-
endorsed consensus standards are in effect one year prior to that date. 
Level 1 refers to the identification and quantification of sequences 
leading to the onset of core damage. Level 2 refers to identification 
and quantification of severe accident progression and containment 
response. Additional information about scope and quality of PRA to meet 
these provisions will be addressed in the NRC documents endorsing the 
standards, or in the standards themselves.
    The one year time period was chosen to allow time for the licensee 
to develop and upgrade its PRA and conduct peer review prior to the 
date when the PRA must be completed (i.e., by the scheduled date for 
initial fuel load). The scheduled fuel load date was selected because 
the COL holder chooses this date, and thus is in a position to 
determine when the ``one-year prior'' requirement comes into effect. 
Note that this provision does not require that this PRA be submitted to 
the NRC for review and approval. The need for any such submittal or 
review would be determined by any risk-informed application for which 
the licensee might wish to use this PRA, such as in support of 
licensing actions.
    Paragraph (h)(2) requires the COL holder to maintain the PRA until 
permanent cessation of operations under Sec.  52.110(a). The Commission 
intends PRA maintenance to be consistent with how it is defined in the 
American Society of Mechanical Engineers (ASME) ``Standard for 
Probabilistic Risk Assessment for Nuclear Power Plant Applications'' 
(ASME-RA-Sb-2005), that is ``the update of the PRA models to reflect 
plant changes, such as modifications, procedure changes or plant 
performance.'' No specific frequency is defined in the rule for such 
maintenance; the Commission expects licensees to follow the ASME (or 
other consensus body) guidance on this aspect.
    The paragraph further provides that the PRA must be upgraded every 
four years, to cover initiating events and operational modes contained 
in NRC-endorsed consensus standards in effect one year prior to each 
required upgrade. The Commission intends PRA upgrade to be consistent 
with how it is defined in consensus standards, such as ASME-RA-Sb-2005, 
that is, ``the incorporation into a PRA model of a new methodology or 
significant changes in scope or capability.'' If no new standards are 
issued during a four-year upgrade cycle, licensees would not be 
required to upgrade their PRAs; however, the requirement to maintain 
the PRA would still be in effect. It should also be noted that there 
may be situations where a PRA upgrade is needed more frequently than 
the four year cycle, as for instance to support a new risk-informed 
application.
    Finally, paragraph (h)(3) specifies that each holder of a combined 
license shall, no later than the date on which the licensee submits an 
application for a renewed license, upgrade the PRA to

[[Page 49406]]

cover all modes and all initiating events. This requirement is not 
premised on the existence of NRC-approved consensus standards, and an 
all-mode, all-initiator PRA must be developed even if standards do not 
yet exist. The requirement to develop and maintain such a PRA by the 
time of license renewal application is intended only to establish a 
timing requirement for completing the upgrade of the PRA, and does not 
have any implications on the current requirements for license renewal. 
The upgraded PRA is not an element of any (i.e., past, present, or 
future) review or approval of a license renewal application.
    In implementing these new requirements, it is the NRC's expectation 
that industry stakeholders will work with the NRC and appropriate codes 
and standard setting bodies to continually upgrade the relevant codes 
and standards, identify potential issues, resolve problems, and create 
relevant guidance to assist in periodically improving the quality and 
comprehensiveness of the PRA.
c. Section 50.72, Immediate Notification Requirements for Operating 
Nuclear Power Reactors
    Section 50.72 currently requires holders of operating licenses 
under part 50 for nuclear power plants to notify the NRC Operations 
Center via the Emergency Notification System of the declaration of any 
of the emergency classes specified in the licensee's approved emergency 
plan and of certain non-emergency events. The NRC's regulatory interest 
in these events also extends to nuclear power plants operating under a 
combined license under subpart C of part 52, but the former language 
did not impose the notification requirements on combined license 
holders. Accordingly, in a conforming change in the final rule, the NRC 
is extending the notification requirements to holders of combined 
licenses under part 52 after the Commission has made the finding under 
Sec.  52.103(g). The NRC did not include a conforming change to this 
section in the proposed rule. However, based on public comments, the 
NRC is including the change in the final rule to make it clear that the 
requirements of Sec.  50.72 only apply to a combined license holder 
after the Commission makes the finding under Sec.  52.103(g). The NRC 
is not extending the notification requirements to other part 52 
processes because the events to be reported under the existing rule 
concern events which can only occur upon fuel load and operation, and 
the remaining part 52 licensing and regulatory approval processes do 
not authorize fuel load or operation.
d. Section 50.73, Licensee Event Report System
    Section 50.73 requires holders of operating licenses under part 50 
for nuclear power plants to submit licensee event reports (LERs) on the 
occurrence of certain operating events to the NRC. LERs facilitate the 
NRC's oversight of operating nuclear power plants, by alerting the NRC 
to the occurrence and underlying causes of events having potential 
safety implications. The NRC's regulatory interest in these events also 
extends to nuclear power plants operating under a combined license 
under subpart C of part 52, but the former language did not impose the 
LER requirement on combined license holders. Accordingly, in a 
conforming change, the NRC is extending the LER reporting requirements 
to holders of combined licenses under part 52 after the Commission has 
made the finding under Sec.  52.103(g). The final rule does not extend 
the LER requirement to other part 52 processes, because the events to 
be reported under the existing rule concern events which can only occur 
upon fuel load and operation, and the remaining part 52 licensing and 
regulatory approval processes do not authorize fuel load or operation.
e. Section 50.75, Reporting and Recordkeeping for Decommissioning 
Planning
    The requirements in Sec.  50.75 are intended to ensure that 
entities who construct and ultimately operate a nuclear power plant 
will have sufficient funds at the end of the operational life of the 
plant to complete the decommissioning of the plant. Section 50.75 
requires a nuclear power plant operating license application to address 
the predicted costs of decommissioning, provide financial assurance by 
one of the means specified in the regulation, and submit evidence that 
one or more of these means has been established. Section 50.75 also 
requires the operating license holder to update the cost estimates for 
decommissioning on an annual basis, and to submit reports to the NRC 
every 2 years describing, inter alia, any adjustments to the amount of 
funds collected annually to reflect any changes in projected 
decommissioning cost. When a plant is within 5 years of its projected 
end of its operation, the reports must be submitted annually, and a 
site-specific decommissioning cost estimate must be submitted. Some of 
these requirements are directed at the two phase licensing process in 
10 CFR part 50, in which the NRC issues a construction permit followed 
by an operating license. These requirements are not well-suited to the 
combined license process under part 52. For example, requiring the 
combined license applicant to comply with the current requirement in 
Sec.  50.75(b)(4) that the operating license applicant submit a copy of 
the financial instrument obtained to satisfy the requirements of Sec.  
50.75(e), would place a more stringent requirement on the combined 
license applicant, inasmuch as that applicant would be required to fund 
decommissioning assurance at an earlier date as compared with the 
operating license applicant.
    To address these discrepancies, the NRC is revising Sec.  50.75 to 
address decommissioning funding assurance for combined licenses. Under 
the final rule, the combined license applicant must submit a 
decommissioning report as required by Sec.  50.33(k), but it need not 
obtain a financial instrument to fund decommissioning or to submit a 
copy to the NRC. Instead, under Sec.  50.75(b)(1) and (4), the combined 
license application must contain a certification that the financial 
assurance will be provided no later than 30 days after the NRC 
publishes notice in the Federal Register under Sec.  52.103(a). See 
Sec.  50.75(b)(1).
    The proposed rule would have required the combined license holder 
to submit, by March 31 of each year until the date that the NRC 
authorizes fuel load under Sec.  52.103(g), an updated certification of 
the information required by paragraph (b)(1). The proposed rule also 
would have required the combined license holder to submit, no later 
than 30 days after the Commission publishes notice in the Federal 
Register under Sec.  52.103(a), a certification that financial 
assurance is being provided in the relevant amount together with a copy 
of the financial instrument obtained to satisfy the requirements of 
Sec.  50.75(e). Once the Commission has made the finding under Sec.  
52.103, the proposed rule would have required the combined license 
holder to be subject to the reporting and updating requirements as an 
operating license holder under part 50, including the requirements 
applicable when the plant is within 5 years of the projected end of 
operation. A commenter objected to the annual reporting requirement, 
arguing that an annual update during the construction period would 
serve no purpose and is unnecessary and unduly burdensome. The 
commenter proposed that the holder be allowed to adjust or update the 
original certification at the time construction is complete and the 
plant is ready to begin operation. Upon

[[Page 49407]]

further consideration, the Commission has decided to modify the final 
rule by eliminating the requirement for annual reports, and instead 
requiring the updating reports 2 years and 1 year before the date 
scheduled for initial loading of fuel load (consistent with the 
schedule required by Sec.  52.99(a)). The Commission's objective is to 
have sufficient time to evaluate the projected costs of 
decommissioning, and any licensee-proposed changes in the financial 
assurance mechanism for funding before fuel is loaded into the reactor 
and operation commences. This will allow the Commission to take any 
necessary regulatory action before fuel loading and commencement of 
operation.
    The final rule requires that no later than 30 days after the 
Commission publishes notice in the Federal Register under Sec.  
52.103(a), the combined license holder must submit a report to the NRC. 
The report must contain a certification that financial assurance is 
being provided in an amount specified in the licensee's most recent 
updated certification (i.e., the certification provided 1 year before 
the scheduled date for initial loading of fuel, in accordance with the 
first sentence of Sec.  50.75(e)(3)). The certification must include a 
copy of the financial instrument obtained to provide decommissioning 
funding assurance. The requirements in paragraph (f)(1) of Sec.  
52.103(a), which are applicable to the combined license holder after 
the Commission has made the finding under Sec.  52.103, are adopted in 
the final rule without change from the proposed rule.
    The Sec.  50.75 decommissioning funding requirements do not apply 
to an applicant for, and holder of, a manufacturing license under part 
52. The NRC did not intend, when it first adopted Sec.  50.75, to 
subject holders of manufacturing licenses to the requirements of that 
section. It is clear from the words of former Sec.  50.33(k)(1) that 
the rule applies only to applications for operating licenses for 
production and utilization facilities. A manufacturing license by 
itself does not authorize either fuel load or operation, which are the 
activities necessitating the expenditure of funds for decommissioning. 
Therefore, there is no need for a holder of a manufacturing license, 
who does not intend to operate the reactor being manufactured to 
provide funding.
7. US/IAEA Safeguards Agreement
a. Section 50.78, Installation Information and Verification
    Since 1980, the U.S./International Atomic Energy Agency (IAEA) 
Safeguards Agreement has allowed IAEA inspection and verification 
activities at U.S. facilities that the IAEA selects from the U.S. 
Eligible Facilities List. The safeguards agreement is implemented under 
the Nuclear Non-Proliferation Treaty, which provides assurance that all 
nuclear materials declared to be in peaceful use are not diverted to 
potential use in nuclear explosives. Although 10 CFR part 75 contains 
most of the NRC requirements intended to implement the installation, 
inspection, and verification provisions of the Safeguards Agreement 
with IAEA, Sec.  50.78 requires each holder of a construction permit to 
submit certain information on Form N-71, permit verification by 
representatives of the IAEA, and take any other action necessary to 
implement the Safeguards Agreement. Inasmuch as combined licenses 
authorize construction of a nuclear power plant at a fixed site, the 
provisions of Sec.  50.78 should also apply to a holder of a combined 
license under part 52. Accordingly, Sec.  50.78 is revised to specify 
that holders of combined licenses must, if requested by the NRC, submit 
installation information on Form N-71, permit verification of that 
information by the IAEA, and take other action as may be necessary to 
implement the Safeguards Agreement, in the manner set forth in Sec.  
75.6, and Sec. Sec.  75.11 through 75.14.
8. Transfers of Licenses--Creditors' Rights--Surrender of Licenses
a. Section 50.80, Transfer of Licenses
    Section 50.80 implements Sections 101 and 184 of the AEA, which 
require Commission approval for the transfer of a license for a 
production or utilization facility, including a nuclear power reactor. 
Section 50.80(a) explicitly refers to transfers of a ``license for a 
production or utilization facility * * *,'' which would include 
construction permits under part 50, as well as all licenses and permits 
issued under part 52. However, to explicitly recognize the 
applicability of Sec.  50.80(a) to both permits under parts 50 and 52 
and all licenses under part 52, Sec.  50.80(a) is revised to explicitly 
refer to permits under parts 50 and 52, and licenses under part 52. The 
proposed rule would have only made these clarifying revisions. A 
commenter on the proposed rule stated that some of the requirements in 
Sec.  50.80 are not relevant to transfers of an ESP. The NRC agrees, 
and has revised the final rule to specify which criteria are applicable 
to transfer of an ESP. Specifically, paragraph (b)(1)(ii) requires an 
application for transfer of an ESP to include as much of the 
information described in Sec. Sec.  52.16 and 52.17 with respect to the 
identity and technical qualifications of the proposed transferee as 
would be required by those sections if the application were for an 
initial license. This change removes the requirement for the applicant 
for transfer of an ESP to address financial qualifications since this 
is not required of an initial ESP applicant. In addition, this change 
removes the provision that the NRC may require additional information 
as part of an ESP transfer with respect to data on proposed safeguards 
against hazards from radioactive materials and the applicant's 
qualifications to protect against such hazards. Information on these 
subject matters is not relevant to an ESP transfer, inasmuch as an ESP 
does not authorize the holder to possess radioactive material.
    The NRC declines to adopt the suggestion of a commenter who 
suggested that the statement of considerations clarify when a transfer 
of an ESP is necessary. The NRC's revision to Sec.  50.80 is a 
conforming change to a procedural regulation, the process by which the 
NRC processes and determines a transfer of a license. Section 50.80 
does not, by itself, specify the circumstances for which a license 
transfer is necessary; it simply addresses what procedures must be 
followed if a license transfer request is received. Therefore, the NRC 
does not believe that it is necessary or desirable to provide such 
guidance in the context of this rulemaking.
b. Section 50.81, Creditor Regulations
    Section 50.81 implements Section 184 of the AEA, which requires the 
consent of the Commission for the creation of any mortgage, pledge or 
other lien upon any Commission-licensed facility or special nuclear 
material. To ensure that the reach of Sec.  50.81 is as broad as the 
statutory requirement, the NRC is revising the definition of license 
and facility. The definition of license in this section is revised to 
explicitly refer to all licenses under 10 CFR, and early site permits 
under part 52. The definition of facility is revised to add a new 
paragraph which explicitly refers to an early site permit under part 
52, and a reactor manufactured under a manufacturing license under part 
52.

[[Page 49408]]

9. Amendment of License or Construction Permit at Request of Holder
a. Section 50.90, Application for Amendment of License or Construction 
Permit; section 50.91, Notice for Public Comment; State Consultation; 
and section 50.92, Issuance of Amendment
    Sections 50.90, 50.91, and 50.92 govern the procedures and criteria 
for NRC consideration and issuance of amendments to licenses and 
construction permits. The regulations do not clearly address early site 
permits, combined licenses, or manufacturing licenses. Accordingly, the 
NRC is making a number of changes in these regulations.
    Section 50.90 provides that applicants for amendment of a license 
or construction permit must file their application with the NRC as 
described in Sec.  50.4, following the form prescribed for the original 
application. Although the term, license, as amended in Sec.  50.2 
includes combined licenses, manufacturing licenses, and early site 
permits under part 52, Sec.  50.92 is revised to explicitly refer to 
these part 52 licenses to eliminate any confusion with respect to the 
applicability of this section to part 52 licenses. A similar change is 
made in the introductory paragraph of Sec.  50.91.
    Sections 50.92 and 50.91(a)(4) implement the Commission's authority 
under Section 189 of the AEA to dispense with the advance publication 
of a Federal Register document requesting a hearing with respect to 
license amendments, and to make operating license and combined license 
amendments immediately effective upon issuance, if the NRC finds that 
the amendment involves no significant hazards consideration. The NRC is 
revising Sec.  50.92(c) to clarify that, consistent with Section 189 of 
the AEA, the NRC may make a no significant hazards consideration 
determination for amendments of combined licenses under part 52. 
Combined licenses are explicitly mentioned in Section 189.a.(2)(A) of 
the AEA with respect to immediate effectiveness following a Commission 
determination of a no significant hazards consideration. In addition, a 
combined license merges into a single license the authority otherwise 
contained in a construction permit and an operating license, and the 
language of Section 189.a.(1)(A) of the AEA which refers to both 
amendments of construction permits and operating licenses, also applies 
to amendments of combined licenses.
    Finally, Sec.  50.92(a) is revised to provide that a separate 
application for a construction permit is not required even where a 
holder of a combined license or a manufacturing license must seek a 
license amendment because of a material alteration. There is no safety 
or regulatory benefit in requiring the licensee to concurrently submit 
an application for a new construction permit in addition to a license 
amendment, inasmuch as NRC review of the alteration is assured.
10. Revocation, Suspension, Modification, Amendment of Licenses and 
Construction Permits, Emergency Operations by the Commission
a. Section 50.100, Revocation, Suspension, Modification of Licenses, 
Permits, and Approvals for Cause
    Section 50.100 is revised to explicitly address the Commission's 
authority to suspend, modify, or revoke any standard design approval 
under subpart E of parts 50 or 52 for any material false statement in 
the application, or because of any statement in any report, record, 
inspection, or condition revealed by the application, or by other 
means, which would warrant the NRC to refuse to grant the design 
approval on an original application. The former language of Sec.  
50.100, which is retained as paragraph (a) in the final rule, applied 
to any license or any license or construction permit issued under part 
50 for any material false statement in the application for the license 
or permit, or because of any statement in any report, record, 
inspection, or condition revealed by the application, or by other 
means, which would warrant the NRC to refuse to grant a license on an 
original application, or for failure to construct or operate a facility 
in accordance with the applicable license or permit. While this 
language applies to early site permits, combined licenses and 
manufacturing licenses, by virtue of their status as licenses under the 
AEA, it does not clearly apply to standard design approvals as these 
are not licenses. Nonetheless, the Commission possesses authority to 
modify, suspend or revoke the regulatory approvals. Accordingly, the 
NRC is revising this section to add a reference to a standard design 
approval.
    The final rule is different than the proposed rule in several ways. 
A reference to part 50 is added in the clause governing revocations, 
suspensions, and modifications of licenses. The word, ``provided * * 
*,'' is revised to read ``provided, however,* * *.'' Finally, a 
reference to a combined license is added to the clause stating that a 
failure to meet the timely completion of proposed construction or 
alteration is subject to Sec.  50.55(b) (which is also revised in this 
final rulemaking to make its provisions applicable to combined 
licenses).
11. Backfitting
a. Section 50.109, Backfitting
    The backfit rule, 10 CFR 50.109, provides certain protection to 
nuclear power plant licensees against changes in the NRC requirements 
and NRC staff positions on those requirements. Prior to the final rule, 
the backfitting provisions in Sec.  50.109 applied to standard design 
approvals, construction permits, and operating licenses, but did not 
address combined licenses or manufacturing licenses. Part 52 contains 
special backfitting requirements on early site permits, design 
certification rules, but prior to this rulemaking, neither Sec.  50.109 
or part 52 addressed backfitting of a combined license, although the 
NRC recognizes that backfitting restraints for an early site permit and 
a design certification rule would apply to a combined license 
referencing either or both. To address these gaps in backfitting, and 
to clarify the application of special backfitting provisions, Sec.  
50.109(a)(1) is revised by establishing the date that backfitting 
protection begins for a manufacturing license, a construction permit 
for a duplicate design license, and a combined license. Moreover, with 
respect to a part 50 construction permit, a part 50 operating license, 
and a part 52 combined license, Sec.  50.109 is revised by listing the 
specific backfitting restrictions that apply if an early site permit, 
standard design approval, or standard design certification rule is 
referenced, or if a nuclear power reactor manufactured under a part 52 
manufacturing license is used.
    In the statement of considerations for the 2006 proposed rule, the 
Commission asked whether, instead of conforming the language of Sec.  
50.109 to reflect the licensing and regulatory approval processes in 
part 52, the Commission should adopt a general backfitting provision, 
analogous to Sec.  50.109, in part 52. Commenters either expressed no 
opinion on the matter, or otherwise indicated that they did not have a 
preference. Accordingly, the Commission has decided to revise Sec.  
50.109 to include the conforming changes, rather than adopting a 
backfitting provision in part 52.

[[Page 49409]]

12. Enforcement
a. Section 50.120, Training and Qualification of Nuclear Power Plant 
Personnel
    This section sets forth the requirements for training and 
qualifying nuclear power plant personnel. In a conforming change, the 
NRC is revising Sec.  50.120 to add applicants for and holders of 
combined licenses as being subject to this provision.
13. Appendices
a. Appendix A to Part 50--General Design Criteria for Nuclear Power 
Plants
    The first paragraph of the Introduction to appendix A to part 50 is 
revised to clarify that the general design criteria in appendix A to 
part 50 apply to applications for combined licenses, design approvals, 
design certification, and manufacturing licenses, as well as for 
construction permits. Also, General Design Criterion (GDC) 19 of 
appendix A to part 50, which sets forth requirements for a main control 
room in a nuclear power plant, is revised to clarify that the radiation 
protection requirements in GDC 19 for applications filed after January 
10, 1997, apply to design approvals and manufacturing licenses issued 
under part 52, in addition to design certifications and combined 
licenses.
b. Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power 
Plants and Fuel Reprocessing Plants
    Appendix B to part 50 states that every applicant for a 
construction permit is required to include in its preliminary safety 
analysis report a description of the quality assurance program to be 
applied to the design, fabrication, construction, and testing of the 
SSCs of the facility and every applicant for an operating license is 
required to include, in its FSAR, information pertaining to the 
managerial and administrative controls to be used to assure safe 
operation. The NRC is revising appendix B to part 50 to clarify that 
these requirements also apply to early site permits, design approvals, 
design certifications, combined licenses, and manufacturing licenses 
under 10 CFR part 52. Specifically, the introduction to appendix B to 
part 50 is revised to state that every applicant for a combined license 
is required by the provisions of Sec.  52.79 to include in its FSAR a 
description of the quality assurance program applied to the design, and 
to be applied to the fabrication, construction, and testing of the SSCs 
of the facility and to the managerial and administrative controls to be 
used to assure safe operation. The introduction also states that, for 
applications submitted after the effective date of the final rule, 
every applicant for an early site permit is required by the provisions 
of Sec.  52.17 to include in its site safety analysis report a 
description of the quality assurance program applied to site activities 
related to the design, fabrication, construction, and testing of the 
SSCs of a facility or facilities that may be constructed on the site. 
The introduction states that every applicant for a design approval or 
design certification is required by the provisions of Sec. Sec.  52.137 
and 52.47, respectively, to include in its FSAR a description of the 
quality assurance program applied to the design of the SSCs of the 
facility. Finally, the introduction states that every applicant for a 
manufacturing license is required by the provisions of 10 CFR 52.157 to 
include in its FSAR a description of the quality assurance program 
applied to the design, and to be applied to the manufacture of, the 
SSCs of the reactor. The wording in appendix B of part 50 and in the 
related provisions in the contents of application sections in 10 CFR 
part 52 is modified slightly in the final rule to reflect that some 
activities have already occurred when the application is submitted 
(e.g., design of SSCs for design certification applicants). Therefore, 
instead of requiring that the application describe the QA program ``to 
be applied'' to these activities, the final rule requires that the 
application describe the QA program ``applied'' to these activities, 
since they have already occurred.
    The NRC is maintaining the current regulatory structure for 
requirements that implement appendix B to part 50 whereby QA for 
construction activities is governed by Sec.  50.55(f), and QA for 
operation is governed by Sec.  50.54(a). Because a combined license 
under part 52 authorizes both construction and operation, a combined 
license holder should be subject to the QA requirements in Sec.  
50.55(f) from the date of issuance of the combined license until the 
Commission makes the finding under Sec.  52.103(g) that allows the 
licensee to load fuel and operate. Thereafter, the combined license 
holder should be governed by the QA requirements in Sec.  50.54(a). The 
manufacture of a nuclear power reactor under a manufacturing license is 
the functional equivalent of construction. Accordingly, the NRC is 
revising Sec.  50.55(f) to refer to holders of manufacturing licenses 
under part 52. Early site permits under subpart A precede construction 
and are considered partial construction permits. Hence the NRC believes 
that they should be subject to QA under Sec.  50.55(f), and Sec.  
50.55(f) is revised accordingly.
    Appendix B to part 50 was formerly applicable to combined licenses 
under the provisions of Sec.  52.83, which states that all provisions 
of 10 CFR part 50 and its appendices applicable to holders of operating 
licenses also apply to holders of combined licenses. Appendix B to part 
50 formerly applied to design certifications by virtue of the provision 
in former Sec.  52.48, which stated that design certification 
applications will be reviewed for compliance with the standards set out 
in 10 CFR part 50 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. Former appendix O to part 52, Section O.3, required 
applicants for design approvals to include the information required by 
Sec. Sec.  50.34(a) and (b), as appropriate, and stated that the 
information required by Sec.  50.34(a)(7) (a description of the quality 
assurance program and a discussion of how the applicable requirements 
of appendix B to part 50 will be satisfied), shall be limited to the QA 
program to be applied to the design, procurement and fabrication of the 
SSCs for which design review has been requested. Appendix B to part 50 
formerly applied to manufacturing licenses by virtue of the provision 
in former appendix M to part 52, Section M.1, which stated that the 
provisions in part 50 applicable to construction permits apply in 
context, with respect to matters of radiological health and safety, 
environmental protection, and the common defense and security, to 
manufacturing licenses.
    Early site permits are considered partial construction permits, 
therefore, the NRC believes that they should be subject to the QA 
requirements of appendix B to part 50. Section 52.39, with certain 
specific exceptions, requires the Commission to treat matters resolved 
in an early site permit proceeding as resolved in making findings for 
issuance of a construction permit, operating license, or combined 
license. Because of this finality, conclusions made during the early 
site permit phase will be relied upon for use in subsequent design, 
construction, fabrication, and operation of a reactor that might be 
constructed on the site for which an early site permit is issued. 
Therefore, the NRC believes that the level of quality used to control 
activities related to safety-related SSCs should be equivalent in the 
early site permit and combined license phases. For these reasons, 
applicants must apply quality

[[Page 49410]]

controls to each early site permit activity associated with the 
generation of design information for safety-related SSCs that meet the 
criteria in appendix B to part 50. Therefore, the NRC is revising 
appendix B to part 50 to make it applicable to early site permits.
c. Appendix C to Part 50--A Guide for the Financial Data and Related 
Information Required To Establish Financial Qualifications for 
Construction Permits and Combined Licenses
    Section 182.a of the AEA requires an applicant for a license for a 
production or utilization facility to submit information in its 
application * * * ``as the Commission, regulation, may determine to be 
necessary to decide such of the technical and financial qualifications 
of the applicant * * * as the Commission may deem appropriate for the 
license.'' The NRC has long determined the need for non-utility 
applicants for nuclear power plant construction permits and operating 
licenses to establish their financial qualifications (see 10 CFR 
50.33(f)), and has set forth the specific information on financial 
qualifications to be provided by applicants for construction permits in 
appendix C to part 50. Inasmuch as holders of combined licenses under 
part 52 are authorized to perform the same construction activities with 
respect to a nuclear power plant as a holder of a construction permit 
under part 50, the NRC believes that applicants for combined licenses 
should be subject to the requirements of appendix C to part 50. 
Accordingly, the title of appendix C is revised to make clear the 
applicability of this appendix to applicants for combined licenses. 
This change constitutes a conforming change to the revision of Sec.  
50.33.
    With the exception of manufacturing licenses, none of the other 
regulatory processes under part 52, e.g., early site permits, standard 
design certifications, and standard design approvals, authorize any 
activities constituting ``construction'' under the AEA and the 
Commission's regulations.\7\ Therefore, the final rule does not refer 
to early site permits, design certifications, or design approvals under 
part 52. With respect to a reactor manufacturing license, the NRC does 
not believe that a financial qualifications review is necessary for 
several reasons. A financial qualifications review at the manufacturing 
license stage would appear to be redundant to the financial 
qualifications review that is already necessary at the construction 
permit and operating license stages, or combined license stage. 
Sufficient safety and quality assurance reviews, including the use of 
ITAAC in the case of a combined license, should be sufficient to 
address any adverse impacts on safety as the result of inadequate 
financial resources to properly manufacture the reactor. Furthermore, 
the NRC notes that manufacture of a reactor is, in many respects, no 
different than fabrication of components and systems by third party 
vendors, who are not required to obtain an NRC license and demonstrate 
financial qualifications. There seems to be no regulatory value to 
mandate a financial qualifications review of manufacturing license 
applicants, when this type of review is not conducted by the NRC for 
fabricators of nuclear power plant systems and components.
---------------------------------------------------------------------------

    \7\ Although early site permit applicants may seek the authority 
to conduct activities allowed under 10 CFR 50.10(e)(1) (but not 
activities allowed under Sec.  50.10(e)(3), see Sec.  52.17(c)), 
these activities are not considered ``construction.''
---------------------------------------------------------------------------

 d. Appendix E to Part 50--Emergency Planning and Preparedness for 
Production and Utilization Facilities
    See discussion in Section V.D.4.f of this document.
e. Appendix I to Part 50--Numerical Guides for Design Objectives and 
Limiting Conditions for Operation To Meet the Criterion ``as Low as is 
Reasonably Achievable'' for Radioactive Material in Light-Water-Cooled 
Nuclear Power Reactor Effluents
    The Commission is revising appendix I to part 50 to conform to the 
changes in Sec. Sec.  50.34a and 50.36a which are being made as part of 
this final rule. Specifically, a statement is added in Section I of 
appendix I to part 50, stating that Sec. Sec.  52.47, 52.79, 52.137, 
and 52.157 provide that applications for design certification, combined 
license, design approval, or manufacturing license, respectively, shall 
include a description of the equipment and procedures for the control 
of gaseous and liquid effluents and for the maintenance and use of 
equipment installed in radioactive waste systems. In addition, Section 
II of appendix I to part 50 is revised to state that the guides on 
design objectives set forth in appendix I to part 50 may be used by an 
applicant for a combined license as guidance in meeting the 
requirements of Sec.  50.34a(d) or by an applicant for a design 
approval, a design certification, or a manufacturing license as 
guidance in meeting the requirements of Sec.  50.34a(e). Section IV of 
appendix I to part 50 is revised to state that the guides on limiting 
conditions for operation for light-water-cooled nuclear power reactors 
in appendix I to part 50 may be used by an applicant for an operating 
license or a design certification or combined license, or a licensee 
who has submitted a certification of permanent cessation of operations 
under Sec.  50.82(a)(1) or Sec.  52.110 as guidance in developing 
technical specifications under Sec.  50.36a(a) to keep levels of 
radioactive materials in effluents to unrestricted areas as low as is 
reasonably achievable. Finally, Section V of appendix I to part 50 is 
revised to state that the guides for limiting conditions for operation 
set forth in appendix I are applicable to any application filed on or 
after January 2, 1971, for a construction permit for a light-water-
cooled nuclear power reactor, or a design certification, a combined 
license, or a manufacturing license for a light-water-cooled nuclear 
power reactor under part 52. Note that the NRC added the phrase ``for a 
light-water-cooled nuclear power reactor'' to Section V in the final 
rule. This phrase was inadvertently left out of the introduction to 
Section V in the proposed rule. The NRC did not intend to change the 
applicability of appendix I in this rulemaking and is, therefore, 
correcting this omission in the final rule. The NRC has also removed 
the conforming change it had proposed to paragraph A.3 of the 
Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-
2) Guides on Design Objectives for Light-Water-Cooled Nuclear Power 
Reactors in appendix I. The design objectives in this staff position 
are only applicable to those light-water-cooled nuclear power reactors 
that applied for a construction permit before January 2, 1971 (per 
Appendix I, Section V, B.2.). Because part 52 did not exist before 
1971, the proposed change is unnecessary.
f. Appendix J to Part 50--Primary Reactor Containment Leakage Testing 
for Water-Cooled Power Reactors
    Section 50.54(o) provides a condition for all operating licenses 
for water-cooled power reactors that primary reactor containments must 
meet the containment leakage test requirements set forth in appendix J 
to part 50. These test requirements provide for preoperational and 
periodic verification by test of the leak-tight integrity of the 
primary reactor containment, and systems and components which penetrate 
containment of water-cooled power reactors, and establish the 
acceptance criteria for these tests. The purpose of the tests are to 
assure that leakage through the primary reactor containment systems and 
components penetrating primary containment shall not exceed allowable 
leakage rate values

[[Page 49411]]

as specified in the technical specifications or associated bases, and 
periodic surveillance of reactor containment penetrations and isolation 
valves is performed so that proper maintenance and repairs are made 
during the service life of the containment, and systems and components 
penetrating primary containment. The Commission is revising appendix J 
to clarify that these requirements also apply to combined licenses 
under 10 CFR part 52. This is consistent with former Sec.  52.83, which 
stated that all provisions of 10 CFR part 50 and its appendices 
applicable to holders of operating licenses also apply to holders of 
combined licenses.
g. Appendices M and O to Part 50 [Removed]
    The NRC has removed appendices M and O from 10 CFR part 50. 
Appendix M provided for issuance of a license authorizing the 
manufacture of a nuclear power reactor to be incorporated into a 
nuclear power plant under a construction permit and operated under an 
operating license at a different location from the place of 
manufacture. Appendix O addressed the approval of standard designs for 
nuclear power reactors. These appendices were transferred to 10 CFR 
part 52 when it was first issued (54 FR 15372; April 18, 1989). 
However, the NRC failed to remove those appendices from 10 CFR part 50, 
though the NRC intended to do so (see 54 FR 15385; April 18, 1989).
h. Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear 
Power Plants
    Appendix S to part 50 provides earthquake engineering criteria for 
nuclear power plants and applies to applicants for a design 
certification or combined license under part 52 or a construction 
permit or operating license under part 50. The final rule revises 
appendix S to clarify that the requirements in appendix S also apply to 
applicants for design approvals and manufacturing licenses issued under 
10 CFR part 52. Although former appendix O to part 52 did not 
explicitly require applicants for design approvals to comply with the 
requirements of appendix S, the NRC is requiring design approval 
holders to comply with appendix S to part 50 because the NRC believes 
that the requirements for a design approval should be the same as the 
requirements for a design certification, given that the reviews 
performed by the NRC staff for the two products are essentially 
identical. Finally, appendix S formerly applied to manufacturing 
licenses by virtue of former appendix M to part 52, Section M.1, which 
stated that the provisions in part 50 applicable to construction 
permits apply in context, with respect to matters of radiological 
health and safety, environmental protection, and the common defense and 
security, to manufacturing licenses. Therefore, the Commission is 
revising the General Information section of appendix S to part 50 to 
state that the appendix applies to applicants for a design 
certification, design approval, combined license, or manufacturing 
license under 10 CFR part 52 or a construction permit or operating 
license under 10 CFR part 50. The NRC also made conforming changes to 
the Introduction, paragraph (a) to appendix S to part 50, and added 
definitions for design approval and manufacturing license to Section 
III of appendix S to part 50, to be consistent with the definitions in 
proposed part 52.

E. Change to 10 CFR Part 1

1. Section 1.43, Office of Nuclear Reactor Regulation
    Section 1.43 describes the responsibilities of the Office of 
Nuclear Reactor Regulation (NRR), which includes the development and 
implementation of regulations, policies, programs and procedures for 
the receipt, possession or ownership of source, byproduct and special 
nuclear material that is used or produced at nuclear power plants. 
Inasmuch as power plants may be licensed under part 52 as well as part 
50, Sec.  1.43(a)(2) is revised to clarify that NRR has authority over 
the development and implementation of regulations, policies, programs 
and procedures for the receipt, possession or ownership of source, 
byproduct and special nuclear material that is used or produced at 
nuclear power plants licensed under part 52. In addition, a correction 
has been made to reference part 54, to clarify that NRR has the same 
authority with respect to renewed operating licenses for nuclear power 
plants.

F. Changes to 10 CFR Part 2

1. Section 2.1, Scope
    The statement of scope for part 2 is revised by adding a reference 
to rulemaking and standard design approvals. Previously, the scope 
statement did not mention rulemakings, even though subpart H of part 2 
applied to rulemakings, nor did it mention standard design approvals 
even though the NRC processed applications for design approvals in 
accordance with the procedures in part 2. Accordingly, the change in 
the statement of scope for part 2 correctly reflects the applicability 
of its procedures to both rulemaking and the processing of standard 
design approvals.
2. Section 2.4, Definitions
    The definitions of contested proceeding, license, and licensee, are 
revised in part 2 by adding conforming references, as appropriate, to 
the licensing processes in part 52. The revised definition of contested 
proceeding clarifies that contested proceedings include those involving 
permits, such as early site permits and construction permits. The 
revised definition of license, ensures that early site permits and 
construction permits, as well as part 52 combined licenses and 
manufacturing licenses, are considered to be licenses for purposes of 
part 2. Similarly, the revised definition of licensee ensures that 
holders of early site permits and construction permits, as well as 
combined licenses and manufacturing licenses, are considered to be 
licensees for purposes of part 2.
3. Section 2.100, Scope of Subpart
    This section is revised by adding conforming references to issuance 
of a standard design approval under subpart E of part 52.
4. Section 2.101, Filing of Application
    This section, which governs the procedures for, and the timing and 
content of applications, has been revised in several respects. 
Paragraphs (a)(1), (a)(2), the introductory paragraph of (a)(3), 
paragraph (a)(3)(iii), and paragraph (a)(4) are revised by adding 
conforming references to combined licenses, early site permits, and 
standard design approvals. The Commission notes that the former 
language of Sec.  2.101 already applied to combined licenses, as well 
as early site permits, inasmuch as they are both licenses. Nonetheless, 
consistent with the revisions to the definitions of license and 
licensee, Sec.  2.101 has been revised to explicitly refer to early 
site permits, as applicable.
    In response to public comment on the proposed rule, paragraph 
(a)(5) of Sec.  2.101 and paragraph (a-1) are revised to allow 
applicants for combined licenses--as well as applicants for 
construction permits as provided under this section--to submit 
applications in parts. Paragraph (a)(5) of the final rule allow 
applicants for combined licenses and construction permits to submit an 
application in two parts, with one part containing the environmental 
report required under Sec.  50.30(f) if the application is for a 
construction permit or Sec.  52.80(b) if the application is for a 
combined license. The other part must

[[Page 49412]]

contain the information required by Sec. Sec.  50.34(a) and 50.34a if 
the application is for a construction permit, or Sec.  52.79 and Sec.  
52.80(a) if the application is for a combined license. In addition, the 
part that is filed first must contain the information required by Sec.  
50.33, Sec.  50.34(a)(1) if the application is for a construction 
permit, Sec.  52.79(a)(1) if the application is for a combined license, 
and Sec.  50.37. There are no considerations unique to combined 
licenses which would weigh against allowing a combined license 
applicant to submit a two part application under paragraph (a)(5) of 
Sec.  2.101. Accordingly, the Commission is adopting this change in the 
final rulemaking. Inasmuch as the revisions are to the Commission's 
rules of procedure and practice, the Commission may adopt them in final 
form without further notice and comment, under the rulemaking 
provisions of the APA, 5 U.S.C. 553(b)(A).
    Paragraph (a-1) of Sec.  2.101 allows applicants for combined 
licenses, as well as applicants for construction permits, to submit an 
application in parts to allow for early consideration and a presiding 
officer's partial initial decision on those site suitability matters 
for which the applicant seeks NRC resolution. The provisions governing 
early consideration of site suitability issues in a combined license 
proceeding are set forth in paragraph (a-1)(2). Under this paragraph, a 
combined license application may be submitted in three parts, with the 
first part containing information on the site suitability issues which 
the applicant wishes to have resolved first. The second and third 
parts, which constitute the remainder of the application as described 
in paragraph (a-1)(2)(ii) and (iii), must be submitted during the 
period that the partial decision on part one is effective, viz., 5 
years under new Sec.  2.627 in subpart F of part 2. There are no 
considerations unique to combined licenses which would weigh against 
allowing a combined license applicant to obtain early consideration of 
site suitability issued under paragraph (a-1). As with the change to 
paragraph (a)(5), this revision to paragraph (a-1) constitutes 
revisions to the Commission's rules of procedure and practice. 
Accordingly, the Commission may adopt them in final form without 
further notice and comment, under the rulemaking provisions of the APA, 
5 U.S.C. 553(b)(A).
5. Section 2.102, Administrative Review of Application
    This section is revised by adding conforming references in Sec.  
2.102(a) to applications for early site permits, standard design 
approvals, combined licenses, and manufacturing licenses under part 52. 
Under the revised section, the NRC staff will establish a review 
schedule for an application for these processes, thereby treating the 
applications the same as applications for construction permits or 
operating licenses.
6. Section 2.104, Notice of Hearing
    Section 2.104 sets forth the NRC's requirements regarding 
publication in the Federal Register of notice of hearings. The former 
rule, as well as the proposed part 52 rule, specified the nature of the 
issues that the presiding officer must address in both uncontested and 
contested proceedings. The NRC has decided, based upon its experience 
in noticing hearings in the last decade (in which the Commission's 
notices for more significant proceedings have varied from requirements 
in this section), as well as its consideration of the nature of 
mandatory hearings under Section 189 of the AEA, that much of this 
detailed prescription of the content of the notice of hearing should be 
removed from Sec.  2.104.
    Accordingly, the language of Sec.  2.104 has been considerably 
truncated from the former rule. Paragraph (a) is largely the same as 
former paragraph (a). However, paragraph (b) has been modified to 
specify only the requirements of the notice of hearing which are common 
to all proceedings. All provisions in the former Sec.  2.104 specifying 
the issues to be addressed by the presiding officer are removed in the 
final rule. Inasmuch as this revision is to the NRC's rules of 
procedure and practice, the NRC may adopt them in final form without 
further notice and comment, under the rulemaking provisions of the APA, 
5 U.S.C. 553(b)(A).
    Paragraph (c), (paragraph (m) in the proposed rule, former 
paragraph (e)) requires the NRC to transmit a notice of hearing on an 
initial application of a license for a production or utilization 
facility to an appropriate state official and the chief executive of 
the municipality or county in which the facility is to be located or an 
activity is to be conducted. In addition to the redesignation, 
paragraph (c) is revised to clarify that the notice must be provided 
for applications for early site permits, combined licenses, but not 
manufacturing licenses. Manufacturing licenses are excluded from the 
notification provisions because the NRC is not licensing any particular 
location or site where manufacturing may occur (see discussion of the 
manufacturing license concept).
7. Section 2.105, Notice of Proposed Action
    Section 2.105 contains the NRC's procedures for notices of proposed 
actions where a hearing is not required by law and if the Commission 
has determined that a hearing is in the public interest. Inasmuch as 
amendments to combined licenses and manufacturing licenses do not 
require a mandatory hearing under the AEA, Sec.  2.105(a)(4) is revised 
to clarify that the procedures in Sec.  2.105 also apply to 
applications for amendments of combined licenses and manufacturing 
licenses. Furthermore, because the AEA does not require a mandatory 
hearing for the initial issuance of manufacturing licenses, paragraph 
(a)(13) is added in the final rule to provide for publication of a 
notice of proposed action in connection with an application for a 
manufacturing license under subpart F of part 52.
    Under Sec.  52.103(a), which implements Section 189.a(1)(B)(i) of 
the AEA, the NRC is required to publish in the Federal Register a 
notice of intended operation and an opportunity to request a hearing 
with respect to compliance of the facility with inspections, tests, and 
acceptance criteria in a part 52 combined license. Accordingly, the NRC 
is revising Sec.  2.105 by adding Sec.  2.105(a)(12) which addresses 
the information to be contained in the Federal Register notice required 
by Sec.  52.103(a).
    Because the Commission's authorization for a combined license 
holder to operate under Sec.  52.103 does not constitute ``issuance'' 
of a license or amendment under Sec.  2.106, Sec.  2.105(b)(3) is added 
indicating that the Commission will publish a notice of intended 
operation in the Federal Register that identifies the proposed Agency 
action as making the finding under Sec.  52.103(g). Paragraph 
(b)(3)(iii) of the proposed rule, which would have required that the 
Commission publish, as part of that Federal Register notice, a finding 
that ITAAC have been met, has not been included in the final rule. This 
is because Commission may not have made, at the time of the Federal 
Register notice, the finding that all ITAAC have been met. After 
careful review of the language of Section 189 of the AEA, the 
Commission concludes that the Federal Register notice required by 
Section 189.a(1)(B)(i) need not include a finding that ITAAC have been 
met. Accordingly, Sec.  2.105(b)(3) of the final rule does not include 
a requirement for such a finding to be

[[Page 49413]]

included in the Federal Register notice of intended operation.
8. Section 2.106, Notice of Issuance
    Section 2.106(a) formerly provided that the NRC will publish in the 
Federal Register a notice of issuance of a license or amendment of a 
license where a notice of proposed action has been previously 
published, and notice of amendment of a nuclear power plant license. 
However, that language did not require publication in the Federal 
Register that the Commission has made the finding under Sec.  
52.103(g). Although the AEA does not require publication of a notice of 
the Commission finding under Sec.  52.103, the Commission believes that 
this publication is desirable as a matter of public transparency and 
consistency with past practice of the Federal Register publication of 
Commission action with similar effects (i.e., the issuance of a nuclear 
power plant operating license). Accordingly, Sec.  2.106(a) is revised 
to require Federal Register publication of the Commission finding under 
Sec.  52.103.
    Section 2.106(b)(2) is also revised to set forth the minimum 
requirements for the contents of a Federal Register notice of action, 
e.g., the manner in which copies of the safety analyses, if any, may be 
obtained and examined, and a finding that the prescribed inspections, 
tests, and analyses have been performed and that the acceptance 
criteria prescribed in the combined license have been met, and that the 
license complies with the requirements of the AEA and the NRC's 
regulations. These provisions are the same as the existing requirements 
with respect to notices of issuance for licenses and license 
amendments, but adds the requirements with respect to ITAAC mandated by 
Section 185 of the AEA and part 52. The NRC disagrees with the 
contention raised by the nuclear industry that Section 185 of the AEA 
limits the NRC to a finding of compliance with respect to ITAAC under 
Sec.  52.103(g). Nothing in the legislative history suggests that by 
adopting Section 185 of the AEA, Congress intended to override the 
NRC's long-standing practice of making findings of compliance with the 
Act and the Commission regulations when issuing nuclear power plant 
licenses.
9. Section 2.109, Effect of Timely Renewal Application
    Section 2.109 is revised to add conforming references to a combined 
license under subpart C of part 52. The revised language clarifies that 
an application for a combined license filed no later than 5 years 
before its expiration will not be deemed to have expired until the 
renewal application has been finally determined.
10. Section 2.110, Filing and Administrative Action on Submittals for 
Standard Design Approval or Early Review of Site Suitability Issues
    In a conforming change, paragraphs (a) and (b) of Sec.  2.110 are 
revised to refer to subpart E of part 52 and appendix Q of part 50. 
Paragraph (c) is corrected by adding Sec.  2.110(c)(2) to address the 
procedures applicable to administrative determinations of submittals 
for early review of site suitability issues; formerly, paragraph (c) 
only refers to standard designs.
11. Section 2.111, Prohibition of Sex Discrimination
    This section prohibits sex discrimination against certain persons 
with respect to, inter alia, a license under the AEA. This section is 
revised to include standard design approvals under part 52, and 
petitions for rulemaking, including an application for a design 
certification under part 52.
12. Section 2.202, Orders
    This section is revised by redesignating Sec.  2.202(e) as Sec.  
2.202(e)(1), and adding Sec. Sec.  2.202(e)(2) through (5), to indicate 
the backfitting provisions in part 52 applicable to the various 
licensing processes under part 52. No provisions were deemed necessary 
to address issuance of orders representing backfitting of NRC approvals 
such as standard design approvals.
13. Section 2.309, Hearing Requests, Petitions To Intervene, 
Requirements for Standing, and Contentions
    Section 2.309, which establishes the NRC requirements governing 
requests for hearing and petitions to intervene--including submission 
of contentions--is revised to add three conforming and clarifying 
changes. First, paragraph (a) is revised, consistent with a change to 
Sec.  52.103(c), to make clear that in a proceeding under Sec.  52.103, 
the Commission itself will act as the presiding officer, will consider 
and act upon a request for a hearing under Sec.  52.103, and will also 
determine whether a period of interim operation may be permitted, as 
provided for under Section 189.a(1)(B)(iii) of the AEA. Inasmuch as the 
Commission itself will make the contention admission determination, 
there should be no need for further Commission review of the contention 
admission decision at the end of the hearing.
    Second, paragraph (f)(1)(i) has been revised to make clear that 
contentions in Sec.  52.103(b) requests for hearing must raise issues 
in law or fact with respect to whether one or more of the acceptance 
criteria in a combined license have not been, or will not be met, and 
that the specific operational consequences of nonconformance would be 
contrary to providing reasonable assurance of adequate protection to 
public health and safety. This is consistent with the statutory 
limitation on the scope of a hearing in Section 189.a(1)(B)(ii) of the 
AEA.
    Third, a new paragraph (f)(1)(vii) has been added to set forth the 
specific requirements for a contention under Section 189.a(1)(B)(ii) 
and 10 CFR 52.103(b). The new paragraph provides that, in a request for 
hearing under Sec.  52.103(b), the information submitted must be 
sufficient and include supporting information showing, prima facie, 
that: (i) One or more of the acceptance criteria in a combined license 
have not been, or will not be met, and (ii) the specific operational 
consequences of nonconformance would be contrary to providing 
reasonable assurance of adequate protection to public health and 
safety. The revision also makes clear that the information in support 
of a contention that an acceptance criterion is not, or will not be 
met, must identify the specific portions of the Sec.  52.99(c) report 
which is inaccurate, incorrect, or incomplete. The terms, 
``inaccurate,'' and ``incorrect,'' while somewhat overlapping, are 
intended to cover a broad range of situations. ``Inaccurate'' is 
intended to address a situation where information contained in, 
referenced by, or relied upon (either explicitly or implicitly) as a 
supporting basis for a representation in a Sec.  52.99(c) report, is 
erroneous (e.g., an erroneous computation, or inaccurate data entry of 
a test result). By contrast, ``incorrect'' focuses on a situation where 
such information is the result of a cognitive inadequacy or failure 
(even if, under the circumstances, the inadequacy or failure is 
justifiable), poor judgement, negligence, or deliberate wrongdoing. By 
``incomplete,'' the NRC means that the report does not provide the 
information which must be provided in the report as required by Sec.  
52.99. Furthermore, if the requestor contends that the Sec.  52.99(c) 
report is incomplete, and the requestor contends that the incomplete 
portion prevents the requestor from making the necessary prima facie 
showing, then the requestor must also, as provided by Sec.  
2.309(f)(1)(vii), explain why the deficiency (viz., the incomplete 
nature of the report) prevents the requestor from making the necessary 
prima facie

[[Page 49414]]

showing. The NRC believes that these changes to Sec.  2.309 will help 
ensure that any 10 CFR 52.103 hearing on whether the acceptance 
criteria in ITAAC have been, or will be met, is focused only on the 
matters which Congress intended to be adjudicated at this juncture, as 
directed by Section 189.a.(1)(B) of the AEA.
    Fourth, paragraph (g) is revised to conform with the change in: (i) 
10 CFR 52.103(c), which now provides that the Commission will act as 
the presiding officer in determining whether to grant or deny a request 
for hearing with respect to whether acceptance criteria in ITAAC have 
been or will be met; and (ii) 10 CFR 2.310, which provides that the 
Commission, acting as the presiding officer, will determine the hearing 
procedures to be utilized in a Sec.  52.103 hearing. Under the revised 
paragraph (g), a request for hearing under Sec.  52.103 shall not 
address the hearing procedures to be utilized.
    Fifth, paragraph (h) is revised to prohibit a reply by a requestor 
for a hearing under Sec.  52.103. The NRC believes that Congress 
intended the Commission's initial decision to grant the hearing and the 
determination of interim operation to be based upon the same set of 
information. The Commission's view is based upon the language of 
Section 189.a.(1)(B)(iii), which refers to a Commission determination 
to allow a period of interim operation based upon the ``petitioner's 
prima facie showing and any answers thereto. * * *'' That the statute 
only refers to a request and the answers thereto suggests that Congress 
did not intend that a reply was necessary. This is understandable given 
Congress'' explicit direction that any hearing granted be completed 
``to the maximum possible extent * * * within 180 days of the 
publication of the notice [of opportunity to request a hearing under 
Section 189.a(10)(B)(i)] or the anticipated date for initial loading of 
fuel into the reactor, whichever is later.'' While the relevant 
statutory language literally applies only to the Commission 
determination of interim operation, the NRC believes that as a matter 
of logic, Congress must have intended that it would also apply to the 
threshold question of granting or denying the hearing request. It is 
unclear why Congress would allow more information to be considered in 
the threshold question of the hearing request, but limit the 
information to be considered in the interim operation determination. 
The NRC concludes that it would be closer to Congress' intention to 
prohibit a requestor for a Sec.  52.103 hearing from replying to any 
answers filed by the applicant and/or the NRC staff.
    Finally, in a conforming change associated with the revision to 
Sec.  52.103(c), paragraph (i) is revised to prohibit any ``appeal'' 
under Sec.  2.311 of a Commission decision to grant or deny a request 
for hearing. Inasmuch as the Commission is acting as a presiding 
officer, there can be no further ``appeal'' to a higher agency 
decisionmaker. Moreover, an adversely affected party may seek 
reconsideration of the Commission's decision under Sec.  2.345, and it 
would be duplicative to afford an adversely-affected party a Sec.  
2.311 ``review'' right in addition to the opportunity to seek 
reconsideration under Sec.  2.345.
    Inasmuch as these revisions are to the NRC's rules of procedure and 
practice, the NRC may adopt them in final form without further notice 
and comment, under the rulemaking provisions of the APA, 5 U.S.C. 
553(b)(A).
14. Section 2.310, Selection of Hearing Procedures
    Section 2.310 is revised, in part to conform with the change in 10 
CFR 52.103(c), which now provides that the Commission will act as the 
presiding officer in determining whether to grant or deny a request for 
hearing with respect to whether acceptance criteria in ITAAC have been 
or will be met. The revised Sec.  2.310 now provides that the 
Commission will determine the hearing procedures to be utilized in its 
determination on a hearing request under Sec.  52.103, as well as the 
hearing procedures to be utilized in resolving admitted contentions 
under Sec.  52.103(c) and (g).\8\
---------------------------------------------------------------------------

    \8\ The NRC notes that 10 CFR 2.309 does not apply, by its 
terms, to petitions to modify the terms and conditions of a combined 
license under 10 CFR 52.103(f). Such petitions must meet the 
requirements of 10 CFR 2.206.
---------------------------------------------------------------------------

    Inasmuch as this revision is to the NRC's rules of procedure and 
practice, the NRC may adopt it in final form without further notice and 
comment, under the rulemaking provisions of the APA, 5 U.S.C. 
553(b)(A).
15. Section 2.340, Initial Decision in Certain Contested Proceedings; 
Immediate Effectiveness of Initial Decisions; Issuance of 
Authorizations, Permits, and Licenses
    Section 2.340 addresses several different matters relating to the 
presiding officer's initial decision and its effect. The final rule 
reorganizes the paragraphs in this section in order to better 
distinguish among these matters, reserves paragraphs (g) and (h) for 
future use by the Commission, and makes substantial changes to these 
matters addressed in this section, as discussed below. These changes 
are to the NRC's rules of procedure and practice, and the NRC is 
adopting the changes in final form without further notice and comment, 
under the rulemaking provisions of the APA, 5 U.S.C. 5, 553(b)(A).

Scope of Presiding Officer's Initial Decision

    Formerly, paragraph (a) limited the scope of the presiding 
officer's findings and conclusions of law in initial decisions in 
contested proceedings for production or utilization facility operating 
licenses to matters put into controversy by the parties. Matters not 
put into controversy by the parties could only be examined by the 
presiding officer by direction of the Commission, either on its own 
initiative or upon the presiding officer's referral of the matter to 
the Commission. In a conforming change, a new paragraph (b) is added to 
apply the limitation in contested hearings under Sec.  52.103(g) with 
respect to whether the acceptance criteria in a combined license ITAAC 
have been, or will be met.
    The Sec.  2.340(a) limitation did not apply to a contested 
utilization facility construction permit proceeding. Although the 
statement of considerations for the original rulemaking adopting this 
limitation (in former Sec.  2.760a) does not directly address the basis 
for this limitation (see January 17, 1975; 40 FR 2973), the underlying 
rationale may be gleaned from the Commission's order in Consolidated 
Edison Co. of New York (Indian Point Nuclear Generating Unit 3), 8 AEC 
7 (1974) which engendered the rulemaking. In explaining that the 
Licensing Board has no obligation at the operating license stage to 
inquire into matters which parties have not raised and the Licensing 
Board itself has no reason to inquire, the Commission stated:

    To have a Licensing Board engage in an idle exercise examining 
issues just for the sake of examination--when the parties have not 
raised such matters, and the Board is satisfied that there is 
nothing to inquire about--would serve no useful purpose. This is 
particularly true since an operating license proceeding is not to be 
used to rehash issues already well ventilated and resolved at the 
construction permit stage. Alabama Power Co. (Joseph M. Farley 
Nuclear Plant, Units 1 and 2), CLI-74-12 (RAI-74-3-203).

    Id. at 8. Thus, the limitation was based, in part, upon the broader 
scope of inquiry for the presiding officer at construction permit 
stage, which is a ``mandatory hearing'' required by

[[Page 49415]]

Section 189.a(1)(A). This rationale continues to apply today, and 
consequently the NRC does not propose to alter the NRC's practice by 
extending the Sec.  2.340(a)/Sec.  2.760a limitation to construction 
permit (including early site permit) proceedings. Nor should the Sec.  
2.340(a)/Sec.  2.760a limitation apply in a part 52 combined license 
proceeding with respect to matters that would otherwise be addressed 
and resolved in a construction permit issuance proceeding.
    The final part 52 rule includes several changes to implement the 
NRC's conclusions in this regard. Section 2.340(a) is revised to 
provide that the presiding officer in a contested operating license 
proceeding shall make findings of fact and conclusions of law to, inter 
alia, those matters put into controversy or otherwise directed by the 
Commission. Paragraphs (b), (c), and (d) are revised to address the 
scope of the presiding officer's initial decision in a combined license 
proceeding (including a renewal or amendment proceeding), in a 
proceeding under Sec.  52.103(g), and in a manufacturing license 
proceeding (including a renewal or amendment proceeding).
    As discussed previously, the former Sec.  2.340(a)/Sec.  2.760a 
limitation applied only to operating license proceedings, and did not 
apply to other contested proceedings which do not require a ``mandatory 
hearing,'' which includes most materials licensing proceedings (with 
the notable exception of the licensing of a uranium enrichment 
facility). The statement of consideration in this document merely 
states that the rule codifies the Commission's Indian Point decision. 
(see January 17, 1975; 40 FR 2973 (first column)). Inasmuch as the 
Indian Point proceeding involved a utilization facility license, it is 
likely that the Commission simply did not consider as part of the 
rulemaking the possibility of applying the limitation to non-production 
or utilization facility proceedings, as opposed to making a deliberate 
decision not to apply the limitation to non-production or utilization 
facility proceedings. Currently, the NRC believes that with 30 
additional years of hearing experience, there is no practical, 
compelling policy-based, or legal reason why the Sec.  2.340(a) 
limitation should not be extended to non-production or utilization 
facility proceedings. Accordingly, the NRC is revising Sec.  2.340 by 
adding a new paragraph (e), which extends the existing limitation on 
the presiding officer's initial decision in contested proceedings to 
all other proceedings not covered by paragraphs (a) or (b) of Sec.  
2.340. Although this change is not related to the part 52 rulemaking 
effort, the NRC is adopting this change as part of the part 52 final 
rule to ensure that stakeholders understand the provisions of Sec.  
2.340 as an integrated whole.

Immediate Effectiveness of Presiding Officer's Initial Decision in 
Production and Utilization Facility Proceedings

    The remainder of former Sec.  2.340 was an amalgam of the 
Commission's original rule (10 CFR 2.764 \9\) a presiding officer's 
initial decision in certain proceedings was immediately effective upon 
issuance, combined with newer provisions--first adopted in 1979 and 
modified in 1981--which suspended the immediate effectiveness rule. The 
``automatic stay'' provisions were adopted following the accident at 
TMI-2, in order to provide for the Commission's direct involvement in 
the issuance of nuclear power plant licenses. The Commission first 
issued an Interim Statement of Policy and Procedure in October 1979, 
which first noted that the TMI-2 accident was being investigated by the 
NRC and may result in ``significant changes in the Commission's 
regulatory policy and in the procedures it employs to license nuclear 
power facilities.'' The Policy Statement then indicated that ``new 
construction permits, limited work authorizations, or operating 
licenses for any nuclear power plants shall be issued only after action 
of the Commission itself.'' (See October 10, 1979; 44 FR 58559.) Soon 
thereafter, on November 9, 1979 (44 FR 65049), the NRC issued a 
Suspension of Sec.  2.764 and Statement of Policy on the Conduct of 
Adjudicatory Proceedings. As part of this final rulemaking, the NRC 
adopted a new appendix B to part 2 addressing the suspension of 
immediate effectiveness provisions in Sec.  2.764, and providing for 
both Atomic Safety and Licensing Appeal Board review and Commission 
review of the presiding officer's initial decision.
---------------------------------------------------------------------------

    \9\ 31 FR 12774 (September 30, 1966).
---------------------------------------------------------------------------

    On May 28, 1981 (46 FR 28627), the NRC issued a final rule which 
removed the need for the Appeal Board review of a presiding officer's 
initial decision, but retained a minimum 60-day period for Commission 
review. The final rule was almost immediately amended to exclude from 
Commission review presiding officer decisions authorizing fuel load and 
low-power testing (September 30, 1981; 46 FR 47764). In 2004, the 
provisions in Sec.  2.764 were transferred without substantive change 
to a new Sec.  2.340 as part of the general revision to 10 CFR part 2 
(January 14, 2004; 69 FR 2182).
    While the NRC's 1979 and 1981 rulemakings were justified in light 
of the circumstances at that time, other factors now lead the NRC to 
believe that the oversight provisions adopted in 1981 are no longer 
necessary or desirable. In the 25 years since the adoption of the 1981 
provisions, the NRC's regulatory framework and requirements for nuclear 
power plants has evolved and strengthened. The NRC's technical 
requirements for nuclear power reactors were substantially augmented in 
the years immediately following the TMI accident, and thereafter have 
evolved to reflect lessons learned, new information, and the increasing 
acceptance of risk-informed methodologies. Similarly, the NRC's 
oversight of nuclear power plants has evolved to reflect lessons 
learned, new information, and the maturation of risk assessment 
methodologies. Thus, the NRC believes its regulations may be revised to 
remove the regulatory requirement for direct Commission involvement in 
all production and utilization licensing proceedings. The Commission's 
words in the May 1981 final rulemaking apply with more force today:

    This amendment does not compromise the Commission's commitment 
to the protection of public health and safety or to a fair hearing 
process. Thorough technical safety reviews of license applications 
by the NRC staff and the Advisory Committee on Reactor Safeguards, 
the availability of public hearings on license applications, and the 
Commission's inherent supervisory authority form the basis of the 
network of procedural safeguards intended to implement this 
commitment to a fair decision process and public health and safety. 
(May 28, 1981; 46 FR 28628 first column)

    The NRC's commitment remains unchanged, and the NRC's safeguards 
have been strengthened since that time, for example, by refocusing the 
regulatory process to include considerations of risk. In addition, the 
NRC's rules of practice in part 2 provide several procedural safeguards 
within the NRC's administrative process, including: (1) A petition for 
presiding officer reconsideration under Sec.  2.345; (2) a petition for 
Commission review under Sec.  2.341; and (3) a motion for a stay with 
the presiding officer or the Commission under Sec.  2.342.
    By removing the ``automatic stay'' provisions in former Sec.  
2.340(f) and (g), the NRC's administrative process will be completed in 
less time, thereby benefitting all parties from the reduction in 
litigation resources without compromising the fairness of the overall 
hearing process. Faster completion of

[[Page 49416]]

the adjudication will also enable aggrieved parties to more quickly 
seek relief via an appeal to a U.S. Circuit Court of Appeals. The NRC 
believes that Congress intends the Commission to conduct fair, but 
efficient, hearings with respect to licensing, and to remove 
unnecessary hearing procedures which do not contribute to such a 
hearing process. This is evidenced by Section 189 of the AEA, as 
amended by the Energy Policy Act of 1992, which directs the Commission 
to issue, ``to the maximum possible extent,'' a final decision on 
issues raised with respect to acceptance criteria by the anticipated 
date for initial loading of fuel. The Commission concludes that the 
changes to Sec.  2.340 are consistent with applicable law, and will 
provide tangible benefits to all parties in NRC adjudications.

Immediate Effectiveness of Presiding Officer's Initial Decision in 
Other, Non-Production or Utilization Facility Proceedings

    As noted previously, the 1981 final rulemaking provided for an 
``automatic stay'' to provide for direct Commission involvement in the 
issuance of nuclear power plant licenses. Since that time, the NRC has 
extended the ``automatic stay'' provisions in Sec.  2.340 to other 
licensing contexts, such as independent spent fuel storage facilities 
(ISFSIs) at sites away from nuclear power reactors, monitored 
retrievable storage (MRO) licenses, and provided for a parallel 
provision in 10 CFR part 61 for low-level waste (LLW) facilities, see 
10 CFR 2.1211. The NRC did not explain the basis for requiring direct 
Commission involvement in the issuance of a part 61 LLW license (see 47 
FR 57446; December 27, 1982), although one could surmise from the 
timing of the rulemaking that the factors underlying the 1981 
rulemakings also were the basis for the 1982 rulemaking's provision 
providing for direct Commission involvement in part 61 license 
issuances. The NRC's original intent in requiring direct Commission 
involvement in the issuance of specific ISFSI licenses and a MRS 
license was the lack of regulatory experience (see, e.g., 60 FR 20879 
and 20883; April 28, 1995), and, therefore, is somewhat different from 
the motivating factors for the 1981 rulemakings. In any event, the NRC 
now has had the benefit of experience in licensing a specific ISFSI, as 
well as several specific ISFSIs located at reactor sites. Thus, the NRC 
has come to a recognition that the safety, security and regulatory 
issues associated with these licenses are of less complexity than those 
associated with nuclear power plants, and that the NRC has greater time 
to respond to potentially adverse situations. Compare 46 FR 47764, 
47765 (issuance of licenses for activities involving minimal risk to 
public health and safety, and greater time to take corrective action, 
do not require Commission involvement). Furthermore, the Commission 
possesses general supervisory authority over the NRC staff and may 
direct the staff to keep the Commission appraised of licensing status 
and issues for such licenses. Accordingly, the NRC concludes that there 
is little regulatory benefit to be provided by a rule requiring direct 
Commission involvement in the issuance of these licenses and that the 
provisions in Sec.  2.340 providing for such involvement should also be 
removed as part of this streamlining of the regulatory process.

Issuances of Authorizations, Permits, Licenses, and Sec.  52.103(g) 
Findings

    Former paragraph (c) of Sec.  2.340 provided that the appropriate 
staff Office Director was authorized to issue certain delineated 
licenses, including license amendments, construction permits, and 
construction authorizations, within 10 days from the date of issuance 
of an initial decision. The former language could be erroneously read 
as requiring the Director to issue a license following an initial 
decision on a contested matter, even if other issues not contested had 
yet to be resolved by the NRC staff. In addition, paragraph (c) did not 
address the issuance of a finding under Sec.  52.103(g). To resolve 
these concerns, new paragraphs (i), (j), and (k) are added to Sec.  
2.340. In general, each paragraph authorizes the appropriate staff 
Office Director to issue the delineated license, permit, authorization 
or finding within 10 days from the issuance of an initial decision, if 
all other safety and environmental findings necessary for issuance of 
the license, permit, authorization or finding have been made, 
notwithstanding the pendency of various petitions or motions for 
reconsideration, review or stay before the presiding officer or the 
Commission.
    Paragraph (i) authorizes the Director of Nuclear Reactor Regulation 
(NRR) or the Director of the Office of New Reactors (NRO), as 
appropriate, to issue nuclear power plant licenses, including 
amendments, permits and authorizations, within 10 days of the initial 
decision. Paragraph (j) authorizes the Commission or the appropriate 
staff Office Director to make the finding under 10 CFR 52.103(g) that 
the acceptance criteria in a combined license have been met. Finally, 
paragraph (k) addresses the issuance of other licenses that are issued 
by the Director of Nuclear Material Safety and Safeguards (NMSS). 
Typical licenses of this type would be materials licenses for, inter 
alia, medical uses, well logging, radiography, irradiators, and 
research.
16. Section 2.341, Review of Decisions and Actions of a Presiding 
Officer
    This section addresses requests for review and appeals to the 
Commission from a presiding officer's decision or actions in a hearing. 
In a conforming change associated with the revision to Sec.  52.103(c), 
paragraph (a)(1) of Sec.  2.341 is revised to explicitly prohibit a 
party from seeking a ``review'' or an ``appeal'' of the Commission's 
determination to allow a period of interim operation under Sec.  
52.103(c), separate from and in addition to a request for 
reconsideration under Sec.  2.345. Inasmuch as the Commission is acting 
as the presiding officer in the Sec.  52.103(c) determination, there 
can be no further ``appeal'' to a higher agency decisionmaker. 
Moreover, it would be duplicative to afford a Sec.  2.341 ``review'' or 
``appeal'' right in addition to the opportunity to seek reconsideration 
under Sec.  2.345.
    Inasmuch as this revision is to the NRC's rules of procedure and 
practice, the NRC may adopt it in final form without further notice and 
comment, under the rulemaking provisions of the APA, 5 U.S.C. 
553(b)(A).
17. Section 2.347, Ex Parte Communications
    Section 2.347, which sets forth the NRC's requirements governing ex 
parte communications with the Commission and its adjudicatory 
employees, is revised in this final rule to address several problems 
with the current rule.
    First, Sec.  2.347 is revised to make clear that ex parte 
communication restrictions are not applicable in uncontested 
proceedings. The APA requirements in 5 U.S.C. 557(d)(1) governing ex 
parte communications apply only to communications ``relevant to the 
merits of the proceeding * * *,'' which are made to and from 
``interested persons outside the agency.'' In an uncontested 
proceeding, there are no ``interested persons outside the agency,'' in 
the sense that there are no persons for which a hearing has been 
requested or intervention in a hearing has been granted. Hence, ex 
parte communication restrictions do not apply. Moreover, as the NRC has 
stated in the 2004 rulemaking revising 10 CFR part 2, Section 189 of 
the AEA does not require NRC hearings under that section to be ``on the 
record.'' See 69 FR 2183-2185, 2192-2193 (January 14, 2004).

[[Page 49417]]

Accordingly, Sec.  2.347 is revised to explicitly provide that ex parte 
restrictions do not apply to uncontested proceedings.
    Second, Sec.  2.347 is revised to exclude undisputed (i.e., 
uncontested) issues in contested proceedings from the application of ex 
parte restrictions. It makes little sense to require the Commission to 
inform parties to the proceeding of the Commission's communications 
with the applicant or licensee on matters for which those parties have 
not been admitted (and may have no interest in litigating). In 
addition, the NRC believes that uncontested matters are not, for 
purposes of applying the ex parte limitations in Section 557(d)(1) of 
the APA, either ``a fact in issue'' or a matter which is ``relevant to 
the merits of the [contested] proceeding.'' The NRC also believes, as 
stated above, that the ex parte limitations in Section 557(d) of the 
APA do not apply to NRC proceedings, and therefore the application of 
ex parte restrictions in NRC proceedings is a matter of discretion on 
the part of the NRC. The NRC believes that it is appropriate to exclude 
undisputed issues from the application of ex parte limitations in 
contested proceedings, inasmuch as there appears to be little, if any, 
public confidence benefit from extending ex parte limitations to 
``undisputed issues,'' i.e., matters which have not been raised by any 
party in the proceeding.
    Finally, Sec.  2.347 is also revised to make clear that ex parte 
restrictions apply to matters which are the subject of a presiding 
officer referral to the Commission under Sec.  2.340(a), and the 
presiding officer's examination of that matter following Commission 
approval under Sec.  2.340(a) (referred to as ``sua sponte'' issues at 
53 FR 10361; March 31, 1988). The application of ex parte restrictions 
to Sec.  2.340(a) ``sua sponte'' matters does not represent a change in 
NRC practice, cf., 53 FR 10360, 10361 (first and second column) (March 
31, 1988). Nonetheless, upon further reflection the NRC believes it is 
inaccurate to treat Sec.  2.340(a) ``sua sponte'' matters as a 
``disputed issue'' for purposes of applying Sec.  2.347. Accordingly, 
the NRC is revising Sec.  2.347 to explicitly state that consideration 
of Sec.  2.340(a) ``sua sponte'' matters are to be subject to ex parte 
restrictions.
    Inasmuch as these Sec.  2.347 revisions are to the NRC's rules of 
procedure and practice, the NRC may adopt them in final form without 
further notice and comment under the rulemaking provisions of the APA, 
5 U.S.C. 553(b)(A).
18. Section 2.348, Separation of Functions
    This section sets forth the NRC's requirements governing separation 
of functions of the Commission and its adjudicatory employees when 
acting in their adjudicatory capacity. The rule prohibits an NRC 
officer or employee engaged in the performance of investigative or 
litigation function in that proceeding from participating in or 
advising the Commission and its adjudicatory employees about ``any 
disputed issue in that proceeding * * *,'' with certain delineated 
exceptions (10 CFR 2.348(a)).
    The NRC believes that there are two problems with the current 
language. First, the rule does not explicitly state that in an 
uncontested proceeding, separation of functions does not apply. More 
importantly, the rule applies separation of functions in circumstances 
where it is not required by Section 554(d), viz., determinations 
involving initial licenses (5 U.S.C. 554(d)(2)(A) of the APA). The NRC 
recognizes that public confidence considerations may favor compliance 
with separation of functions restrictions in contested initial 
licensing proceedings. However, there is little apparent value in 
applying separation of functions to the NRC's resolution of uncontested 
(i.e., ``undisputed'') issues in contested proceedings. The NRC also 
notes that (as in the case of the APA restrictions on ex parte 
communications) the APA separation of functions requirements apply only 
to adjudications which are required to be ``on the record.'' As 
discussed above, NRC licensing proceedings are not required by the AEA 
or any other statute to be on the record. Thus, there is no legal 
requirement to apply separation of functions in initial licensing 
proceedings. Although the NRC could voluntarily, as a matter of 
discretion, apply separation of functions in circumstances where it is 
not required by law, such a course of action seems unjustified in view 
of the lack of a clear public confidence benefit--which is the primary 
objective of separation of functions restrictions. For these reasons, 
the final part 52 rule revises Sec.  2.348 to make explicit that 
separation of functions requirements do not apply to either uncontested 
proceedings, or to an undisputed issue in contested initial licensing 
proceedings.
    Section 2.348 is also revised to make clear that separation of 
functions applies to matters which are the subject of a presiding 
officer referral to the Commission under Sec.  2.340(a), and the 
presiding officer's examination of that matter following Commission 
approval under Sec.  2.340(a). As with the change in Sec.  2.347 with 
respect to ex parte restrictions, this change in Sec.  2.348 does not 
depart from the NRC's current practice of applying separation of 
function restrictions to ``sua sponte'' matters under Sec.  2.340(a). 
The NRC believes that it is more accurate to explicitly state that sua 
sponte matters under Sec.  2.340(a) are subject to separation of 
functions restrictions, rather than characterizing such matters as 
``disputed issues.''
    Inasmuch as these Sec.  2.348 revisions are to the NRC's rules of 
procedure and practice, the NRC may adopt them in final form without 
further notice and comment under the rulemaking provisions of the APA, 
5 U.S.C. 553(b)(A).
19. Section 2.390, Public Inspections, Exemptions, Requests for 
Withholding
    Section 2.390 governs the availability of NRC records and documents 
regarding a license, permit or order, and implements the Freedom of 
Information Act (FOIA). This section is revised to make clear that its 
provisions also applies to NRC records and documents regarding standard 
design approvals under part 52.
20. Subpart D--Additional Procedures Applicable to Proceedings for the 
Issuance of Licenses To Construct and/or Operate for Nuclear Power 
Plants of Identical Design at Multiple Sites
    Formerly, subpart D of part 2 set forth the Commission's 
administrative and hearing procedures for proceedings for issuance of 
construction permits and operating licenses under part 52 for nuclear 
power plants of ``duplicate'' design at multiple sites. The 
requirements governing the content of such applications and the 
technical consideration of such applications are set forth in 10 CFR 
part 50, appendix N, which was ``transferred'' to part 52 as part of 
the 1989 part 52 rulemaking. However, the 1989 rulemaking did not 
remove appendix N from part 50, nor did the NRC make conforming changes 
to appendix N in part 52 to make its provisions applicable to combined 
licenses under subpart C of part 52. As discussed elsewhere, in the 
March 2006 proposed rule the NRC proposed deleting appendix N in part 
52, and retaining these provisions in part 50. Although no comment was 
received on this proposal, the NRC has decided to withdraw its proposal 
to delete appendix N in part 52. Instead, the NRC is revising appendix 
N in part 52 to apply only to proceedings for combined licenses under 
subpart C of part 52

[[Page 49418]]

(appendix N in part 50 will continue to address proceedings for 
construction permits and operating licenses under that part).
    To reflect the expanded scope of appendix N of part 52 and to 
ensure that all of the NRC's regulations use consistent terminology, 
the NRC is revising subpart D of part 2 as part of this final 
rulemaking. Inasmuch as the changes to the provisions in subpart D 
constitute revisions to the NRC's rules of procedure and practice, the 
NRC may adopt them in final form without further notice and comment, 
under the rulemaking provisions of the APA, 5 U.S.C. 553(b)(A).
21. Section 2.400, Scope of Subpart
    This section is revised to refer to both appendix N of both part 50 
and part 52, in order to reflect the Commission's determination that 
the appendix should be retained in both parts, and that the procedures 
in the appendices (both of which refer to this subpart) should apply to 
applications for construction permits, operating reactors, and combined 
licenses of identical design. In addition, Sec.  2.400 is revised to 
use the term ``identical design,'' instead of the former ``essentially 
the same design,'' so that subpart D and appendix N of part 50 and part 
52 use identical terminology.
22. Section 2.401, Notice of Hearing on Construction Permit or Combined 
License Applications Pursuant to Appendix N of 10 CFR Parts 50 or 52
    Paragraph (a) of Sec.  2.401 is revised to indicate that notices of 
hearing will be published for both construction permits under part 50 
and combined licenses under part 52. Notices of the issuance of 
operating licenses is addressed, as was the case under the former 
provisions of subpart D, in Sec.  2.403. No other substantive changes 
are intended by this revision. Paragraph (b) remains unchanged.
23. Section 2.402, Separate Hearings on Separate Issues; Consolidation 
of Proceedings
    Both paragraphs of this section are revised to refer to 
applications under part 50 and part 52. No other substantive changes 
are intended by this revision.
24. Section 2.403, Notice of Proposed Action on Applications for 
Operating Licenses Pursuant to Appendix N of 10 CFR Part 50
    This section is revised to refer to operating licenses issued under 
part 50, rather than part 52. This reflects the Commission's 
determination that appendix N of part 50 applies to construction 
permits and operating licenses, whereas appendix N of part 52 applies 
to combined licenses under subpart C of part 52.
25. Section 2.404, Hearings on Applications for Operating Licenses 
Pursuant to Appendix N of 10 CFR Part 50
    This section is revised to make clarifying changes by adding 
references to a presiding officer, correctly referring to the Chief 
Administrative Judge, and removing a reference to the atomic safety and 
licensing board. No substantive changes are intended by this revision.
26. Section 2.405, Initial Decisions in Consolidated Hearings
    This section is revised by requiring the presiding officer to issue 
a separate partial initial decision on the common design. Section 2.405 
is also revised by clarifying that the presiding officer may, if 
otherwise determined under the consolidation provisions of Sec.  
2.317(b), issue a consolidated decision for those proceedings. No other 
substantive changes are intended by this revision.
27. Section 2.406, Finality of Decisions on Separate Issues
    This section is revised to refer to both appendix N of both part 50 
and part 52. No other substantive changes are intended by this 
revision.
28. Section 2.407, Applicability of Other Sections
    This section is revised to correctly reference subparts C, L, and N 
of part 2. No other substantive changes are intended by this revision.
29. Section 2.500, Scope of Subpart
    This section is revised by adding a conforming reference to subpart 
F of part 52 on manufacturing licenses.
30. Section 2.501, Notice of Hearing on Application Under Subpart F of 
Part 52 for a License To Manufacture Nuclear Power Reactors
    This section is revised by adding a conforming reference to subpart 
F of part 52 on manufacturing licenses. In addition, paragraph (b) of 
this section is revised by removing the detailed requirements governing 
the content of the notice of hearing published in the Federal Register, 
and instead referencing proposed Sec.  2.104(f). As previously 
discussed, the Commission is consolidating in Sec.  2.104 the 
requirements governing the content of a notice of hearing with respect 
to part 52 licensing and regulatory approval processes (with the 
exception of standard design certifications, which are addressed in 
subpart H of part 2).
31. Sections 2.502, 2.503, and 2.504
    The text of these sections is removed, and their places are 
reserved in the final rule, because the matters addressed in these 
sections, regarding finality and the referencing of a manufactured 
reactor in a combined license, are addressed with greater specificity 
in the revisions to subpart F of part 52.
32. Subpart F, Additional Procedures Applicable to Early Partial 
Decisions on Site Suitability Issues in Connection with an Application 
for a Construction Permit or Combined License for Certain Utilization 
Facilities
    Subpart F provides special procedures for the acceptance, 
docketing, administrative consideration, the conduct of hearings, and 
the presiding officer's issuance of a partial initial decision in 
licensing proceedings where there is early submittal of site 
suitability information in connection with an application for a 
construction permit or operating license, as described in Sec.  
2.101(a-1). As discussed earlier, the NRC has revised Sec.  2.101(a-1) 
to allow applicants for combined licenses under part 52, as well as 
applicants for construction permits under part 50, to submit their 
applications in two parts, and to allow for early consideration and 
presiding officer's partial initial decision on those site suitability 
matters for which the applicant seeks early resolution in accordance 
with subpart F of part 2.
    The NRC has reorganized subpart F in an attempt to improve its 
usability (the reorganization is reflected in the provisions of Sec.  
2.600, Scope of subpart). Requirements applicable to partial decisions 
in construction permit proceedings continue to be addressed in 
Sec. Sec.  2.602 through 2.606; a new subheading is added before Sec.  
2.602 to reflect the subject matter of these sections. The new 
requirements applicable to partial decisions in combined license 
proceedings are in Sec. Sec.  2.621 through 2.629; a new subheading is 
also added before Sec.  2.621 to reflect the subject matter covered by 
these sections. Section 2.629, which has no analogous provisions in 
Sec. Sec.  2.602 through 2.606, is added by the NRC to ensure that the 
finality of a presiding officer's partial initial decision in a 
combined license proceeding is clearly addressed using regulatory 
language similar to that used in the finality provisions in part 52, 
e.g., Sec. Sec.  52.39, 52.63, 52.98.

[[Page 49419]]

    Section 2.601 is revised to correctly list subparts A, C, G, L, and 
N of part 2 as subparts which are either applicable to or may be 
utilized in proceedings under subpart F.
33. Section 2.800, Scope and Applicability
    Subpart B of part 52 sets out the requirements applicable to 
Commission issuance of regulations granting standard design 
certification for nuclear power facilities. Standard design 
certifications are approved through a rulemaking proceeding, and, in 
concept, the applicant for a design certification may be considered as 
a petitioner for rulemaking. However, subpart H of part 2, which sets 
forth the Commission's procedures governing rulemaking, including 
petitions for rulemaking, did not specifically address design 
certification. Furthermore, based upon the Commission's experience with 
three final design certification rules and a proposed design 
certification rule, it is clear that some of the procedural 
requirements applicable to petitions for rulemaking are not well-suited 
to the administrative process for determining a design certification 
application, e.g., the existing prohibition against pre-application 
consultation with the NRC. These consultations between potential 
license applicants and the NRC staff are not currently prohibited and 
indeed are encouraged by the Commission to enhance NRC resource 
planning and to facilitate early identification and resolution of 
technical and regulatory issues. An application for design 
certification is more like a license application than a traditional 
petition for rulemaking, and the current prohibition against pre-
application consulting appears to be inconsistent with the Commission's 
strategic objectives of safety, effectiveness, and management 
excellence. The Commission also believes, based upon its experience, 
that administrative provisions ordinarily applied in the context of 
licensing (e.g., docketing and acceptance review, denial of application 
for failure to supply information), should also be available for 
application as appropriate in its determination of design certification 
applications.
    For these reasons, the Commission is revising subpart H of part 2 
to address standard design certifications. Section 2.800 is revised to 
delineate which provisions of subpart H are applicable to all petitions 
for rulemaking, and which provisions are applicable only to initial 
applications for design certification and applications for amendments 
to existing design certification rules filed by the original applicant 
(or successors in interest). The title of Sec.  2.800 is revised to 
reflect the additional function of this section. New Sec. Sec.  2.811 
through 2.819 are added to address initial applications for design 
certification as well as applications for amendments to existing design 
certifications filed by the original applicant (or successors in 
interest), and are based upon Sec. Sec.  2.101, 2.107, and 2.109. 
Petitions for amendment of existing design certification, which are 
filed by third parties other than the original applicant for that 
design certification (or successor in interest), will be treated as an 
amending petition for rulemaking under the provisions of Sec. Sec.  
2.801 through 2.810.
34. Section 2.801, Initiation of Rulemaking
    In a conforming change, Sec.  2.801 is revised to refer to 
applications for standard design certification rulemaking.
35. Section 2.811, Filing of Standard Design Certification Application; 
Required Copies
    New Sec.  2.811 clarifies the requirements that are related to the 
filing of applications for standard design certifications. The 
requirements in this section are derived from procedural requirements 
for license applications located in several different regulations in 
part 50. Section 2.811(a), which is analogous to Sec.  50.4(a), 
identifies the NRC addresses where an application for a standard design 
certification must be filed, and provides the requirements for 
electronic submission of a design certification application. Section 
2.811(b), which is analogous to Sec.  50.30(a)(1) and (3), provides 
that a standard design certification application must meet the written 
communications requirements in Sec.  2.813. Section 2.811(c), which is 
analogous to Sec.  50.30(a)(2), requires the applicant to have the 
capability to make and supply additional copies of the application upon 
NRC request. Section 2.811(d), which is analogous to the requirement in 
Sec.  50.30(a)(4), requires the applicant to make a copy of the updated 
application for use by any party in a hearing conducted under subpart O 
of part 2 (a legislative-style hearing). Section 2.811(e), which 
addresses pre-application consultation with the NRC staff, provides 
that the potential applicant for a design certification may consult 
with the NRC on the subject matters listed in Sec.  2.802(a)(1)(i) 
through (iii), including the procedure and process for filing and 
processing an application for a design certification. However, Sec.  
2.811(e) also allows the prospective standard design certification 
applicant to consult with the NRC staff on substantive technical and 
regulatory matters relevant to the design certification; the 
prohibitions in Sec.  2.802(a)(2) do not apply to these consultations.
36. Section 2.813, Written Communications
    New Sec.  2.813 contains procedural and ``housekeeping'' 
requirements governing written communications with the NRC, and are 
derived from analogous requirements located in several different 
regulations in part 50. Section 2.813(a) is analogous to Sec.  50.4(a). 
Section 2.813(b) is analogous to Sec.  50.4(c), and sets forth the 
requirement that written copies be submitted in permanent form on 
unglazed paper. Section 2.813(c) is analogous to Sec.  50.4(d), and 
expresses the Commission's preference that the upper right corner of 
the first page of the applicant's submission set forth the specific 
regulation or other basis which instigated the written communication.
37. Section 2.815, Docketing and Acceptance Review
    New Sec.  2.815 is analogous to Sec.  2.101(a)(2), and permits the 
NRC to conduct a review to determine whether the application is 
complete (i.e., addresses all matters specifically required by NRC 
regulation to be addressed in an application) and acceptable for 
docketing. Section 2.815(a) provides that the NRC may determine, in its 
discretion, the acceptability for docketing of an application based on 
the technical adequacy of the application, not just on the completeness 
of the application.
38. Section 2.817, Withdrawal of Application
    New Sec.  2.817 is analogous to Sec.  2.107, and addresses the 
procedures that the NRC will follow if a design certification applicant 
withdraws its application. Section 2.817 also provides for a notice of 
action on the withdrawal on the NRC Web site if the notice of 
application was published on the NRC Web site.
39. Section 2.819, Denial of Application for Failure To Supply 
Information
    New Sec.  2.819 is analogous to Sec.  2.108, and states in 
paragraph (a) that the NRC may deny an application for a standard 
design certification if the applicant fails to respond to an NRC 
request for additional information concerning its application within 30 
days of the request. Section 2.819(b) provides that the NRC will 
publish in the Federal Register a document denying the application. 
Section 2.819(b) also states that the NRC will publish a notice on

[[Page 49420]]

the NRC's Web site denying the application if the NRC previously 
published a notice of receipt of the application on the NRC Web site.
40. Section 2.1202, Authority and Role of NRC Staff
    Paragraph (a) of Sec.  2.1202 acknowledges and confirms the 
authority of the NRC staff to take regulatory (including licensing) 
action during the pendency of a hearing, with several delineated 
exceptions in numbered paragraphs (a)(1) through (5). Most of these 
exceptions are mandated by Section 189.a.(1)(A) of the AEA, which 
requires that the NRC hold a ``mandatory hearing,'' after 30 days 
notice and publication once in the Federal Register, on any application 
for a construction permit for a facility to be licensed under Section 
103 or 104b. Paragraph (a)(1) is revised by adding specific references 
to applications for limited work authorizations and combined licenses 
under 10 CFR part 52. A limited work authorization is considered to be 
a partial construction permit, and a combined license under part 52 
includes a construction permit. Therefore, they are both subject to the 
strictures of Section 189.a.(1)(A). Paragraphs (2), (3), and (4) are 
redesignated as paragraphs (4), (5), and (6), and a new paragraph (2) 
is added for early site permits applications. An early site permit is 
considered to be a partial construction permit, and therefore is also 
subject to Section 189.a(1)(A). A new paragraph (3) is added for 
manufacturing licenses, as a matter of NRC discretion. The Section 
189.a.(1)(A) requirement for a mandatory hearing applies only to 
construction permits; a manufacturing license is not a construction 
permit. Hence, the remaining provisions of Section 189.a.(1)(A), 
including the NRC's authority to issue an operating license or 
amendment to a construction permit without a hearing but only upon 30 
days notice and publication once in the Federal Register of the NRC's 
intent to do so, are inapposite and do not constrain the NRC's 
authority to issue manufacturing licenses despite a pending hearing. 
Nonetheless, as a matter of discretion, the NRC has decided to treat 
manufacturing licenses similar to construction permits in this regard, 
although the NRC reserves the right to change its practice in the 
future.

G. Changes to 10 CFR Part 10

1. Section 10.1, Purpose; and Sec.  10.2, Scope
    Part 10, which contains the NRC's requirements and procedures for 
determining eligibility for granting access to Restricted Data and 
National Security Information, did not reflect the licensing and 
approval processes in part 52. Accordingly, the Commission made two 
changes to ensure that there are defined criteria and procedures 
governing requests for access to Restricted Data and National Security 
Information by individuals with respect to a license or approval under 
part 52.
    Section 10.1 is revised by adding a new paragraph (a)(3), which 
refers to the eligibility of individuals for employment with NRC 
licensees and applicants, and holders of standard design approvals 
under part 52. Section 10.2(b) is revised so that it refers to standard 
design approvals under part 52 and applicants for consultants. This 
change will address the provision of services associated with design 
approvals, who may not, per se, be ``employees.''

H. Changes to 10 CFR Part 19

    Part 19, entitled Notices, Instructions and Reports to Workers: 
Inspection and Investigations, establishes the NRC's requirements for 
notices, instructions and reports to persons participating in NRC 
licensed and other regulated activities. For example, it requires 
licensees and applicants for licenses to post a copy of, inter alia, 
the regulations in 10 CFR parts 19 and 20, and NRC Form 3. NRC Form 3 
provides a statement of rights and responsibilities to employees with 
respect to NRC requirements. Part 19 also establishes the rights and 
responsibilities of the NRC and individuals during interviews compelled 
by subpoena as part of a NRC inspection or investigation under Section 
161.c of the AEA. Finally, part 19 prohibits, on the grounds of sex, 
the exclusion from participation in, or being subjected to 
discrimination under any program or activity licensed by the NRC. The 
regulatory authority for part 19 stems from Sections 211 and 401 of the 
Energy Reorganization Act of 1974, as amended (1974 ERA).
    The NRC has identified a number of weaknesses with the former 
regulatory language in part 19. Formerly, part 19's regulatory 
requirements and proscriptions applied only to licensees who receive, 
possess, use or transfer material licensed under the NRC's regulations, 
including persons licensed to operate a production or utilization 
facility under 10 CFR part 50, but did not cover holders of 10 CFR part 
52 licenses such as combined licenses, early site permits, and 
manufacturing licenses. Moreover, part 19 applied only to licensees who 
receive, possess, use or transfer materials licensed under 10 CFR parts 
30 through 36, 39, 40, 60, 61, 63, 70 or 72 (including persons licensed 
to operate a production or utilization facility under part 50). Thus, 
the former regulations did not appear to address discrimination against 
an employee during ``non-operational'' activities such as manufacturing 
or construction of a nuclear power plant. Because the NRC's regulatory 
scheme relies upon the proper design, manufacture, siting, and/or 
construction of a production or utilization facility; discrimination 
against an employee at any of these stages could have significant 
adverse public health and safety or common defense and security 
implications and effects. One would therefore expect that part 19 would 
apply to such non-operational activities. Finally, part 19 applied only 
to a ``licensee'' and activities authorized by a ``license'' (see, 
e.g., Sec. Sec.  19.1, 19.2, 19.11, 19.20, 19.32), and did not extend 
to part 52's non-licensing regulatory approvals, i.e., standard design 
approvals and standard design certifications. Inasmuch as these non-
licensing activities regulated under part 52 are not different in kind 
from the licensing which are currently subject to part 19 requirements, 
the NRC concludes that they should also be subject to the requirements 
in part 19.
    Accordingly, the NRC is amending various provisions in part 19 to 
ensure that its provisions extend to applicants for and holders of part 
50 construction permits, and combined licenses, early site permits and 
manufacturing licenses under part 52. In addition, the NRC extends part 
19 to cover applicants for and holders of standard design approvals and 
standard design certifications. The NRC believes that its regulatory 
authority under Section 211 and Section 401 of the 1974 ERA is much 
broader than the former scope of part 19. The anti-discrimination 
proscriptions in Section 211 of the ERA apply to any ``employer,'' 
which the NRC regards as including non-licensee entities otherwise 
regulated by the NRC, such as applicants for and holders of standard 
design approvals, and applicants for standard design certifications. 
The Commission believes that the use of the term, ``includes,'' in 
paragraph (a)(2) of Section 211 of the 1974 ERA was not intended to be 
an exclusive list of the persons and entities subject to the anti-
discrimination provisions in that section. The House Report on H.R. 
776, which was adopted by Congress as the Energy Policy Act of 1992, 
states:

    [Title V] also broadens the coverage of existing whistle blower 
protection provisions to include * * * any other employer engaged

[[Page 49421]]

in any activity under the Energy Reorganization Act of the Atomic 
Energy Act of 1954. (H. Rep. No. 102-474, part 8, 102d Congress, 2d 
Sess., at 78-79 (1992) (emphasis added))

    There was no discussion of the statutory language in the conference 
report. (H.R. Conf. Rep. No. 102-1018, 102d Cong., 2d Sess. (1992)). 
The provisions in Section 401 of the ERA, prohibiting sex 
discrimination apply to ``any program or activity carried on * * * 
under any title of this Act.'' Accordingly, the NRC concludes that it 
has the authority to extend the former scope of part 19 to address the 
non-licensing regulatory approvals in part 52.
    To implement the NRC's broadening of the scope of part 19, 
Sec. Sec.  19.1 and 19.2 are revised to explicitly refer to: (1) 
applicants for and holders of licenses and permits under part 52; (2) 
applicants for and holders of final design approvals; and (3) 
applicants for standard design certifications. The NRC notes that the 
existing provision in Sec.  19.2 excluding part 19 from applying to NRC 
employees and NRC contractors remains unchanged in the final rule. To 
provide a convenient term for referring to persons and entities 
applying for, or granting non-licensed regulatory approvals in part 52, 
as well as any future regulatory processes, the NRC is amending Sec.  
19.3 to the terms, regulated activities, and regulated entities. 
Regulated entities are defined to include (but not be limited to) 
applicants for and holders of standard design approvals under subpart E 
of part 52, and applicants for standard design certifications under 
subpart B of part 52.
    Section 19.11 establishes requirements for posting of notices to 
workers. Because Sec. Sec.  19.11(a)(2) and (a)(4) contain posting 
requirements which are not relevant to early site permits, 
manufacturing licenses, standard design approvals, and standard design 
certifications, the NRC delineated in Sec.  19.11(b) the applicable 
posting requirements for those regulatory processes. Section 19.11(c) 
is reserved for future Commission use.
    Sections 19.14 and 19.20 are revised to apply to regulated 
entities, as well as licensees.
    Section 19.31, governing exemptions from part 19, is revised to use 
language consistent with Sec.  50.12 and Sec.  52.7. Unlike the former 
regulation, which limits a request for exemption to a ``licensee,'' the 
final rule allows ``interested persons,'' as well as licensees to 
request an exemption from one or more provisions of part 19. This will 
allow applicants for and holders of non-license regulatory vehicles in 
part 52 (standard design approvals and design certifications) to 
request exemptions from part 19. The broadened scope of persons that 
will be allowed to request an exemption is consistent with most of the 
exemption provisions throughout the NRC's regulations in Title 10 of 
the CFR, including the specific exemption provision in part 50 (i.e., 
Sec.  50.12).
    Section 19.32 is revised to more closely track the broad scope of 
statutory language in Section 401 of the 1974 ERA, which is not limited 
to licensing, but extends the sex discrimination prohibition to ``any * 
* * activity carried on * * * under any title'' of the ERA. By using 
the statutory language in the proposed rule, the NRC believes that the 
regulations cover not only the existing non-license regulatory vehicles 
in part 52, but any other regulatory approaches that the NRC may adopt 
in the future (Section 401 of the 1974 ERA applies to NRC regulatory 
activities under the AEA, inasmuch as the 1974 ERA transferred the AEA 
regulatory authority from the old AEC to the NRC, see 1974 ERA, Sec. 
104(c)).

I. Changes to 10 CFR Part 20

1. Section 20.1002, Scope
    10 CFR part 20 applies to persons licensed by the NRC to receive, 
possess, use, transfer, or dispose of byproduct, source, or special 
nuclear material or to operate a production or utilization facility. 
Accordingly, Sec.  20.1002 is revised by adding a conforming reference 
to part 52, which sets forth a process for licensing a utilization 
facility.
2. Section 20.1401, General Provisions and Scope
    This section on decommissioning of facilities is revised to add a 
conforming reference to facilities licensed under 10 CFR part 52.
3. Section 20.1406, Minimization of Contamination
    Section 20.1406 requires applicants for licenses, other than 
renewals, after August 20, 1997, to describe in the application how 
facility design and procedures for operation will minimize, to the 
extent practicable, contamination of the facility and the environment, 
facilitate eventual decommissioning, and minimize, to the extent 
practicable, the generation of radioactive waste. The NRC is adding 
conforming changes to Sec.  20.1406 in the final rule. These conforming 
changes to address part 52 were inadvertently overlooked in the 
proposed rule. Section 20.1406 contains requirements that relate both 
to design and operation of a facility and therefore applies in whole or 
in part to design approvals, design certifications, manufacturing 
licenses, and combined licenses. The final rule divides Sec.  20.1406 
into two paragraphs. Paragraph (a) addresses applicants for licenses, 
other than early site permits and manufacturing licenses, and contains 
all of the requirements in former Sec.  20.1406. Paragraph (b) 
addresses applicants for standard design certifications, standard 
design approvals, and manufacturing licenses and only contains the 
requirements related to design. If a combined license applicant 
references a design approval, design certification, or a reactor 
manufactured under a manufacturing license that has addressed the 
design portion of this requirement under paragraph (b), then it would 
only need to address the remaining ``operational'' requirements under 
paragraph (a).
4. Section 20.2203, Reports of Exposures, Radiation Levels, and 
Concentrations of Radioactive Material Exceeding the Constraints or 
Limits
    Sections 20.2203(c) and (d) are revised to add a reference to 
holders of combined licenses to the procedures on submitting reports.

J. Changes to 10 CFR Part 21

    Part 21 implements the reporting requirements in Section 206 of the 
ERA. The proposed part 52 rule published in 2003 set forth the NRC's 
proposals as to how Section 206 reporting and, therefore, part 21 
applicability should be extended to early site permits, standard design 
certifications, and combined licenses. However, the 2003 proposed rule 
did not address Section 206 reporting requirements with respect to 
standard design approvals or manufacturing licenses. Moreover, the 
proposals were developed without the benefit of the NRC's in-depth 
consideration of the issues as applied in the context of the early site 
permit applications that are currently before the NRC. Accordingly, NRC 
withdrew its earlier proposal and developed a more complete and 
integrated rule on Section 206 reporting under part 21 and Sec.  
50.55(e). As discussed previously, Sec.  50.55(e) sets forth the 
Section 206 reporting requirements applicable to holders of 
construction permits, combined licenses, and manufacturing licenses.
Key Principles of Reporting Under Section 206 of the ERA
    The NRC believes that the extension of NRC's reporting requirements

[[Page 49422]]

implementing Section 206 of the ERA to part 52 licensing and approval 
processes should be consistent with three key principles. First, NRC 
regulatory requirements implementing Section 206 of the ERA should be a 
legal obligation throughout the entire ``regulatory life'' of an NRC 
license, a standard design approval, or standard design certification. 
Second, reporting of defects or failures to comply associated with 
substantial safety hazards should occur whenever the information on 
potential defects would be most effective in ensuring the integrity and 
adequacy of the NRC's regulatory activities under part 52 and the 
activities of entities \10\ subject to the part 52 regulatory regime. 
Third, each entity conducting activities within the scope of part 52 
should develop and implement procedures and practices to ensure that it 
fulfills its Section 206 of the ERA reporting obligation in an accurate 
and timely manner.
---------------------------------------------------------------------------

    \10\ Throughout this discussion, reference to entities, 
licensees and/or applicants includes the contractors and 
subcontractors of those entitles, licensees and/or applicants.
---------------------------------------------------------------------------

First Principle--Section 206 of the ERA Applies Throughout ``Regulatory 
Life''
    The first principle, that NRC regulatory requirements implementing 
Section 206 must extend throughout the entire ``regulatory life'' of a 
part 52 process, reflects the regulatory pattern inherent in part 52, 
whereby certain designated licenses or approvals--e.g., an early site 
permit, nuclear power reactor manufactured under a manufacturing 
license, or a design certification--are capable of being referenced in 
a subsequent nuclear power plant licensing application. Under the part 
52 regulatory scheme, a referenced NRC approval constitutes the NRC's 
basis for the licensing action within the scope of the prior Commission 
approval, and becomes part of the ``licensing basis'' for that plant. 
However, if Section 206 of the ERA reflects that effective NRC 
decision-making and regulatory oversight require accurate and timely 
information about defects and failures to comply associated with 
substantial safety hazards, then Section 206 of the ERA should apply 
whenever necessary to support effective NRC decision-making and 
regulatory oversight of the referencing licenses and regulatory 
approvals. To put it in different terms, if the NRC decision that it 
may safely issue a license depends in part upon an earlier NRC safety 
determination for a referenced license, standard design approval, or 
standard design certification, it follows that a safety issue with 
respect to the referenced license, design approval, or design 
certification has safety implications for the referencing license or 
design certification, and the continuing validity of the NRC's 
licensing decision. Thus, the NRC concludes that the need for Section 
206 reporting should not be limited to those licenses and approvals 
under part 52 which are referenced or ``relied upon'' in a subsequent 
nuclear power plant licensing application (viz., early site permits, 
standard design approvals, standard design certifications, and 
manufacturing licenses), but rather should extend to licenses and 
approvals that are capable of being referenced in a future licensing 
application. In other words, they must extend until there can be no 
further potential safety implications for a referencing license or 
approval.
    The NRC believes that the beginning of the ``regulatory life'' of a 
referenced license, standard design approval, or standard design 
certification under part 52 occurs when an application for a license, 
design approval, or design certification is docketed. Docketing of an 
application marks the start of the NRC's formal safety and 
environmental review of the application, and therefore the initiation 
of the NRC's need for accurate and timely information to support its 
regulatory review and approval. However, the NRC cautions that this 
does not mean that an applicant is without Section 206 responsibilities 
for pre-application activities. As the NRC staff discussed in a June 
22, 2004, letter to the Nuclear Energy Institute (NEI) (ML040430041) in 
the context of an early site permit, there are two aspects, namely, a 
``backward looking'' or retrospective aspect with respect to existing 
information, and a ``forward looking'' or prospective aspect with 
respect to future information. The retrospective obligation is that the 
early site permit holder and its contractors, must report all known 
defects or failures to comply in ``basic components,'' as defined in 
part 21. The prospective obligation is that the early site permit 
holder and its contractors must report all defects or failures to 
comply in basic components discovered subsequent to early site permit 
issuance. The early site permit holder and its contractors are required 
to meet these requirements, and must continue to meet them throughout 
the term of the early site permit. Accordingly, safety-related design 
and analysis or consulting services should be procured and controlled, 
or dedicated, in a manner sufficient to allow the early site permit 
holder and its contractors, as applicable, to comply with the above 
described reporting requirements of Section 206, as implemented by 10 
CFR 50.55(e) and part 21.
    The NRC believes that the end of regulatory life occurs at the 
later of: (1) The termination or expiration of the referenced license, 
standard design approval, or standard design certification; or (2) the 
termination or expiration of the last of the license or design 
certification directly or indirectly referencing the (referenced) 
license, design approval, or design certification. For example, if the 
NRC approves a standard design approval, which is subsequently 
referenced in a final standard design certification rule, and that 
standard design certification is, in turn referenced in a combined 
license issued by the NRC, the ``end'' of the regulatory life occurs 
when the authorization to operate under the combined license is 
terminated (ordinarily, under the provisions of Sec.  52.110). As long 
as a referenced combined license continues to be effective, the 
``regulatory life'' of a referenced license, standard design approval, 
standard design certification, or manufactured reactor (as applicable) 
must also continue and cannot be deemed to have ended.
    Some commenters argued that the NRC's regulatory interests would be 
met if reporting under Section 206 of the ERA were limited to the 
referencing applicant/licensee, and that there should be no ongoing 
part 21 reporting obligation imposed on the early site permit holder, 
original applicant for a standard design certification, or holder of a 
part 52 regulatory approval. Under this proposal the referencing 
applicant and licensee would satisfy its obligation by an appropriate 
contractual provision between the referencing applicant/licensee and 
the entity ``supplying'' the referenced license or regulatory approval. 
Although this could be a viable alternative for some combined licenses, 
early site permits, and standard design approvals, the approach would 
not be effective in the following contexts. This approach would not 
result in reporting of defects to the NRC by the applicant of the early 
site permit or standard design certification, which violates the NRC's 
second principle (discussed more fully in the next section). In 
addition, this approach would not result in reporting where there is no 
contractual relationship between the combined license applicant/
licensee and the original applicant of the standard design 
certification. Because the approach suggested by these commenters does 
not

[[Page 49423]]

satisfy the NRC's regulatory objectives, it is not adopted.
    One of the original applicants for the current standard design 
certifications stated that any arguable Section 206 requirements must 
logically end upon expiration of the standard design certification, 
inasmuch as expiration marks the end time that the standard design 
certification may be referenced. The NRC disagrees with this position. 
Under Sec.  52.55(b) of the current regulations, a standard design 
certification continues to be effective in a hearing for a combined 
license or operating license docketed before the expiration date, and 
in a hearing under Sec.  52.103 for authority to load fuel and operate. 
At minimum, the original standard design certification applicant should 
be subject to Section 206 requirements until the proceeding is 
completed. Beyond the minimum requirements, the NRC also believes that 
the original design certification applicant's Section 206 obligations 
should continue until operation is no longer authorized in accordance 
with Sec.  50.82(a)(2) for the last operating license or combined 
license referencing that standard design certification. The NRC 
believes that the regulatory need for information concerning defects in 
a standard design certification continues throughout the operating life 
of a license referencing that design certification; the relevance of 
and the NRC's need for this information, if subsequently discovered by 
the original design certification applicant, does not diminish simply 
because the standard design certification may no longer be referenced.
Second Principle--Notification Occurs When Information Is Needed
    The second principle is focused on ensuring that the NRC, its 
licensees, and license applicants receive information on defects at the 
time when the information would be most useful to the NRC in carrying 
out its regulatory responsibilities under the AEA, and to the licensee 
or applicant when engaging in activities regulated by the NRC. A result 
of this principle is that reporting may be delayed if there is no 
immediate consequence or regulatory interest in prompt reporting, and 
that delayed reporting will actually occur when necessary to support 
effective, efficient, and timely action by the NRC, its licensees and 
applicants. Applying the second principle and its result to part 52 
processes, the NRC believes that immediate reporting is required 
throughout the period of pendency of an application, be it a license, a 
standard design approval, or a standard design certification. Allowing 
an applicant to delay the reporting of a defect would appear to be 
inconsistent with the NRC's statutory mandate to provide adequate 
protection to public health and safety and common defense and security. 
Even if delayed reporting would allow the NRC an opportunity to modify 
its prior safety finding with respect to the license, design approval, 
or design certification, the delayed consideration is inconsistent with 
one of the fundamental purposes of part 52, viz., to provide for early 
consideration and resolution of issues in a manner that avoids the 
potential for delay during licensing of a facility. Accordingly, the 
Commission has determined that the NRC's requirements implementing 
Section 206 of the ERA must extend to applicants (and their contractors 
and subcontractors) for all part 52 processes (licenses, early site 
permits, design approvals, and design certifications). Some commenters 
stated that part 21 should not apply to applicants and claimed that the 
NRC's proposal was contrary to the ERA. For the reasons stated 
previously, the Commission does not agree with that position. However, 
once an application has been granted, the Commission has decided that 
immediate reporting of subsequently-discovered defects is not necessary 
in certain circumstances. For those part 52 processes which do not 
authorize continuing activities required to be licensed under the AEA, 
but are intended solely to provide early identification and resolution 
of issues in subsequent licensing or regulatory approvals, the 
reporting of defects or failures to comply associated with substantial 
safety hazards may be delayed until the time that the part 52 process 
is first referenced. The Commission's view is based upon its 
determination that a defect with respect to part 52 processes should 
not be regarded as a ``substantial safety hazard,'' because the 
possibility of a substantial safety hazard becomes a tangible 
possibility necessitating NRC regulatory interest only when those part 
52 processes are referenced in an application for a license, such as a 
combined license or manufacturing license.
    Some commenters believe that these reporting requirements should 
not apply to a holder of an early site permit or a vendor of a standard 
design until the ESP or standard design is referenced in a COL 
application. As stated previously, the Commission agrees that reporting 
may be delayed until the approval, certification, or permit is 
referenced. After referencing, the holder (or in the case of a design 
certification, the applicant who submitted the application leading to 
the final design certification regulation) must make the necessary 
notifications to the NRC as well as provide final engineering. The 
notification must address the period from the Commission adoption of 
the final design certification regulation up to the filing of the 
application referencing the final design certification regulations. 
Thereafter, notice must be made in the ordinary manner. The 
notification obligation ends when the last license referencing the 
design certification is terminated.
Third Principle--Procedures and Practices Must Be Implemented To Ensure 
Accurate and Timely Reporting
    The third principle (viz., each entity conducting activities under 
the purview of part 52, should develop and implement procedures and 
practices to ensure that the entity accurately and timely fulfils its 
reporting obligation as delineated in the NRC's regulations), is 
intended to ensure the effectiveness of each entity's reporting 
processes. This is especially true where there is a potential for 
substantial passage of time between the discovery of a defect and the 
reporting of the defect, as may be allowed by the NRC consistent with 
the second principle. For example, following issuance of a final 
standard design certification regulation, if the original applicant 
determines that there is a substantial safety hazard, that applicant 
need not report the discovery until the time that the design 
certification rule is referenced--which may be as long as 15 years from 
the date of the final rule. Given the substantial time that may pass 
between the time of discovery and the date of reporting, it is 
imperative that the original standard design certification applicant 
develop and implement procedures from the time of effectiveness of the 
final design certification regulations.
    The result of the third principle, consistent with part 21's 
current requirements, is that licensees, license applicants, and other 
entities seeking a design approval or design certification, must have 
contractual provisions with their contractors, subcontractors, 
consultants, and other suppliers which notify them that they are 
subject to the NRC's regulatory requirements on reporting and the 
development and implementation of reporting procedures. This result is 
set forth in Sec. Sec.  21.31 and 50.55(e)(7).

[[Page 49424]]

Division of Implementing Requirements Between Part 21 and Sec.  
50.55(e)

    Under the Commission's current regulatory structure, persons and 
entities engaged in construction (or the functional equivalent of 
construction) are subject to reporting requirements under Sec.  
50.55(e). Persons and entities engaged in all other activities within 
the purview of Section 206 of the ERA are subject to the requirements 
in part 21 and/or Sec.  50.55(e). The revised part 21 and Sec.  
50.55(e) reflect the Commission's determination to retain this divided 
regulatory structure. The NRC believes that the only part 52 processes 
that authorize ``construction'' or its functional equivalent are 
manufacturing licenses and combined licenses before the Commission 
makes the finding under Sec.  52.103(g). Therefore, the reporting 
requirements with respect to Section 206 of the ERA for manufacturing 
licenses and combined licenses before the Commission makes the finding 
under Sec.  52.103(g) are contained in Sec.  50.55(e). The requirements 
in part 21 apply after the Commission makes the finding under Sec.  
52.103(g) for a combined license. Part 21 was revised to explicitly 
apply to the remaining part 52 processes, i.e., early site permits, 
standard design approvals, and standard design certifications. Table A-
1 provides a summary of the applicability of part 21 and Sec.  50.55(e) 
to each of the various approvals under part 52.

 Table A-1.--Applicability of NRC Requirements Implementing Section 206 of the Energy Reorganization Act to Part
                                       52 Licensing and Approval Processes
----------------------------------------------------------------------------------------------------------------
                                                      Applicable NRC requirement               Sanctions
     Part 52 licensing or approval processes        implementing section 206 of the  ---------------------------
                                                                  ERA                     Civil       Criminal
----------------------------------------------------------------------------------------------------------------
Early Site Permit (ESP)
    Application.................................  part 21...........................        21.61         21.62
    Issuance of ESP.............................  part 21...........................        21.61         21.62
Standard Design Approval (SDA)
    Application.................................  part 21...........................        21.61         21.62
    Issuance of SDA.............................  part 21...........................        21.61         21.62
Standard Design Certification Rule (DCR)
    Application.................................  part 21...........................        21.61         21.62
    Final DCR Rulemaking........................  part 21...........................        21.61         21.62
Combined License (COL)
    Application.................................  50.55(e)..........................        50.110        50.111
    COL before Sec.   52.103 Authorization......  50.55(e)..........................        50.110        50.111
    COL after Sec.   52.103 Authorization.......  part 21...........................        21.61         21.62
Manufacturing License (ML)
    Application.................................  50.55(e)..........................        50.110        50.111
    Issuance of ML..............................  50.55(e)..........................        50.110        50.111
----------------------------------------------------------------------------------------------------------------

Reporting Requirements for Early Site Permits

    If the ESP holder becomes aware of a significant safety concern 
with respect to its site (e.g., that the specified site characteristics 
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 any future application referencing 
the ESP. As stated previously, the reporting may be delayed until the 
ESP is referenced. This reporting attains special importance given the 
NRC's proposal not to impose an updating requirement for ESP 
information other than that related to emergency preparedness. In order 
for the applicant for an ESP to have the capability to report to the 
NRC any known significant safety concerns with respect to its site, or 
any safety concerns of which it may subsequently become aware (i.e., to 
be able to report any defects or failures to comply associated with 
substantial safety hazards under part 21) the ESP applicant would have 
to have a program in place for implementing the requirements of 10 CFR 
part 21. The applicant's program may be inspected by the NRC as part of 
the application review. Approval of the ESP application would be 
subject to approval of the part 21 program.
    Some commenters claimed that there is no practicable method for ESP 
applicants or holders to determine whether an error in siting 
information creates a substantial safety hazard and, therefore, part 21 
should not be applicable to ESP applicants or holders. The Commission 
does not agree with this position. As stated previously, the ESP holder 
and its contractors can determine defects or failures to comply with 
``basic components,'' as defined in part 21. This information is 
necessary in order to support effective NRC decisionmaking and 
regulatory oversight of the referencing licenses and approvals.

Applicability of Part 21 to Contractors or Subcontractors of an ESP 
Applicant or Holder

    In accordance with 10 CFR 21.31, the purchaser of a basic component 
must state in the procurement documents for the basic component that 
part 21 is applicable to that procurement. As explained previously, 
services that are required to support an early site permit application 
(e.g., geologic or seismic analyses, etc.) that are safety-related and 
could be relied upon in the siting, design, and construction of a 
nuclear power plant, are to be treated as basic components as defined 
in part 21. Therefore, these services must be either purchased as basic 
components, requiring the service provider to have an appendix B to 
part 50 QA program, as well as its own part 21 program, or the early 
site permit applicant could dedicate the service in accordance with 
part 21, which requires the dedication process itself to be controlled 
under an appendix B to part 50 QA program.

Reporting Requirements for Standard Design Approvals

    A standard design approval represents the NRC staff's determination 
regarding the acceptability of the design for a nuclear power reactor 
(or major portions thereof). Although a standard design approval does 
not represent the NRC's final determination as to the acceptability of 
the design, it nonetheless represents a substantial expenditure of 
agency resources in reviewing the design. A standard design

[[Page 49425]]

approval may be referenced in a subsequent application for a design 
certification, construction permit, operating license, combined 
license, or manufacturing license. Accordingly, consistent with the 
first principle, the final rule imposes requirements implementing 
Section 206 of the ERA on applicants for and holders of standard design 
approvals.
    A standard design approval does not authorize construction of a 
nuclear power plant; it merely constitutes the NRC staff's approval of 
the design of a nuclear power reactor (or major portion thereof). 
Therefore, the requirements implementing Section 206 of the ERA, which 
are applicable to standard design approvals, were placed in part 21, as 
opposed to Sec.  50.55(e).

Reporting Requirements for Standard Design Certification Regulations

    A standard design certification represents the NRC's approval by 
rulemaking of an acceptable nuclear power reactor design, which may 
then be referenced in a subsequent combined license or manufacturing 
license application. Consistent with the first principle, the 
Commission imposed Section 206 of the ERA reporting requirements on 
applicants for design certifications, including applicants whose 
designs are certified in a final design certification rulemaking. As 
with a standard design approval, a design certification does not 
actually authorize construction. Accordingly, the NRC revised 
Sec. Sec.  21.2, 21.3, 21.21, 21.51, and 21.61 to explicitly refer to 
an applicant for a standard design certification, rather than Sec.  
50.55(e).
    Some commenters have asserted that because there is no ``holder'' 
or licensee, the NRC is without authority under Section 206 of the ERA 
to impose part 21 and/or Sec.  50.55(e) evaluation and reporting 
requirements on applicants for standard design certification. The NRC 
disagrees with this assertion. The statute by its terms does not limit 
its reach to licensees; rather, the statute applies to any individual 
or responsible officer of a firm ``constructing, owning, operating, or 
supplying the components of any facility or activity which is licensed 
or otherwise regulated * * *.'' The NRC believes that an applicant for 
a standard design certification, by submitting its application, is 
constructively ``supplying'' a ``component'' (the nuclear power plant) 
for use in a future ``facility * * * licensed'' by the NRC. One of the 
consequences of the design certification provisions in part 52 is the 
ability of the applicant to subsequently offer its design with 
additional, value-added services. Thus, applying for and facilitating 
NRC adoption of a final standard design certification regulation is 
simply a partial step in the overall activity of ``supplying'' the 
certified design to potential nuclear power plant license applicants. 
Alternatively, one could treat the standard design certification 
applicant as supplying a component of an ``activity'' which is 
``otherwise regulated'' by the NRC. Under this interpretation, the 
``activity * * * otherwise regulated by the NRC'' can be viewed as the 
design certification rulemaking, and/or the entire part 52 regulatory 
regime whereby a design certification rule is referenced in a 
subsequent licensing application. The NRC concludes that under either 
interpretation, Section 206 of the ERA provides ample statutory 
authority for the NRC to impose regulations implementing Section 206 on 
design certification applicants, during the pendency of the application 
before the NRC, as well as after NRC adoption of a final design 
certification regulation (for those applicants whose application is 
granted).
    As with standard design approvals, a standard design certification 
does not authorize construction of a nuclear power plant; it 
constitutes the NRC's approval of the design of a nuclear power plant. 
Therefore, the requirements implementing Section 206 of the ERA which 
are applicable to design certifications were placed in part 21, as 
opposed to Sec.  50.55(e).

Reporting Requirements for Combined Licenses

    A combined license authorizes both construction of a nuclear power 
plant, and loading of fuel and operation if the NRC makes the findings 
specified in Sec.  52.103. As such, the application of the first and 
second principles to combined licenses is the most straightforward of 
all the part 52 processes. Under the final rule, the NRC's requirements 
implementing Section 206 of the ERA would apply throughout the 
regulatory life of the combined license, i.e., from docketing of the 
application until termination of the combined license.
    To maintain the current division between Sec.  50.55(e) and part 21 
with respect to NRC requirements implementing Section 206 of the ERA, 
the NRC revised Sec.  50.55(e) to make its provisions applicable to 
each holder of a combined license under part 52 before the effective 
date of the NRC's finding under Sec.  52.103(g), and to revise part 21 
to clarify that its provisions apply to each holder of a combined 
license on the effective date of the Commission's authorization under 
Sec.  52.103(g).

Reporting Requirements for Manufacturing Licenses

    Under subpart F of part 52, a manufacturing license would 
constitute both the NRC's approval of a final nuclear power reactor 
design, as well as approval to manufacture one or more reactors in 
accordance with approved programs and procedures. The manufactured 
reactors would then be transported offsite and incorporated into 
nuclear power facilities by holders of combined licenses--who may be 
different entities than the holder of a manufacturing license. Given 
the possibility that the manufacturing license holder is different from 
the combined license holder whose facility uses the manufactured 
reactor, the NRC believes that the combined license holder must be kept 
informed of any significant issue with design or manufacture of the 
reactor, to ensure that they evaluate the significance of these matters 
for their facility and undertake any necessary action to assure public 
health and safety and common defense and security. Furthermore, unlike 
a standard design certification, the financial resources necessary to 
obtain a manufacturing license will, as a practical matter, result in 
manufacturing beginning immediately after issuance of the manufacturing 
license. There will be no interim period similar to a design 
certification where there is no activity occurring under the 
manufacturing license. Accordingly, in compliance with the first and 
second principles, the NRC proposes that Section 206 of the ERA 
requirements should apply continuously from the filing of the 
application, until the manufacturing license expires or is otherwise 
terminated by the NRC.
    A manufacturing license holder would essentially be conducting the 
same activities as a construction permit holder, albeit with several 
differences.\11\ Nonetheless, the NRC believes that manufacturing is 
similar to construction such that the NRC's requirements implementing 
Section 206 of the ERA which are applicable to manufacturing licenses, 
are contained in Sec.  50.55(e).

[[Page 49426]]

Accordingly, the NRC revised Sec.  50.55(e) to specifically apply its 
provisions to holders of manufacturing licenses.
---------------------------------------------------------------------------

    \11\ These key differences are, first, the design of the 
manufactured plant would be approved before manufacturing commences, 
unlike the historical practice with construction permits. Second, a 
single manufacturing license may authorize the manufacture of 
multiple reactors, with the manufacturing process to be accomplished 
in a controlled setting rather than as a ``field'' operation. This 
is unlike the historical approach where non-standardized nuclear 
power facilities were constructed onsite using a ``roving'' 
workforce. Third, the manufacturing license will specify the 
inspections, tests, and acceptance criteria for determining 
successful manufacturing.
---------------------------------------------------------------------------

K. Change to 10 CFR Part 25

1. Section 25.35, Classified Visits
    Part 25 sets forth the NRC's requirements governing the granting of 
access authorization to classified information to certain individuals. 
Section 52.35, which requires that licensees and certificate holders 
minimize the number of classified visits, did not, by its terms, apply 
to applicants for standard design certifications, and applicants for or 
holders of standard design approvals. Accordingly, Sec.  25.35 is 
revised to refer to an applicant for a standard design certification 
under part 52 (including the applicant after the NRC adopts a final 
standard design certification rule), and the applicant for or holder of 
a standard design approval under part 52.

L. Changes to 10 CFR Part 26

1. Section 26.2, Scope, Sec.  26.10, General Performance Objectives; 
and Appendix A to Part 26
    Part 26, which sets forth the NRC's requirements governing fitness-
for-duty, currently uses a two-part regulatory regime for the 
application of fitness-for-duty requirements. A holder of an operating 
license for a nuclear power plant is required to implement all of the 
provisions in part 26. By contrast, a holder of a construction permit 
is required to comply with Sec. Sec.  26.10, 26.20, 26.23, 26.70, and 
26.73, and also implement a chemical testing program, including random 
tests, and make provisions for employee assistance programs, imposition 
of sanctions, appeals procedures, the protection of information, and 
record keeping.
    The NRC has extended the applicability of parts 26 to 52, in 
keeping with the existing two-part regulatory regime, so that the full 
array of requirements in part 26 apply to a combined license holder 
after the date that the NRC authorizes makes the finding under Sec.  
52.103(g), analogous to holder of an operating license under part 50. 
By contrast, holders of combined licenses, before the date that the NRC 
makes the Sec.  52.103(g) findings, are required to comply with the 
part 26 provisions currently applicable to construction permit holders. 
Similarly, holders of manufacturing licenses under subpart F of part 52 
are treated the same as holders of construction permits. Finally, 
persons authorized to conduct the limited construction activities 
allowed under Sec.  50.10(e)(3) are also treated the same as a 
construction permit holder. The final rule accomplishes this by: (1) 
Revising Sec.  26.2(a) to refer to combined license holders after the 
date that the NRC makes the finding under Sec.  52.103(g); (2) revising 
Sec.  26.2(c) to refer to a holder of a combined license before the 
date that the NRC makes the finding under Sec.  52.103(g), a holder of 
a manufacturing license under subpart F of part 52, and a person 
authorized to conduct the activities under Sec.  50.10(e)(3); (3) 
revising Sec.  26.10(a) to refer to the personnel of a holder of a 
manufacturing license and those authorized to conduct the activities 
under Sec.  50.10(e)(3); and (4) revising appendix A to part 26, 
paragraph 1.1(1) to include a reference to a holder of combined license 
after the date that the NRC makes the finding under Sec.  52.103(g).
    The NRC believes that part 26 need not be extended to cover 
applicants for and holders of early site permits, standard design 
approvals, and applicants for standard design certifications. These 
activities present less of a concern with respect to public health and 
safety, and common defense and security, as compared with construction 
permits, manufacturing licenses, operating licenses, and combined 
licenses. None of these regulatory approvals or design certification 
regulations authorize the construction, manufacture, or operation of a 
facility, nor do they authorize possession of special nuclear material 
(SNM). The adverse impacts on public health and safety or common 
defense and security attributable to any fitness-for-duty issues are 
likely to be of a much lower level of significance, as compared to 
issues that may occur during construction, manufacture, operation, or 
possession of SNM. The NRC believes that the potential benefits of 
imposing the fitness-for-duty requirements are not justified in view of 
the regulatory burden to be imposed upon such applicants and holders. 
Accordingly, these requirements will not be imposed on applicants for 
and holders of standard design approvals and applicants for standard 
design certifications under part 52.

M. Changes to 10 CFR Part 51

    The NRC is making several conforming changes to part 51 to clarify 
the environmental protection regulations applicable to the various part 
52 licensing processes.
NEPA Compliance for Design Certifications
    For each of the four design certification rules in appendices A, B, 
C, and D of part 52, the NRC prepared an environmental assessment 
which: (1) Provides the bases for a Commission finding of no 
significant environmental impact (FONSI) for issuance of the design 
certification regulation; and (2) identifies and addresses the need for 
incorporating SAMDAs into the design certification rule. Based upon 
this experience, the NRC is making changes to part 51 to accomplish two 
objectives.
    First, the NRC is eliminating the need for the NRC to prepare 
essentially repetitive discussions in environmental assessments 
supporting a FONSI on issuance of a final standard design certification 
regulation. Each of the environmental assessments and FONSIs prepared 
to date conclude that there is no significant environmental impact 
associated with NRC issuance of a final design certification regulation 
because a design certification does not authorize either the 
construction or operation of a nuclear power facility. Design 
certification represents the NRC's pre-approval of the design for the 
nuclear power facility, but does not authorize manufacture or 
construction. For the design certification to have practical effect, it 
must be referenced in an application for a combined license. The NRC is 
revising part 51 to eliminate the need for the NRC to make repetitive 
findings of no significant environmental impact for future design 
certifications and amendments to design certifications.
    Second, the NRC is requiring that SAMDAs be addressed at the design 
certification stage. SAMDAs are alternative design features for 
preventing and mitigating severe accidents, which may be considered for 
incorporation into the proposed design. The SAMDA analysis is that 
element of the severe accident mitigation alternatives analysis dealing 
with design and hardware issues. At the design certification stage, the 
NRC's review is directed at determining if there are any cost 
beneficial SAMDAs that should be incorporated into the design, and if 
it is likely that future design changes would be identified and 
determined to be cost-justified in the future based on cost/benefit 
considerations. It is most cost effective to incorporate SAMDAs into 
the design at the design certification stage. Retrofitting a SAMDA into 
a design certification once site-specific design and engineering for a 
nuclear power facility have been completed would increase the cost of 
implementing a SAMDA. The retrofitting costs continue to increase in 
ensuing stages of facility construction and operation. For these 
reasons, the NRC believes that environmental

[[Page 49427]]

assessments for design certifications should address SAMDAs. However, 
under the former provisions of part 51, both the environmental 
information submitted by the design certification applicant, and the 
environmental assessment prepared by the NRC, are directed either at 
determining whether an EIS must be prepared, or that a FONSI is 
justified. Accordingly, the NRC is requiring that SAMDAs be addressed 
in environmental reports and environmental assessments for design 
certifications.
    The NRC is making a number of changes to accomplish these two 
objectives. The NRC is redesignating existing Sec.  51.55 as Sec.  
51.58, and is adding new Sec.  51.55 to indicate that an environmental 
report submitted by the design certification applicant must be directed 
towards addressing the costs and benefits of possible SAMDAs, and 
presenting the bases for not incorporating identified SAMDAs into the 
design to be certified. The environmental report for an applicant 
seeking to amend an existing design certification would be somewhat 
narrower by focusing on if the design change which is the subject of 
the amendment, renders a SAMDA previously rejected to become cost-
beneficial, and if the design change results in the identification of 
new SAMDAs that may be reasonably incorporated into the design 
certification.
    The NRC is revising Sec.  51.30 to provide for a new Sec.  51.30(d) 
establishing the scope of an environmental assessment for a design 
certification. The NRC is adding Sec. Sec.  51.32(b)(1) and (2) to set 
forth the NRC's generic determination of no significant environmental 
impact associated with issuance of a final or amended design 
certification rule. This is, essentially, the legal equivalent of a 
categorical exclusion. The NRC is including an explicit statement of no 
significant environmental impact in Sec.  51.32. The NRC believes that 
external stakeholders will better understand the nature of the 
Commission's action by doing so. The NRC is modifying Sec.  51.31 by 
adding Sec.  51.31(b) specifying the information on the environmental 
assessment to be included in the proposed rulemaking on the design 
certification published in the Federal Register.
    The NRC is revising Sec.  51.50(c)(2) to indicate that if a 
combined license application references a design certification then the 
combined license applicant's environmental report may reference the 
SAMDA discussion in the design certification environmental assessment 
as part of its SAMDA analysis, but must contain information 
demonstrating that the site characteristics for the combined license 
site falls within the site parameters in the design certification 
environmental assessment.\12\
---------------------------------------------------------------------------

    \12\ The design certification applicant may have chosen to 
specify site parameters for the design certification safety review 
under Sec.  52.79 which differ from the site parameters specified in 
the environmental report for its design. If such a design 
certification is referenced in a combined license application, the 
combined license applicant must demonstrate that the two differing 
sets of site parameters are met, in order for the full panoply of 
issue finality provisions in Sec.  52.63 to apply in the combined 
license proceeding.
---------------------------------------------------------------------------

    Finally, the NRC is adding Sec.  51.75(c)(2) to provide that if a 
combined license application references a design certification, then 
the combined license EIS will incorporate by reference the design 
certification environmental assessment, and summarize the SAMDA 
analysis and conclusions of the environmental assessment.
NEPA Compliance for Manufacturing Licenses
    The NRC believes that its current approach for meeting the 
Commission's NEPA responsibilities for standard design certifications 
should be extended to manufacturing licenses for nuclear power 
reactors. Under subpart F to part 52, a manufacturing license is 
similar to a standard design certification in that a final nuclear 
power reactor design would be approved. Therefore, the NRC is requiring 
that the environmental effects of construction and operation of a 
nuclear power facility using a manufactured reactor would be addressed 
in the EIS for the combined license application for a nuclear power 
facility using a manufactured reactor, rather than in an environmental 
assessment or EIS at the manufacturing license stage.
    Further, the NRC does not believe that NEPA requires the NRC to 
address the environmental impacts of actually manufacturing a nuclear 
power reactor licensed under subpart F of part 52, either at the 
manufacturing license stage or at the combined license stage where an 
application proposes to use a manufactured reactor. The manufacturing 
license approves the final design of the manufactured reactor, the 
organization and technical procedures for designing and manufacturing 
the reactor, and the ITAAC that are to be used by the licensee in 
determining whether the reactor has been properly manufactured in 
accordance with NRC requirements and the manufacturing license, and the 
possession (but not the use or transport offsite) of the manufactured 
reactor. The manufacturing license does not approve any specific 
location, building, or facility where the actual manufacture of the 
reactors may occur,\13\ and the NRC does not require the applicant for 
the manufacturing license to submit any information on these matters as 
part of its application. These matters are commercial matters generally 
unrelated to the NRC's regulatory jurisdiction. The Federal Aviation 
Administration (FAA) does not prepare an EIS when issuing a production 
certificate under 14 CFR part 21, subpart G, authorizing the production 
of an aircraft or component in conformance with a type certificate. See 
Federal Aviation Agency Order 1050.1E, Sec. 308c (June 8, 2004). 
Because the NRC does not approve any specific location or facility in 
which to manufacture any component of or the reactor licensed under the 
manufacturing license, it would be speculative for the NRC to describe 
and assess the environmental impacts of manufacturing. NEPA does not 
require that an EIS address speculative impacts. The NRC also notes 
that EISs prepared in the past for construction permits and operating 
licenses under part 50, as well as current environmental assessments 
for nuclear power plant license amendments, have never considered the 
offsite environmental impacts of fabricating systems and components by 
vendors and subcontractors, even for circumstances where the 
fabrication activities are subject to NRC regulatory jurisdiction 
(e.g., under applicable provisions of parts 19 and 21). For these 
reasons, the NRC concludes that NEPA does not require the NRC to 
address, either at the manufacturing license stage or at the combined 
license stage where the application proposes to use a manufactured 
reactor, the speculative impacts of manufacturing a reactor offsite at 
a location or in a facility not specified or approved in the 
manufacturing license.
---------------------------------------------------------------------------

    \13\ A reactor manufactured outside of the United States would 
not be within the scope of a manufacturing license under subpart F 
of part 52, by virtue of proposed Sec.  52.9, which states that no 
license shall be deemed to have been issued for activities which are 
not under or within the jurisdiction of the United States.
---------------------------------------------------------------------------

    The NRC is making a number of changes to part 51, in some cases 
parallel to those described previously with respect to design 
certifications, consistent with its views on manufacturing licenses. 
The NRC is revising existing Sec.  51.54 to clarify that an 
environmental report for a manufacturing license must address the costs 
and benefits of SAMDAs and the bases for not incorporating SAMDAs

[[Page 49428]]

into the design of the reactor to be manufactured, and to state that 
the environmental report need not address the impacts of manufacturing 
a reactor under the manufacturing license. The NRC is removing both 
Sec.  51.20(b)(6), which formerly required preparation of an EIS for 
issuance of a manufacturing license, and Sec.  51.76, which formerly 
addressed the subject matter of an EIS for a manufacturing license, 
from part 51.
    The NRC is revising Sec.  51.30(e) to establish the scope of an 
environmental assessment prepared for a manufacturing license. The NRC 
is adding Sec. Sec.  51.32(b)(3) and (4) to state the NRC's generic 
determination of no significant environmental impact associated with 
issuance of a final or amended manufacturing license. As with the 
parallel provisions governing design certifications in Sec.  
50.32(b)(1) and (2), the NRC is including an explicit statement of no 
significant environmental impact for manufacturing licenses in Sec.  
51.32(b)(3) and (4) to facilitate external stakeholders' understanding 
of the nature of the Commission's action. The NRC is adding Sec.  
51.31(c) to describe the NRC's process for determining the 
manufacturing license with respect to environmental issues covered by 
NEPA.
    The NRC is adding Sec.  51.50(c)(3) to provide that if a combined 
license application proposes using a manufactured reactor, then the 
combined license environmental report may incorporate by reference the 
environmental assessment for the manufacturing license under which the 
reactor is to be manufactured and, if so, must include information 
demonstrating that the site characteristics for the combined license 
site fall within the site parameters specified in the manufacturing 
license environmental assessment. This section also states that the 
environmental report need not address the environmental impacts 
associated with manufacturing the reactor under the manufacturing 
license.
    Finally, the NRC is adding Sec.  51.75(c)(3) to indicate that if 
the proposed combined license application to use a manufactured reactor 
and the site characteristics of the combined license's site fall within 
the site parameters specified in the manufacturing license 
environmental assessment,\14\ then the combined license EIS must 
incorporate by reference the manufacturing license environmental 
assessment. As in the case where the combined license application 
references a design certification, Sec.  51.75(c)(3) requires the 
combined license EIS to summarize the findings and conclusions of the 
environmental assessment with respect to SAMDAs. Finally, Sec.  
51.75(c)(3) explicitly provides that the combined license EIS will not 
address the environmental impacts of manufacturing the reactor under 
the manufacturing license.
---------------------------------------------------------------------------

    \14\ Analogous to design certifications, it is possible that an 
applicant for a manufacturing license may have chosen to specify 
site parameters for the manufacturing license safety review under 
Sec.  52.79 which differ from the site parameters specified in the 
environmental report for its design. If the combined license 
application proposes to use such a manufactured reactor, then the 
combined license applicant must demonstrate that the two differing 
sets of site parameters are met, in order for the full division of 
issue finality provisions in Sec.  52.171 to apply in the combined 
license proceeding.
---------------------------------------------------------------------------

NEPA Obligations Associated With Sec.  52.103(g) Findings on ITAAC
    Formerly, neither part 51 nor subpart C of part 52 explicitly 
addressed whether an environmental finding under NEPA is needed in 
connection with an NRC finding under Sec.  52.103(g) that combined 
license ITAAC have been met. Nor does part 51 or subpart C of part 52 
explicitly address whether contentions on environmental matters may be 
admitted in a hearing under Sec.  52.103(b). The NRC never intended to 
make an environmental finding in connection with the Sec.  52.103(g) 
finding on ITAAC, and the NRC does not believe that NEPA requires such 
a finding. The Sec.  52.103(g) finding that ITAAC have been met is not 
a ``major Federal action significantly affecting the environment.'' The 
major Federal action occurs when the NRC issues the combined license, 
which includes the authority to operate the nuclear power plant--
subject to an NRC finding of successful completion of ITAAC. This is 
the reason why the environmental impacts of operation under the 
combined license are evaluated and considered by the NRC in determining 
whether to issue the combined license even under the former provisions 
of part 52, see Sec.  52.89. By contrast, the scope and nature of the 
NRC finding that ITAAC have been met is constrained by the ITAAC itself 
(indeed, the NRC has always recognized the possibility that ITAAC could 
be written such that the ``inspections and tests'' exception in Section 
554(a)(3) of the APA could be invoked to preclude the need to provide 
an opportunity for hearing on Sec.  52.103(g) findings). The safety 
consequences of operation are not considered when making the Sec.  
52.103(g) findings; these issues are addressed by the NRC in 
determining whether to issue the combined license in the first place. 
Therefore, the NRC does not view the Sec.  52.103(g) finding as 
constituting a ``major Federal action,'' and makes no environmental 
findings in connection with that finding. It, therefore, follows that 
no contentions on environmental matters should be admitted in any 
hearing under Sec.  52.103(b).
    Accordingly, the NRC is adding Sec.  51.108 to clarify that: (1) 
The Commission will not make any environmental findings in connection 
with the finding under Sec.  52.103(g); and (2) contentions on any 
environmental matters, including the adequacy of the combined license 
EIS and any referenced environmental assessment, may not be admitted 
into any Sec.  52.103(b) hearing on compliance with ITAAC. Those issues 
are essentially challenges to the continuing validity of the combined 
license or any referenced design certification or manufacturing 
license. Accordingly, these challenges should be raised with the 
Commission using relevant Commission-established processes for 
requesting Commission action. A challenge on environmental grounds with 
respect to the combined license or manufacturing license must be filed 
under the provisions of Sec.  2.206. A challenge to an existing design 
certification on environmental grounds must be filed as a petition for 
rulemaking to modify the existing design certification under subpart H 
of part 2.
NEPA Compliance for Combined Licenses Referencing an Early Site Permit
    The NRC has made several changes in the final rule based on public 
comments regarding the requirements for a combined license application 
referencing an early site permit and further consideration of the NRC's 
obligations under NEPA for such actions. Several commenters believed 
that an ESP and COL met the definition of ``connected actions,'' under 
NEPA case law and Council on Environmental Quality (CEQ) regulations, 
and should therefore not require the preparation of a new EIS for the 
second of the two connected actions, or a revalidation of previous 
findings if neither the applicant nor others identify new and 
significant information. Commenters stated that under applicable NEPA 
case law, there was no requirement to prepare a new EIS for the latter 
of the two connected actions that were previously evaluated together in 
a single EIS. The commenters stated that the EIS prepared at the ESP 
stage serves as the EIS for issuance of both the ESP and COL. 
Commenters stated that the ESP EIS included an evaluation of the 
environmental impacts related to

[[Page 49429]]

issuance of a COL inasmuch as it considered the environmental impact of 
plant construction and operation.
    The NRC continues to believe that it is not necessary to require 
that all topics be covered in a single EIS at the ESP stage, and that 
topics such as alternative energy sources and need for power may be 
treated in an EIS supplement at the COL application stage when the 
detailed planning for the project is completed. As the commenters note, 
new and significant information may also prompt the preparation of a 
supplement to the ESP EIS in connection with the COL application. Since 
the NRC believes that some issues may not be ripe for consideration at 
the ESP stage, and an ESP EIS need not address such issues, the 
Commission is declining to take a position on whether the granting of 
an ESP and the granting of a COL referencing that ESP are connected 
actions. Nevertheless, the Commission believes that, inasmuch as an 
early site permit and a combined license are major Federal actions 
significantly affecting the quality of the human environment, both 
actions require the preparation of an EIS. However, 10 CFR part 52 does 
provide finality for previously resolved issues. Under NEPA, the 
combined license environmental review is informed by the EIS prepared 
at the ESP stage and the NRC staff intends to incorporate by reference 
the ESP EIS in the combined license supplemental EIS. A description of 
what the combined license applicant must address in this situation can 
be found under the discussion of changes to Sec.  51.50(c)(1).
    More specific changes to individual sections in part 51 are 
discussed as follows:
1. Section 51.20, Criteria for and Identification of Licensing and 
Regulatory Actions Requiring Environmental Impact Statements
    The NRC is revising Sec.  51.20(b) to identify the part 52 
licensing processes that require an EIS or a supplement to an EIS. 
Specifically, the NRC is revising Sec.  51.20(b)(1) to indicate that 
issuance of an early site permit requires an EIS. The NRC is revising 
Sec.  51.20(b)(2) to indicate that issuance of a combined license 
requires an EIS. Also, paragraph (b)(6) is being removed and reserved 
because, under the Commission's proposed revision to the requirements 
for manufacturing licenses, only an environmental assessment is 
required at this stage.
2. Section 51.22, Criterion for Categorical Exclusion; Identification 
of Licensing and Regulatory Actions Eligible for Categorical Exclusion 
or Otherwise Not Requiring Environmental Review
    The NRC is revising Sec.  51.22(c) to identify part 52 licensing 
processes that are eligible for categorical exclusion or otherwise do 
not require environmental review.
3. Section 51.23, Temporary Storage of Spent Fuel After Cessation of 
Reactor Operation--Generic Determination of No Significant 
Environmental Impact
    The NRC is revising Sec. Sec.  51.23(b) and(c) to indicate that the 
provisions of these paragraphs also apply to combined licenses.
4. Section 51.26, Requirement To Publish Notice and Conduct Scoping 
Process
    The NRC is adding a new paragraph (d) to this section to provide 
requirements for publication of a notice of intent when the NRC 
determines that a supplement to an EIS will be prepared. This new 
provision also states that, in such cases, the NRC staff need not 
conduct a scoping process, provided, however, that if scoping is 
conducted, then the scoping must be directed at matters to be addressed 
in the supplement. If scoping is conducted in a proceeding for a 
combined license referencing an ESP under part 52 , then the scoping 
must be directed at matters to be addressed in the supplement as 
described in Sec.  51.92(e).
5. Section 51.27, Notice of Intent
    The NRC is adding a new paragraph (b) to this section to provide 
requirements for the contents of a notice of intent when the NRC 
determines that a supplement to an EIS will be prepared. Paragraph (b) 
states that the notice of intent will, among other things, describe the 
matters to be addressed in the supplement to the final EIS and describe 
any proposed scoping process that the NRC staff may conduct.
6. Section 51.29, Scoping-Environmental Impact Statement and Supplement 
to Environmental Impact Statement
    The NRC is revising paragraph (a)(1) of this section in the final 
rule to include requirements for supplements to an ESP EIS prepared for 
a combined license application.
7. Section 51.45, Environmental Report
    The NRC is revising Sec.  51.45(c) to indicate that the analysis in 
an environmental report prepared for an ESP need not include 
consideration of the economic, technical, and other benefits and costs 
of the proposed action and of energy alternatives. This change is being 
made for consistency with the provisions of Sec.  51.50(b), which state 
that an environmental report included in an ESP application need not 
include an assessment of the benefits (e.g., need for power) of the 
proposed action and with the Commission's denial of a Petition for 
Rulemaking (See PRM-52-02 (October 28, 2003; 68 FR 55905)).
8. Section 51.50, Environmental Report--Construction Permit, Early Site 
Permit, or Combined License Stage
    The NRC is revising the title of Sec.  51.50 to ``Environmental 
Report Construction Permit, Early Site Permit, or Combined License 
Stage,'' and including separate paragraphs with specific requirements 
for environmental reports for early site permit and combined license 
applications which are based on existing requirements in part 51 for 
construction permits and operating licenses and requirements for early 
site permits and combined licenses in part 52.
    The NRC is revising the requirements from former Sec.  52.17(a)(2) 
to clarify that an early site permit applicant has the flexibility of 
either addressing the matter of alternative energy sources in the 
environmental report supporting its early site permit application, or 
deferring consideration of alternative energy sources to the time that 
the early site permit is referenced in a licensing application. The NRC 
believes the former regulations already afforded the early site permit 
applicant such flexibility, inasmuch as former Sec.  52.17(a)(2) stated 
that the environmental report submitted in support of an early site 
permit 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 
constitute part of the discussion of reasonable alternatives to the 
proposed action, which is required by Section 102(2)(C)(iii) of NEPA. 
[See 10 CFR 51.71(e) n.4; 46 FR 39440 (August 3, 1981) (proposed rule 
that would eliminate consideration of need for

[[Page 49430]]

power and alternative energy sources at operating license stage), at 
39441 (first column) (final rule published March 26, 1982; 47 FR 
12940).] See Exelon Generation Company, LLC et al., CLI-05-17, 62 NRC 
5, where the Commission ruled that:

    [T]he ``reasonable alternatives'' issue does not apply with full 
force to ESP (or ``partial'' construction permit) cases. At the ESP 
stage of the construction permit process, the boards' ``reasonable 
alternatives'' responsibilities are limited because the proceeding 
is focused on an appropriate site, not the actual construction of a 
reactor. Thus, boards must merely weigh and compare alternative 
sites, not other types of alternatives (such as alterative energy 
sources). (Id. at 48 (citations omitted).)

    Accordingly, the NRC believes that former Sec.  52.17(a)(2) already 
provided the early site permit applicant the flexibility of choosing to 
defer consideration of alternative energy sources to the time that the 
early site permit is referenced in a combined license or a construction 
permit application. The revisions in Sec.  51.50(b) clarify that the 
early site permit applicant may either include a discussion of 
alternative energy sources in its environmental report, or defer 
consideration of the matter. The NRC made conforming amendments 
elsewhere in part 51 to clarify that the NRC's EIS need not address the 
need for power or alternative energy sources (and therefore these 
matters may not be litigated) if the early site permit applicant 
chooses not to address these matters in its environmental report. The 
environmental report and EIS for an early site permit must address the 
benefits associated with issuance of the early site permit (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 characteristics, 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 early site 
permit must always be addressed in the environmental report and EIS for 
an early site permit, regardless of whether the early site permit 
applicant chooses to defer consideration of the benefits associated 
with the construction and operation of a nuclear power plant that may 
be located at the early site permit site. This is because the 
``benefits * * * of the proposed action'' for which the discussion may 
be deferred are the benefits associated with the construction and 
operation of a nuclear power plant that may be located at the early 
site permit site; the benefits which may be deferred are entirely 
separate from the benefits of issuing an early site permit. The 
proposed action of issuing an early site permit 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.  51.50(b).
    The NRC is further modifying Sec.  51.50(b) in the final rule based 
on public comments. This section addresses requirements for 
environmental reports at the early site permit stage. In the proposed 
rule, Sec.  51.50(b) stated that environmental reports ``must focus on 
the environmental effects of construction and operation of a reactor, 
or reactors, which have characteristics that fall within the postulated 
site parameters.'' Commenters pointed out that the use of ``postulated 
site parameters'' was not consistent with the terminology the NRC had 
used elsewhere in the proposed rule. Consequently, the NRC is revising 
this provision in the final rule to require that the environmental 
report ``must focus on the environmental effects of construction and 
operation of a reactor, or reactors, which have design characteristics 
that fall within the site characteristics and design parameters for the 
early site permit application.'' A similar change is being made to the 
same language in final rule Sec.  51.75(b) [proposed Sec.  51.71(d)].
    The NRC is making additional changes to Sec.  51.50(b) to further 
clarify the scope of the environmental review at the early site permit 
stage. Final Sec.  51.50(b)(2) states that an early site permit 
environmental report may address one or more of the environmental 
effects of construction and operation of a reactor, or reactors, which 
have design characteristics that fall within the site characteristics 
and design parameters for the early site permit application, but that 
the environmental report must address all environmental effects of 
construction and operation necessary to determine whether there is any 
obviously superior alternative to the site proposed. The purpose of 
this change is to clearly delineate that the scope of the environmental 
review at the early site permit stage is, at a minimum, to address all 
issues needed for the NRC to perform its evaluation of the alternative 
sites. In addition, the applicant may choose to address one or more 
issues related to construction and operation of the facility with the 
goal of achieving finality on those issues at the early site permit 
stage.
    In addition, the NRC is modifying Sec. Sec.  51.50(b) and 51.50(c) 
in the final rule to reflect comments made at the NRC's public 
workshops during the public comment period on the proposed rule. These 
discussions related to the requirement to include a proposed list of 
activities and a redress plan in license applications that request 
authority to perform activities under Sec.  50.10(e). The NRC concluded 
that it is preferable to include both the list of proposed activities 
and the redress plan as separate documents in the application, outside 
of both the final safety analysis report (or site safety analysis 
report in the case of an early site permit) and the environmental 
report. The NRC's conclusion is based on the fact that the requirements 
in Sec.  50.10(e) address both safety and environmental issues. 
Additional changes were made to Sec. Sec.  52.17(c), 52.79(a), and 
52.80 to implement this concept.
    The NRC is also revising Sec.  51.50(c) based on public comments in 
the final rule. These revisions address the situation where a combined 
license applicant is referencing an early site permit and provide for a 
clearer link to the finality provisions in Sec.  52.39, eliminate 
language that attempted to define ``new and significant,'' and provide 
greater consistency with related requirements elsewhere in part 51. The 
revisions also provide requirements for addressing environmental terms 
and conditions. The discussion that follows reflects the language in 
the final rule.
    The NRC is adding a requirement in Sec.  51.50(c)(1) that the 
applicant's environmental report need not contain information or 
analyses submitted to the Commission in the early site permit 
environmental report or resolved in the Commission's early site permit 
environmental impact statement, but must contain, in addition to the 
environmental information and analyses otherwise required: (1) 
Information to demonstrate that the design of the facility falls within 
the site characteristics and design parameters specified in the early 
site permit; (2) information to resolve any significant environmental 
issue that was not resolved in the early site permit proceeding; (3) 
any new and significant information for issues related to the impacts 
of construction and operation of the facility that were resolved in the 
early site permit proceeding; (4) a description of the process used to 
identify new and significant information regarding the NRC's 
conclusions in the early site permit environmental impact statement, 
including a requirement that the process use a reasonable

[[Page 49431]]

methodology for identifying such new and significant information; and 
(5) a demonstration that all environmental terms and conditions that 
have been included in the early site permit will be satisfied by the 
date of issuance of the combined license. Any terms or conditions of 
the early site permit that cannot be met by the time the combined 
license is issued must be set forth as terms or conditions of the 
combined license.
    For an early site permit, the NRC prepares an EIS that resolves 
numerous issues within certain bounding conditions. These issues have 
issue preclusion at the combined license or CP stage provided certain 
conditions are met. A combined license or CP application must 
demonstrate that the design of the facility falls within the site 
characteristics and design parameters specified in the early site 
permit. In addition, the application must include any new and 
significant information for issues related to the impacts of 
construction and operation of the facility (i.e., the issue being 
addressed at the combined license stage) that were resolved in the 
early site permit proceeding. Documentation related to the applicant's 
search for new information and its determination about the significance 
of the new information should be maintained in an auditable form by the 
applicant. The NRC staff may also use the environmental scoping process 
to assist it in determining if there is new and significant information 
regarding issues that were resolved in the early site permit 
proceeding. Although the NRC is ultimately responsible for completing 
any required NEPA review under 10 CFR 51.70(b), for example, an 
evaluation of the impact of new and significant information on the 
conclusions for a resolved early site permit environmental issue, the 
combined license applicant must identify whether there is new and 
significant information on such an issue. A combined license applicant 
should have a reasonable process to ensure it becomes aware of new and 
significant information that may have a bearing on the earlier NRC 
conclusion, and should document the results of this process in an 
auditable form. The NRC staff will verify that the applicant's process 
for identifying new and significant information is effective.
    The NRC, in the context of a combined license application that 
references an early site permit, has defined the term ``new'' in the 
phrase ``new and significant information'' as any information that was 
both (1) not considered in preparing the ESP environmental report or 
EIS (as may be evidenced by references in these documents, applicant 
responses to NRC requests for additional information, comment letters, 
etc.) and (2) not generally known or publicly available during the 
preparation of the EIS (such as information in reports, studies, and 
treatises). For new information to be ``significant,'' it must be 
material to the issue being considered, that is, it must have the 
potential to affect the finding or conclusions of the NRC staff's 
evaluation of the issue. The COL applicant need only provide 
information about a previously resolved environmental issue if it is 
both new and significant.
    The combined license applicant referencing an early site permit is 
also required to provide information sufficient to resolve any other 
significant environmental issue not considered in the early site permit 
proceeding (e.g., need for power) and the information contained in the 
application should be sufficient to aid the staff in its development of 
an independent analysis (see 10 CFR 51.45).
    Finally, the combined license applicant referencing an early site 
permit must demonstrate that all environmental terms and conditions 
included in the early site permit will be satisfied by the date of 
issuance of the combined license. In some cases, this may require 
adding a condition to the combined license to adequately address the 
environmental issue raised in the early site permit condition. Note 
that this provision was added to Sec.  51.50(c)(1) in the final rule. 
Requirements to include environmental conditions in an early site 
permit environmental report were addressed in the proposed rule in 
Sec.  51.50(b), but the associated provision to ensure any conditions 
included in the permit would be met was inadvertently left out of Sec.  
51.50(c)(1).
    In the past, the NRC staff has attempted to explain the 
relationship between the environmental review of an early site permit 
application to that of a combined license application referencing the 
early site permit by analogy to the license renewal environmental 
review process. The NRC believes the analogy especially useful because 
the license renewal process is well-established and clearly understood. 
Because there appears to be some confusion regarding this analogy, NRC 
believes a brief explanation of the similarities of the two processes 
is warranted.
    For license renewal, the NRC prepared a generic EIS (GEIS) that 
resolved more than 60 issues for all plants based on certain bounding 
assumptions. These were termed Category 1 issues. If a license renewal 
applicant identifies new and significant information with respect to a 
Category 1 issue, it documents its assessment of that information in 
its application. If the applicant determines that this new information 
is not significant, or that there is no new information, the applicant 
documents the bases for these determinations in an auditable form and 
makes the documentation available for staff inspection. If there is new 
and significant information on a Category 1 issue, the NRC staff limits 
its inquiry to determine if this information changes the Commission's 
earlier conclusion set forth in the GEIS. The NRC staff may inquire if 
the applicant has a reasonable process for identifying new and 
significant information on Category 1 issues.
    Similarly, in the NRC environmental review process for a combined 
license application, the combined license EIS brings forward the 
Commission's earlier conclusions from the early site permit EIS and 
articulates the activities undertaken by the NRC staff to ensure that 
an issue that was resolved can remain resolved. If there is new and 
significant information on a previously resolved issue, then the staff 
will limit its inquiry to determine if the information changes the 
Commission's earlier conclusion. Environmental matters subject to 
litigation in a combined license proceeding mainly include (1) those 
issues that were not considered in the previous proceeding on the site 
or the design; (2) those issues for which there is new and significant 
information; and (3) those issues subject to the change or exemption 
processes in 10 CFR part 52.
    Notwithstanding that, in the context of renewal, the GEIS resolves 
Category 1 issues through rulemaking and an early site permit resolves 
environmental issues through an individual licensing proceeding, the 
staff believes that the license renewal practice is similar to the part 
52 process in which a combined license application references an early 
site permit.
    The NRC has determined that a combined license is a major Federal 
action significantly affecting the quality of the human environment 
and, in accordance with 10 CFR 51.20, the NRC must prepare an EIS on 
that action. If there is no new and significant information for matters 
resolved at the ESP stage, then the staff will rely upon (``tier off'') 
the ESP EIS at the combined license stage and disclose the NRC 
conclusion for matters covered in the early site permit review. Such 
matters

[[Page 49432]]

will not be subject to litigation at the combined license stage.
9. Section 51.51, Uranium Fuel Cycle Environmental Data--Table S-3
    The NRC is revising Sec.  51.51 to require that every environmental 
report prepared for the early site permit stage or combined license 
stage of a light-water-cooled nuclear power reactor use Table S-3, 
Table of Uranium Fuel Cycle Environmental Data, as the basis for 
evaluating the contribution of the environmental effects of the uranium 
fuel cycle to the environmental costs of licensing light-water-cooled 
nuclear power reactors. If the application for a combined license 
references an early site permit in which the environmental impacts and 
costs related to the uranium fuel cycle were already evaluated and 
resolved, then the repetition of this information in the environment 
report for the combined license is not required unless the applicant 
has identified new and significant information regarding these 
environmental impacts and costs.
10. Section 51.52, Environmental Effects of Transportation of Fuel and 
Waste--Table S-4
    The NRC is revising Sec.  51.52 to require that every environmental 
report prepared for the early site permit stage or combined license 
stage of a light-water-cooled nuclear power reactor contain a statement 
concerning transportation of fuel and radioactive wastes to and from 
the reactor. If the application for a combined license references an 
early site permit in which the transportation of fuel and radioactive 
wastes to and from the reactor has already been evaluated and resolved, 
then the repetition of this information in the environment report for 
the combined license is not necessary unless the applicant has 
identified new and significant information regarding the associated 
environmental impacts.
11. Section 51.53, Postconstruction Environmental Reports
    The NRC is revising Sec.  51.53(a) to clarify that any 
postconstruction environmental report may incorporate by reference any 
information contained in a prior environmental report or supplement 
thereto that relates to the site or any information contained in a 
final environmental document previously prepared by the NRC staff that 
relates to the site. This change reflects the recognition that 
environmental documents will be prepared at the early site permit stage 
and may be referenced in environmental documents for future licensing 
actions. The NRC is also revising Sec.  51.53(a) to clarify that 
documents that may be referenced in post-construction environmental 
reports include those prepared in connection with an early site permit 
or a combined license. In addition, the NRC is revising Sec.  
51.53(c)(3) to clarify that the requirements for the content of 
environmental reports submitted in applications for renewal of a 
combined license are the same as those for renewal of an operating 
license.
12. Section 51.54, Environmental Report--Manufacturing License
    The NRC is revising this section by adding two paragraphs to 
delineate the difference in the matters with respect to SAMDAs that 
must be addressed in an environmental report for issuance of a 
manufacturing license under subpart F of part 52, versus that for an 
amendment to the manufacturing license. Section 51.54(a) provides that 
the environmental report for the manufacturing license must address the 
costs and benefits of SAMDAs, and the bases for not incorporating into 
the design of the manufactured reactor any SAMDAs identified during the 
applicant's review. Section 51.54(b) reflects the narrower scope of an 
environmental report submitted in connection with a proposed amendment 
to a manufacturing license, by providing that the report need only 
address whether the design change which is subject of a proposed 
amendment either renders a SAMDA previously identified and rejected to 
become cost beneficial, or results in the identification of new SAMDAs 
that may be reasonably incorporated into the design of the manufactured 
reactors.
    As discussed earlier, the environmental impacts of manufacturing a 
reactor under a manufacturing license are not considered by the NRC, 
and Sec.  51.54 indicates that the environmental report need not 
include a discussion of the environmental impacts of manufacturing a 
reactor.
13. Section 51.55, Environmental Report--Standard Design Certification
    The NRC is transferring the provisions in current Sec.  51.55 to a 
new Sec.  51.58 (discussed in Sec.  51.58), and the NRC is revising 
this section to address the contents of environmental reports for 
design certifications under subpart B of part 52. The structure of new 
Sec.  51.55 is similar to that of Sec.  51.54, reflecting the fact that 
the environmental review for either manufacturing licenses or design 
certifications is limited to SAMDAs. Section 51.55(a) provides that the 
environmental report for the design certification must address the 
costs and benefits of SAMDA, and the bases for not incorporating into 
the design certification any SAMDAs identified during the applicant's 
review. Section 51.55(b) provides that the environmental report 
submitted in support of a request to amend a design certification need 
only address whether the design change which is the subject of a 
proposed amendment either renders a SAMDA previously identified and 
rejected to become cost beneficial, or results in the identification of 
new SAMDAs that may be reasonably incorporated into the design 
certification.
14. Section 51.58, Environmental Report--Number of Copies; Distribution
    The matters previously addressed in Sec.  51.55 are addressed in a 
new Sec.  51.58. The NRC is adding conforming references to Sec.  
51.58(a) for early site permits and combined licenses. Section 51.58(b) 
contains a conforming reference to subpart F of part 52.
15. Section 51.71, Draft Environmental Impact Statement--Contents
    The NRC is revising Sec.  51.71(d) to include a reference to Sec.  
51.75 in the first sentence because Sec.  51.75 also includes 
exceptions to the provisions in Sec.  51.71(d). This represents a 
change the NRC is making in the final rule to move the specific 
discussions on early site permits and combined licenses from Sec.  
51.71(d) to their associated paragraphs in Sec.  51.75. The NRC is also 
revising associated footnote 3 to include references to early site 
permits and combined licenses.
16. Section 51.75, Draft Environmental Impact Statement--Construction 
Permit, Early Site Permit, or Combined License
    The NRC is adding Sec. Sec.  51.75(b) and (c) to include separate 
requirements for the preparation of draft EISs at the early site permit 
and combined license stages. In the final rule, the NRC is also moving 
information related to early site permits that was contained in 
proposed Sec.  51.71(d) to Sec.  51.75(b). In addition, the NRC is 
providing further clarification in the final rule on the scope of the 
environmental review at the early site permit stage. Section 51.75 
requires that the draft environmental impact statement must include an 
evaluation of alternative sites to determine whether there is any 
obviously superior alternative to the site proposed. The draft 
environmental impact statement must also include an evaluation of the 
environmental effects of construction

[[Page 49433]]

and operation of a reactor, or reactors, which have design 
characteristics that fall within the site characteristics and design 
parameters for the early site permit application, but only to the 
extent addressed in the early site permit environmental report or 
otherwise necessary to determine whether there is any obviously 
superior alternative to the site proposed. The purpose of this change 
is to clearly delineate that the scope of the environmental review at 
the early site permit stage is, at a minimum, to address all issues 
needed for the NRC to perform its evaluation of the alternative sites. 
In addition, the applicant may choose to address one or more issues 
related to construction and operation of the facility with the goal of 
achieving finality on those issues at the early site permit stage. The 
NRC also notes that, where the early site permit application identifies 
a specific nuclear power reactor design (i.e., a standard design 
certification or manufacturing license) under Sec.  52.17(a)(1)(i), the 
environmental report for an early site permit may address the 
applicability of the severe accident mitigation design alternatives 
(SAMDA) evaluation for that reactor design to the proposed site. In 
this situation, the early site permit EIS must determine whether the 
site characteristics bound the site parameters relevant to the SAMDA 
analysis, as specified in the environmental assessment for the 
identified nuclear power reactor design.
    The requirements for combined licenses are organized into separate 
paragraphs (c)(1), (c)(2), and (c)(3) which address the contents of the 
combined license environmental impact statement if the combined license 
application references an early site permit or standard design 
certification, or proposes to use a manufactured reactor. For example, 
Sec.  51.75(c)(3) provides that the combined license EIS will not 
address the environmental impacts associated with manufacturing the 
reactor under the manufacturing license.
    In the final rule, Sec.  51.75(c)(1) states that if a combined 
license application references an early site permit, then the NRC staff 
shall prepare a supplement to the early site permit EIS. Paragraph 
(c)(1) also requires that the supplement be prepared in accordance with 
Sec.  51.92. Section 51.92 contains the requirements for the content of 
a supplemental EIS prepared for a combined license application that 
references an early site permit.
17. Section 51.92, Supplement to the Final Environmental Impact 
Statement
    The NRC is revising Sec.  51.92 in the final rule to provide 
requirements for NRC staff preparation of a supplement to the final 
environmental impact statement for an early site permit as required by 
Sec.  51.75(c)(1). Paragraph (b) of Sec.  51.92 states that, in a 
proceeding for a combined license application referencing an early site 
permit, the NRC staff shall prepare a supplement to the final 
environmental impact statement for the referenced early site permit in 
accordance with Sec.  51.92(e). In the final rule, the NRC is moving 
information related to combined licenses that was contained in proposed 
Sec.  51.71(d) to Sec.  51.92(e) and is revising the wording of this 
provision. In the proposed rule, Sec.  51.71(d) stated that the draft 
supplemental environmental impact statement prepared at the combined 
license stage when an early site permit is referenced need not include 
detailed information or analyses that were resolved in the final 
environmental impact statement prepared by the Commission in connection 
with the early site permit, provided that the design of the facility 
falls within the design parameters specified in the early site permit, 
the site falls within the site characteristics specified within the 
early site permit, and there is no new and significant environmental 
issue or information not considered on the site or the design only to 
the extent that they differ from that discussed in the final 
environmental impact statement prepared by the Commission in connection 
with the early site permit. In the final rule, the NRC has modified 
these provisions and moved them to Sec.  51.92(e). The revised language 
in paragraph (e) provides a clearer link to the finality provisions in 
Sec.  52.39, eliminates language in the proposed rule that attempted to 
define ``new and significant,'' and provides greater consistency with 
related requirements elsewhere in part 51. Specifically, paragraph (e) 
requires that a supplement to an early site permit final environmental 
impact statement must: (1) Identify the proposed action as the issuance 
of a combined license for the construction and operation of a nuclear 
power plant as described in the combined license application at the 
site described in the early site permit referenced in the combined 
license application; (2) incorporate by reference the final 
environmental impact statement prepared for the early site permit; (3) 
contain no separate discussion of alternative sites; (4) include an 
analysis of the economic, technical, and other benefits and costs of 
the proposed action, to the extent that the final environmental impact 
statement prepared for the early site permit did not include an 
assessment of these benefits and costs; (5) include an analysis of 
other energy alternatives, to the extent that the final environmental 
impact statement prepared for the early site permit did not include an 
assessment of energy alternatives; (6) include an analysis of any 
environmental issue related to the impacts of construction or operation 
of the facility that was not resolved in the proceeding on the early 
site permit; and (7) include an analysis of the issues related to the 
impacts of construction and operation of the facility that were 
resolved in the early site permit proceeding for which new and 
significant information has been identified, including, but not limited 
to, new and significant information demonstrating that the design of 
the facility falls outside the site characteristics and design 
parameters specified in the early site permit.
18. Section 51.95, Postconstruction Environmental Impact Statements
    The NRC is revising Sec.  51.95(a) to indicate that documents that 
may be referenced in a supplement to a final environmental impact 
statement include documents prepared in connection with an early site 
permit or combined license. In addition, the NRC is revising Sec.  
51.95(c) to add provisions for renewal of combined licenses and to 
correct the address for the NRC Public Document Room. The NRC is 
revising Sec.  51.95 to indicate that the NRC will prepare a 
supplemental environmental impact statement in connection with the 
amendment of a combined license authorizing decommissioning activities 
or with the issuance, amendment, or renewal of a license to store spent 
fuel at a nuclear power reactor after expiration of the combined 
license, and that the supplement may incorporate by reference any 
information contained in the final environmental impact statement for 
the combined license or in the records of decision prepared in 
accordance with an early site permit or combined license. Finally, the 
NRC is revising Sec.  51.95(d) to indicate that, unless otherwise 
required by the Commission, in accordance with the provisions of Sec.  
51.23(b), a supplemental environmental impact statement for the post 
combined license stage will address the environmental impacts of spent 
fuel storage only for the term of the license, amendment, or renewal 
applied for.

[[Page 49434]]

19. Section 51.105, Public Hearings in Proceedings for Issuance of 
Construction Permits or Early Site Permits
    The NRC is revising the section heading and Sec.  51.105(a) to 
indicate that the requirements for presiding officers in public 
hearings on construction permits also apply to public hearings on early 
site permits. In addition, the NRC is adding Sec.  51.105(b) to 
indicate that the presiding officer in an early site permit hearing 
shall not admit contentions concerning the benefits assessment (e.g., 
need for power), or alternative energy sources if the applicant did not 
address those issues in the early site permit application. This change 
is being made for consistency with the provisions of Sec.  51.50(b), 
which state that an environmental report included in an early site 
permit application need not include an assessment of the benefits 
(e.g., need for power) of the proposed action, and with the 
Commission's denial of a Petition for Rulemaking (See PRM-52-02 
(October 28, 2003; 68 FR 55905)). The NRC notes that the environmental 
report and EIS for an early site permit must address the benefits 
associated with issuance of the early site permit (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 characteristics, 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 early site 
permit must always be addressed in the environmental report and EIS for 
an early site permit, regardless of whether the early site permit 
applicant chooses to defer consideration of the benefits associated 
with the construction and operation of a nuclear power plant that may 
be located at the early site permit site. This is because the 
``benefits * * * of the proposed action'' for which the discussion may 
be deferred are the benefits associated with the construction and 
operation of a nuclear power plant that may be located at the early 
site permit site; the benefits which may be deferred are entirely 
separate from the benefits of issuing an early site permit. The 
presiding officer needs to be mindful of whether the applicant has 
addressed only the benefits of issuing the early site permit or whether 
the applicant has also addressed all of the benefits of construction 
and operation of the facility. This is because the presiding officer, 
in accordance with Sec.  51.105(a)(3), must determine, after weighing 
the environmental, economic, technical, and other benefits against 
environmental and other costs, and considering reasonable alternatives, 
whether the early site permit should be issued, denied, or 
appropriately conditioned to protect environmental values. If the 
applicant has addressed all of the costs and benefits associated with 
construction and operation of the facility in its environmental report, 
the final balancing between costs and benefits needs to occur at the 
early site permit stage.
    The NRC also notes that, where the early site permit application 
identifies a specific nuclear power reactor design (i.e., a standard 
design certification or manufacturing license) under Sec.  
52.17(a)(1)(i), the environmental report for an early site permit may 
address the applicability of the severe accident mitigation design 
alternatives evaluation for that reactor design to the proposed site. 
In this situation, the early site permit EIS must determine whether the 
site characteristics bound the site parameters relevant to the SAMDA 
analysis, as specified in the environmental assessment for the 
identified nuclear power reactor design. In addition, in accordance 
with Section 52.107(c), the presiding officer shall not admit 
contentions proffered by any party concerning severe accident 
mitigation design alternatives unless the contention demonstrates that 
the site characteristics fall outside of the site parameters in the 
standard design certification or underlying manufacturing license for 
the manufactured reactor.
20. Section 51.105a, Public Hearings in Proceedings for Issuance of 
Manufacturing Licenses
    The NRC is adding Sec.  51.105a to provide requirements for public 
hearings in proceedings for issuance of manufacturing licenses. 
Specifically, Sec.  51.105a establishes that the presiding officer in a 
proceeding for a manufacturing license will determine whether the 
manufacturing license should be issued as proposed by the appropriate 
NRC staff director.
21. Section 51.107, Public Hearings in Proceedings for Issuance of 
Combined Licenses
    The NRC is adding Sec.  51.107 to set out the requirements for 
public hearings in proceedings for issuance of combined licenses. The 
requirements parallel the associated requirements for public hearings 
on construction permits and operating licenses, as appropriate, and 
provide requirements unique to the combined license process that are 
derived from various provisions in part 52, namely Sec. Sec.  52.39 and 
52.103. The NRC is making changes to the language in Sec.  51.107 in 
the final rule to more clearly define the role of the presiding officer 
in a proceeding for the issuance of a combined license where an early 
site permit is being referenced. Specifically, paragraph (b) addresses 
the situation where a combined license application references an early 
site permit and a supplement to the early site permit environmental 
impact statement is prepared in accordance with Sec.  51.75(c)(1) and 
Sec.  51.92(e). In such cases, the presiding officer in the combined 
license hearing shall not admit any contention proffered by any party 
on environmental issues which have been accorded finality under Sec.  
52.39 unless the contention: (1) Demonstrates that the nuclear power 
reactor proposed to be built does not fit within one or more of the 
site characteristics or design parameters included in the early site 
permit; (2) raises any significant environmental issue that was not 
resolved in the early site permit proceeding; or (3) raises any issue 
involving the impacts of construction and operation of the facility 
that was resolved in the early site permit proceeding for which new and 
significant information has been identified.

N. Changes to 10 CFR Part 54

1. Section 54.1, Purpose
    This part applies to renewed operating licenses for nuclear power 
plants. A conforming change is made to this section to include renewed 
combined licenses.
2. Section 54.3, Definitions
    The definition for renewed combined license is added to explain the 
meaning of the new phrase as it is used in this part.
3. Section 54.17, Filing of Application
    Section 54.17(c) is revised to add a conforming reference to 
combined licenses issued under 10 CFR part 52.
4. Section 54.27, Hearings
    This section is revised to include a conforming reference to 
renewed combined license issued under 10 CFR part 52.
5. Section 54.31, Issuance of a Renewed License
    Sections 54.31(a), (b), and (c) are revised to include conforming 
references to combined licenses in this procedure on issuance of 
renewed licenses.

[[Page 49435]]

6. Section 54.35, Requirements During Term of Renewed License
    This section is revised to include conforming references to holders 
of combined licenses and the regulations in part 52 into the 
requirements for a renewed license.
7. Section 54.37, Additional Records and Recordkeeping Requirements
    Section 54.37(a) is revised to include a conforming reference to a 
renewed combined license.

O. Changes to 10 CFR Part 55

    Part 55 establishes the NRC's requirements for licensing of 
operators of utilization facilities in accordance with the statutory 
requirements in Section 202 of the ERA. Formerly, the provisions in 
part 55 referred only to utilization facilities licensed under part 50, 
and therefore, do not address utilization facilities licensed for 
operation under a combined license issued under subpart C of part 52. 
Section 202 of the ERA, however, does not limit its mandate to 
operators of facilities licensed under part 50; the statutory 
requirement would also appear to apply to operators of facilities 
licensed under part 52 (i.e., combined licenses under subpart C of part 
52).
    Accordingly, Sec. Sec.  55.1 and 55.2 are revised by adding a 
reference to part 52. This clarifies that each operator of a nuclear 
power reactor licensed under a part 52 combined license or renewed 
under part 54 must first obtain an operator's license under part 55. In 
addition, the conforming changes clarify that these operators, as well 
as holders of combined licenses issued under part 52 or renewed under 
part 54, are subject to the requirements in part 55 (e.g., part E of 
part 55, Written Examinations and Operating Tests, set forth 
requirements which are directed, for the most part, at the holders of 
operating licenses for utilization facilities).

P. Changes to 10 CFR Part 72

1. Section 72.210, General License Issued
    Part 72 sets forth the requirements for independent spent fuel 
storage facilities. This section is revised to include a conforming 
reference to persons authorized to operate nuclear power reactors under 
10 CFR part 52 (i.e., a combined license holder).
2. Section 72.218, Termination of Licenses
    Section 72.218(b) is revised to include a conforming reference to 
combined licenses issued under part 52.

Q. Changes to 10 CFR Part 73

    Part 73 establishes the NRC's requirements for the physical 
protection of production and utilization facilities licensed by the 
NRC. It provides requirements for the physical protection of licensed 
activities, for personnel access authorization, and for criminal 
history checks of individuals granted unescorted access to a nuclear 
power facility or access to Safeguards Information. Formerly, the 
language of Sec.  73.1, Purpose and scope, Sec.  73.2, Definitions, 
Sec.  73.50, Requirements for physical protection of licensed 
activities, Sec.  73.56, Personnel access authorization requirements 
for nuclear power plants, and Sec.  73.57, Requirements for criminal 
history checks of individuals granted unescorted access to a nuclear 
power facility or access to Safeguards Information by power reactor 
licensees, and Appendix C, Licensee Safeguards Contingency Plans, did 
not refer to combined licenses issued under part 52. However, part 73 
was formerly applicable to combined licenses under the provisions of 
Sec.  52.83, Applicability of part 50 provisions, which states that all 
provisions of 10 CFR part 50 and its appendices applicable to holders 
of operating licenses also apply to holders of combined licenses. 
Accordingly, Sec.  73.1 is revised to clarify that the regulations in 
part 73 apply to persons who receive combined licenses under part 52, 
and Sec.  73.2 is revised to state that terms defined in part 52 have 
the same meaning when used in part 73. The NRC has addressed combined 
licenses in Sec.  73.57 by making the provisions that are required 
before receiving an operating license under part 50 applicable before 
the date that the Commission makes the finding under Sec.  52.103 for a 
combined license. Additional conforming changes to include part 52 
licenses are made for Sec. Sec.  73.50 and 73.56, and appendix C to 
part 73.

R. Change to 10 CFR Part 75

1. Section 75.6, Maintenance of Records and Delivery of Information, 
Reports, and Other Communications
    Part 75 sets forth NRC requirements intended to implement the 
agreement between the United States and the International Atomic Energy 
Agency (IAEA) with respect to safeguards of nuclear material. Various 
provisions throughout part 75 require certain licensees and other 
individuals and entities regulated by the NRC to submit to the NRC 
various reports and communications. Section 75.6 specifies the NRC 
officials to whom these reports and communications are to be sent. 
However, Sec.  75.6(b)--the provision applying to, inter alia, nuclear 
power plants--refers only to holders of a construction permit or an 
operating license, and does not include holders of combined licenses. 
Accordingly, Sec.  75.6(b) is revised to reference combined licenses. 
The NRC notes that early site permits and manufacturing licenses need 
not be referenced, inasmuch as the U.S.-IAEA Safeguards Agreement does 
not extend to early site permits or manufacturing licenses.

S. Changes to 10 CFR Part 95

    The following discussion explains the requirements in part 95 
generically and covers Sec. Sec.  95.5, 95.13, 95.19, 95.20, 95.23, 
95.31, 95.33 through 95.37, 95.39, 95.43, 95.45, 95.49, 95.51, 95.53, 
95.57, and 95.59.
    Part 95 sets forth the NRC requirements governing what individuals 
and entities may be provided access to National Security Information 
(NSI) and/or Restricted Data (RD) received or developed in connection 
with activities licensed, certified, or regulated by the NRC, and how 
this information and data is to be protected by these individuals and 
entities against unauthorized disclosure.
    Although requirements for protection of NSI and RD must, by 
statute, apply to all individuals and entities provided access to such 
information, various sections in part 95 use slightly different wording 
to delineate the relevant set of individuals and entities. To ensure 
consistency, the Commission is revising its regulations to refer to 
``licensee, certificate holder, or other person,'' to describe the 
individuals and entities subject to the applicable requirements. In 
adopting this phrase, the NRC intends to ensure that its regulatory 
requirements for protection of NSI and RD in part 95 extend as broadly 
as the NRC's authority provided under applicable law. The term, 
``licensee,'' includes both holders of all NRC licenses, including (but 
not limited to) combined licenses, as well as holders of permits such 
as construction permits and early site permits. The term, ``certificate 
holder,'' includes (but is not limited to) all certificates of approval 
that the Commission may issue, such as a certificate of compliance for 
spent fuel casks under 10 CFR part 72. Finally, the term, ``or other 
person,'' is intended to include individuals and entities who are 
subject to the regulatory authority of the Commission, including 
applicants for standard design approvals and standard design 
certifications under part 52. For the same reasons, the Commission is 
revising Sec.  95.39 to use the phrase, ``NRC

[[Page 49436]]

license, certificate, or standard design approval or standard design 
certification under part 52.''

T. Changes to 10 CFR Part 140

    Part 140 addresses the NRC requirements applicable to nuclear 
reactor licensees with respect to financial protection and indemnity 
agreements to implement Section 170 of the AEA, commonly referred to as 
the Price-Anderson Act. In general, the indemnification and financial 
protection requirements in part 140 become applicable when a holder of 
a 10 CFR part 50 construction permit who also possesses a materials 
license under 10 CFR part 70 brings fuel onto the site. However, part 
140 did not address the indemnification and financial protection 
requirements of combined license holders. Accordingly, the final rule 
revises various sections in part 140 to address combined licenses under 
part 52.
    The NRC does not believe that part 140 must be revised to address 
any part 52 licensing process other than a combined license. Neither an 
early site permit nor a manufacturing license authorizes the possession 
or use of nuclear fuel or other nuclear materials, and the NRC would 
not issue these licenses with a materials license under part 70. The 
NRC also believes that part 140 need not be revised to address standard 
design approvals or standard design certifications, because neither of 
these processes authorize the possession or use of nuclear fuel or 
other nuclear materials.

U. Changes to 10 CFR Part 170

    Part 170 sets out the fees charged for licensing services performed 
by the NRC. The NRC is revising Sec.  170.2(g) and (k) to add 
conforming references to manufacturing licenses and standard design 
approvals issued under part 52, revise the existing reference to 
appendix Q to part 52 to be a reference to appendix Q to part 50, and 
delete the reference to a manufacturing license issued under part 50 
(which is being removed from part 50 because of its transfer to part 52 
in the 1989 rulemaking adopting part 52).

V. Changes to 10 CFR Part 171

    Part 171 sets out the annual fees charged to persons who hold 
licenses issued by the NRC. The NRC is revising Sec.  171.15 to add 
conforming references to combined licenses issued under part 52. Note 
that for combined licenses, the requirements of part 171 are not 
applicable until after the Commission has made the finding under Sec.  
52.103(g). This section also provides fee requirements for each person 
holding a part 50 power reactor license that is in decommissioning or 
possession only status and each person holding a part 72 license who 
does not hold a part 50 license. The NRC also added conforming changes 
to include references in part 52 in these provisions.

VI. Section-by-Section Analysis

Part 52, General Provisions

Section 52.0 Scope; Applicability of 10 CFR Chapter I Provisions
    This section, formerly designated as Sec.  52.1, has been expanded 
to: (1) address all licensing and regulatory processes covered in part 
52; and (2) more clearly define the relationship between part 52 and 
remaining provisions of 10 CFR Chapter I. Paragraph (a), which 
establishes the scope of part 52, is revised by referring to all 
licensing and regulatory processes covered in part 52. In addition, 
paragraph (a) is revised to give notice to contractors, subcontractors 
or consultants of applicants for or holders of licenses or regulatory 
approvals under part 52 that they are subject to NRC enforcement action 
for violations of the deliberate misconduct proscriptions in Sec.  
52.4. The Commission notes, as discussed below in the section-by-
section analysis of Sec.  52.4, that deliberate misconduct under Sec.  
52.4 may occur as the result of a violation of any Commission rule and 
regulation throughout 10 CFR Chapter I, not just a violation of a 
requirement in part 52.
    Paragraph (b) is a new provision that supersedes former Sec.  
52.83. The first sentence of paragraph (b) is intended to make clear 
that the Commission's regulations in 10 CFR Chapter I apply to 
applicants and holders of licenses, permits and other regulatory 
approvals in part 52 (e.g., design approvals and standard design 
certifications). Accordingly, applicants, licensees and holders of 
regulatory approvals under part 52 should review the regulations in 10 
CFR Chapter I to ensure that they are in compliance with applicable 
Commission requirements throughout 10 CFR Chapter I. The second 
sentence of paragraph (b) reinforces the applicability of the 
Commission's requirements throughout 10 CFR Chapter I to part 52 
licenses, permits, and other regulatory approvals. As part of this 
final rule, the Commission is making conforming changes as necessary 
throughout Chapter I to ensure that relevant regulations clearly set 
forth their applicability to part 52 licenses and approvals, and to 
part 52 entities such as applicants, licensees, and holders. 
Nonetheless, the Commission is adopting paragraph (b) in order to 
clearly and unambiguously impose applicable regulatory requirements 
that exist throughout 10 CFR Chapter I.
Section 52.1 Definitions
    This section, formerly designated as Sec.  52.2, has been 
supplemented by: (1) adding definitions of terms that are used in part 
52 but were undefined in the previous rule; and (2) providing 
definitions of new terms that were added in this rulemaking to provide 
greater clarity and precision. New definitions which are noteworthy are 
discussed individually as follows.
    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). This special 
provision for modular designs was added to 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 three or four 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. The 
Commission believes that the term ``modular design'' needs to be 
defined to aid future use of the current Sec.  52.103(g) by 
distinguishing the intended definition from other currently used 
definitions for ``modular design.'' Also, future combined license 
applicants for a multi-unit site that would be similar to current 
multi-unit sites (where each unit is similar in design but independent 
of all other units) could use this provision.
    Definitions of the terms design characteristics, design parameters, 
site characteristics, and site parameters were added to Sec.  52.1 to 
clarify their meaning and use in the licensing and approval processes 
of part 52. Design characteristics are defined as the actual features 
of a nuclear reactor or reactors. Design characteristics are specified 
in the final safety analysis report for a standard design approval, a 
standard design certification, a combined license application, or a 
manufacturing license. Design parameters are defined as the postulated 
features of a nuclear reactor or reactors that could be built at the 
proposed site. Design parameters are specified in an early site permit 
application. Site characteristics are defined as the actual physical, 
environmental, and demographic

[[Page 49437]]

features of a site. Site characteristics are specified in an early site 
permit or combined license application. Site parameters are defined as 
the postulated physical, environmental, and demographic features of an 
assumed site. Site parameters are specified in a standard design 
approval, standard design certification, or manufacturing license.
    The values for the characteristics and parameters will be used in 
the NRC's review of combined license applications that reference design 
approvals, design certifications, manufacturing licenses, or early site 
permits. For example, Sec.  52.79(b) requires that a combined license 
application referencing an early site permit contain information 
sufficient to demonstrate that the actual design characteristics of the 
nuclear facility fall within the design parameters and site 
characteristics specified in the early site permit. Also, Sec.  
52.79(d) requires that a combined license application referencing a 
design certification rule must contain information sufficient to 
demonstrate that the actual site characteristics fall within the site 
parameters specified in the design certification.
    The above terms are also used in Sec. Sec.  52.39 and 52.93. 
Because the NRC is relying on certain design parameters specified in 
the early site permit applications to reach its conclusions on site 
suitability, these design parameters will be included in any early site 
permit issued. The NRC believes that its review of a combined license 
application that references an early site permit will involve a 
comparison to ensure that the actual characteristics of the design 
chosen by the combined license applicant fall within the design 
parameters specified in the early site permit. A combined license 
application that references a design certification will involve a 
comparison to ensure that the actual characteristics of the site chosen 
by the combined license applicant fall within the site parameters in 
the design certification. Similarly, if a combined license applicant 
references both an early site permit and a design certification, the 
NRC will review the application to ensure that the site characteristics 
in the early site permit fall within the site parameters in the 
referenced design certification and that the actual design 
characteristics fall within the design parameters in the early site 
permit.
    A new definition of major features of the emergency plans is added 
to explain what aspects of emergency preparedness--short of full and 
integrated emergency plans--an early site permit applicant may seek 
approval of under Sec.  52.17(b)(2)(i). A major feature may consist of 
a specific aspect of a plan necessary to address in whole or part 1 or 
more of the 16 planning standards in 10 CFR 50.47(b). Additional 
requirements for each of the planning standards are set forth in part 
50, appendix E, and the applicant may choose to demonstrate compliance 
with one or more provisions in appendix E, either in addition to or 
without a full demonstration of compliance with a planning standard in 
Sec.  50.47(b), when seeking approval of part of a major feature. A 
major feature may also be a description of one or both of the emergency 
planning zones (EPZs) required by 10 CFR 50.33(g). Regulatory 
considerations governing EPZs are set forth in Sec.  50.33(g); a major 
feature need not address all of these considerations.
    A definition of prototype plant is added to explain the type of 
nuclear power plant that the Commission intended in the former Sec.  
52.47(b) (new Sec.  50.43), and Sec.  52.157(e)). A prototype plant is 
a licensed nuclear reactor test facility that is similar to and 
representative of either the first-of-a-kind or standard 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 safety features for the first-of-a-kind nuclear plant 
design, as well as being used as a commercial nuclear power facility.
Section 52.2 Interpretations
    This section, formerly designated as Sec.  52.5, remains unchanged. 
It provides that the only interpretations of part 52 that are legally 
binding on the Commission are interpretations provided by the General 
Counsel. These written interpretations, which are rarely provided by 
the General Counsel, are set forth in 10 CFR part 8.
Section 52.3 Written Communications
    This new section, which is analogous to Sec.  50.4, sets forth 
administrative requirements regarding written communications with the 
NRC, including the addressing of such communications, and listings of 
the various NRC offices and officials who must receive copies of 
different types of communications (e.g., applications for licenses and 
license amendments, security plan and related submissions, quality 
assurance related submissions). The administrative requirements 
themselves are identical to those in Sec.  50.4; they are reproduced in 
Sec.  52.3 to make clear that they apply to applicants for and holders 
of permits, licenses, and regulatory processes that are contained in 
part 52.
Section 52.4 Deliberate Misconduct
    This section, formerly designated as Sec.  52.9, has been 
substantially rewritten in order to more clearly delineate the 
applicability of the proscriptions against deliberate misconduct to all 
delineated part 52 entities, including applicants for and holders of 
standard design approvals, and applicants for standard design 
certifications (including those applicants whose designs are certified 
by the Commission in a standard design certification rulemaking). 
Although the regulatory language in Sec.  52.4 differs from former 
Sec.  52.9, no substantive change in any aspect of the Commission law 
or the underlying policy considerations is being made by the 
Commission's adoption of Sec.  52.4. The relevant law and policy 
considerations for former Sec.  52.9 are merely clarified and extended 
in Sec.  52.4 to cover applicants for and holders of permits, licenses, 
and regulatory processes that are contained in part 52.
Section 52.5 Employee Protection
    This new section, which is analogous to Sec.  50.7, prohibits 
discrimination against employees for engaging in protected activities 
established in Section 211 of the Energy Reorganization Act of 1974, as 
amended (1974 ERA). These protected activities, which are listed in 
Sec.  52.5(a)(1), include (but are not limited to) providing the 
Commission or the employer information about alleged violations of the 
AEA or 1974 ERA, of any of the Commission's regulations. No substantive 
change in any aspect of the Commission law or the underlying policy 
considerations with respect to employee protection is being made by the 
Commission adoption of Sec.  52.5; the relevant law and policy 
considerations for former Sec.  50.7 are merely clarified and extended 
in Sec.  52.5 to cover applicants for and holders of permits, licenses, 
and regulatory processes that are contained in part 52 (currently, 
standard design approvals and standard design certifications).
Section 52.6 Completeness and Accuracy of Information
    This new section, which is analogous to Sec.  50.9, requires that 
all information submitted to the NRC by the delineated part 52 entities 
be complete and accurate, and imposes a reporting requirement on such 
entities with respect to information with respect to the regulated 
activity having a significant implication for public health and safety 
or common defense and security. No substantive change in any aspect of 
the Commission law or the

[[Page 49438]]

underlying policy considerations is being made by the Commission 
adoption of Sec.  52.6; the relevant law and policy considerations 
underlying Sec.  50.9 are merely clarified and extended to cover 
applicants for and holders of permits, licenses and regulatory 
processes that are contained in part 52. For example, Sec.  50.9 does 
not impose a positive obligation on licensees to seek out new 
information meeting the reporting thresholds in the rule. In applying 
Sec.  52.6, the Commission would extend this interpretation to part 52 
entities such as combined license holders and standard design 
certification applicants (including applicants whose applications were 
approved, for the regulatory life of the certification rule).
Section 52.7 Specific Exemptions
    This new section, which is analogous to Sec.  50.12, provides for 
specific procedures and criteria for Commission grants of exemptions 
from the provisions of part 52. No substantive change in any aspect of 
the Commission law or the underlying policy considerations is being 
made by the Commission adoption of Sec.  52.7; the relevant law and 
policy considerations underlying Sec.  50.12 are merely extended to 
part 52.
    The NRC notes that the exemption provisions in Sec.  52.7 do not 
supercede or otherwise diminish more specific exemption provisions that 
are in part 52, such as the provision of a specific design 
certification rule or Sec.  52.63(b)(1) governing exemptions from one 
or more elements of a design certification rule. An applicant or 
licensee referencing a standard design certification rule who wishes to 
obtain an exemption from one or more elements must meet the criteria in 
the specific design certification rule or Sec.  52.63(b)(1). If the 
applicant or licensee is unable to demonstrate compliance with those 
criteria, then it may request an exemption under the more encompassing 
authority of Sec.  52.7. However, the exemption request must then 
demonstrate compliance with the additional criteria in Sec.  52.7.
    The Commission also notes that Sec.  52.7 does not supercede the 
applicability of more specific dispensation provisions in other parts 
of Chapter I. For example, a holder of a combined license would not 
require a separate part 52 exemption in order to obtain approval of an 
alternative to a provision of an applicable ASME Code provision that is 
otherwise required under 10 CFR 50.55a; the licensee need only satisfy 
the criteria in Sec.  50.55a(a)(3). However, in the absence of a more 
specific dispensation provision, the Commission intends to utilize 
Sec.  52.7 as a means for granting dispensation from compliance with 
Commission requirements in other parts of 10 CFR Chapter I. The person 
requesting an exemption need only address the Sec.  52.7 criteria as 
applied to the underlying requirement for which dispensation from 
compliance is sought, and need not also address dispensation from 
compliance with the relevant part 52 requirement. For example, the 
holder of the combined license who wishes dispensation from compliance 
with a fire protection requirement in 10 CFR 50.48 need only address 
the relevant criteria in Sec.  52.7 with respect to the reasons for 
dispensation from compliance with Sec.  50.48. The holder need not 
address dispensation from compliance with Sec.  52.0, which otherwise 
makes applicable the provisions of Sec.  50.48 on the licensee. Any 
exemption granted by the Commission would address the reasons for 
dispensation with the underlying requirement--in this case, Sec.  
50.48, and would also provide dispensation from compliance with Sec.  
52.0.
Section 52.8 Combining Licenses; Elimination of Repetition
    This new section includes provisions analogous to Sec. Sec.  50.31, 
50.32, and 50.52 and is added to clarify that these regulatory 
provisions also apply to part 52 licenses. Paragraph (a), which is 
analogous to Sec.  50.31, is added to make clear that an applicant for 
a license under part 52 may combine in one application, several 
applications for different kinds of licenses under various regulations 
in 10 CFR Chapter I. Section 50.31 currently provides that an applicant 
may combine in one application, several applications for different 
kinds of licenses under various regulations in 10 CFR Chapter I. The 
plain reading of this language, given that this provision is located in 
part 50, is that a part 50 application may contain in one application 
other applications for different licenses in other parts of 10 CFR 
Chapter I. Thus, Sec.  50.31 would not appear to allow a part 52 
application (as for a combined license) to combine in one application 
other applications for different license in other parts of 10 CFR 
Chapter I. Accordingly, paragraph (a) makes clear that a part 52 
application may be combined with application for different licenses in 
other parts of 10 CFR Chapter I.
    Paragraph (b), which is analogous to Sec.  50.32, is added to make 
clear that an applicant for a license, standard design certification, 
or design approval under part 52 may incorporate by reference in its 
application information contained in other documents provided to the 
Commission, but that such incorporation must clearly specify the 
information to be incorporated.
    Paragraph (c), which is analogous to Sec.  50.52, is added to 
clarify the Commission's authority under Section 161.h of the AEA to 
combine NRC licenses, such as a special nuclear materials license under 
part 70 for the reactor fuel, with a combined license under part 52. 
Analogous to the situation with respect to Sec.  50.31, the language in 
Sec.  50.52 would not appear to allow the Commission to combine into a 
single part 52 license, other non-part 52 licenses. No substantive 
change in any aspect of the Commission law or the policy considerations 
underlying Sec. Sec.  50.31, 50.32, and 50.52 is being made by the 
Commission adoption of Sec.  52.8; the relevant law and policy 
considerations underlying Sec. Sec.  50.31, 50.32, and 50.52 are merely 
extended to part 52.
Section 52.9 Jurisdictional Limits
    This new section, which is analogous to Sec.  50.53, makes clear 
that no approval provided by the Commission under part 52 addresses or 
approves in any manner activities which are not under or within the 
territorial jurisdiction of the United States. As a practical matter, 
this means that an approval or license issued by the NRC under part 52 
has no legal effect outside the territorial jurisdiction of the United 
States. No substantive change in any aspect of the Commission law or 
the policy considerations underlying Sec.  50.53 is being made by the 
Commission adoption of Sec.  52.9; the relevant law and policy 
considerations are merely extended to part 52.
Section 52.10 Attacks and Destructive Acts
    This new section, which is analogous to Sec.  50.13, applies the 
existing Commission law and policy that a licensee need not provide for 
design features or other measures to protect against certain attacks 
and destructive acts, or the use or deployment of weapons incident to 
U.S. defense activities, to the applicants for and holders of permits, 
licenses and other approvals under part 52. No substantive change in 
any aspect of the Commission law or the underlying policy 
considerations is being made by the Commission adoption of Sec.  52.10; 
the relevant law and policy considerations for the Sec.  50.13 
exclusion are merely extended to cover applicants for and holders of 
permits, licenses, and regulatory processes that are contained in part 
52.

[[Page 49439]]

Section 52.11 Information Collection Requirements: OMB Approval
    This section, formerly designated as Sec.  52.8, remains unchanged. 
It gives notice that all information collection and reporting 
requirements in part 52 have been approved by the Office of Management 
and Budget. No requirement, action or responsibility is imposed on part 
52 entities by this section.

Subpart A--Early Site Permits

Section 52.12 Scope of Subpart
    This section describes the scope of this licensing process. Under 
this subpart an applicant can request pre-approval of a site (so-called 
site banking), separate from other licensing actions, and subsequently 
reference that early site permit in a future application to build a 
nuclear power plant. This process was created for proposed sites that 
the applicant may not plan to use in the near term.
Section 52.13 Relationship to Other Subparts
    This section explains the relationship of the early site permit 
process to the construction permit process under 10 CFR part 50 and to 
the combined license process under part 52.
Section 52.15 Filing of Applications
    This section explains who can file, how to file, and the fees for 
NRC review of an application for an early site permit.
Section 52.16 Contents of Applications; General Information
    This section sets forth the type of general information that is 
required to be included in an early site permit application, namely, 
the information required by 10 CFR 50.33(a) through (d) and (j). 
Section 50.33 requires that the application include information such as 
the name and address of the applicant, a description of the business or 
occupation of the applicant, and citizenship information of the 
applicant. Section 50.33 also provides requirements for the handling of 
Restricted Data or other defense information in an application.
Section 52.17 Contents of Applications; Technical Information
    The purpose of this section is to set forth the type of technical 
information to be included in an application for an early site permit. 
Paragraph (a)(1) identifies the information needed for the site safety 
review, excluding emergency planning information. The site safety 
information is a subset of the information required of applicants for 
construction permits. Although an ESP applicant does not need to 
specify a particular nuclear plant design, as in construction permit 
applications, it does need to provide sufficient surrogate design 
information (developed to bound the nuclear plant design(s) that are 
being considered by the applicant) so that the NRC can make a 
determination on the acceptability of the site and the environmental 
impacts, and determine whether designs bounded by the surrogate design 
information provided by the applicant can be qualified for the proposed 
site. The application must contain, among other things, the specific 
number, type (e.g., pressurized-water reactor), and thermal power level 
of the facilities, or range of possible facilities, for which the site 
may be used; the anticipated maximum levels of radiological and thermal 
effluents each facility will produce; the type of cooling systems, 
intakes, and outflows that may be associated with each facility; the 
boundaries of the site; and the proposed general location of each 
facility on the site. As part of the description of the proposed 
general location of each facility on the site (Sec.  52.17(a)(1)(v)), 
the applicant should describe the foot print for all structures and 
external safety-related design features proposed for the site.
    The application must also include the seismic, meteorological, 
hydrologic, and geologic characteristics of the proposed site with 
appropriate consideration of the most severe of the natural phenomena 
that have been historically reported for the site and surrounding area 
and with sufficient margin for the limited accuracy, quantity, and 
period of time in which the historical data have been accumulated. This 
information is to ensure that future plants built at the site would be 
in compliance with General Design Criterion 2 from appendix A to part 
50, which requires that structures, systems, and components important 
to safety be designed to withstand the effects of natural phenomena 
such as earthquakes, tornadoes, hurricanes, floods, tsunami, and 
seiches without loss of capability to perform their safety functions.
    The application must also include the location and description of 
any nearby industrial, military, or transportation facilities and 
routes, and the existing and projected future population profile of the 
area surrounding the site. The application 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 from 
a radiological safety standpoint. In addition, the application must 
demonstrate that adequate security plans and measures can be developed 
for the site and must provide a description of the quality assurance 
program applied to site-related activities.
    Paragraph (a)(2) identifies that the application must include an 
environmental report that meets the requirements of Sec.  51.50(b). 
Environmental reports must focus on the environmental effects of 
construction and operation of a nuclear reactor, or reactors, which 
have characteristics that fall within the design parameters postulated 
in the early site permit. Environmental reports must also include an 
evaluation of alternative sites to determine whether there is any 
obviously superior alternative to the site proposed. Environmental 
reports submitted in an early site permit application are not required 
to but may include an assessment of the economic, technical, and other 
benefits and costs of the proposed action or an analysis of other 
energy alternatives.
    Paragraph (b) identifies the emergency planning information to be 
included in the application. All ESP applicants are required to 
identify in the site safety analysis report (SSAR) physical 
characteristics unique to the proposed site that could pose a 
significant impediment to the development of emergency plans, e.g., a 
physical characteristic or combination of physical characteristics that 
could pose major difficulties for evacuation or the taking of other 
protective actions. In addition, if the applicant identifies such 
physical characteristics, the application must identify measures that 
would, when implemented, mitigate or eliminate the significant 
impediment. After meeting this mandatory requirement, paragraph (b) 
allows applicants the option of either submitting major features of 
emergency plans or complete and integrated emergency plans for approval 
by the NRC, in consultation with the Department of Homeland Security 
(DHS). For complete and integrated emergency plans, the applicant must 
include the proposed inspections, tests, and analyses that the holder 
of a combined license referencing the early site permit shall perform, 
and the acceptance criteria that are necessary and sufficient to 
provide reasonable assurance that, if the inspections, tests, and 
analyses are performed and the acceptance criteria met, the facility 
has been constructed and will operate in conformity with the license, 
the provisions of the Atomic Energy Act,

[[Page 49440]]

and the NRC's regulations. The inclusion of such inspections, tests, 
analyses, and acceptance criteria (ITAAC) is necessary to allow the NRC 
to make the finding that the plans submitted by the applicant provide 
reasonable assurance that adequate protective measures can and will be 
taken in the event of a radiological emergency. Paragraph (b) also 
allows applicants proposing major features of emergency plans to 
include proposed ITAAC. Where the applicant is submitting a complete 
and integrated emergency plan, a utility plan must be submitted if any 
offsite agencies elect not to participate in the development of 
emergency planning information.
    If the applicant plans to perform the preparations for construction 
activities identified in 10 CFR 50.10(e)(1), then paragraph 52.17(c) 
requires the applicant to describe the activities it is requesting to 
perform and propose a redress plan that, if carried out, would achieve 
a ``self-maintaining, environmentally stable, and aesthetically 
acceptable site'' that conforms to local zoning laws. Redress plans are 
expected to be modeled on the redress requirements imposed on the 
Clinch River Breeder Reactor project (see In the Matter of the U.S. 
Department of Energy, et al., LBP-85-7, 21 NRC 507 (1985)). By 
containing a redress plan, the ESP will constitute assurance that, if 
site preparation activities are conducted but the site is never used 
for a nuclear power plant, the site will be returned to an acceptable 
and stable condition.
Section 52.18 Standards for Review of Applications
    This section identifies the regulations that the NRC staff will use 
in performing its review of an application for an early site permit, 
including the standards that the NRC staff will use in performing its 
assessment of emergency preparedness information provided in the ESP 
application.
Section 52.21 Administrative Review of Applications; Hearings
    This section identifies the procedural requirements that apply to 
the mandatory hearing for the early site permit licensing process. This 
section also clarifies that the applicant's environmental report is not 
required to but may include an assessment of the benefits of 
construction and operation of the reactor or reactors, or an analysis 
of alternative energy sources. In addition, the presiding officer in an 
ESP hearing is prohibited from admitting contentions on these matters 
if those issues were not addressed in the early site permit 
application.
Section 52.23 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS)
    This section states that the ACRS will report on those portions of 
the application which concern safety which is the same role the ACRS 
had with respect to construction permits in the past.
Section 52.24 Issuance of Early Site Permit
    The purpose of this section is to set forth the timing of issuance 
of an ESP and the findings that the Commission must make to issue the 
ESP, including that issuance of the permit will not be inimical to the 
common defense and security or to the health and safety of the public, 
that the applicant is technically qualified to engage in activities 
necessary to prepare the ESP application and any site preparation 
activities that the applicant is seeking approval to perform, and that 
the findings required by subpart A of 10 CFR part 51 regarding the NRC 
staff's assessment of the environmental impact have been made.
    This section also requires that the early site permit specify the 
site characteristics, design parameters, and terms and conditions of 
the early site. Before issuance of either a construction permit or a 
combined license referencing an early site permit, the Commission must 
find that any relevant terms and conditions of the early site permit 
have been met. Any terms or conditions that could not be met by the 
time of issuance of the construction permit or combined license must be 
set forth as terms or conditions of the construction permit or combined 
license. Finally, this section requires that the early site permit 
specify the site preparation activities under Sec.  52.17(c) that the 
permit holder is authorized to perform.
Section 52.25 Extent of Activities Permitted
    This section specifies that, if the construction preparation 
activities authorized by Sec.  52.24(c) are performed and the site is 
not referenced in a application for a construction permit or a combined 
license while the permit remains valid, then the early site permit 
remains in effect for the purpose of site redress with the goal of 
achieving an environmentally stable and aesthetically acceptable site.
Section 52.27 Duration of Permit
    The purpose of paragraph (a) of this section is to specify the 
duration of an early site permit. The applicant can request a duration 
of up to 20 years. Paragraph (b) describes the conditions under which 
an ESP can continue to be valid beyond its expiration date. Paragraph 
(c) allows an applicant for a construction permit or combined license, 
at its own risk, to reference an ESP that is under review by the NRC 
but not yet granted. Paragraph (d) explains that, upon issuance of a 
construction permit or combined license, a referenced early site permit 
is subsumed, to the extent referenced, into the construction permit or 
combined license. By ``subsumed'' the NRC means that the information 
that was contained in the early site permit SSAR becomes part of the 
referencing combined license FSAR upon issuance of the combined 
licenses in the same manner as if the combined license applicant had 
not referenced an early site permit. The NRC is including the phrase 
``to the extent referenced,'' to indicate that it is not all of the 
information submitted in the early site permit application that is 
subsumed into the combined license, but, rather, only that information 
that is contained in the SSAR and identified by the applicant as being 
referenced in the combined license application. This subsumption of the 
early site permit into the referencing license affects the way changes 
to the early site permit information will be handled because it breaks 
the tie to the finality provisions in Sec.  52.39. After issuance of 
the construction permit or combined license, Sec.  52.39 no longer 
applies to the early site permit information and such information will 
be covered by the same finality provisions as the rest of the 
information in the FSAR (with the exception of any referenced design 
certification information), as outlined in Sec.  52.98 (e.g., in 
accordance with Sec. Sec.  50.54, 50.59, etc.).
Section 52.28 Transfer of Early Site Permit
    This section specifies the requirements to be followed if a holder 
of an early site permit wants to transfer the ESP to another person or 
company.
Section 52.29 Application for Renewal
    Paragraph (a) of this section explains the contents and timing of 
an application for renewal of an early site permit. Paragraph (b) sets 
forth the procedure for requesting a hearing on the application for 
renewal. Paragraph (c) explains that an ESP may remain in effect beyond 
its expiration under

[[Page 49441]]

certain circumstances. Specifically, an ESP for which a timely 
application for renewal has been filed remains in effect until the 
Commission has determined whether to renew the permit. If an ESP is not 
renewed, it continues to be valid in any proceeding on an application 
for a construction permit or a combined license which references the 
ESP and was docketed prior to the expiration of the ESP. Finally, 
paragraph (d) identifies the responsibilities of the ACRS on an ESP 
renewal application.
Section 52.31 Criteria for Renewal
    Paragraph (a) of this section sets forth the criteria for granting 
a renewal of an early site permit and provides that, if the NRC wants 
to impose new requirements, it must demonstrate that the new 
requirements meet the backfit standard from Sec.  50.109. Paragraph (b) 
explains that even if an application for renewal of an ESP is denied by 
the NRC, the applicant can submit a new application for an ESP that 
corrects the problems with the application for renewal.
Section 52.33 Duration of Renewal
    This section specifies the duration of a renewed early site permit. 
An ESP may, upon application, be extended for periods of up to 20 years 
beyond the previously approved duration, provided the criteria in Sec.  
52.31 are met.
Section 52.35 Use of Site for Other Purposes
    The purpose of this section is to explain how the holder of an 
early site permit could use the site for other activities. An approved 
site may be used for purposes not related to the construction of a 
nuclear power facility, e.g., a fossil-fueled station or a park, 
provided that the Commission is informed of all significant non-nuclear 
uses prior to actual construction or site modification activities. A 
permit may be revoked if a non-nuclear use would interfere with a 
nuclear use, or would so alter the site that important assumptions 
underlying the issuance of the permit were called into question.
Section 52.39 Finality of Early Site Permit Determinations
    This section specifies the special backfit requirements that apply 
to an early site permit. Paragraph (a) provides requirements regarding 
finality of ESP issues as they relate to the Commission. Paragraph 
(a)(1) states that, notwithstanding any provision in 10 CFR 50.109 
(Backfitting), while an early site permit or renewed early site permit 
is in effect, the Commission may not change or impose new site 
characteristics, design parameters, or terms and conditions, including 
emergency planning requirements, on the early site permit unless the 
Commission meets one of four conditions. Those conditions are that the 
Commission either determines that a modification is necessary to bring 
the permit or the site into compliance with the Commission's 
regulations and orders applicable and in effect at the time the permit 
was issued; determines that a modification is necessary to assure 
adequate protection of the public health and safety or the common 
defense and security; determines that a modification is necessary based 
on an update under Sec.  52.39(b); or issues a variance requested under 
Sec.  52.39(d).
    Paragraph (a)(2) addresses the finality of an early site permit for 
a license that references the early site permit and requires that the 
Commission treat as resolved those matters resolved in the proceeding 
on the application for issuance or renewal of the early site permit, 
except as provided for in Sec. Sec.  52.39(b), (c), and (d). This 
paragraph also addresses finality of changes to an early site permit 
approved emergency plan (or major features thereof).
    Paragraph (b) requires a license applicant that references an ESP 
to update and correct the emergency preparedness information that was 
provided in the ESP and to discuss whether the new information 
materially changes the bases for compliance with the applicable NRC 
requirements. 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, and (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 by the NRC under Sec.  
52.39(a)(1) would be subject to litigation during the licensing 
proceedings in accordance with Sec.  52.39(c).
    Section 52.39(c) provides requirements for the submittal of 
contentions in a proceeding for the issuance of a license referencing 
an early site permit and for the filing of petitions requesting that an 
early site permit be modified, suspended, or revoked. Paragraph (c)(1) 
states that contentions on several matters may be litigated in the 
proceeding on a combined license that references an early site permit. 
Matters that may be litigated include contentions related to the 
following: (1) The nuclear power reactor proposed to be built does not 
fit within one or more of the site characteristics or design parameters 
included in the early site permit; (2) one or more of the terms and 
conditions of the early site permit have not been met; (3) a variance 
requested under Sec.  52.39(d) is unwarranted or should be modified; 
(4) new or additional information is provided in the application that 
substantially alters the bases for a previous NRC conclusion or 
constitutes a sufficient basis for Commission to modify or impose new 
terms and conditions related to emergency preparedness; or (5) any 
significant environmental issue that was not resolved in the early site 
permit proceeding, or any issue involving the impacts of construction 
and operation of the facility that was resolved in the early site 
permit proceeding for which significant new information has been 
identified. An issue related to the impacts of construction and 
operation of the facility resolved in the early site permit proceeding 
is afforded finality at the combined license stage provided that there 
is no ``new and significant'' information on the issue. If an 
environmental issue was not resolved at the early site permit stage, 
either because information was not sufficient to resolve it or because 
the early site permit applicant was permitted to defer it (e.g., need 
for power analysis), then the combined license applicant would need to 
address the issue in its combined license application. The NRC, in the 
context of a combined license application that references an early site 
permit, has defined the term ``new'' in the phrase ``new and 
significant information'' as any information that was both (1) not 
considered in preparing the ESP environmental report or EIS (as may be 
evidenced by references in these documents, applicant responses to NRC 
requests for additional information, comment letters, etc.) and (2) not 
generally known or publicly available during the preparation of the EIS 
(such as information in reports, studies, and treatises). This new 
information may or may not be significant. For an issue to be 
significant, it must be material to the issue being considered, i.e., 
it must have the potential to affect the NRC staff's evaluation of the 
issue. The COL applicant need only provide information about a 
previously resolved

[[Page 49442]]

environmental issue if it is both new and significant.
    Paragraph (c)(2) allows any person to file a petition requesting 
that the site characteristics, design parameters, or terms and 
conditions of the early site permit be modified, or that the permit be 
suspended or revoked. The petition will be considered in accordance 
with Sec.  2.206. Section 2.206 provides that any person may file a 
request to institute a proceeding to modify, suspend, or revoke a 
license, or for any other action as may be proper. Section 52.39(c)(2) 
addresses the Commission's required action on such a petition and 
states that construction under the construction permit or combined 
license will not be affected by the granting of the petition unless the 
Commission makes the order immediately effective.
    Paragraph (d) provides that an applicant for a license or an 
amendment to such a license who has filed an application referencing an 
early site permit may request a variance from one or more site 
characteristics, design parameters, or terms and conditions of the 
early site permit, or from the SSAR. This paragraph also states that, 
once a construction permit or combined license referencing an early 
site permit is issued, a variance from the early site permit will not 
be granted for that construction permit or combined license. At that 
point, the early site permit is subsumed into the combined license and 
any request for a change to the terms or conditions of the combined 
license is a request for a license amendment that must be filed under 
the provisions of Sec.  50.90.
    The NRC is adding new paragraph (e) in the final rule in response 
to public comments expressing support for adding provisions to provide 
an early site permit holder with the option of requesting an amendment 
to the early site permit in order to resolve issues that were not 
addressed in the original early site permit review or to achieve 
finality on updated early site permit information. Paragraph (e) states 
that the holder of an early site permit may not make changes to the 
early site permit, including the SSAR, without prior Commission 
approval. The request for a change to the early site permit must be in 
the form of an application for a license amendment, and must meet the 
requirements of 10 CFR 50.90 and 50.92. The NRC considers an early site 
permit SSAR to be equivalent to a combined license FSAR; therefore, 
when an early site permit is amended, the SSAR must be revised 
consistent with the ESP amendments. In addition, the SSAR retains 
continuing viability for early site permits that are for multiple units 
after it is referenced in the first combined license. However, unlike 
an FSAR, there is no change process for the SSAR that does not require 
NRC review and approval.
    Finally, the Commission is adding a new paragraph (f) (proposed 
paragraph (e)) to the ``finality'' section in each subpart of part 52, 
including Sec.  52.39, entitled ``Information requests,'' which 
delineates the restrictions on the NRC for information requests to the 
holder of the early site permit. This provision is analogous to the 
former provision on information requests in paragraph 8 of appendix O 
to parts 50 and 52, and is based upon the language of Sec.  50.54(f). 
For early site permits, this provision is contained in Sec.  52.39(f), 
and requires the NRC to evaluate each information request on the holder 
of an early site permit to determine that the burden imposed by the 
information request is justified in light of the potential safety 
significance of the issue to be addressed in the information request. 
The only exceptions would be for information requests seeking to verify 
compliance with the current licensing basis of the early site permit. 
If the request is from the NRC staff, the request would first have to 
be approved by the Executive Director for Operations (EDO) or his or 
her designee.

Subpart B--Standard Design Certifications

Section 52.41 Scope of Subpart
    This section describes the scope of this licensing process for 
certification of standard nuclear power plant designs. Under this 
subpart, an applicant may request pre-approval of either an 
evolutionary light-water or advanced nuclear power plant design, 
separate from a site review or other licensing action, and subsequently 
reference that certified design in an application to build a nuclear 
power plant. The requirements for the type of plant to be certified 
were moved from Sec.  52.45 to this section. The scope of the standard 
plant design must be essentially complete as described in Sec.  
52.47(c).
Section 52.43 Relationship to Other Subparts
    The purpose of this section is to explain the relationship of the 
design certification process to the processes set forth in subparts C, 
E, and F of 10 CFR part 52, which provide for combined licenses, 
standard design approvals, and manufacturing licenses. The requirement 
to hold a final design approval under former appendix O to part 52 as a 
prerequisite to design certification was deleted from Sec.  52.45. 
However, applicants for design certification have the option of also 
applying for a standard design approval under subpart E. Also, 
applicants for a manufacturing license may reference a certified 
design.
Section 52.45 Filing of Applications
    This revised section is similar to the ``filing of applications'' 
sections in subparts A and C of this part. This section explains how to 
file an application for design certification and how the fees for NRC's 
review of the application will be assessed. Because design 
certification is a rule and not a license, the applicant for design 
certification does not need to be a U.S. citizen or company (AEA, 
Section 103).
Section 52.46 Contents of Applications; General Information
    This is a new section and it is similar to the ``general 
information'' sections in subparts A and C of this part. It identifies 
the general information that must be included in all applications.
Section 52.47 Contents of Applications; Technical Information
    The purpose of this section is to identify the technical 
information that must be included in an application for design 
certification. This section was revised to provide a comprehensive list 
of requirements for a design certification application. Paragraphs (a) 
and (c) describe the information that must be included in the FSAR, 
which is included in the application, and paragraph (b) describes the 
information that must also be included in the application but does not 
need to be included in the FSAR. Paragraph (c) describes additional 
requirements for particular types of applications. This section also 
specifies the level of detail for the design information that must be 
provided in an application.
    Many of the requirements in this section were taken from 10 CFR 
50.34 or are pointers to technical requirements in parts 20, 50, 51, 
and 73 that must be addressed in the application. The requirements 
taken from Sec.  50.34 are a subset of the information required of 
applicants for construction permits and operating licenses. Other 
requirements came from the original version of 10 CFR 52.47 or were 
developed by the Commission during the initial design certification 
reviews (e.g., SECY-93-087, ML003708021).
    Although an applicant for design certification does not need to 
specify a particular site for the nuclear power plant, as in a combined 
license application, it does need to identify the site parameters, 
under paragraph (a)(1),

[[Page 49443]]

that the standard nuclear power plant is designed to meet, e.g., 
postulated values for the safe-shutdown earthquake response spectra and 
maximum tornado wind speed. These parameters are usually selected to 
envelop a large portion of existing nuclear plant sites in the United 
States. Once the design is certified by the NRC, conformance of the 
actual site with the established site parameters must be demonstrated 
by the applicant for a combined license and verified by the NRC when 
the application is submitted.
    Paragraph (a)(7) requires the applicant for design certification to 
describe its qualifications to design and analyze a standard nuclear 
power plant, which may become part of the bases for a future license.
    Paragraph (a)(13) requires the applicant to provide the electric 
equipment list required by Sec.  50.49(d). The NRC understands that the 
applicant may not be able to establish qualification files for all 
applicable components.
    In its staff requirements memorandum (SRM) on SECY-90-377, 
``Requirements for Design Certification under 10 CFR part 52,'' dated 
February 15, 1991, the Commission directed the staff to ensure that the 
design certification process preserves operating experience insights in 
the certified design. Therefore, for plant designs that are based on or 
are evolutions of nuclear plants that have operated in the United 
States, paragraph (a)(22) requires the applicant to demonstrate how 
relevant operating experience insights, from NRC's generic letters and 
bulletins issued after the most recent revision of the applicable SRP 
and 6 months before the docket date of the application, have been 
incorporated into the plant design. Operating experience includes 
consideration of operating events and the reliability and performance 
of structures, systems, and components. If the application is for a 
design that is not based on or is not an evolution of a nuclear plant 
that operated in the United States, the applicant must demonstrate how 
insights from any relevant international operating experience have been 
incorporated into that plant design.
    In its SRMs, dated June 26, 1990, and July 21, 1993, on SECY-90-16, 
``Evolutionary Light-Water Reactor Certification Issues and their 
Relationship to Current Regulatory Requirements,'' and SECY-93-087, 
``Policy, Technical, and Licensing Issues Pertaining to Evolutionary 
and Advanced Light-Water Reactor Designs,'' respectively, the 
Commission approved NRC staff recommendations for selected preventative 
and mitigative design features for future light-water reactor designs. 
Paragraph (a)(23) requires the applicant to provide a description and 
analysis of those design features discussed in SECY-90-16 and SECY-93-
087. Postulated severe accidents are not design-basis accidents (DBAs) 
and the severe accident design features do not have to meet the 
requirements for DBAs. However, the severe accident design features are 
part of a plant's design bases information.
    Paragraph (a)(24) requires the applicant to provide a conceptual 
design for those design features that are outside the scope of the 
certified design, e.g., service water intake structure or ultimate heat 
sink.
    Paragraph (a)(25) requires the applicant to describe the interface 
requirements for those design features that are outside the scope of 
the certified design, e.g., service water intake structure or ultimate 
heat sink. Paragraph (a)(26) requires justification that the interface 
requirements can be verified with the ITAAC for the plant.
    Paragraph (a)(27) requires the applicant to provide a description 
of the design-specific PRA and its results. Guidance on how to meet the 
PRA information requirement will be provided in separate regulatory 
guidance documents.
    Paragraph (b)(1) requires the applicant to provide the ITAAC that 
are necessary and sufficient to demonstrate that a facility that 
references the design certification has been constructed and will be 
operated in conformity with the design certification, the Atomic Energy 
Act of 1954, as amended, and the Commission's rules and regulations. 
These ITAAC will be a part of the Commission's verification program and 
must cover all of the design information that is within the scope of 
the certified design. ITAAC for the remaining design features that are 
outside of the scope of the certified design will be provided in a 
combined license application that references the design certification 
rule.
    In its SRM on SECY-91-229, ``Severe Accident Mitigation Design 
Alternatives for Certified Standard Designs,'' dated October 25, 1991, 
the Commission approved the staff's recommendation that design 
certification applicants assess SAMDAs for their standard plant 
designs. The Commission required SAMDA evaluations in order to achieve 
greater finality for the design features that are resolved in design 
certification rulemakings. For further explanation, see discussion in 
SECY-93-087, dated April 2, 1993. In order to implement this 
requirement, paragraph (b)(2) requires the applicant to provide a SAMDA 
evaluation for the standard plant design. This assessment is distinct 
from, and in addition to, the requirement in paragraph (a)(23) to 
provide a description and analysis of severe accident design features.
    Paragraph (c)(1) requires an essentially complete scope of design 
in applications for evolutionary nuclear power plants. These plants are 
improved versions of light-water reactor designs that were in operation 
when part 52 was originally codified. Examples of evolutionary designs 
include General Electric's U.S. Advanced Boiling Water Reactor and 
Westinghouse's SP/90 and System 80+ designs. Evolutionary designs do 
not have to meet the design qualification testing requirements set 
forth in 10 CFR 50.43(e).
    Paragraph (c)(2) requires applications for ``advanced'' nuclear 
power plants to provide an essentially complete scope of design and 
meet the design qualification testing requirements in 10 CFR 50.43(e). 
Advanced designs differ significantly from evolutionary light-water 
reactor designs or incorporate, to a greater extent than evolutionary 
designs do, simplified, inherent, passive, or other innovative means to 
accomplish their safety functions. Examples of advanced nuclear power 
plant designs include General Atomic's Modular High Temperature Gas-
Cooled Reactor, General Electric's Simplified Boiling Water Reactor, 
and Westinghouse's AP600.
    Paragraph (c)(3) requires applications for modular nuclear power 
plant designs to describe and analyze the possible operating 
configurations of reactor modules. Modular designs are defined in Sec.  
52.1. Modular plant designs are not portions of a single nuclear plant, 
rather they are separate nuclear power reactors with some shared or 
common systems.
Section 52.48 Standards for Review of Applications
    This section sets forth the parts of 10 CFR that contain applicable 
requirements for the technical review of design certification 
applications. The applicability of these requirements to the design 
certification process is specified in the identified parts. The 
Commission recognizes that new designs may incorporate design features 
that are not addressed by the current standards set out in 10 CFR parts 
20, 50 and its appendices, 51, 73, or 100, and that new standards may 
be required to address these new design features. The Commission will 
determine whether additional rulemakings are needed or appropriate to 
resolve generic safety

[[Page 49444]]

issues that are applicable to multiple designs. On the other hand, new 
design features that are unique to a particular design could be 
addressed in the design certification rulemaking for that particular 
design.
Section 52.51 Administrative Review of Applications
    This section sets forth the procedures for performing a notice and 
comment rulemaking for design certification. Paragraph (b) states that 
the Commission will determine, at its sole discretion, whether to hold 
a legislative hearing on the proposed design certification rule under 
the procedures in subpart O of 10 CFR part 2. Paragraph (c) states that 
proprietary information contained in an application for design 
certification will be given the same treatment that such information 
would be given in a proceeding on an application for a construction 
permit or an operating license under 10 CFR part 50. This gives the 
design certification applicant (vendor) an opportunity to treat 
elements of its design as trade secrets.
Section 52.53 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS)
    This section states that the application for design certification 
shall be sent to the ACRS for its review of safety issues.
Section 52.54 Issuance of Standard Design Certification
    Paragraph (a) of this section sets forth the findings that the 
Commission must make in order to issue a design certification rule. 
Paragraph (b) requires that site parameters, design characteristics, 
and any additional requirements and restrictions be specified in the 
design certification rule. Previous DCRs set forth the additional 
requirements and restrictions in Section IV of the rule. Site 
parameters and design characteristics are defined in Sec.  52.1 and can 
be specified in the design control document. These values will be used 
during the review of a combined license application that references the 
design certification rule to verify that the standard plant design 
conforms with the characteristics of the actual site and the design 
parameters used in the early site permit.
    Section 52.54 was amended to include a new paragraph (c) which 
requires that every DCR contain a provision stating that, after the 
Commission has adopted the final DCR, the applicant for that design 
certification will not permit any individual to have access to, or any 
facility to possess, Restricted Data or classified National Security 
Information until the individual and/or facility has been approved for 
access under the provisions of 10 CFR parts 25 and/or 95. The NRC 
believes that this amendment, along with the changes to parts 25, 95, 
and Sec.  50.37, are necessary to ensure that access to classified 
information is adequately controlled by all entities applying for NRC 
certifications.
Section 52.55 Duration of Certification
    The purpose of this section is to specify the duration that a 
standard design certification is valid for referencing in a combined 
license application.
Section 52.57 Application for Renewal
    The purpose of this section is to set forth the process for 
applying for renewal of an existing design certification rule. 
Paragraph (a) specifies the time period for submitting an application 
for renewal and states that any person can apply for renewal. However, 
if the applicant for renewal is not the same person or entity that 
applied for the existing design certification, as identified in Section 
I of the DCR, then the new applicant is required to demonstrate that 
they have the capability to provide the detailed design for that 
certified nuclear power plant under Sec.  52.63(c) or Sec.  52.73(b).
Section 52.59 Criteria for Renewal
    The purpose of this section is to identify the regulations that 
will be used to determine if an existing design certification should be 
renewed. Paragraph (a) states that the Commission will grant a request 
for renewal if the design complies with the regulations in effect at 
the time the certification was originally issued (see Section V of an 
existing design certification rule) and imposition of any new safety 
requirements on the design during a renewal proceeding will be governed 
by the backfit standards in paragraph (b).
    Under paragraph (c), the applicant for renewal may request an 
amendment to the existing certified design to make some design changes 
provided that the new design meets the regulations in effect at the 
time that the amended, renewed design certification rule is issued and 
the changes do not require a major review or reanalysis of the new 
design. If the changes to the original design certification are so 
extensive that the NRC concludes an essentially new standard design is 
being proposed, then the applicant must submit an application for a new 
design certification under Sec.  52.45.
    Under paragraph (d), denial by the NRC of a request for renewal of 
a design certification does not prevent an applicant from submitting a 
new application for certification under Sec.  52.45.
Section 52.61 Duration of Renewal
    This section specifies the duration that a renewed design 
certification is valid for referencing in a combined license 
application.
Section 52.63 Finality of Standard Design Certifications
    The purpose of this section is to set forth the process for 
amending or backfitting existing design certification rules (DCRs) or 
issuing orders to nuclear plants that referenced a DCR. This section 
also describes the finality of issue resolution under a design 
certification and the process for plant-specific departures from a 
certified design. This amendment process places a nuclear plant 
designer on the same footing as the Commission or any other member of 
the public (see 54 FR 15377, first column, April 18, 1989). Therefore, 
it cannot be said that this section makes it easier for a designer to 
amend design certification information than for the NRC to backfit the 
certified design. The amendment and backfitting process uses the phrase 
``certification information'' in order to distinguish the rule language 
in the DCRs from the design certification information (e.g., Tier 1 and 
Tier 2) that is incorporated by reference in the DCRs.
    No matter who proposes it, a generic change under Sec.  52.63(a)(1) 
will not be made to a DCR while it is in effect unless the change: (1) 
is necessary for compliance with Commission regulations applicable and 
in effect at the time the certification was issued; (2) is necessary to 
provide adequate protection of the public health and safety or common 
defense and security; (3) reduces unnecessary regulatory burden and 
maintains protection to public health and safety and common defense and 
security; (4) provides the detailed design information necessary to 
resolve selected design acceptance criteria; (5) corrects material 
errors in the certification information; (6) substantially increases 
overall safety, reliability, or security of a facility and the costs of 
the change are justified; or (7) contributes to increased 
standardization of the certification information.
    Paragraphs (a)(1)(i) and (a)(1)(ii) did not change in the final 
rule. Paragraph (a)(1)(i) provides the compliance exception to the 
NRC's backfit process. Paragraph (a)(1)(ii) sets forth the special

[[Page 49445]]

backfit criteria, which uses the adequate protection standard rather 
than the backfit standard in 10 CFR 50.109. The remaining paragraphs 
permit amendments of design certification information without meeting 
the special backfit requirement in Sec.  52.63(a)(1)(ii).
    Paragraph (a)(1)(iii) allows the Commission to change the design 
certification rule language to reduce unnecessary regulatory burdens, 
i.e., incorporate the revised Sec.  50.59 change criteria, or change 
the 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, i.e., 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 RG 1.174).
    Paragraph (a)(1)(iv) allows for generic resolutions of design 
acceptance criteria (DAC) by amending DCRs. The DAC are a special type 
of ITAAC that are used to verify the resolution of design issues where 
sufficient design information was not provided in the design 
certification application. By generically resolving DAC with the 
amendment process, the Commission achieves resolution of additional 
design issues, achieves finality for those issue resolutions, and 
avoids repetitive consideration of those design issues in individual 
combined license proceedings. Also, the amendments will enhance 
standardization by further completing the certification information. 
The NRC staff will review the amendment application to ensure that the 
DAC are met and that the new design information conforms with the 
applicable regulations.
    Paragraph (a)(1)(v) allows for generic resolutions of material 
errors in the certification information. This provision is only to be 
used to correct a material error, which is an error that significantly 
and adversely affects a design function or analysis conclusion 
described in the design control document (certification information). 
The Commission wants to correct material errors so that these errors 
will not have to be addressed in individual licensing proceedings.
    Paragraph (a)(1)(vi) allows for generic amendments of certification 
information that will substantially increase the overall safety, 
reliability, or security of facility design, construction, or operation 
provided that the direct and indirect costs of implementation of the 
amendment are justified in view of this increased safety, reliability, 
or security. This amendment process will function similar to the 
backfitting process in 10 CFR 50.109.
    Finally, paragraph (a)(1)(vii) allows for generic amendments that 
would increase the standardization of certification information in 
referencing applications. The Commission is still committed to 
achieving and maintaining the benefits of standardization. Therefore, 
the final rule allows for generic amendments of certification 
information through this additional process, provided that the 
amendment is applied to all plants that reference the DCR. This 
paragraph will allow applicants and licensees to request corrections or 
changes to certification information through a generic process rather 
than through individual licensing actions. In determining whether to 
codify a proposed amendment under this paragraph, the Commission will 
give special consideration to comments from applicants or licensees who 
referenced the DCR regarding whether they want to backfit their plants 
with these additional changes.
    The process for amending DCRs will be a rulemaking with opportunity 
for public comment under paragraph (a)(2). As part of the rulemaking 
under Sec.  52.63(a)(1), except for Sec.  52.63(a)(1)(ii), the 
Commission will give consideration to whether the benefits justify the 
costs for plants that are already licensed or for which an application 
for a permit or license is under consideration. The duration of the 
amended DCR will be for the same period of time as the original DCR and 
have the same expiration date.
    Once a DCR is amended by rulemaking, under paragraph (a)(3) the 
changes will apply to all future applications referencing the DCR as 
well as all current plants referencing the design certification, unless 
the change has been rendered ``technically irrelevant'' through other 
action taken under paragraphs (a)(4) or (b)(1) of this section. Thus, 
standardization is maintained by ensuring that any amendment to a DCR 
is imposed upon all nuclear power plants referencing the design 
certification rule.
    Paragraph (a)(4) sets forth the criteria that must be met before 
the Commission can impose new requirements by plant-specific order on a 
nuclear plant that references a DCR. Under this paragraph, the 
Commission must meet either the compliance or adequate protection 
backfit criteria and cite one or more special circumstances as defined 
in Sec.  52.7. In addition, the Commission shall consider whether the 
special circumstances that justify the plant-specific order outweigh 
any decrease in safety that may result from the reduction in 
standardization caused by the plant-specific order. This additional 
requirement was added to ensure that the benefits of standardization 
will be preserved.
    Paragraph (a)(5) sets forth the finality of matters that are 
resolved as part of a design certification rulemaking. Each of the DCRs 
have detailed provisions on the issues that were resolved for that 
plant design and detailed processes for changes to and departures from 
certification information (refer to Sections VI and VIII of appendices 
A, B, C, or D to part 52).
    Paragraphs (b)(1) and (b)(2) provide processes for requesting 
exemptions and departures from certification information. As part of 
its adoption of a two-tiered rule structure (refer to SRM on SECY-90-
377, dated February 15, 1991), the Commission codified detailed 
processes for changes to and departures from certification information 
in each of the design certification rules (refer to Section VIII of 
appendices A, B, C, or D to part 52). The processes for a specific 
certified design must be used when requesting exemptions and departures 
from certification information.
    Paragraph (c) identifies the detailed design information that an 
applicant for a combined license must have completed and available for 
audit by the NRC. The NRC expects that design certification applicants 
(vendors) will have this information available during the review of a 
combined license application that references the certified design. 
Because a rule certifying a standard plant design does not belong to 
the designer (vendor), an applicant for a combined license that 
references the DCR could use a vendor other than the applicant that 
achieved the design certification. In that situation, the combined 
license applicant must acquire the detailed design information 
identified in paragraph (c) in order to demonstrate that the new vendor 
has the ability to provide the certified design and that the combined 
license applicant's design information is consistent with the design 
information for the DCR.

Subpart C--Combined Licenses

Section 52.71 Scope of Subpart
    This section describes the scope of the requirements in this 
subpart. Under this subpart an applicant can request a combined 
construction permit and operating license with conditions (combined 
license) for a nuclear power

[[Page 49446]]

facility. The combined license is essentially a combination of a 
construction permit, which requires consideration and resolution of 
many of the issues currently considered at the operating license stage, 
and a conditional operating license. Operation is allowed only after 
the Commission has made the finding that all acceptance criteria in 
ITAAC have been met.
    The combined license application could describe a site and a custom 
design, or it could reference an early site permit (subpart A of part 
52), a standard design certification (subpart B of part 52), a standard 
design approval (subpart E of part 52), or a reactor manufactured under 
a manufacturing licenses (subpart F of part 52) or a combination 
thereof. Although a pre-approved site and certified standard design 
need not be referenced for the combined license, maximum efficiency 
will result if site-related issues, as well as design-related issues, 
have been resolved before commencement of the combined license 
proceeding.
Section 52.73 Relationship to Other Subparts
    The purpose of this section is to explain the relationship of the 
combined license process to the licensing processes in subparts A, B, 
E, and F of 10 CFR part 52.
Section 52.75 Filing of Applications
    This section explains who can file, how to file, and the fees for 
NRC review of an application for a combined license.
Section 52.77 Contents of Applications; General Information
    This section sets forth the type of general information that is 
required to be included in an combined license application, namely, the 
information required by 10 CFR 50.33. Section 50.33 requires that the 
application include information such as the name and address of the 
applicant, a description of the business or occupation of the 
applicant, citizenship information of the applicant, the class of 
license applied for, the use to which the facility will be put, the 
time for which the license is sought, financial qualification 
information, State and local emergency response plans, the earliest and 
latest dates for the completion of construction, and information about 
decommissioning funding. Section 50.33 also provides requirements for 
the handling of Restricted Data or other defense information in an 
application.
Section 52.79 Contents of Applications; Technical Information in Final 
Safety Analysis Report
    The purpose of this section is to identify specific technical 
information to be included in the final safety analysis report as part 
of an application for a combined license. This generally includes the 
same information required of applicants for construction permits and 
operating licenses under 10 CFR part 50.
    This section specifies the complete set of FSAR information needed 
for a combined license that is a stand-alone application, but also 
takes into account that certain information may already have been 
submitted and reviewed in those instances where the application 
references an early site permit (subpart A), a certified design 
(subpart B), a standard design approval (subpart E), a manufacturing 
license (subpart F), or some combination. The required FSAR information 
also includes requirements for descriptions of operational programs 
that need to be included in the FSAR to allow a reasonable assurance 
finding of acceptability. These additional requirements are in support 
of the Commission's direction to the staff in SRM-SECY-02-0067 dated 
September 11, 2002, ``Inspections, Tests, Analyses, and Acceptance 
Criteria for Operational Programs (Programmatic ITAAC),'' that a 
combined license applicant was not required to have ITAAC for 
operational programs if the applicant fully described the operational 
program and its implementation in the combined license application. In 
this SRM, the Commission stated:

[a]n ITAAC for a program should not be necessary if the program and 
its implementation are fully described in the application and found 
to be acceptable by the NRC at the COL stage. The burden is on the 
applicant to provide the necessary and sufficient programmatic 
information for approval of the COL without ITAAC.

    The Commission clarified its definition of fully described in SRM-
SECY-04-0032, ``Programmatic Information Needed for Approval of a 
Combined License Application Without Inspections, Tests, Analyses, and 
Acceptance Criteria,'' dated May 14, 2004, as follows:

    In this context, fully described should be understood to mean 
that the program is clearly and sufficiently described in terms of 
the scope and level of detail to allow a reasonable assurance 
finding of acceptability. Required programs should always be 
described at a functional level and at an increased level of detail 
where implementation choices could materially and negatively affect 
the program effectiveness and acceptability.

    Accordingly, this section contains requirements for descriptions of 
operational programs and their implementation.
    Paragraph (b) describes the information that is needed if the 
application references an early site permit. Although a combined 
license applicant referencing a certified design need not resubmit 
information or analyses submitted in connection with the early site 
permit, the combined license application FSARs must either include or 
incorporate by reference the SSAR for the early site permit. The SSAR 
must be included or incorporated into the combined license FSAR to 
ensure that matters addressed in the SSAR legally become part of the 
FSAR upon issuance of the combined license. This will also ensure that 
the information in the SSAR is subject to control under Sec.  50.59 
after issuance of the combined license. This provision is meant to 
convey that the combined license applicant referencing the early site 
permit does not need to resubmit, for NRC review, information or 
analyses that were already reviewed and resolved in the early site 
permit proceeding (such as information provided in responses to NRC 
requests for additional information). At the same time, this provision 
provides combined license applicants guidance as to what the combined 
license application must contain to be considered complete, including a 
requirement that it contain or incorporate the early site permit SSAR.
    Because an early site permit applicant need not specify a 
particular nuclear plant design, the combined license application must 
demonstrate that the design of the facility falls within the site 
characteristics and postulated design parameters specified in the early 
site permit. If the application does not demonstrate that design of the 
facility falls within the site characteristics and design parameters of 
the early site permit, then, the applicant must request for a variance 
from the early site permit. Paragraph (b) requires that the application 
demonstrate that all terms and conditions in the early site permit, 
excluding terms and conditions imposed under Sec.  50.36b, be satisfied 
by the date of issuance of the combined license. Any terms or 
conditions of the early site permit that could not be met by the time 
of issuance of the combined license must be set forth as terms or 
conditions of the combined license. Early site permit conditions 
imposed under Sec.  50.36b are to be addressed in the environmental 
report and not in the FSAR.

[[Page 49447]]

    Paragraph (b) also addresses emergency planning information 
submitted in a referenced early site permit and requires that the 
combined license application include any new or additional information 
to update or correct information provided with the early site permit 
and to discuss whether the new information may materially change the 
bases for compliance with the applicable NRC requirements. 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 that substantially alters the bases for a previous NRC 
conclusion or constitutes a sufficient basis for Commission to modify 
or impose new terms and conditions related to emergency preparedness 
would be subject to litigation during the combined license proceeding 
in accordance with Sec.  52.39(c). This paragraph also addresses 
referenced early site permit emergency plans that incorporate existing 
emergency plans and requires the combined license application to 
identify changes to the emergency plans that constitute a decrease in 
effectiveness under 10 CFR 50.54(q). This requirement ensures that the 
NRC can review such changes to assess their impact on the emergency 
plans for the proposed combined license facility.
    Paragraph (c) and (d) provide application requirements for a 
combined license that is referencing a standard design approval or a 
standard design certification, respectively. Similar to a combined 
license application referencing an early site permit, a combined 
license application referencing a design approval or design 
certification must either include or incorporate by reference the 
design approval or design certification FSAR. Because a design approval 
or design certification applicant need not specify a particular site, 
the combined license application must demonstrate that characteristics 
of the site fall within the site parameters specified in the design 
approval or design certification. In addition, the plant-specific PRA 
information must use the PRA information for the design certification 
and must be updated to account for site-specific design information and 
any design changes or departures. An applicant referencing a design 
certification must demonstrate that the interface requirements 
established for the design have been met. Applicants referencing either 
a design approval or a design certification must demonstrate that any 
terms and conditions in the design approval or requirements and 
restrictions in the referenced design certification rule will be 
satisfied by the date that the combined license is issued. Any terms or 
conditions of the design approval that cannot be met or satisfied by 
the time of issuance of the combined license must be set forth as terms 
or conditions of the combined license. Likewise, any requirements or 
restrictions of the design certification that cannot be met or 
satisfied by the time of issuance of the combined license must be set 
forth as terms or conditions of the combined license.
    Paragraph (e) describes the information that is needed if the 
combined license application references one or more manufactured 
reactors. Similar to a combined license application referencing an 
early site permit, design approval, or design certification, a combined 
license application referencing one or more manufactured nuclear power 
reactors under subpart F or part 52 must either include or incorporate 
by reference the manufacturing license FSAR. Because a manufacturing 
license applicant need not specify a particular site for the 
installation of a manufactured reactor, the combined license 
application must demonstrate that the site parameters for the 
manufactured reactor are bounded by the site where the manufactured 
reactor is to be installed and used. In addition, the plant-specific 
PRA information must use the PRA information for the manufactured 
reactor and must be updated to account for site-specific design 
information and any design changes or departures. The combined license 
application must also demonstrate that the interface requirements 
established for the design have been met and that any terms and 
conditions in the manufacturing license will be satisfied by the date 
that the combined license is issued. Any terms or conditions of the 
manufacturing license that could not be met by the time of issuance of 
the combined license must be set forth as terms or conditions of the 
combined license.
Section 52.80 Contents of Applications; Additional Technical 
Information
    This section covers the required technical contents of a combined 
license application that are not contained in the FSAR. These 
application contents include the proposed ITAAC, the environmental 
report, and information to address an applicant's request to perform 
activities at the site allowed by 10 CFR 50.10(e) before issuance of 
the combined license.
    Paragraph (a) requires the application to include the proposed 
ITAAC and, if the application references an early site permit with 
ITAAC or a design certification, requires the applicant to use the 
ITAAC contained in the early site permit or design certification for 
the applicable portion of the combined license application. ITAAC that 
must be included are those that are necessary and sufficient to 
demonstrate that the facility has been constructed and will be operated 
in conformity with the combined license, the provisions of the Atomic 
Energy Act of 1954 and the Commission's rules and regulations. In 
addition, under Section 52.103(g), the Commission must find that all 
acceptance criteria specified in the license are met before facility 
operation. Because ITAAC are the sole source of acceptance criteria for 
subsequent resolution of items which cannot be fully evaluated prior to 
issuance of a combined license, it is essential that the combined 
license ITAAC include all significant issues that require satisfactory 
resolution before fuel loading.
    This paragraph also provides an applicant for a combined license 
with a process for resolving certain acceptance criteria in one or more 
of the ITAAC before issuance of the combined license. This provision is 
included mainly to allow for completion of DAC at the combined license 
application stage 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. Many DAC were originally written to be verified as part 
of the normal, post-combined license, ITAAC verification process. 
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

[[Page 49448]]

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. Accordingly, 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 process would also allow findings on successful completion of 
inspections or tests of components procured before the issuance of the 
combined license.
    Paragraph (b) requires a complete environmental report in 
accordance with 10 CFR 51.50(c).
    Paragraph (c) requires that, if the applicant is requesting to 
perform any activities at the site allowed by 10 CFR 50.10(e), then the 
applicant must identify and describe the activities and propose a plan 
for redress of the site in the event that the activities are performed 
and either construction is abandoned or the combined license is 
revoked. This paragraph also requires the applicant to 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. These requirements attempt to limit, to the extent 
practicable, the environmental impact of any site work done in the case 
where construction of the nuclear power facility is not completed.
Section 52.81 Standards for Review of Applications
    This section identifies the regulations that the NRC staff will use 
in performing its review of an application for a combined license.
Section 52.83 Finality of Referenced NRC Approvals; Partial Initial 
Decision of Site Suitability
    This section describes the finality of regulatory products that may 
be referenced in a combined license application. Specifically, 
paragraph (a) states that the finality of matters resolved in a 
referenced early site permit, design certification, design approval, or 
manufacturing license are governed by the finality provisions in the 
respective subparts that address each of these regulatory processes. 
Paragraph (b) states that, while a partial decision on site suitability 
is in effect under 10 CFR 2.617(b)(2), the finality provisions in 10 
CFR 2.629 govern the scope and nature of matters resolved in the 
proceeding.
Section 52.85 Administrative Review of Applications; Hearings
    This section identifies the procedural requirements that apply to 
the mandatory combined license hearing. This section also identifies 
that, 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.
Section 52.87 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS)
    This section states that the ACRS will report on those portions of 
the application which concern safety.
Section 52.91 Authorization To Conduct Site Activities
    The purpose of this section is to outline the activities that can 
be performed at the site by a combined license applicant. Paragraph (a) 
of this section discusses the authorization a combined license 
applicant needs to obtain in order to perform limited work activities 
at the site while the NRC is considering the combined license 
application in the case where a combined license applicant does not 
reference an early site permit that contains a redress plan. The 
requirements contained in paragraph (a) discuss work commonly referred 
to as a limited work authorization 1 (LWA-1) that is allowed in 
accordance with the requirements contained in 10 CFR 50.10(e)(1). These 
requirements do not allow the applicant to perform LWA-1 activities 
without first submitting a redress plan and obtaining the separate 
authorization required by 10 CFR 50.10(e)(1). Plans are expected to be 
modeled on the Midland Site Stabilization Report that was submitted on 
October 2, 1986 (ML061710504).
    Paragraph (a) recognizes this possibility and notes that 
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 redress carried 
out under the site redress plan will return the site to an 
aesthetically acceptable and environmentally stable condition.
    Paragraph (b) contains requirements for work commonly referred to 
as an LWA-2. An LWA-2 allows structural work 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. Because the design must be known to obtain authorization for 
LWA-2 activities, an LWA-2 is an option for a combined license 
applicant but not an option for an early site permit holder. A combined 
license applicant may request LWA-2 authority prior to the combined 
license being granted. Paragraph (b) recognizes this possibility and 
notes that authorization may be granted only after the presiding 
officer in the combined license makes the additional finding required 
by 10 CFR 50.10(e)(3)(ii), namely, that there are no unresolved safety 
issues relating to the LWA-2 activities.
    Paragraph (c) of this section clarifies that, if work is performed 
either under an LWA-1, or LWA-2 or both, and the combined license 
application is subsequently withdrawn by the applicant or denied by the 
NRC, then the combined license applicant must redress the site in 
accordance with the terms of the site redress plan. Paragraph (c) of 
this section also provides the combined license applicant with the 
ability to redress the site for an alternate use that was not 
considered at the time that the original redress plan was prepared.
Section 52.93 Exemptions and Variances
    The purpose of this section is to describe the process for combined 
license applicants to obtain exemptions and variances. If the request 
is for an exemption from any part of a referenced design certification 
rule, the Commission can grant the request only if it determines that 
the exemption complies with any exemption provisions in the referenced 
design certification rule, or with Sec.  52.63 if there are no 
applicable exemption provisions in the referenced design certification 
rule. A request for an exemption that is outside the scope of a design 
certification rule must be processed in accordance with the 
requirements contained in Sec.  52.7.
    For the General Electric ABWR, Westinghouse System 80+, 
Westinghouse AP600, and Westinghouse AP1000 designs, these requirements 
are contained in Section VIII, ``Processes for Changes and 
Departures,'' of appendices A, B, C, and D respectively, of 10 CFR part 
52. Section VIII of these appendices discusses the process for 
exemptions from different portions of the design certification rule. 
The section-by-section analysis for these respective rules

[[Page 49449]]

discuss requirements regarding processing of exemptions that are 
expected to be carried forward to future design certification 
rulemakings. Therefore, if applicable, the applicant should refer to 
the respective section-by-section analysis in the portion of the design 
certification rule that discusses exemptions for additional 
information. Exemptions requested in accordance with this section are 
subject to litigation in the same manner as other issues in the 
licensee hearing.
    Paragraph (b) of this section sets forth the process for requesting 
variances from an early site permit if one is referenced in the 
combined license. Paragraph (c) sets forth the process for requesting 
variances from one or more design characteristics, site parameters, 
terms and conditions, or approved design of a manufactured reactor. 
Issuance of a variance is subject to litigation during the combined 
license proceeding in the same manner as other issues material to that 
proceeding.
Section 52.97 Issuance of Combined Licenses
    The purpose of this section is to set forth the process for issuing 
a combined license. Paragraph (a)(1) of this section sets forth the 
requirements relative to the Commission findings that must be made for 
granting of a combined license.
    Paragraph (a)(2) of this section allows for completion of certain 
acceptance criteria in one or more of the ITAAC in a combined license 
being met prior to granting of the combined license. This paragraph 
could apply to DAC found in the applicable design certification rules. 
DAC set forth processes and criteria for completing certain design 
information, such as information about the digital instrumentation and 
control system. Paragraph (a)(2) would allow the Commission to make a 
finding of successful completion of DAC when a combined license is 
issued, if the combined license applicant demonstrates that the DAC 
have been successfully completed. This process would also allow 
findings on successful completion of inspections or tests of components 
procured before the issuance of a combined license. Paragraph (a)(2) 
notes that such a finding will preclude any required finding under 
Sec.  52.103(g) with respect to that ITAAC.
    Paragraph (b) requires the Commission to identify the ITAAC within 
the combined license 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. This definition of 
what ITAAC are intended to accomplish is consistent with that contained 
in Sec.  52.17 regarding early site permits, Sec.  52.47 regarding 
design certifications and Sec.  52.80, which are discussed above. If 
the combined license application references an early site permit with 
ITAAC related to emergency planning information, then the applicant 
must use these ITAAC in the emergency planning information submitted 
with the combined license application. If a combined license applicant 
references a design certification rule, the ITAAC contained in the 
license would be those contained in the design certification rule plus 
any additional ITAAC that were identified during the combined license 
review that were outside the scope of the certified design. If the 
Commission wishes to identify additional ITAAC that fall within the 
scope of the review of the referenced certified design it needs to meet 
the requirements contained in the design certification rule itself (see 
Section VIII.A.3 of appendix A, B, C, and D for the ABWR, System 80+, 
AP600, and AP1000) and the requirements contained in Sec.  52.63. If a 
combined license applicant does not reference an early site permit or a 
certified design, then the ITAAC that are identified by the Commission 
for paragraph (b) of this section are those that were identified during 
the combined license review.
Section 52.98 Finality of Combined Licenses; Information Requests
    This section covers the finality of combined license provisions and 
sets forth the requirements to modify the combined license after it has 
been issued. After issuance of a combined license, the Commission may 
not modify, add, or delete any term or condition of the combined 
license, the design of the facility, the inspections, tests, analyses, 
and acceptance criteria contained in the license which are not derived 
from a referenced standard design certification or manufacturing 
license, except in accordance with the backfit provisions of Sec. Sec.  
52.103 or 50.109, as applicable.
    Paragraphs (b), (c), and (d) outline the applicability of the 
change processes in 10 CFR part 50, Section VIII of the design 
certification rules, and subpart F of 10 CFR part 52 to a combined 
license. The change processes in 10 CFR part 50 apply to a combined 
license that does not reference a design certification rule or a 
reactor manufactured under a manufacturing license. Section 52.98(c) 
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. If the combined license references a reactor 
manufactured under a subpart F manufacturing license, then changes to 
or variances from information within the scope of the manufactured 
reactor's design are subject to the change processes in Sec.  52.171.
    Paragraph (e) was added in 1992, and discussed in the section-by-
section analysis (57 FR 60976; December 23, 1992), as following:

    This section has been amended with regard to making amendments 
to a combined license immediately effective under the so-called 
``Sholly Amendment.'' Under the Energy Policy Act, an amendment to a 
combined license can be made immediately effective if the Commission 
determines there are no significant hazards considerations. This 
section of the rule has been revised to incorporate the statutory 
provisions and previously issued Commission regulations implementing 
the ``Sholly'' amendment. The Commission, however, stresses that it 
will not look with favor upon license amendments to a combined 
license filed shortly before planned operation that could have the 
effect of undermining standardization or changing the scope of 
imminent or pending hearings on conformance issues.

    Paragraph (f) states that any modification to a combined license is 
an amendment to the license and that there must be an opportunity for 
hearing on these amendments. Such amendments would be processed in 
accordance with the requirements contained in 10 CFR 50.90 and 50.91. 
In addition, if the applicant has referenced a certified design, or a 
reactor manufactured under a manufacturing license, additional 
requirements may apply. For example, a combined license that references 
an ABWR certified design may request an exemption from Tier 1 material 
in accordance with the provisions contained in Section VIII.A.4 of 
appendix A of 10 CFR part 52. In such a case, the licensee would have 
to process an exemption in accordance with the requirements contained 
in appendix A to part 52 and 10 CFR 52.63(b)(1) and a license amendment 
in accordance with paragraph (f) of this section.
    Paragraph (g) which is analogous to Sec. Sec.  52.39(f), 52.145(c), 
and 52.171(c),

[[Page 49450]]

provides that NRC information requests must be evaluated before 
issuance to ensure that the burden to be imposed by the information 
request is justified in view of the potential safety significance of 
the issue to be addressed, except when the information requests seeks 
to verify compliance with the current licensing basis of the combined 
license. Information requests may be in the form of a new rule 
requiring submission of information (i.e., a new information collection 
and reporting requirement), or in the form of a NRC staff request for 
information. Information requests by the staff must be in accordance 
with 10 CFR 50.54(f) and must be approved by the EDO or his or her 
designee before the request may be issued.
Section 52.99 Inspection During Construction
    The purpose of this section is to set forth the requirements to 
support the NRC's inspections during construction. A new Sec.  52.99(a) 
has been added to require that the licensee submit to the NRC, no later 
than 1 year after issuance of the combined license or at the start of 
construction as defined in 10 CFR 50.10, whichever is later, its 
schedule for completing the inspections, tests, or analyses in the 
ITAAC. This provision also requires the licensee to submit updates to 
the ITAAC schedule every 6 months thereafter and, within 1 year of its 
scheduled date for initial loading of fuel, licensees must submit 
updates to the ITAAC schedule every 30 days until the final 
notification is provided to the NRC under Sec.  52.99(c). The 
information provided by the licensee will be used by NRC in developing 
the NRC's inspection activities and activities necessary to support the 
Commission's finding whether all of the ITAAC have been met prior to 
the licensee's scheduled date for fuel load. Even in the case where 
there were no changes to a licensee's ITAAC schedule during an update 
cycle, the NRC expect the licensee to notify the NRC that there have 
been no changes to the schedule.
    Section 52.99 has also been amended 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 regulations. The purpose of the 
requirement in Sec.  52.99(b) is to clarify 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 met.
    Section 52.99(c)(1) requires the licensee to notify the NRC that 
the prescribed inspections, tests, and analyses have been performed and 
that the prescribed acceptance criteria have been met. Section 
52.99(c)(1) further requires that the notification contain sufficient 
information to demonstrate that the prescribed inspections, tests, and 
analyses have been performed and that the prescribed acceptance 
criteria have been met.
    Section 52.99(c)(2) requires that, if the licensee has not 
provided, by the date 225 days before the scheduled date for initial 
loading of fuel, the notification required by paragraph (c)(1) of this 
section for all ITAAC, then the licensee shall notify the NRC that the 
prescribed inspections, tests, or analyses for all uncompleted ITAAC 
will be performed and that the prescribed acceptance criteria will be 
met prior to operation (consistent with the Section 185.b requirement 
that the Commission, ``prior to operation,'' find that the acceptance 
criteria in the combined license are met). The notification must be 
provided no later than the date 225 days before the scheduled date for 
initial loading of fuel, and must provide sufficient information to 
demonstrate that the prescribed inspections, tests, or analyses will be 
performed and the prescribed acceptance criteria for the uncompleted 
ITAAC will be met.
    Section 52.99(c) ensures that: (1) The NRC has sufficient 
information to complete all of the activities necessary for the 
Commission to make a determination as to whether all of the ITAAC have 
been or will be met prior to initial operation; and (2) interested 
persons will have access to information on both completed and 
uncompleted ITAAC at a level of detail sufficient to address the AEA 
Section 189.a(1)(B) threshold for requesting a hearing on acceptance 
criteria. It is the licensee's burden to demonstrate compliance with 
the ITAAC and the NRC expects the information submitted under paragraph 
(c)(1) to contain more than just a simple statement that the licensee 
believes the ITAAC has been completed and the acceptance criteria met. 
The NRC expects the notification to be sufficiently complete and 
detailed for a reasonable person to understand the bases for the 
licensee's representation that the inspections, tests, and analyses 
have been successfully completed and the acceptance criteria have been 
met. The term ``sufficient information'' requires, at a minimum, a 
summary description of the bases for the licensee's conclusion that the 
inspections, tests, or analyses have been performed and that the 
prescribed acceptance criteria have been met. Furthermore, with respect 
to uncompleted ITAAC, it is the licensee's burden to demonstrate that 
it will comply with the ITAAC and the NRC expects the information that 
the licensee submits under paragraph (c)(2) to be sufficiently detailed 
such that the NRC can determine what activities it will need to 
undertake to determine if the acceptance criteria for each of the 
uncompleted ITAAC have been met, once the licensee notifies the NRC 
that those ITAAC have been successfully completed and their acceptance 
criteria met. The term ``sufficient information'' requires, at a 
minimum, a summary description of the bases for the licensee's 
conclusion that the inspections, tests, or analyses will be performed 
and that the prescribed acceptance criteria will be met. In addition, 
``sufficient information'' includes, but is not limited to, a 
description of the specific procedures and analytical methods to be 
used for performing the inspections, tests, and analyses and 
determining that the acceptance criteria have been met.
    The NRC notes that, even though it did not include a provision 
requiring the completion of all ITAAC by a certain time prior to the 
licensee's scheduled fuel load date, the NRC staff will require some 
period of time to perform its review of the last ITAAC once the 
licensee submits its notification that the ITAAC has been successfully 
completed and the acceptance criteria met. In addition, the Commission 
itself will require some period of time to perform its review of the 
staff's conclusions regarding all of the ITAAC and the staff's 
recommendations regarding the Commission finding under Sec.  52.103(g). 
Therefore, licensees should structure their construction schedules to 
take into account these time periods.
    A new paragraph (d) 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. If an activity is subject to an ITAAC 
derived from a referenced standard design certification and the 
licensee has not demonstrated that the ITAAC has been met, the licensee 
may take corrective actions to

[[Page 49451]]

successfully complete that ITAAC or request an exemption from the 
standard design certification ITAAC, as applicable. A request for an 
exemption must also be accompanied by a request for a license amendment 
under Sec.  52.98(f). Also, if an activity that is subject to an ITAAC 
is not derived from a referenced standard design certification and the 
licensee has not demonstrated that the ITAAC has been met, the licensee 
may take corrective actions to successfully complete that ITAAC or 
request a license amendment under Sec.  52.98(f).
    Paragraph (e)(1) of this section indicates that the NRC is 
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. When part 52 was 
first adopted by the Commission in 1989 (April 18, 1989; 54 FR 15372), 
the rule 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 Commission 
revised the language in this portion of the rule in 1992 (December 23, 
1992; 57 FR 60975) 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 this portion of the rule in 1992 simply 
reflected 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 prohibit the Commission's long-standing practice of delegating to 
the NRC staff the responsibility for performing the necessary 
activities, including audits and inspections, to ensure that ``the 
required inspections, tests, and analyses in the ITAAC are performed.'' 
Accordingly, the language from the 1992 rule change is retained in this 
final rule.
    Paragraph (e)(1) requires the NRC to publish, at appropriate 
intervals until the last date for submission of requests for hearing 
under Sec.  52.103(a), notices in the Federal Register of the NRC 
staff's determination of the successful completion of inspections, 
tests, and analyses. Paragraph (e)(2) provides that the NRC shall make 
publicly available the licensee notifications under paragraphs (c)(1) 
and (c)(2). In general, the NRC expects to make the paragraph (c)(1) 
notifications availability shortly after the NRC has received the 
notifications and concluded that they are complete and detailed. 
Furthermore, by the date of the Federal Register notice of intended 
operation and opportunity to request a hearing on whether acceptance 
criteria have been or will be met (under Sec.  52.103(a)), the NRC will 
make available the notifications under paragraph (c)(2), and the 
notifications under paragraph (c)(2) for all ITAAC for which paragraph 
(c)(1) notifications have not been provided by the licensee.
Section 52.103 Operation Under a Combined License
    The purpose of this section is to set forth the requirements for 
operation under a combined license. This section has been previously 
discussed in a section-by-section analysis for the 1992 revisions to 
part 52 (57 FR 60976; December 23, 1992) which the NRC adopted in 
response to the Energy Policy Act of 1992. The 1992 section-by-section 
analysis states:

    In an effort to adhere as closely as possible to the new 
statutory requirements of the Energy Policy Act, the NRC has 
replaced most of its old Sec.  52.103 with the text of section 2802 
of that Act. Under the revised language, any request for a post-
construction hearing must show, prima facie, both that one or more 
of the acceptance criteria are not or will not be met, and those 
specific operational consequences of nonconformance that would be 
contrary to providing reasonable assurance that the public health 
and safety will be adequately protected. The Commission may permit 
interim operation of a facility pending a hearing if it determines 
that this assurance exists. The Commission has the discretion to 
decide if any post-construction hearing will use formal or informal 
hearing procedures, and it must state publicly the reasons for 
choosing either set of procedures. The Commission must find, prior 
to operation of the facility, that the acceptance criteria have been 
met.

    Paragraph (a) of this section is revised to require licensees to 
notify the NRC of its schedule date for initial loading of fuel no 
later than 270 days before the scheduled date and to notify the NRC of 
updates to its schedule every 30 days thereafter. This information will 
be used by the NRC to develop the notice of intended operation in the 
Federal Register, which must be published not less than 180 days before 
the licensee's initial fuel load date, as required by Section 
189.a.(1)(B) of the AEA. In addition, paragraph (a) addresses the 
possibility that an applicant for a combined license may choose to 
resolve certain acceptance criteria in one or more of the ITAAC 
required by Sec.  52.80 before issuance of the combined license. In 
such a case, if the Commission makes a finding in accordance with Sec.  
52.97 associated with these ITAAC at the time that a combined license 
is granted, these ITAAC would not be subjected to a hearing opportunity 
again under paragraph (a) of this section. The section-by-section 
analysis for Sec.  52.97 discusses this issue in more detail.
    Paragraph (b) provides the criteria that must be met for any 
request for a hearing on whether the facility complies or will comply 
with the acceptance criteria. The petitioner must set forth with 
reasonable specificity the facts and arguments which form the basis for 
the request. These provisions are designed to accord finality to the 
Commission's earlier decisions regarding the facility and to ensure 
that any proceeding is focused on significant safety issues.
    Paragraph (c) requires the Commission to expeditiously either deny 
or grant any request for a hearing under this section. If a request is 
granted, the Commission must determine whether to allow interim 
operation of the facility based on reasonable assurance of adequate 
protection of the public health and safety.
    Paragraph (d) provides that the Commission will determine the 
appropriate hearing procedures in accordance with 10 CFR part 2 for any 
hearing under paragraph (a) of this section. Under Sec.  2.309, as 
adopted by the Commission in 2004 (69 FR 2182; January 14, 2004), such 
a hearing would ordinarily be conducted under subpart L of part 2. 
However, the Commission may direct, in the notice of required by 
paragraph (a) or in a subsequent order, that any hearing that may be 
conducted in a particular combined license proceeding under paragraph 
(a) use other, less formal hearing procedures, consistent with the 
requirements of the AEA. Any such Commission direction is consistent 
with the Commission's statement in the SOC for the 1989 final part 52 
rulemaking (54 FR 15372, 15383; April 18, 1989) that any hearing held 
under former Sec.  52.103(b)(2)(i) (Sec.  52.103(b) in this final rule) 
will use informal procedures to the maximum extent practical and 
permissible under law.
    Paragraph (e) states that the Commission will, to the maximum 
extent possible, render a decision on issues raised in any hearing 
request within 180 days of the publication of the notice or by the 
anticipated date for initial fuel load, whichever is later.

[[Page 49452]]

    Paragraph (f) provides requirements related to the submittal of 
petitions to modify the terms and conditions of a combined license and 
states that fuel loading and operation under a combined license will 
not be affected by the granting of a petition unless the Commission 
makes an order immediately effective.
    Paragraph (g) prohibits the licensee from operating the facility 
until the Commission makes a finding that the acceptance criteria in 
the combined license are met (except for acceptance criteria that the 
Commission found were met when the combined license was issued). The 
NRC believes that the rule should reflect, as closely as possible, the 
statutory requirement in Section 185.b of the AEA. Although the NRC has 
historically viewed ``operation'' as including loading of fuel into the 
reactor, the NRC believes it is not necessary to change the language of 
Sec.  52.103(g) to continue the historical practice.
    Paragraph (h) of this section incorporates 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). 
This paragraph states that ITAAC do not, by virtue of their inclusion 
in the design certification rule or combined license, constitute 
regulatory requirements after the licensee has received authorization 
to load fuel or for any renewal of the license. However, subsequent 
modifications to the facility or procedures described in the FSAR must 
comply with the requirements in Sec.  52.98.
Section 52.104 Duration of Combined License
    This section addresses the duration of a combined license which is 
a period not to exceed 40 years from the date that the Commission makes 
the finding that the acceptance criteria in the license are met, in 
accordance with Sec.  52.103(g). Where the Commission has allowed 
operation during an interim period under Sec.  52.103(c), the period of 
operation is not to exceed 40 years from the date allowing operation 
during the interim period. This provision implements Section 621 of the 
Energy Policy Act of 2005 which amended Section 103c. of the AEA. The 
AEA provided that the 40 year duration started on the date that the 
Commission authorized construction of the facility (i.e., the date of 
issuance of the combined license).
Section 52.105 Transfer of Combined License
    This section states that a combined license may by transferred in 
accordance with 10 CFR 50.80, ``Transfer of licenses.'' Section 50.80 
provides the requirements regarding application for a license transfer. 
All license transfers must be approved by the Commission.
Section 52.107 Application for Renewal
    This section states that an application to renew a combined license 
must be in accordance with 10 CFR part 54, ``Requirements for Renewal 
of Operating Licenses for Nuclear Power Plants.''
Section 52.109 Continuation of Combined License
    This section, which is analogous to Sec.  50.51, provides 
requirements for a combined license facility that has permanently 
ceased operations and states that the license continues in effect 
beyond the expiration date until the Commission notifies the licensee 
in writing that the license is terminated. During this period, the 
licensee is required to decommission and decontaminate the facility; 
maintain the facility, including the spent fuel, in a safe condition; 
and continue to follow the NRC's regulations and the provisions of the 
combined license.
Section 52.110 Termination of License
    This section, which is analogous to Sec.  50.82, provides 
requirements the termination of a combined license. These provisions 
include a requirement to notify the NRC within 30 days when a licensee 
has decided to permanently cease operations and to submit a 
certification to the NRC once fuel has been permanently removed from 
the reactor vessel. This section also requires decommissioning of the 
facility within 60 years of permanent cessation of operations and 
outlines requirements regarding decommissioning activities.

Subpart E--Standard Design Approvals

Section 52.131 Scope of Subpart
    This section describes the scope of this process for design 
approvals of standard nuclear power plants or major portions thereof, 
i.e., a nuclear steam supply system or balance of plant. Under this 
subpart an applicant may request pre-approval of a standard nuclear 
power plant design, separate from a site review or other licensing 
action, and subsequently have that design approval referenced in an 
application to build a nuclear power plant. This licensing process was 
first adopted by the Commission in 1975 and has been used many times.
Section 52.133 Relationship to Other Subparts
    The purpose of this section is to explain the relationship of the 
standard design approval process to the processes set forth in subparts 
B, C, and F of 10 CFR part 52, which provide for design certifications, 
combined licenses, and manufacturing licenses. The Commission continues 
to believe that the best approach for obtaining early resolution of 
design issues is through the design certification process in subpart B 
of this part. Applicants for a design approval have the option of also 
applying for design certification. Applicants for a combined license or 
a manufacturing license may reference a design approval.
Section 52.135 Filing of Applications
    This section explains how to file an application for a standard 
design approval and how the fees for NRC's review of the application 
will be assessed. Applications are limited to final design information, 
in order to remove the unpredictability of issuing a construction 
permit that references only preliminary design information and 
initiating construction while the final design information is being 
completed. Approval of a final standard design ensures early 
consideration and resolution of technical matters by the NRC staff 
before there is any substantial commitment of resources, which will 
greatly enhance regulatory stability and predictability.
Section 52.136 Contents of Applications; General Information
    This section identifies the general information that must be 
included in all applications.
Section 52.137 Contents of Applications; Technical Information
    The purpose of this section is to identify the technical 
information that must be included in an application for a design 
approval. Paragraphs (a) and (c) describe information that must be 
included in the FSAR, which is included in the application, and 
paragraph (b) describes the information that must also be included in 
the application but does not need to be included in the FSAR. 
Applications for a major portion of the plant design, such as the 
nuclear steam supply system, only need to contain the technical 
information that is applicable to the major portion of the plant for 
which NRC staff approval is requested.

[[Page 49453]]

    Many of the requirements in this section were taken from 10 CFR 
50.34 or are pointers to technical requirements in parts 20, 50, and 73 
that must be addressed in the application. The requirements taken from 
Sec.  50.34 are a subset of the information required of applicants for 
construction permits and operating licenses. Other requirements came 
from appendix O to part 50 or were created by the Commission during its 
simultaneous reviews of applications for design approvals and design 
certifications.
    Although an applicant for design approval does not need to specify 
a particular site for the nuclear power plant, which is required in a 
combined license application, it does need to identify the site 
parameters that the standard nuclear power plant or major portion 
thereof is designed to meet, e.g., postulated values for the safe 
shutdown earthquake response spectra and maximum tornado wind speed. 
These parameters are usually selected to envelop a large portion of 
nuclear plant sites in the United States. Once the design is approved 
by the NRC, conformance of the actual site characteristics with the 
established site parameters must be demonstrated by an applicant 
referencing the design approval and verified by the NRC staff at the 
time that the referencing application is submitted, i.e., combined 
license application.
    Paragraph (a)(7) requires the applicant for design approval to 
describe its qualifications to design and analyze a standard nuclear 
power plant.
    In its staff requirements memorandum (SRM) on SECY-90-377, 
``Requirements for Design Certification under 10 CFR part 52,'' dated 
February 15, 1991, the Commission stated that information submitted in 
an application should incorporate the experience from operating events 
in current designs which we want to prevent in the future. Therefore, 
for plant designs that are based on or are evolutions of nuclear plants 
that have operated in the United States, paragraph (a)(22) requires the 
applicant to demonstrate how relevant operating experience insights, 
from NRC's generic letters and bulletins issued after the most recent 
revision of the applicable SRP and 6 months before the docket date of 
the application, have been incorporated into the plant design. 
Operating experience includes consideration of operating events and the 
reliability and performance of structures, systems, and components. If 
the application is for a design that is not based on or is not an 
evolution of a nuclear plant that operated in the United States, the 
applicant must demonstrate how insights from any relevant international 
operating experience have been incorporated into that plant design.
    In its SRMs, dated June 26, 1990, and July 21, 1993, on SECY-90-16, 
``Evolutionary Light-Water Reactor Certification Issues and their 
Relationship to Current Regulatory Requirements,'' and SECY-93-087, 
``Policy, Technical, and Licensing Issues Pertaining to Evolutionary 
and Advanced Light-Water Reactor Designs,'' respectively, the 
Commission approved NRC staff recommendations for selected preventative 
and mitigative design features for future light-water reactor designs. 
Paragraph (a)(23) requires the applicant to provide a description and 
analysis of those design features discussed in SECY-90-16 and SECY-93-
87.
    Paragraph (a)(U0 ) requires the application to describe the 
interfaces for those design features that are outside the scope of the 
approved design, e.g., service water intake structure or ultimate heat 
sink or, if the application is for approval of a major portion of the 
plant design, the interfaces between the nuclear steam supply system 
and the balance of plant.
    Paragraph (a)(25) requires the applicant to provide a description 
of the design-specific PRA and its results. Guidance on meeting the PRA 
information requirements will be provided in separate regulatory 
guidance documents.
    Paragraph (b) requires applications for ``advanced'' nuclear power 
plants to meet the design qualification testing requirements in 10 CFR 
50.43(e). Advanced designs differ significantly from evolutionary 
light-water reactor designs or incorporate, to a greater extent than 
evolutionary designs do, simplified, inherent, passive, or other 
innovative means to accomplish their safety functions. Examples of 
advanced nuclear power plant designs include General Atomic's Modular 
High Temperature Gas-Cooled Reactor, General Electric's Simplified 
Boiling Water Reactor, and Westinghouse's AP600.
Section 52.139 Standards for Review of Applications
    This section sets forth the parts of 10 CFR that contain applicable 
requirements for the technical review of applications for a design 
approval. The applicability of these requirements is specified in the 
identified parts. The Commission recognizes that new designs may 
incorporate design features that are not addressed by the current 
standards in 10 CFR parts 20, 50 and its appendices, 73, or 100 and 
that new standards may be required to address these new design 
features. The Commission will determine whether rulemakings are needed 
or appropriate to resolve generic safety issues that are applicable to 
multiple designs.
Section 52.141 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS)
    This section states that the application for design approval shall 
be sent to the ACRS for its review of safety issues.
Section 52.143 Staff Approval of Design
    This section states that upon completion of the NRC staff's review 
of the standard design and receipt of a letter report from the ACRS, 
the staff shall issue a final safety evaluation report (FSER) and make 
that report available on the NRC's Web site. Also, if the FSER 
demonstrates that the standard design is acceptable, the Director of 
the Office of New Reactors or the Office of Nuclear Reactor Regulation 
may issue a final design approval with appropriate terms and 
conditions. The NRC's approval of a standard design is commonly 
referred to as an FDA because it is an approval of final design 
information.
Section 52.145 Finality of Standard Design Approvals; Information 
Requests
    This section states that a valid FDA must be relied upon by the 
ACRS and NRR in any review of a license application that references the 
FDA unless significant new information substantially affects the 
staff's FSER. The Commission, Atomic Safety Licensing Board Panel, or 
presiding officers are not bound by NRC staff determinations in the FDA 
or FSER for the standard plant design. Therefore, there is no issue 
preclusion in the mandatory hearing for a combined license that 
references an FDA. Generic changes to the standard design can be made 
as a compliance backfit or under the backfit process in 10 CFR 50.109. 
Under paragraph (c), the justification for requests for information to 
FDA holders must be approved by the EDO or his or her designee, in 
accordance with the process set forth in 10 CFR 50.54(f).
Section 52.147 Section Duration of Design Approval
    The purpose of this section is to specify the time period that an 
FDA can be referenced in a construction permit, operating license, 
combined license, or manufacturing license application.

[[Page 49454]]

Subpart F--Manufacturing Licenses

Section 52.151 Scope of Subpart
    This new section is analogous to the ``scope of subpart'' sections 
in subparts A through C of part 52 (e.g., Sec. Sec.  52.13, 52.41, 
52.71). Section 52.151 describes the general subject matter of subpart 
F as the requirements and procedures applicable to NRC issuance of 
licenses authorizing the manufacture of nuclear power reactors to be 
installed at sites not identified in the manufacturing license 
application. This subpart does not cover the manufacture of 
subcomponents (e.g., a pump or a reactor pressure vessel) or major 
subassemblies (e.g., an integrated module consisting of a pump, piping 
and instrumentation and control) for installation in a nuclear power 
plant, either on a specific site, or being delivered for integration 
into a nuclear power plant under a manufacturing license issued under 
this subpart. For purposes of this subpart, a manufactured ``nuclear 
power reactor'' would not include site-specific SSCs such as the site 
foundation or SSCs related to the ultimate heat sink.
Section 52.153 Relationship to Other Subparts
    This new section is analogous to the ``relationship to other 
subpart'' sections in subparts A through C of part 52 (e.g., Sec. Sec.  
52.13, 52.43, 52.73). Section 52.153 explains how this subpart relates 
to other licensing processes in parts 50 and 52, as well as to the 
regulatory approvals in part 52.
    A manufactured reactor may only be transported to and installed at 
a site for which either a construction permit under part 50 or a 
combined license under part 52 has been issued to a licensee, as stated 
in paragraph (a). However, the licensing requirements associated with 
transport of a manufactured reactor from its place of manufacture to 
the site where it is to be installed and operated are not addressed in 
this rulemaking.
    The NRC will issue a manufacturing license only if it approves the 
final design of the reactor to be manufactured. Paragraph (b) provides 
that the manufacturing license applicant may reference either a 
standard design certification rule or a standard design approval, in 
order to speed the NRC's review of the manufacturing license 
application. The language of paragraph (b) has been corrected in the 
final rule by deleting the reference to ``preliminary or final'' design 
approvals, inasmuch as the final part 52 rule does not provide for 
preliminary design approvals.
Section 52.155 Filing of Applications
    This new section is analogous to the ``filing of applications'' 
sections in subparts A through C of part 52 (e.g., Sec. Sec.  52.15, 
52.45, 52.75). Section 52.155 addresses who may file an application for 
a manufacturing license, the administrative requirements with respect 
to filing (referring to Sec. Sec.  52.3 and 50.30), and the fees for 
filing and review of the application (referring to 10 CFR part 170). 
With respect to these matters, a manufacturing license application is 
no different than any other license application under parts 50 or 52, 
and the applicant shall comply with all of these administrative 
requirements (which have been revised as part of the final rule to 
refer, as necessary, to manufacturing licenses).
Section 52.156 Contents of Applications; General Information
    This new section is analogous to the ``contents of application; 
general information'' sections in subparts A through C of part 52 
(e.g., Sec. Sec.  52.16, 52.46, 52.77). Section 52.156 requires that 
the applicant include the information set forth in Sec.  50.33(a) 
through (d) and (j), which are the same information required to be 
supplied by applicants of construction permits, early site permits, 
operating licenses, and combined licenses. Paragraphs (a) through (d) 
of Sec.  50.33 require an application to include information 
identifying the applicant, including its name, address, business or 
occupation, and certain corporate information, including whether it is 
owned, controlled, or dominated by an alien, foreign corporation, or 
foreign government. Paragraph (j) of Sec.  50.33 requires the applicant 
to segregate and protect any Restricted Data or other defense 
information from unclassified information. Manufacturing license 
applicants should note that there are other NRC requirements governing 
Restricted Data or National Security Information in other parts of 10 
CFR Chapter I, including 10 CFR parts 10, 50, and 95.
Section 52.157 Contents of Applications; Technical Information in Final 
Safety Analysis Report
    This new section is analogous to the ``contents of application; 
technical information'' sections in subparts A through C of part 52 
(e.g., Sec. Sec.  52.17, 52.47, 52.79). Section 52.157 identifies the 
technical information that must be included in an application for a 
manufacturing license. These requirements were modeled on those 
subparts, in particular subpart B's provisions dealing with standard 
design certifications, because of the commonality with respect to the 
nature and scope of NRC approval of the design in both regulatory 
processes. As with the existing part 50 licensing process, and part 
52's combined license and standard design certification processes, the 
manufacturing license application must include an FSAR. The FSAR 
contains the information necessary for the NRC to determine the safety 
of the reactor design to be manufactured and the adequacy of the 
applicant's proposed means of assuring that the manufacturing conforms 
to the design. The FSAR must contain a level of detail sufficient to 
permit preparation of construction and installation specifications by 
an applicant who seeks to use the manufactured reactor, and for the NRC 
to prepare acceptance and inspection requirements.
    The information required to be included in the manufacturing 
license FSAR is largely the same as what is required for a design 
certification or combined license, but the requirements have been 
modified as necessary to reflect the fact that the design and 
manufacture of a reactor is being approved by license, but that the 
reactor must be transported to a site and integrated into site specific 
plant elements in order to operate. In addition, unlike the case with a 
design certification, the NRC is not distinguishing between 
evolutionary plants versus more advanced plants with respect to the 
level of detail required to be developed to support the license 
application. The NRC expects that the designs of all manufactured 
plants will be completed at a level of detail sufficient for: (1) The 
holder of the manufacturing license to develop procurement, 
construction and installation specifications; and (2) the NRC to 
develop acceptance and inspection requirements.
    Paragraph (a) requires that the FSAR contain the principal design 
criteria for the reactor to be manufactured, and references appendix A 
to 10 CFR part 50 as establishing minimum requirements for the 
principal design criteria for water-cooled nuclear power plants. The 
NRC expects to develop technology-neutral design criteria for non-light 
water cooled reactor designs in the future. This requirement was drawn 
from Sec.  50.34(a)(3)(i).
    Paragraph (b) requires that the FSAR describe the design bases and 
the relation of the design bases to the principal design criteria that 
are identified in accordance with paragraph (a). This requirement was 
drawn from Sec.  50.34(a)(3)(ii).

[[Page 49455]]

    Paragraph (c) requires that the FSAR describe and analyze the 
structures, systems, and components of the reactor to be manufactured, 
with the objective of demonstrating that the necessary safety functions 
will be accomplished. This requirement was drawn from Sec.  50.34(a)(1) 
and (b)(2), but modified to reflect the fact that a manufacturing 
license represents approval of a final reactor design.
    Paragraph (d) requires that the FSAR describe the safety features 
that are engineered into the reactor. This requirement was drawn from 
Sec.  50.34(a)(1)(ii)(D), but modified to reflect the fact that a 
manufacturing license represents approval of a final reactor design.
    Paragraph (e) requires the FSAR to describe the kinds and 
quantities of radioactive materials expected to be produced in the 
operation and the means for controlling and limiting radioactive 
effluents and radiation exposures within the limits set forth in part 
20.
    Paragraph (f) requires that the FSAR include that information 
necessary to establish that the design of the reactor to be 
manufactured complies with 18 delineated technical requirements in 10 
CFR part 50. Applicants and licensees should note that the part 50 
requirements listed in paragraph (f) do not constitute the sum total of 
requirements in part 50 for which either an applicant for or holder of 
a manufacturing license must comply with in its application and 
throughout the life of its license. Rather, the listed requirements in 
paragraph (f) simply represents the minimum necessary content of the 
FSAR for a manufacturing license. The part 50 requirements listed in 
paragraph (e) are mainly applicable to LWRs. Potential applicants and 
licensees should also note that the NRC may, in the future, adopt 
additional technical requirements in part 50 applicable to LWRs. If the 
NRC believes that future manufacturing license holder's compliance with 
that new requirement must be documented and controlled through the 
FSAR, the NRC will make a conforming change in Sec.  52.157 to refer to 
the new part 50 requirement. A similar course would also be followed if 
the NRC backfits, in accordance with the finality provisions in Sec.  
52.171, the new requirement on existing manufacturing licenses.
    Paragraph (f)(19) requires that the FSAR include the site 
parameters postulated for the design of the manufactured reactor. 
Although an applicant for a manufacturing license does not need to 
specify a particular site where the manufactured reactor will be 
integrated into a nuclear power plant, as in a combined license 
application, it does need to identify the site parameters, under 
paragraph (f)(20), that the manufactured reactor is designed to meet, 
e.g., postulated values for the safe-shutdown earthquake response 
spectra and maximum tornado wind speed. These parameters are usually 
selected to envelop a large portion of nuclear plant sites in the 
United States. Once the manufacturing license is issued by the NRC, 
conformance of the actual site with the established site parameters 
must be demonstrated by the applicant referencing the use of the 
manufactured reactor.
    Paragraph (f)(20) requires the FSAR to describe the interface 
requirements for those design features that are outside the scope of 
the design of the manufactured reactor, e.g., service water intake 
structure or ultimate heat sink, and paragraph (f)(21) requires 
justification that compliance with the interface requirements in 
paragraph (g) can be verified through inspections or tests (which may 
be conducted at the plant where the manufactured reactor is utilized, 
or elsewhere, e.g., the place of manufacture) or analysis. This 
paragraph does not require, however, that the FSAR contain ``acceptance 
criteria'' for determining whether the interface requirements have been 
met.
    Paragraph (f)(22) requires the FSAR to include a representative 
conceptual design for the nuclear power facility using the manufactured 
reactor. This will be used by the NRC in its review of the FSAR, to 
assess the adequacy of the interface requirements in paragraph (g) of 
this section, and to help the Commission in determining the adequacy of 
the site parameters and design characteristics to be included in the 
manufacturing license. The conceptual design will not, however, be 
approved as part of the manufacturing license and the Commission does 
not anticipate directly requiring a nuclear power plant utilizing the 
manufactured reactor to use the conceptual design. Instead, the 
Commission intends to use site parameters, design characteristics, 
ITAAC, and interface requirements to ensure that the manufactured 
reactor will be utilized safely at a specific nuclear power plant.
    Paragraph (f)(23) requires the applicant to provide a description 
and analysis of design features to address prevention and mitigation of 
severe accidents, consistent with the Commission's SRM on SECY-91-229, 
``Severe Accident Mitigation Design Alternatives for Certified Standard 
Designs,'' dated October 25, 1991.
    Paragraph (f)(U0 ) is reserved to accommodate any new requirement 
for the contents of an FSAR submitted as part of an application for a 
manufacturing license which the Commission may adopt in the future.
    Paragraph (f)(25) requires FSARs for modular nuclear power plant 
designs to describe and analyze the various options for the 
configuration of the multi-reactor nuclear power plant. Modular nuclear 
power plant designs are defined in Sec.  52.1. Modular designs are not 
portions of a single nuclear plant, rather they are separate nuclear 
reactors with some shared or common systems.
    Paragraphs (f)(26)(i), (ii), (iii), and (v) focus on FSAR 
information necessary to demonstrate applicants technical, managerial, 
and organizational capability and resources to design and manufacture a 
nuclear power reactor consistent with the approved design, and in 
accordance with all applicable requirements.
    Paragraph (f)(26)(iv) requires the FSAR to include proposed 
procedures for the preparation of the manufactured reactor for 
shipping, the conduct of shipping, and for verifying the condition of 
the manufactured reactor upon receipt at the site. However, the holder 
of the manufacturing license need not be responsible for implementing 
the procedures for verifying the condition of the reactor upon receipt 
at the site. The NRC will require the licensee whose application 
referenced the use of the manufactured reactor to implement the 
approved verification procedures (this could be done as a license 
condition). With respect to shipping, the holder of the manufacturing 
license may use an agent (e.g., a shipping company) to transport the 
reactor. To ensure that the shipping requirements in the manufacturing 
license are complied with by the third party transporter, the NRC has 
included a provision in Sec.  52.167(c)(2) requiring the manufacturing 
license holder to include, in any contract governing the transport of a 
manufactured reactor from the place of manufacture to any other 
location, a provision requiring that the person or entity transporting 
the manufactured reactor to comply with all NRC-approved shipping 
requirements in the manufacturing license.
    For plant designs that are based on or are evolutions of nuclear 
plants that have operated in the United States, paragraph (f)(29) 
requires the applicant to demonstrate how relevant operating experience 
insights, from NRC's generic letters and bulletins issued after the 
most recent revision of the applicable SRP and 6 months before the 
docket

[[Page 49456]]

date of the application, have been incorporated into the design of the 
reactor to be manufactured. Operating experience includes consideration 
of operating events and the reliability and performance of structures, 
systems, and components. If the application is for a design that is not 
based on or is not an evolution of a nuclear plant that operated in the 
United States, the applicant must demonstrate how insights from any 
relevant international operating experience have been incorporated into 
that manufactured reactor design.
    Paragraph (f)(31) requires that the FSAR include a description of 
the design--specific probabilistic risk assessment and its results.
Section 52.158 Contents of Application; Additional Technical 
Information
    This new section is analogous, in organizational structure, to 
Sec.  52.80, ``Contents of application; additional technical 
information'' in subpart C of part 52.
    Paragraph (a) requires that the application include inspections, 
tests, and analyses that the licensee who will be placing the 
manufactured reactor on a site and operating the reactor shall perform 
and their associated acceptance criteria. The purpose of these ITAAC 
are to ensure that: (1) The reactor has been manufactured in 
conformance with applicable requirements; and (2) the manufactured 
reactor, as emplaced at the site and integrated into any site-specific 
portions of the nuclear power plant, will operate in conformance with 
the design characteristics in the manufacturing license, the license 
authorizing operation of the manufactured reactor, and applicable 
requirements. Paragraph (a)(3), which is analogous to Sec.  
52.80(a)(3), provides that if the manufacturing license references a 
standard design certification, the manufacturing license application 
may include a notification that one or more ITAAC in the referenced 
design certification rule has been met. In such a situation, the 
Federal Register notice of docketing a hearing required by Sec.  52.163 
must specifically indicate that the application includes such a 
notification.
    Paragraph (b)(1) requires that the application include an 
environmental report meeting the requirements in 10 CFR 51.54, which 
specifies the environmental information that must be submitted by a 
manufacturing license applicant to support the NRC's NEPA review. The 
Commission notes that environmental report need not include a 
discussion of assessment of the benefits and impacts of constructing 
and operating the manufactured reactor or an evaluation of alternative 
energy sources, under Sec.  52.163 and Sec.  51.54.
    Under Sec.  51.54, the environmental report for a manufacturing 
license must address the costs and benefits of SAMDAs that could be 
incorporated into the design, and the bases for not including SAMDAs 
into the design. The SAMDA information that must be included is 
essentially the same information that must be provided to support an 
application for a standard design certification. However, if the 
application references a standard design certification, Sec.  51.54 
provides that the manufacturing license's environmental report need not 
include the SAMDA evaluation. In such a case, the SAMDA determination 
in the EA for the referenced design certification would have finality 
in the manufacturing license proceeding, in accordance with Sec.  
52.63.
Section 52.159 Standards for Review of Applications
    This new section is analogous to the ``standards for review of 
applications'' sections in subparts A through C of part 52 (e.g., 
Sec. Sec.  52.18, 52.48, 52.81). Section 52.159 identifies the 
regulations that the NRC will use in reviewing an application for a 
manufacturing license. The NRC recognizes that reactors to be 
manufactured under a manufacturing license may incorporate design 
features which are inconsistent with current requirements in 10 CFR 
Chapter I, and may require exemptions from current requirements. Such 
exemptions would be granted as part of the NRC's issuance of the 
manufacturing license, together with alternative requirements 
(analogous to the ``applicable regulations'' provisions in the current 
design certifications rules, 10 CFR part 52, appendices A-D, Section 
V).
Section 52.161 Reserved
    This section is reserved to accommodate any new requirements on the 
application process for manufacturing license which the NRC may adopt 
in the future.
Section 52.163 Administrative Review of Applications; Hearings
    This new section is analogous to the ``administrative review of 
applications'' sections in subparts A through C of part 52 (e.g., 
Sec. Sec.  52.21, 52.51, 52.85). Section 52.163 specifies that the 
procedural requirements in 10 CFR part 2 apply to the NRC's processing 
of an application for a manufacturing license, including docketing of 
the initial application.
    Section 52.163 reiterates the Sec.  2.105 requirement that the NRC 
publish in the Federal Register a notice of proposed action on the 
application. Apart from the required Federal Register notice, the 
Commission also expects to publish on the NRC's Web site notice of 
docketing of the application and the opportunity to intervene in the 
proceeding, consistent with the Commission's discussion in the 2004 
final part 2 rulemaking (January 14, 2004; 69 FR 2182, 2198-99). The 
section makes clear, consistent with Sec.  51.54, that the 
environmental report submitted by the manufacturing license applicant 
need not contain an assessment of the benefits of constructing and/or 
operating the manufactured reactor or an evaluation of alternative 
energy sources.
    Finally, this section indicates that the hearing on the 
manufacturing license application will be governed by the procedures in 
part 2, subparts C, G, L, and N. The Commission notes that although 
subpart G is listed in this paragraph, it is unlikely that there would 
be contentions meeting the criteria in Sec.  2.310 (and reiterated in 
Sec.  2.700) for conduct of the hearing under subpart G. This is 
because the primary focus of the manufacturing license proceeding is on 
the adequacy of the design to be manufactured, and the nature of issues 
which are most likely to be raised on the design would not ordinarily 
involve issues of material fact relating to either: (1) The occurrence 
of a past activity, where the credibility of an eyewitness may 
reasonably be expected to be at issue; or (2) issues of motive or 
intent of the party or eyewitness which are material to the resolution 
of the contested matter.
Section 52.165 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS)
    This new section is analogous to the ``Referral to the Advisory 
Committee on Reactor Safeguards'' sections in subparts A through C of 
part 52 (e.g., Sec. Sec.  52.21, 52.53, 52.87). It provides that the 
ACRS will have the same role with respect to manufacturing licenses 
that it has for other nuclear power plant licenses, in that it will 
report on those portions of the application which concern safety.
Section 52.167 Issuance of Manufacturing License
    This new section is analogous to the ``issuance'' sections in 
subparts A through C of part 52 (e.g., Sec. Sec.  52.24, 52.54, 52.97). 
Paragraph (a) sets forth the timing of issuance of a manufacturing 
license and the findings that the Commission must make in

[[Page 49457]]

order to issue the manufacturing license. The findings that must be 
made are similar to those necessary to issue a construction permit, 
inasmuch as construction is analogous to manufacturing. The Commission 
notes that it reserves the right to withhold issuance of the 
manufacturing license, even if all the rules and regulations of the 
Commission have been satisfied, based on public health and safety or 
common defense and security information or considerations not 
adequately addressed in the Commission's rules and regulations.
    Paragraph (b) identifies the specific limitations that the 
Commission will include in each manufacturing license. They include 
technical specifications for the operation of each manufactured 
reactor, site parameters, design characteristics, and interface 
requirements, which are to be used by the applicant for and holder of 
the license referencing the use of the manufactured reactor(s). 
Ordinarily, the limitations to be included in the manufacturing license 
would be derived from the manufacturing license application, but the 
NRC may modify the proposed limitations based upon the NRC's review.
    Paragraph (c) restricts the holder of the manufacturing license 
from transporting or allowing to be removed from the place of 
manufacture the manufactured reactor except to the site of a licensee 
who holds either a construction permit or combined license referencing 
the use of that manufactured reactor.
Section 52.169 Reserved
    This section is reserved to accommodate any new requirements on 
either the issuance of, or activities authorized under a manufacturing 
license which the Commission may adopt in the future. Any new 
requirements adopted after issuance of a manufacturing license, which 
are made applicable to that manufacturing license, would have to 
satisfy the finality restrictions in Sec.  52.171.
Section 52.171 Finality of Manufacturing Licenses; Information Requests
    This new section is analogous to the variously entitled sections 
addressing finality and special backfitting protections which are in 
subparts A through C of part 52 (e.g., Sec. Sec.  52.39, 52.63, 
52.98),\15\ but is more generally modeled on the finality provision for 
standard design certifications. In general, paragraph (a) addresses 
backfitting and finality restrictions on the NRC, paragraph (b) 
addresses finality and standardization restrictions applicable to the 
licensee (i.e., the holder the manufacturing license), and paragraph 
(c) establishes restrictions on certain NRC information collections 
with respect to the manufacturing license.
---------------------------------------------------------------------------

    \15\ The finality provision in Sec.  52.83 performs a different 
function than the finality sections cited above, in that it points 
back to, and thereby re-emphasizes, the primary finality provisions 
for each license or regulatory approval mechanism in part 52, e.g., 
the finality provision in Sec.  52.39 for early site permits.
---------------------------------------------------------------------------

    Paragraph (a)(1) states that the Commission may not modify, 
rescind, or impose new requirements on the design of a nuclear power 
reactor being manufactured, or new requirements for the manufacture of 
the nuclear power reactor, unless the Commission determines that a 
modification is necessary to either bring the design or the manufacture 
of the reactor into compliance with the Commission's requirements 
applicable and in effect at the time the manufacturing license was 
issued, or to provide reasonable assurance of adequate protection to 
public health and safety or common defense and security. This 
restriction on the Commission applies, inter alia, in construction 
permit, operating license, and combined license proceedings which 
reference the use of the manufactured reactor. It also applies in any 
enforcement proceeding initiated by the NRC, or in a rulemaking which 
proposes to apply new or changed requirements to reactors which have 
already been manufactured, as well as any reactors yet to be 
manufactured under the manufacturing license. However, the restrictions 
in paragraph (a)(1) do not apply to NRC information requests directed 
at either the manufacturing license holder, or to any holder of a 
license referencing the use of a manufactured reactor; such information 
requests are governed by paragraph (c) of this section.
    Paragraph (a)(2) provides that any modification to the design of a 
manufactured nuclear power reactor which is imposed by the Commission 
under paragraph (a)(1) of this section will be applied to all reactors 
manufactured under the license, including those that have already been 
manufactured, transported, sited, and are in operation. The only 
exception would be for those reactors to which the Commission-ordered 
modification had been rendered technically irrelevant by action taken 
under paragraph (b) of this section, i.e., either the holder of the 
manufacturing license has requested a change to the design approved in 
the manufacturing license (which ordinarily would apply only to 
reactors manufactured after Commission approval of the change), or the 
holder of a license referencing the use of the manufactured reactor has 
obtained Commission approval for a change to the design of the specific 
manufactured reactor(s) utilized by that licensee.
    Paragraph (a)(3) delineates the nature of finality associated with 
the referencing of a manufactured reactor in subsequent NRC licensing 
proceedings. This paragraph provides that finality is accorded to those 
matters resolved in the proceeding on the issuance or renewal of the 
manufactured reactor. These matters resolved include the adequacy of 
the design of the manufactured reactor and the acceptability and 
completeness of the ITAAC required by Sec.  52.158(a)(1) to be 
performed by the licensee operating the reactor. The matters resolved 
also include the SAMDA evaluation prepared by the Commission in 
compliance with its obligations under NEPA. This finality extends to 
both the Commission's determinations with respect to specific SAMDA 
features included in the design of the manufactured reactor, as well as 
the Commission's determinations regarding the lack of need for any 
other SAMDA features. Finality is accorded in the following situations: 
(1) Issuance of a construction permit, operating license, combined 
license; (2) any hearing under Sec.  52.103; and (3) enforcement 
hearings other than those proceedings initiated by the Commission under 
paragraph (a)(1).
    Paragraph (b)(1) requires the holder of a manufacturing license to 
seek a prior NRC review and approval for any change to the design of 
the nuclear power plant authorized to be manufactured. The holder of 
the manufacturing license may not make a change to the approved design 
for manufacture through the provisions of Sec.  50.59. A request for a 
change to the approved design must be in the form of a license 
amendment application, and the application will be processed in 
accordance with Sec. Sec.  50.90 through 50.92. The Commission notes, 
however, that the procedures for no significant hazards consideration 
(NSHC) are not applicable to manufacturing licenses, inasmuch as 
Section 189.a.(2) of the AEA, which is the statutory authority for 
these procedures, does not apply to manufacturing licenses.
    Paragraph (b)(2) requires a holder of a license referencing the use 
of a manufactured reactor, who wishes to depart from the design 
characteristics, site parameters, terms and conditions,

[[Page 49458]]

or approved design of the manufactured reactor, to seek a departure 
from the NRC. The manner in which a departure is granted depends upon 
the timing of the request. If a departure is requested as part of the 
initial combined license application, the departure would be treated as 
part of the application and issued as part of the combined license. By 
contrast, if the same departure were sought after the combined license 
had been issued, then the licensee must apply for the departure in the 
form of a license amendment. The criteria for granting the departure is 
the exemption criterion in Sec.  52.7; however, the departure itself is 
not considered an exemption (unless, of course, the departure also 
involves a non-compliance with an underlying Commission regulatory 
requirement in 10 CFR Chapter I). Thus, the Commission will not approve 
a departure unless the Commission finds, in addition to the routine 
exemption criteria in Sec.  52.7, that special circumstances outweigh 
any decrease in safety that may result from the reduction in 
standardization caused by the departure. As explained earlier, these 
limitations are intended to maintain the standardization of 
manufactured reactors in operation to the extent practicable. The 
licensee may not depart from the design characteristics, site 
parameters, terms and conditions, or approved design of the 
manufactured reactor through the provisions of Sec.  50.59.
    Paragraph (c), which is analogous to Sec. Sec.  52.39(d), 52.98(g), 
and 52.145(c), provides that NRC information requests must be evaluated 
before issuance to ensure that the burden to be imposed by the 
information request is justified in view of the potential safety 
significance of the issue to be addressed, except when the information 
requests seeks to verify compliance with the current licensing basis of 
either the manufacturing license or the manufactured reactor. This 
paragraph applies to information requests directed at either the holder 
of the manufacturing license or the holder of a license referencing the 
use of a manufactured reactor. Information requests may be in the form 
of a new rule requiring submission of information (i.e., a new 
information collection and reporting requirement), or in the form of a 
NRC staff request for information. Information requests by the staff 
must be in accordance with 10 CFR 50.54(f) and must be approved by the 
EDO or his or her designee before the request may be issued.
Section 52.173 Duration of Manufacturing License
    This new section is analogous to the variously-entitled sections 
addressing duration (term) of each regulatory process in subparts A 
through C of part 52 (e.g., Sec. Sec.  52.33, 52.61, 52.104). Under 
Sec.  52.173, a manufacturing license may be issued for not less than 5 
nor more than 15 years. Manufacturing of a new reactor may not commence 
less than 3 years before the expiration of the manufacturing license, 
even though a timely application for renewal has been filed in 
accordance with Sec.  52.177. However, if a timely application for 
renewal of the manufacturing license has been docketed, manufacturing 
of uncompleted reactors whose manufacture commenced 3 years or more 
before the expiration date, may continue past the date of expiration of 
the license until the NRC acts upon the renewal application, consistent 
with the ``Timely Renewal'' doctrine of the Administrative Procedures 
Act. The NRC believes that timely renewal protection should only be 
provided to those applications which are of sufficient quality to be 
docketed. This is consistent with the requirement in Sec.  2.109(b) 
requiring filing of a ``sufficient'' application for renewal of 
operating licenses as a prerequisite for the applicability of the 
timely renewal protection.
Section 52.175 Transfer of Manufacturing License
    This new section is analogous to the variously entitled transfer 
sections in subparts A and C of part 52 (e.g., Sec. Sec.  52.28, 
52.105).\16\ Section 52.175 provides that a manufacturing license may 
be transferred in accordance with Sec.  50.80, which constitutes the 
Commission's common procedures and criteria governing transfers of 
nuclear power plant licenses. The matters to be addressed in a transfer 
are limited to the matters identified in Sec.  50.80(b), and the 
transfer would not be an opportunity for the Commission to reconsider 
safety and environmental matters previously resolved, or to address new 
safety matters other than the narrow scope of matters identified in 
Sec.  50.80(b).
---------------------------------------------------------------------------

    \16\ A standard design certification is a rule, rather than a 
license. Accordingly, there is no ``holder'' of a standard design 
certification rule and no need for a provision addressing 
``transfer'' of a standard design certification rule.
---------------------------------------------------------------------------

Section 52.177 Application for Renewal
    This new section is analogous to the ``application for renewal'' 
sections in subparts A through C of part 52 (e.g., Sec. Sec.  52.29, 
52.57, 52.107). Section 52.177 sets forth the content of an application 
for renewal, specifies the administrative requirements governing the 
application, addresses the effectiveness of a manufacturing license 
during the period of NRC's consideration of the renewal application, 
summarizes how an interested person may request a hearing on the 
renewal, and addresses the referral of the renewal application to the 
ACRS and the Commission's expectations with respect to the ACRS report 
on the application.
Section 52.179 Criteria for Renewal
    This new section is analogous to the ``criteria for renewal'' 
sections in subparts A and B of part 52 (e.g., Sec. Sec.  52.31, 
52.59).\17\ Section 52.179 provides that the Commission may grant 
renewal of a manufacturing license if the Commission determines that 
the license complies with the relevant provisions of the AEA, the 
Commission's regulations applicable and in effect at the time the 
manufacturing license was originally issued, and any new requirements 
which the Commission imposes which: (1) Are necessary for reasonable 
assurance of adequate protection to public health and safety or common 
defense and security; (2) are necessary for compliance with 
Commission's regulations and orders applicable and in effect at the 
time the manufacturing license was originally issued; or (3) represent 
a substantial increase in overall protection of the public health and 
safety or common defense and security and the direct and indirect costs 
of implementation are justified in light of the increased protection. 
These ``backfitting'' restrictions are similar to--if somewhat narrower 
than--the backfitting restrictions applicable to renewal of standard 
design certification rules under subpart B of this part.
---------------------------------------------------------------------------

    \17\ Subpart C does not contain a ``criteria for renewal'' 
provision, inasmuch as the renewal would be governed by 10 CFR part 
54, see Sec.  52.107. Part 54 contains a provision, Sec.  54.29, 
setting forth the standards for issuance of renewed licenses.
---------------------------------------------------------------------------

    Reasonable assurance of adequate protection to public health and 
safety and common defense and security is provided under this 
regulatory approach, inasmuch as paragraph (b) allows the Commission to 
impose new requirements which are necessary for common defense and 
security, or are necessary for compliance with the Commission's 
regulations and orders applicable and in effect at the time the 
manufacturing license was originally issued.

[[Page 49459]]

Section 52.181 Duration of Renewal
    This new section is analogous to the ``duration of renewal '' 
sections in subparts A and B of part 52 (e.g., Sec. Sec.  52.33, 
52.61).\18\ Section 52.181 specifies the term of a renewed 
manufacturing license as not less than 5 nor more than 15 years from 
the date of expiration of the prior manufacturing license. Thus, a 
holder of a manufacturing license with an original term of 15 years, 
who is granted a 15-year renewal of the manufacturing license 4 years 
before expiration of the license, will obtain a renewed manufacturing 
license of 19 years, representing a 15-year term of the renewed license 
plus the 4 years remaining on its original license.
---------------------------------------------------------------------------

    \18\ Subpart C does not contain a ``duration of renewal'' 
provision, inasmuch as the renewal would be governed in all respects 
by 10 CFR part 54, see Sec.  52.107. Part 54 contains a provision, 
Sec.  54.31, governing the duration of renewed licenses.
---------------------------------------------------------------------------

Subpart G--Reserved

    This subpart is reserved for future use by the Commission.

Subpart H--Enforcement

    This subpart contains two provisions, Sec.  52.301 and Sec.  
52.303, which are comparable to former Sec.  52.111 and Sec.  52.113, 
and are analogous to provisions contained in other parts of 10 CFR 
Chapter I imposing requirements on regulated entities.
    Section 52.301 reiterates, and provides notice to licensees and 
applicants under part 52 of the Commission's authority to obtain 
injunctions or other court orders for the violations enumerated in this 
paragraph.
    Section 52.303 provides notice to all persons and entities subject 
to part 52 that they are subject to criminal sanctions for willful 
violations, attempted violations, or conspiracy to violate certain 
regulations under part 52. The regulations for which criminal penalties 
apply are limited to those which establish either a regulatory 
obligation or prohibition. Most of the regulations in part 52 are 
procedural or administrative in nature, and therefore were listed in 
Sec.  52.113 as not being subject to criminal sanctions. The 
regulations in part 52 which are subject to criminal sanctions are 
Sec. Sec.  52.4 (Deliberate misconduct), 52.5 (Employee protection), 
52.6 (Completeness of information), 52.25 (Extent of activities 
permitted), 52.35 (Use of site for other purpose), 52.91 (Authorization 
to conduct site activities), and 52.110 (Termination of license).

Appendix A--U.S. Advanced Boiling Water Reactor

    Refer to the section-by-section discussion in the final rule dated 
May 12, 1997 (62 FR 25800).

Appendix B--The System 80+ Design

    Refer to the section-by-section discussion in the final rule dated 
May 21, 1997 (62 FR 27840).

Appendix C--The AP600 Design

    Refer to the section-by-section discussion in the final rule dated 
December 23, 1999 (64 FR 72002).

Appendix D--The AP1000 Design

    Refer to the section-by-section discussion in the final rule dated 
January 27, 2006 (71 FR 4464).

Appendix N--Combined Licenses for Nuclear Power Reactors of Identical 
Design

    Appendix N of part 52 contains the Commission's procedures which 
may be used by one or more applicants for combined licenses under part 
52, where the applications seek to construct and operate nuclear power 
reactors of identical design to be located at multiple sites. The 
comparable procedures governing applications for construction permits 
and operating licenses using identical nuclear power reactor designs 
remain in appendix N of 10 CFR part 50. Hearings for applications filed 
under appendix N in part 52, as well as part 50, are governed by 
subpart D of part 2. Thus, appendix N and subpart D of part 2 are 
integral to each other.
    The regulations in appendix N of part 52 apply in two situations: 
(1) Where the same applicant seeks combined licenses at different sites 
utilizing the identical reactor design; and (2) where two or more 
different applicants each seek combined licenses at different sites 
utilizing the identical reactor design. In either situation, there is 
an identical reactor design. The Commission has deliberately used the 
term, ``nuclear power reactor,'' in appendix N and subpart D of part 
2--as distinguished from the term, ``nuclear power plant''--to make 
clear that the site-specific elements, such as the service water intake 
structure or the ultimate heat sink, need not be identical in order for 
appendix N and subpart D to apply.
    The Commission has conformed appendix N and subpart D of part 2 to 
use the term, ``identical'' nuclear power reactor design, and removed 
references to ``duplicate'' and ``essentially identical.'' For purposes 
of appendix N and subpart D of part 2, designs for reactors are 
``identical,'' even if individual licensees request plant-specific 
departures or exemptions from a referenced standard design 
certification (or application). However, those plant-specific 
departures or exemptions are not part of the ``common design.'' 
Therefore, the NRC's review of those departures and exemptions, as well 
as NRC hearings on those departures and exemptions, would be conducted 
separately as part of the safety review of each individual application, 
and would not be part of the hearing on the common design which would 
be conducted under subpart D of part 2.
Section 1
    This is a new section specifying that its provisions apply to 
applicants for combined licenses under subpart C of part 52. Appendix N 
of part 50 would apply to applicants for construction permits and 
operating licenses who use identical reactor designs.
Section 2
    This section, which is analogous to and derived from former Sec.  2 
of appendix N, specifies that each application submitted under this 
appendix must be submitted in accordance with the delineated Commission 
filing requirements. In addition, to ensure that the NRC is clearly 
informed that the applicants wish to have their application processed 
under appendix N and subpart D of part 2, this section requires: (1) 
That each application state the applicant's intent that the application 
be processed by the NRC under appendix N; and (2) that all of the 
applications to be treated together under this appendix be listed in 
each application. All of the applications must be filed simultaneously, 
which will facilitate NRC's administrative handling and technical 
review of the applications, as well as efficient conduct of the hearing 
process.
Section 3
    This section, which is analogous to and derived from former Sec.  3 
of appendix N, specifies that combined license applications submitted 
under this appendix must include all of the information required to be 
submitted in a combined license application in Sec. Sec.  52.77, 52.79, 
and 50.80(a) and (b), but makes clear that each of the applications 
must identify the common design. The common design may be (but is not 
limited to) a standard design certification under subpart B of part 52, 
a standard design approval, a ``common custom design,'' or a 
manufactured reactor.

[[Page 49460]]

    The FSAR for each application must either incorporate by reference 
or include the FSAR for the common design, including, as applicable, 
the FSAR for the referenced design certification or manufactured 
reactor. ``Include,'' means that the FSAR may not simply reference the 
common FSAR; the information from the referenced FSAR must be included 
within each application's FSAR.
Section 4
    This is a new section specifying that each application must submit 
an environmental report which complies with the applicable provisions 
of part 51 with respect to the content of environmental reports. As an 
alternative, this section provides that one or more of the applicants' 
environmental reports may incorporate by reference a single 
environmental report describing the environmental impacts of the common 
design at each of the sites.
Section 5
    This is a new section specifying that, upon a determination that 
each application is acceptable for docketing, each application will be 
docketed and a notice of docketing will be published in the Federal 
Register in accordance with 10 CFR 2.104. The notice of docketing must 
state that the application will be processed under the provisions of 
appendix N. Separate notices of docketing are contemplated, so that a 
problem with acceptance review of one application will not prevent the 
docketing and initiation of the NRC's technical review of the other 
applications determined to be sufficient and acceptable for docketing. 
This could occur, for example, if information, submitted by an 
applicant which is unrelated to the common design, is determined by the 
NRC to be insufficient. However, if the applications are determined to 
be acceptable for docketing, Sec.  5 provides the Commission with the 
discretion to publish a single notice of docketing for those 
applications.
Section 6
    This is a new section which provides that the NRC will prepare a 
separate draft and final EIS for each of the applications. Scoping may 
be conducted simultaneously but need not be conducted jointly (e.g., 
scoping for an application at site 1 need not be conducted as part of 
the same process as the scoping for an application for site 2), at 
least with respect to site-specific environmental issues. However, for 
environmental issues related to the common design, the NRC has the 
discretion to conduct joint scoping. The NRC staff is not, however, 
required to prepare a joint environmental impact statement for the 
common design.
    This section also addresses the content of an EIS when the 
applications reference either a standard design certification or the 
use of a manufactured reactor of common design. In either case, the NRC 
has already prepared and finalized an EA which addresses SAMDAs. This 
SAMDA analysis is accorded finality under the provisions of Sec. Sec.  
52.63 and 52.171, respectively. Therefore, the EIS for each of the 
applications must reference the relevant environmental assessment 
containing the SAMDA analysis.
Section 7
    This section, which is analogous to and derived from former Sec.  1 
of appendix N, provides direction to the ACRS with respect to their 
report on each of the combined license applications. The ACRS must 
issue a separate report on the safety of the common design, except in 
those instances where the applications are referencing either a 
standard design certification or manufactured reactor (of common 
design). In addition, the ACRS must issue a separate report for each 
application. This report must be limited to those matters which are not 
relevant to the common design. This will facilitate the NRC's licensing 
process by eliminating overlap and ensuring that the ACRS reports are 
carefully focused on the relevant safety issues.
Section 8
    This is a new section, which provides that the Commission shall 
designate a presiding officer to conduct the proceeding with respect to 
the health and safety, common defense and security, and environmental 
matters (i.e., SAMDAs) relating to the common design. The presiding 
officer will conduct the hearing in accordance with subpart D of part 
2. The presiding officer is required to issue a separate partial 
initial decision on matters relevant to the common design, consistent 
with 10 CFR 2.405 in subpart D of part 2. Appeals of the partial 
initial decision are governed by 10 CFR 2.341, as provided by 10 CFR 
2.405. The NRC also notes that issues on the contested design may not 
be relitigated in a different 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. See 10 
CFR 2.406.

VII. 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 Web site (Web). The NRC's interactive rulemaking Web 
site is located at http://ruleforum.llnl.gov. These documents may be 
viewed and downloaded electronically via this Web site.
    NRC's Public Electronic Reading Room (EPDR). The NRC's electronic 
public reading room is located at http://www.nrc.gov/reading-rm.html.
    The NRC staff contact. Nanette V. Gilles, Mail Stop O-4D9A, 
Washington, DC 20555-0001, 301-415-1180.

----------------------------------------------------------------------------------------------------------------
                    Document                            PDR             Web            EPDR          NRC staff
----------------------------------------------------------------------------------------------------------------
Part 52 Rule, Cross-Reference Tables............  ..............               X     ML062550246               X
Comments received...............................               X               X               X
Comment Summary Report..........................  ..............  ..............     ML063450216
Regulatory Analysis.............................               X               X     ML071490350               X
Regulatory History Index for the proposed July    ..............  ..............     ML032810026
 2003 rule......................................
Regulatory History Index for the March 13, 2006,  ..............  ..............     ML062080575
 proposed rule..................................
----------------------------------------------------------------------------------------------------------------

VIII. Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' which became effective on September 3, 1997 
(62 FR 46517), NRC program elements (including regulations) are placed 
into compatibility categories A, B, C, D, NRC, or adequacy category, 
Health and Safety (H&S). Category A includes program elements that are 
basic radiation protection standards or related

[[Page 49461]]

definitions, signs, labels, or terms necessary for a common 
understanding of radiation protection principles and should be 
essentially identical to those of NRC. Category B includes program 
elements that have significant direct transboundary implications and 
should be essentially identical to those of the NRC. Compatibility 
Category C includes program elements that do not meet the criteria of 
Category A or B, but the essential objectives of which an Agreement 
State should adopt to avoid conflict, duplication, gaps, or other 
conditions that would jeopardize an orderly pattern in the regulation 
of agreement material on a nationwide basis. Compatibility Category D 
includes those program elements that do not meet any of the criteria of 
Category A, B, or C, and do not need to be adopted by Agreement States. 
Compatibility Category NRC includes program elements that address areas 
reserved to the Commission and cannot be relinquished to Agreement 
States pursuant to the Atomic Energy Act or provisions of Title 10 of 
the Code of Federal Regulations. An Agreement State may inform its 
licensees of certain of these NRC provisions through a mechanism that 
is appropriate under the State's administrative procedure laws as long 
as the State adopts these provisions solely for the purposes of 
notification, and does not exercise any regulatory authority pursuant 
to them. Category H&S include program elements that are not required 
for compatibility, but have a particular health and safety role in the 
regulation of agreement material and the State should adopt the 
essential objectives of the NRC program elements. In addition, a State 
should not adopt provisions that would preclude, or effectively 
preclude, a practice authorized by the Atomic Energy Act, and in the 
national interest. The proposed revisions are categorized as follows:

                                 List of Changes 10 CFR Part 52 Final Rulemaking
----------------------------------------------------------------------------------------------------------------
                                       Description new,        Compatibility            Comments regarding
             Sections                      changes              designation          compatibility designation
----------------------------------------------------------------------------------------------------------------
10 CFR Part 1.....................  Statement of           D....................  This provision is designated
                                     Organization and                              Category D because it does
                                     General Information.                          not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt similar
                                                                                   provisions to reflect their
                                                                                   organizational structure and
                                                                                   may wish to inform its
                                                                                   licensees of the provisions
                                                                                   of this part through a
                                                                                   mechanism that is appropriate
                                                                                   under the State's
                                                                                   administrative procedure
                                                                                   laws.
10 CFR Part 2--Rules of Practice
 for Domestic Licensing
 Proceedings and Issuance of
 Orders
    2.1...........................  Scope................  D, except portions of  These provisions are
                                                            these provisions are   designated Compatibility
                                                            NRC.                   Category D because they do
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt similar
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR Part
                                                                                   52 standard design approvals,
                                                                                   are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.
    2.4--Definitions..............  Contested proceeding.  D, except portions of  This definition is designated
                                                            the definition are     Category D because it does
                                                            NRC.                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt a similar
                                                                                   definition that is compatible
                                                                                   with the orderly pattern of
                                                                                   regulation established by the
                                                                                   Atomic Energy Act, as amended
                                                                                   (Act) and is consistent with
                                                                                   their regulatory authority.
                                                                                   Those portions of the
                                                                                   definition that address areas
                                                                                   reserved to the NRC, e.g., 10
                                                                                   CFR Part 52 activities, are
                                                                                   designated as a Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
                                    License..............  NRC..................  This definition is designated
                                                                                   Compatibility Category NRC
                                                                                   because it addresses areas
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
                                                                                   For purposes of
                                                                                   compatibility, States should
                                                                                   use the language of the 10
                                                                                   CFR 20.1003 definition,
                                                                                   except those portions of the
                                                                                   definition that reference
                                                                                   areas reserved to the NRC,
                                                                                   e.g., 10 CFR Parts 50, 60,
                                                                                   63, and 72, are designated as
                                                                                   a Compatibility Category NRC.
                                    Licensee.............  [D]..................  This definition also appears
                                                                                   in 10 CFR 20.1003. For
                                                                                   purposes of compatibility,
                                                                                   the language of the Part 20
                                                                                   definition should be used
                                                                                   where it is assigned to
                                                                                   Compatibility Category D.

[[Page 49462]]

 
    2.100 thru 2.390..............  All of the sections    D, except portions of  These provisions are
                                     covered by Subparts    these provisions are   designated Compatibility
                                     A, B, and C.           NRC.                   Category D because they do
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt similar
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR
                                                                                   Parts 50, 51, 52, 53, 54, 55,
                                                                                   60, 63, 72, 73, and 76, are
                                                                                   designated as a Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Act, 10 CFR 8.4, 10 CFR Part
                                                                                   150, and other Federal laws,
                                                                                   regulations, or provisions.
    2.400 thru 2.629..............  All of the sections    NRC, for all of the    These provisions are
                                     covered by Subparts    sections.              designated Compatibility
                                     D, E, and F.                                  Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.800.........................  Scope and              D, except portions of  These provisions are
                                     applicability.         these provisions are   designated Compatibility
                                                            NRC.                   Category D because they do
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt similar
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR Part
                                                                                   52, are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.
    2.801.........................  Initiation of          D, except portions of  These provisions are
                                     rulemaking.            these provisions are   designated Compatibility
                                                            NRC.                   Category D because they do
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt similar
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR Part
                                                                                   52, are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.
    2.811.........................  Filing of standard     NRC..................  This provision is designated
                                     design certification                          Compatibility Category NRC
                                     application,                                  because it addresses an area
                                     required copies.                              reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.813.........................  Written                NRC..................  This provision is designated
                                     communications.                               Compatibility Category NRC
                                                                                   because it addresses an area
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.815.........................  Docketing and          NRC..................  This provision is designated
                                     acceptance review.                            Compatibility Category NRC
                                                                                   because it addresses an area
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.

[[Page 49463]]

 
    2.817.........................  Withdrawal of          NRC..................  This provision is designated a
                                     application.                                  Compatibility Category NRC
                                                                                   because it addresses an area
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.819.........................  Denial of application  NRC..................  This provision is designated
                                     for failure to                                Compatibility Category NRC
                                     supply information.                           because it addresses an area
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.1202........................  Authority and role of  NRC..................  This provision is designated
                                     NRC staff.                                    Compatibility Category NRC
                                                                                   because it addresses an area
                                                                                   reserved to the NRC. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
    2.1211--[Removed].............
10 CFR Part 10....................  Criteria and           NRC for all sections.  These provisions are
                                     procedures for                                designated Compatibility
                                     determining                                   Category NRC because they
                                     eligibility for                               address areas reserved to the
                                     access to restricted                          NRC. A State should not adopt
                                     data or national                              provisions that would confer
                                     security information                          regulatory authority to the
                                     or an employment                              State in an area of exclusive
                                     clearance.                                    NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 19--Notices,
 Instructions and Reports to
 Workers: Inspection and
 Investigations
    19.1..........................  Purpose..............  D....................  This provision is designated
                                                                                   Category D because it does
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt a similar
                                                                                   provision that is compatible
                                                                                   with the orderly pattern of
                                                                                   regulation established by the
                                                                                   Atomic Energy Act, as amended
                                                                                   (Act) and are consistent with
                                                                                   their regulatory authority.
    19.2..........................  Scope................  D, except portions of  This provision is designated
                                                            the provisions in      Compatibility Category D
                                                            (a)(1), (a)(2),        because it does not meet any
                                                            (a)(3), and (a)(4)     of the criteria of Category
                                                            are designated as      A, B, or C. A State may adopt
                                                            NRC.                   similar provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR
                                                                                   Parts 50, 51, 52, 53, 54, 60,
                                                                                   63, 72, and 76, are
                                                                                   designated as a Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Act, 10 CFR 8.4, 10 CFR Part
                                                                                   150, and other Federal laws,
                                                                                   regulations, or provisions.
    19.3--Definitions.............  License..............  D, except portions of  This definition is designated
                                                            the definition are     Category D because it does
                                                            NRC.                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt a similar
                                                                                   definition that is compatible
                                                                                   with the orderly pattern of
                                                                                   regulation established by the
                                                                                   Atomic Energy Act, as amended
                                                                                   (Act) and is consistent with
                                                                                   their regulatory authority.
                                                                                   Those portions of the
                                                                                   definition that address areas
                                                                                   reserved to the NRC, e.g., 10
                                                                                   CFR Parts 50, 51, 52, 53, 54,
                                                                                   55, 60, 63, 72, 73, and 76,
                                                                                   are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
                                                                                   This definition appears in 10
                                                                                   CFR 20.1003. For purposes of
                                                                                   compatibility, States should
                                                                                   use the language of the Part
                                                                                   20 definition, which is
                                                                                   assigned a Compatibility
                                                                                   Category D.

[[Page 49464]]

 
                                    Regulated activities.  D....................  This definition is designated
                                                                                   Category D because it does
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt a similar
                                                                                   definition that is compatible
                                                                                   with the orderly pattern of
                                                                                   regulation established by the
                                                                                   Atomic Energy Act, as amended
                                                                                   (Act) and is consistent with
                                                                                   their regulatory authority.
                                    Regulated entities...  D, except portions of  This definition is designated
                                                            the definition are     Category D because it does
                                                            NRC.                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt a similar
                                                                                   definition that is compatible
                                                                                   with the orderly pattern of
                                                                                   regulation established by the
                                                                                   Atomic Energy Act, as amended
                                                                                   (Act) and is consistent with
                                                                                   their regulatory authority.
                                                                                   Those portions of the
                                                                                   definition that address areas
                                                                                   reserved to the NRC are
                                                                                   designated Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
                                    Worker...............  C....................  This definition is designated
                                                                                   Compatibility Category C
                                                                                   because of its role in
                                                                                   effective communication, dose
                                                                                   monitoring, and commerce
                                                                                   (transboundary). A State
                                                                                   should adopt definitions that
                                                                                   are compatible with the
                                                                                   orderly pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. The
                                                                                   essential objectives of this
                                                                                   definition should be adopted.
    19.11.........................  Posting of Notices to  C, except portions of  This provision is designated
                                     workers.               paragraph (a), and     Compatibility Category C
                                                            all of paragraphs      because it is needed to
                                                            (b) and (e) are        provide a minimum level of
                                                            designated as NRC.     information to workers and to
                                                                                   assure that this information
                                                                                   is consistent from one
                                                                                   jurisdiction to another since
                                                                                   workers may work in multiple
                                                                                   jurisdictions. A State should
                                                                                   adopt provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. The
                                                                                   essential objectives of this
                                                                                   definition should be adopted.
                                                                                   Those portions of paragraph
                                                                                   (a) that reference 10 CFR
                                                                                   Part 52 activities, and
                                                                                   paragraphs (b) and (e)
                                                                                   address areas reserved to the
                                                                                   NRC, and are designated
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.
    19.14.........................  Presence of            C, except paragraph    This provision is designated
                                     representatives of     (a) is designated as   Compatibility Category C
                                     licensees and          NRC.                   because it is needed to
                                     workers during                                provide a minimum level of
                                     inspections.                                  consistency from one
                                                                                   jurisdiction to another since
                                                                                   workers may work in multiple
                                                                                   jurisdictions. A State should
                                                                                   adopt provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority.
                                                                                   Paragraph (a) addresses areas
                                                                                   reserved to the NRC, and is
                                                                                   designated Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Act, 10 CFR 8.4, 10 CFR Part
                                                                                   150, and other Federal laws,
                                                                                   regulations, or provisions.
    19.20.........................  Employee protection..  D, except portions of  This provision is designated
                                                            the provision are      Compatibility Category D
                                                            NRC.                   because it does not meet any
                                                                                   of the criteria of Category
                                                                                   A, B, or C. A State may adopt
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR
                                                                                   Parts 50, 52, 54, 60, 63, 72,
                                                                                   and 76, are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.

[[Page 49465]]

 
    19.31.........................  Application for        D....................  This provision is designated
                                     exemptions.                                   Category D because it does
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt provisions
                                                                                   that are compatible with the
                                                                                   orderly pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority.
    19.32.........................  Discrimination         D....................  This provision is designated
                                     prohibited.                                   Category D because it does
                                                                                   not meet any of the criteria
                                                                                   of Category A, B, or C. A
                                                                                   State may adopt provisions
                                                                                   that are compatible with the
                                                                                   orderly pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority.
10 CFR Part 20--Standards of
 Protection
    20.1002.......................  Scope................  D, except portions of  This provision is designated
                                                            the provision are      Compatibility Category D
                                                            designated as NRC.     because it does not meet any
                                                                                   of the criteria of Category
                                                                                   A, B, or C. A State may adopt
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. Those
                                                                                   portions of the provision
                                                                                   that address areas reserved
                                                                                   to the NRC, e.g., 10 CFR
                                                                                   Parts 50, 52, 54, 60, 63, 72,
                                                                                   and 76, are designated as a
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.
    20.1401.......................  General provisions     C, except portions of  This provision is designated
                                     and scope.             the provision are      Compatibility Category C
                                                            designated as NRC.     because it is needed to
                                                                                   provide a minimum level of
                                                                                   consistency regarding
                                                                                   decommissioning activities. A
                                                                                   State should adopt provisions
                                                                                   that are compatible with the
                                                                                   orderly pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. The
                                                                                   essential objectives of these
                                                                                   provisions should be adopted
                                                                                   by States. Those portions of
                                                                                   the provision that address
                                                                                   areas reserved to the NRC,
                                                                                   e.g., 10 CFR Parts 50, 52,
                                                                                   54, 60, 63, and 72, are
                                                                                   designated as a Compatibility
                                                                                   Category NRC. A State should
                                                                                   not adopt provisions that
                                                                                   would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Act, 10 CFR 8.4, 10 CFR Part
                                                                                   150, and other Federal laws,
                                                                                   regulations, or provisions.
    20.1406.......................  Minimization of        C, except portions of  This provision is designated
                                     contamination.         paragraph (a) and      Compatibility Category C
                                                            all of paragraph (b)   because it is needed to
                                                            are designated as      provide a minimum level of
                                                            NRC.                   safety regarding
                                                                                   decommissioning activities. A
                                                                                   State should adopt provisions
                                                                                   that are compatible with the
                                                                                   orderly pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. The
                                                                                   essential objectives of these
                                                                                   provisions should be adopted
                                                                                   by States. Those portions of
                                                                                   paragraph (a) that reference
                                                                                   10 CFR Part 52 activities,
                                                                                   and paragraphs (b) address
                                                                                   areas reserved to the NRC,
                                                                                   and are designated
                                                                                   Compatibility Category NRC. A
                                                                                   State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Act, 10 CFR 8.4, 10 CFR
                                                                                   Part 150, and other Federal
                                                                                   laws, regulations, or
                                                                                   provisions.

[[Page 49466]]

 
    20.2203.......................  Reports of exposures,  C paragraphs (a) and   Paragraphs (a) and (b) are
                                     etc., exceeding the    (b).                   designated Compatibility
                                     limits.               NRC paragraphs (c)      Category C, because they are
                                                            and (d).               needed to provide a common
                                                                                   understanding in collecting
                                                                                   and reporting information on
                                                                                   the regulation of agreement
                                                                                   material on a nationwide
                                                                                   basis. A State should adopt
                                                                                   provisions that are
                                                                                   compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority. The
                                                                                   essential objectives of these
                                                                                   provisions should be adopted
                                                                                   by States. Paragraphs (c) and
                                                                                   (d) address NRC exclusive
                                                                                   areas of authority are
                                                                                   designated Compatibility
                                                                                   Category NRC, and should not
                                                                                   be adopted by States. A State
                                                                                   should not adopt provisions
                                                                                   that would confer regulatory
                                                                                   authority to the State in an
                                                                                   area of exclusive NRC
                                                                                   jurisdiction pursuant to the
                                                                                   Act, 10 CFR 8.4, 10 CFR Part
                                                                                   150, and other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 21....................  Reporting of Defects   Not applicable for     The provisions in Part 21 are
                                     and Noncompliance.     all sections.          derived from statutory
                                                                                   authority in the Energy
                                                                                   Reorganization Act, not the
                                                                                   Atomic Energy Act, which does
                                                                                   not apply to Agreement
                                                                                   States. Therefore, this part
                                                                                   cannot be addressed under
                                                                                   either compatibility or
                                                                                   adequacy. While it may be
                                                                                   argued that there are health
                                                                                   and safety reasons to require
                                                                                   States to adopt the
                                                                                   provisions of Part 21, States
                                                                                   may not have the statutory
                                                                                   authority to do so. States
                                                                                   that have the statutory
                                                                                   authority to implement
                                                                                   provisions similar to those
                                                                                   in Part 21 may adopt similar
                                                                                   provisions consistent with
                                                                                   their regulatory authority
                                                                                   but should not address areas
                                                                                   of exclusive NRC
                                                                                   jurisdiction.
10 CFR Part 25....................  Access Authorization.  NRC for all sections.  These provisions are
                                                                                   designated a Compatibility
                                                                                   Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 26....................  Fitness for Duty       NRC for all sections.  These provisions are
                                     Programs.                                     designated a Compatibility
                                                                                   Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 50....................  Domestic Licensing of  NRC for all sections.  These provisions are
                                     Production and                                designated a Compatibility
                                     Utilization                                   Category NRC because they
                                     Facilities.                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 51....................  Environmental          NRC for all sections.  These provisions are
                                     Protection                                    designated a Compatibility
                                     Regulation for                                Category NRC because they
                                     Domestic Licensing                            address areas reserved to the
                                     and Related                                   NRC. A State should not adopt
                                     Regulatory Functions.                         provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 52....................  Licenses,              NRC for all sections.  These provisions are
                                     Certifications, and                           designated a Compatibility
                                     Approvals For                                 Category NRC because they
                                     Nuclear Power Plants.                         address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 54....................  Requirements for       NRC for all sections.  These provisions are
                                     Renewal of Operating                          designated a Compatibility
                                     License for Nuclear                           Category NRC because they
                                     Power Plants.                                 address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.

[[Page 49467]]

 
10 CFR Part 55....................  Operators License....  NRC for all sections.  These provisions are
                                                                                   designated a Compatibility
                                                                                   Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 72....................  Licensing              NRC for all sections.  These provisions are
                                     Requirements for                              designated a Compatibility
                                     Independent Storage                           Category NRC because they
                                     of Spent Nuclear                              address areas reserved to the
                                     Fuel and High-level                           NRC. A State should not adopt
                                     Radioactive Waste                             provisions that would confer
                                     and Greater than                              regulatory authority to the
                                     Class C.                                      State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 73....................  Physical Protection    NRC for all sections.  These provisions are
                                     of Plants and                                 designated a Compatibility
                                     Materials.                                    Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 75....................  Safeguards on Nuclear  NRC for all sections.  These provisions are
                                     Material--Implementa                          designated a Compatibility
                                     tion of US/IAEA                               Category NRC because they
                                     Agreement.                                    address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 95....................  Facility Security      NRC for all sections.  These provisions are
                                     Clearance and                                 designated a Compatibility
                                     Safeguarding of                               Category NRC because they
                                     National Security                             address areas reserved to the
                                     Information and                               NRC. A State should not adopt
                                     Restricted Data.                              provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 140...................  Financial Protection   NRC for all sections.  These provisions are
                                     Requirements and                              designated a Compatibility
                                     Indemnity Agreements.                         Category NRC because they
                                                                                   address areas reserved to the
                                                                                   NRC. A State should not adopt
                                                                                   provisions that would confer
                                                                                   regulatory authority to the
                                                                                   State in an area of exclusive
                                                                                   NRC jurisdiction pursuant to
                                                                                   the Atomic Energy Act, 10 CFR
                                                                                   8.4, 10 CFR Part 150, and
                                                                                   other Federal laws,
                                                                                   regulations, or provisions.
10 CFR Part 170...................  Fees for Facilities,   D....................  These provisions are
                                     Materials, Import                             designated a Category D
                                     and Export Licenses,                          because they do not meet any
                                     and Other Regulatory                          of the criteria of Category
                                     Services under the                            A, B, or C. A State may adopt
                                     Atomic Energy Act of                          similar provisions that are
                                     1954, as Amended.                             compatible with the orderly
                                                                                   pattern of regulation
                                                                                   established by the Atomic
                                                                                   Energy Act, as amended (Act)
                                                                                   and are consistent with their
                                                                                   regulatory authority.
10 CFR Part 171...................  Annual Fees: For       D....................  These provisions are
                                     Reactor Licenses and                          designated a Category D
                                     Fuel Cycle Licenses                           because they do not meet any
                                     and Material                                  of the criteria of Category
                                     Licenses, Including                           A, B, or C. A State may adopt
                                     Holders of                                    similar provisions that are
                                     Certificates of                               compatible with the orderly
                                     Compliance,                                   pattern of regulation
                                     Registrations, and                            established by the Atomic
                                     Quality Assurance                             Energy Act, as amended (Act)
                                     Program Approvals                             and are consistent with their
                                     and Government                                regulatory authority.
                                     Agencies Licensed by
                                     NRC.
----------------------------------------------------------------------------------------------------------------

IX. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, Pub. 
L. 104-113, requires that Federal agencies use technical standards that 
are developed or adopted by voluntary consensus standards bodies unless 
using such a standard is inconsistent with applicable law or is 
otherwise impractical. In this rule, the NRC is revising the procedural 
requirements for early site permits, standard design approvals, 
standard design certifications, combined licenses, and manufacturing 
licenses to make certain corrections and changes based on the 
experience of the previous design certification reviews and on 
discussions

[[Page 49468]]

with stakeholders on these licensing processes. These procedural 
requirements for rulemaking do not establish standards or substantive 
requirements with which all applicants and licensees must comply. In 
addition, portions of this rulemaking make conforming changes to 
regulatory requirements throughout 10 CFR Chapter I, such as access to 
national security information and the procedures governing the conduct 
of hearings in proceedings. These changes also do not establish 
standards or substantive requirements with which all applicants and 
licensees must comply. Finally, portions of this rulemaking make 
conforming changes to technical requirements throughout 10 CFR Chapter 
I, in order to make clear their applicability to applicants and 
licensees under part 52. Inasmuch as the purpose of this rulemaking was 
not to establish or fundamentally alter these technical requirements, 
the Commission considers it impractical to perform a reassessment of 
the fundamental nature of these technical requirements in this 
rulemaking. In addition, this rule amends certain portions of the three 
design certification regulations 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 applicants and 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.

X. Environmental Impact--Categorical Exclusion

    The NRC has determined that these amendments fall within the types 
of actions described as 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 
regulation.\19\
---------------------------------------------------------------------------

    \19\ When 10 CFR part 52 was issued 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. To ensure that future changes 
in part 52 are categorically excluded, this rule contains an 
appropriate change to Sec.  51.22(c)(3).
---------------------------------------------------------------------------

XI. Paperwork Reduction Act Statement

    This final rule contains new or amended information collection 
requirements contained in 10 CFR parts 21, 25, 50, 51, 52, and 54 that 
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These requirements were approved by the Office of Management and 
Budget approval numbers 3150-0035, 3150-0046, 3150-0011, 3150-0021, 
3150-0151, and 3150-0155. The changes to 10 CFR parts 19, 20, 26, 55, 
72, 73, 75, 95, and 140 do not contain new or amended information 
collection requirements. Existing requirements were approved by the 
Office of Management and Budget, approval numbers 3150-0044, 3150-0014, 
3150-0146, 3150-0018, 3150-0132, 3150-0002, 3150-0055, 3150-0047, and 
3150-0039.
    The burden to the public for the information collections in 10 CFR 
part 52 is estimated to average 11,277 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. Send comments on any aspect of 
these information collections, including suggestions for reducing the 
burden to the records and FOIA/Privacy Services Branch (T-5 F53, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001), or by 
Internet electronic mail to <a href="/cdn-cgi/l/email-protection#80c9cec6cfc3cfccccc5c3d4d3c0ced2c3aec7cfd6"><span class="__cf_email__" data-cfemail="70393e363f333f3c3c35332423303e22335e373f26">[email protected]</span></a>; and to the Desk 
Officer, Office of Information and Regulatory Affairs, NEOB-10202 
(3150-0035, 3150-0046, 3150-0011, 3150-0151, and 3150-0155 with revised 
information collection requirements), Office of Management and Budget, 
Washington, DC 20503.

Public Protection Notification

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

XII. Regulatory Analysis

    The Commission has prepared a regulatory analysis on this final 
rule. Consistent with the Regulatory Analysis Guidelines, the NRC 
performed an aggregate analysis of the rule. The analysis is based on 
the assumption that the NRC will receive 19 COL applications during the 
next 3 years and 1 COL application per year over the next 17 years. The 
net present value of the part 52 rule modifications are estimated to 
result in costs to the industry of $58,992 K and $30,952 K using a 3-
percent and a 7-percent discount rate, respectively. The provisions of 
the rule relating to part 21 are estimated to result in net present 
value costs of $3,873 K and $2,363 K to the industry, using a 3-percent 
and a 7-percent discount rate, respectively. The net present value of 
the entire rule is estimated to result in net costs to the industry of 
$29,726 K and $204 K at a 3-percent and a 7-percent discount rate, 
respectively. In addition, the rule is estimated to be a one time net 
present value savings to the NRC of $10,443 K.

XIII. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act (5 U.S.C. 
605(b)), the Commission certifies that this rule will not have a 
significant economic impact on a substantial number of small entities. 
This rule affects 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 
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).

XIV. Backfit Analysis

    The NRC has determined that the backfit rule does not apply to this 
rule and, therefore, a backfit analysis is not required, because the 
rule does not contain any provisions that would impose backfitting as 
defined in the backfit rule, 10 CFR 50.109.
    There are no current holders of combined licenses or manufacturing 
licenses that are protected by the backfitting restrictions in 
Sec. Sec.  50.109, 52.39, 52.98, or 52.171. To the extent that this 
rule revises the requirements for future early site permits, standard 
design certifications, combined licenses, standard design approvals and 
manufacturing licenses for nuclear power plants, these revisions do not 
constitute backfits because they are prospective in nature and the 
backfit rule is not intended to apply to every NRC action which 
substantially changes the expectations of future applicants.

[[Page 49469]]

    The NRC issued the first early site permits prior to the effective 
date of this final part 52 rule. In addition, there are applications 
for early site permits currently being considered by the NRC. As 
discussed elsewhere, the NRC has included a ``grandfathering 
provision'' in the final part 52 rulemaking which provides that the 
early site permit provisions in subpart A of part 52 do not apply to 
early site permits whose applications were docketed before the 
effective date of the final part 52 rulemaking, unless requested by the 
early site permit applicant. This grandfathering provision prohibits 
any backfitting for these early site permits.
    Other provisions in this rule would apply to currently-approved 
standard design approvals and certifications, but they are not 
protected by the backfitting restrictions in Sec.  50.104 or Sec.  
52.63 because they are either corrections, administrative changes, or 
provide additional flexibility to applicants or licensees who might 
reference the design approvals or certifications, and thus constitute a 
voluntary alternative or relaxation.
    Finally, some of the provisions in this rule represent conforming 
changes throughout 10 CFR Chapter I which are being made to reflect 
Commission adoption of design approvals and design certification 
processes which should have been made at the time the Commission first 
adopted these processes by rulemaking. While these conforming changes 
may, in some cases, affect the way in which a current design 
certification or design approval may be referenced, they do not 
directly affect the design approval nor are the conforming changes 
result in any inconsistency with the finality provisions in the design 
certifications or in part 52. Accordingly, the Commission believes that 
these conforming changes with respect to design approvals and design 
certifications do not raise new backfitting considerations.

XV. Congressional Review Act

    Under the Congressional Review Act of 1996, the NRC has determined 
that this action is not a major rule and has verified this 
determination with the Office of Information and Regulatory Affairs of 
OMB.

List of Subjects

10 CFR Part 1

    Organization and functions (Government Agencies).

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 10

    Administrative practice and procedure, Classified information, 
Government employees, Security measures.

10 CFR Part 19

    Criminal penalties, Environmental protection, Nuclear materials, 
Nuclear power plants and reactors, Occupational safety and health, 
Radiation protection, Reporting and recordkeeping requirements, Sex 
discrimination.

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 
recordkeeping requirements, Source material, Special nuclear material, 
Waste treatment and disposal.

10 CFR Part 21

    Nuclear power plants and reactors, Penalties, Radiation protection, 
Reporting and recordkeeping requirements.

10 CFR Part 25

    Classified information, Criminal penalties, Investigations, 
Reporting and recordkeeping requirements, Security measures.

10 CFR Part 26

    Alcohol abuse, Alcohol testing, Appeals, Chemical testing, Drug 
abuse, Drug testing, Employee assistance programs, Fitness for duty, 
Management actions, Nuclear power reactors, Protection of information, 
Reporting and recordkeeping requirements.

10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Emergency 
Planning, Fire protection, Intergovernmental relations, Nuclear power 
plants and reactors, Radiation protection, Reactor siting criteria, 
Reporting and recordkeeping requirements.

10 CFR Part 51

    Administrative practice and procedure, Environmental impact 
statement, Nuclear materials, Nuclear power plants and reactors, 
Reporting and recordkeeping 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 recordkeeping requirements, 
Standard design, Standard design certification.

10 CFR Part 54

    Administrative practice and procedure, Age-related degradation, 
Backfitting, Classified information, Criminal penalties, Environmental 
protection, Nuclear power plants and reactors, Reporting and 
recordkeeping requirements.

10 CFR Part 55

    Criminal penalties, Manpower training programs, Nuclear power 
plants and reactors, Reporting and recordkeeping requirements.

10 CFR Part 72

    Administrative practice and procedure, Criminal penalties, Manpower 
training programs, Nuclear materials, Occupational safety and health, 
Penalties, Radiation protection, Reporting and recordkeeping 
requirements, Security measures, Spent fuel, Whistleblowing.

10 CFR Part 73

    Criminal penalties, Export, Hazardous materials transportation, 
Import, Nuclear materials, Nuclear power plants and reactors, Reporting 
and recordkeeping requirements, Security measures.

10 CFR Part 75

    Criminal penalties, Intergovernmental relations, Nuclear materials, 
Nuclear power plants and reactors, Reporting and recordkeeping 
requirements, Security measures.

10 CFR Part 95

    Classified information, Criminal penalties, Reporting and 
recordkeeping requirements, Security measures.

10 CFR Part 140

    Criminal penalties, Extraordinary nuclear occurrence, Insurance, 
Intergovernmental relations, Nuclear materials, Nuclear power plants 
and reactors, Reporting and recordkeeping requirements.

10 CFR Part 170

    Byproduct material, Import and export licenses, Intergovernmental

[[Page 49470]]

relations, Non-payment penalties, Nuclear materials, Nuclear power 
plants and reactors, Source material, Special nuclear material.

10 CFR Part 171

    Nuclear power plants and reactors.

0
For the reasons set forth in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting 
the following amendments to 10 CFR parts 1, 2, 10, 19, 20, 21, 25, 26, 
50, 51, 52, 54, 55, 72, 73, 75, 95, 140, 170, and 171.

PART 1--STATEMENT OF ORGANIZATION AND GENERAL INFORMATION

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

    Authority: Secs. 23, 161, 68 Stat. 925, 948, as amended (42 
U.S.C. 2033, 2201); sec. 29, Pub. L. 85-256, 71 Stat. 579, Pub. L. 
95-209, 91 Stat. 1483 (42 U.S.C. 2039); sec. 191, Pub. L. 87-615, 76 
Stat. 409 (42 U.S.C. 2241); secs. 201, 203, 204, 205, 209, 88 Stat. 
1242, 1244, 1245, 1246, 1248, as amended (42 U.S.C. 5841, 5843, 
5844, 5845, 5849); 5 U.S.C. 552, 553; Reorganization Plan No. 1 of 
1980, 45 FR 40561, June 16, 1980.


0
2. In Sec.  1.43, paragraph (a)(2) is revised to read as follows:


Sec.  1.43  Office of Nuclear Reactor Regulation.

* * * * *
    (a) * * *
    (2) Receipt, possession, and ownership of source, byproduct, and 
special nuclear material used or produced at facilities licensed under 
10 CFR parts 50, 52, and 54;
* * * * *

PART 2--RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND 
ISSUANCE OF ORDERS

0
3. 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; sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 
note).
    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(o)), 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, 104, 105, 163, 183i, 189, 68 
Stat. 936, 937, 938, 954, 955, as amended (42 U.S.C. 2132, 2133, 
2134, 2135, 2233, 2239). Sections 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). Subpart C also issued under sec. 189, 68 Stat. 955 (42 U.S.C. 
2239). Sections 2.600-2.606 also issued under sec. 102, Pub. L. 91-
190, 83 Stat. 853, as amended (42 U.S.C. 4332). Section 2.700a also 
issued under 5 U.S.C. 554. Sections 2.343, 2.346, 2.754, 2.712 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 3.790 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). Subpart N also issued 
under sec. 189, 68 Stat. 955 (42 U.S.C. 2239). Appendix A also 
issued under sec. 6, Pub. L. 91-550, 84 Stat. 1473 (42 U.S.C. 2135).


0
4. In Sec.  2.1, paragraphs (c) and (d) are revised and a new paragraph 
(e) is added to read as follows:


Sec.  2.1  Scope.

* * * * *
    (c) Imposing civil penalties under Section 234 of the Act;
    (d) Rulemaking under the Act and the Administrative Procedure Act; 
and
    (e) Standard design approvals under part 52 of this chapter.

0
5. In Sec.  2.4, the definitions of contested proceeding, license and 
licensee are revised to read as follows:


Sec.  2.4  Definitions.

* * * * *
    Contested proceeding means--
    (1) A proceeding in which there is a controversy between the NRC 
staff and the applicant for a license or permit concerning the issuance 
of the license or permit or any of the terms or conditions thereof;
    (2) A proceeding in which the NRC is imposing a civil penalty or 
other enforcement action, and the subject of the civil penalty or 
enforcement action is an applicant for or holder of a license or 
permit, or is or was an applicant for a standard design certification 
under part 52 of this chapter; and
    (3) A proceeding in which a petition for leave to intervene in 
opposition to an application for a license or permit has been granted 
or is pending before the Commission.
* * * * *
    License means a license, including an early site permit, 
construction permit, operating license, combined license, manufacturing 
license, or renewed license issued by the Commission.
    Licensee means a person who is authorized to conduct activities 
under a license.
* * * * *

0
6. The heading of Subpart A is revised to read as follows:

Subpart A--Procedure for Issuance, Amendment, Transfer, or Renewal 
of a License, and Standard Design Approval

0
7. Section 2.100 is revised to read as follows:


Sec.  2.100  Scope of subpart.

    This subpart prescribes the procedure for issuance of a license; 
amendment of a license at the request of the licensee; transfer and 
renewal of a license; and issuance of a standard design approval under 
subpart E of part 52 of this chapter.

0
8. In Sec.  2.101, paragraphs (a)(1), (a)(2), the introductory 
paragraph of (a)(3), paragraph (a)(3)(ii), paragraph (a)(4), paragraph 
(a)(5), and paragraph (a-1) are revised to read as follows:


Sec.  2.101  Filing of application.

    (a)(1) An application for a permit, a license, a license transfer, 
a license amendment, a license renewal, or a standard design approval, 
shall be filed with the Director of New Reactors or Director of Nuclear 
Reactor Regulation or Director of Nuclear Material Safety and 
Safeguards, as prescribed by the applicable provisions of this chapter. 
A prospective applicant may confer informally with the NRC staff before 
filing an application.
    (2) Each application for a license for a facility or for receipt of 
waste radioactive material from other persons for the purpose of 
commercial disposal by the waste disposal licensee will be assigned a 
docket number. However, to allow a determination as to whether an 
application for a construction permit, operating license, early site 
permit, standard design approval, combined license, or manufacturing 
license for a production or utilization facility is complete and 
acceptable for docketing, it will be initially treated as a tendered 
application. A copy of the tendered application will be available for 
public inspection at the NRC Web site, http://www.nrc.gov, and/or at 
the NRC

[[Page 49471]]

Public Document Room. Generally, the determination on acceptability for 
docketing will be made within a period of 30 days. However, in selected 
applications, the Commission may decide to determine acceptability 
based on the technical adequacy of the application as well as its 
completeness. In these cases, the Commission, under Sec.  2.104(a), 
will direct that the notice of hearing be issued as soon as practicable 
after the application has been tendered, and the determination of 
acceptability will be made generally within a period of 60 days. For 
docketing and other requirements for applications under part 61 of this 
chapter, see paragraph (g) of this section.
    (3) If the Director of New Reactors, Director of Nuclear Reactor 
Regulation, or Director of Nuclear Material Safety and Safeguards, as 
appropriate, determines that a tendered application for a construction 
permit, operating license, early site permit, standard design approval, 
combined license, or manufacturing license for a production or 
utilization facility, and/or any environmental report required under 
subpart A of part 51 of this chapter, or part thereof as provided in 
paragraphs (a)(5) or (a-1) of this section are complete and acceptable 
for docketing, a docket number will be assigned to the application or 
part thereof, and the applicant will be notified of the determination. 
With respect to the tendered application and/or environmental report or 
part thereof that is acceptable for docketing, the applicant will be 
requested to:
* * * * *
    (ii) Serve a copy on the chief executive of the municipality in 
which the facility or site which is the subject of an early site permit 
is to be located or, if the facility or site which is the subject of an 
early site permit is not to be located within a municipality, on the 
chief executive of the county, and serve a notice of availability of 
the application or environmental report on the chief executives of the 
municipalities or counties which have been identified in the 
application or environmental report as the location of all or part of 
the alternative sites, containing the following information, as 
applicable: Docket number of the application, a brief description of 
the proposed site and facility; the location of the site and facility 
as primarily proposed and alternatively listed; the name, address, 
telephone number, and email address (if available) of the applicant's 
representative who may be contacted for further information; 
notification that a draft environmental impact statement will be issued 
by the Commission and will be made available upon request to the 
Commission; and notification that if a request is received from the 
appropriate chief executive, the applicant will transmit a copy of the 
application and environmental report, and any changes to these 
documents which affect the alternative site location, to the executive 
who makes the request. In complying with the requirements of this 
paragraph, the applicant should not make public distribution of those 
parts of the application subject to Sec.  2.390(d). The applicant shall 
submit to the Director of New Reactors or the Director of Nuclear 
Reactor Regulation an affidavit that service of the notice of 
availability of the application or environmental report has been 
completed along with a list of names and addresses of those executives 
upon whom the notice was served; and
* * * * *
    (4) The tendered application for a construction permit, operating 
license, early site permit, standard design approval, combined license, 
or manufacturing license will be formally docketed upon receipt by the 
Director of New Reactors, Director of Nuclear Reactor Regulation, or 
Director of Nuclear Material Safety and Safeguards, as appropriate, of 
the required additional copies. Distribution of the additional copies 
shall be deemed to be complete as of the time the copies are deposited 
in the mail or with a carrier prepaid for delivery to the designated 
addresses. The date of docketing shall be the date when the required 
copies are received by the Director of New Reactors, Director of 
Nuclear Reactor Regulation, or Director of Nuclear Material Safety and 
Safeguards, as appropriate. Within 10 days after docketing, the 
applicant shall submit to the Director of New Reactors, Director of 
Nuclear Reactor Regulation, or Director of Nuclear Material Safety and 
Safeguards, as appropriate, an affidavit that distribution of the 
additional copies to Federal, State, and local officials has been 
completed in accordance with the requirements of this chapter and 
written instructions furnished to the applicant by the Director of New 
Reactors, Director of Nuclear Reactor Regulation, or Director of 
Nuclear Material Safety and Safeguards, as appropriate. Amendments to 
the application and environmental report shall be filed and distributed 
and an affidavit shall be furnished to the Director of New Reactors, 
Director of Nuclear Reactor Regulation, or Director of Nuclear Material 
Safety and Safeguards, as appropriate, in the same manner as for the 
initial application and environmental report. If it is determined that 
all or any part of the tendered application and/or environmental report 
is incomplete and therefore not acceptable for processing, the 
applicant will be informed of this determination, and the respects in 
which the document is deficient.
    (5) An applicant for a construction permit under part 50 of this 
chapter or a combined license under part 52 of this chapter for a 
production or utilization facility which is subject to Sec.  51.20(b) 
of this chapter, and is of the type specified in Sec.  50.21(b)(2) or 
(3) or Sec.  50.22 of this chapter or is a testing facility may submit 
the information required of applicants by part 50 or part 52 of the 
chapter in two parts. One part shall be accompanied by the information 
required by Sec.  50.30(f) of this chapter, or Sec.  52.80(b) of this 
chapter, as applicable. The other part shall include any information 
required by Sec.  50.34(a) and, if applicable, Sec.  50.34a of this 
chapter, or Sec. Sec.  52.79 and 52.80(a), as applicable. One part may 
precede or follow other parts by no longer than 6 months. If it is 
determined that either of the parts as described above is incomplete 
and not acceptable for processing, the Director of New Reactors, 
Director of Nuclear Reactor Regulation, or Director of Nuclear Material 
Safety and Safeguards, as appropriate, will inform the applicant of 
this determination and the respects in which the document is deficient. 
Such a determination of completeness will generally be made within a 
period of 30 days. Whichever part is filed first shall also include the 
fee required by Sec. Sec.  50.30(e) and 170.21 of this chapter and the 
information required by Sec. Sec.  50.33, 50.34(a)(1) or 52.79(a)(1), 
as applicable, and Sec.  50.37 of this chapter. The Director of New 
Reactors, Director of Nuclear Reactor Regulation, or Director of 
Nuclear Material Safety and Safeguards, as appropriate, will accept for 
docketing an application for a construction permit under part 50 or a 
combined license under part 52 for a production or utilization facility 
which is subject to Sec.  51.20(b) of this chapter, and is of the type 
specified in Sec.  50.21(b)(2) or (3) or Sec.  50.22 of this chapter or 
is a testing facility where one part of the application as described 
above is complete and conforms to the requirements of part 50 of this 
chapter. The additional parts will be docketed upon a determination by 
the Director of New Reactors, Director of Nuclear Reactor Regulation, 
or Director of Nuclear Material Safety and Safeguards, as appropriate, 
that it is complete.
    (a-1) Early consideration of site suitability issues. An applicant 
for a

[[Page 49472]]

construction permit under part 50 of this chapter or a combined license 
under part 52 of this chapter for a utilization facility which is 
subject to Sec.  51.20(b) of this chapter and is of the type specified 
in Sec.  50.21(b)(2) or (3) or Sec.  50.22 of this chapter or is a 
testing facility, may request that the Commission conduct an early 
review and hearing and render an early partial decision in accordance 
with subpart F of this part on issues of site suitability within the 
purview of the applicable provisions of parts 50, 51, 52, and 100 of 
this chapter.
    (1) Construction permit. The applicant for the construction permit 
may submit the information required of applicants by the provisions of 
this chapter in three parts:
    (i) Part one shall include or be accompanied by any information 
required by Sec. Sec.  50.34(a)(1) and 50.30(f) of this chapter which 
relates to the issue(s) of site suitability for which an early review, 
hearing, and partial decision are sought, except that information with 
respect to operation of the facility at the projected initial power 
level need not be supplied, and shall include the information required 
by Sec. Sec.  50.33(a) through (e) and 50.37 of this chapter. The 
information submitted shall also include:
    (A) Proposed findings on the issues of site suitability on which 
the applicant has requested review and a statement of the bases or the 
reasons for those findings,
    (B) A range of postulated facility design and operation parameters 
that is sufficient to enable the Commission to perform the requested 
review of site suitability issues under the applicable provisions of 
parts 50, 51, and 100, and
    (C) Information concerning the applicant's site selection process 
and long-range plans for ultimate development of the site required by 
Sec.  2.603(b)(1).
    (ii) Part two shall include or be accompanied by the remaining 
information required by Sec. Sec.  50.30(f), 50.33, and 50.34(a)(1) of 
this chapter.
    (iii) Part three shall include the remaining information required 
by Sec. Sec.  50.34a and (in the case of a nuclear power reactor) 
50.34(a) of this chapter.
    (iv) The information required for part two or part three shall be 
submitted during the period the partial decision on part one is 
effective. Submittal of the information required for part three may 
precede by no more than 6 months or follow by no more than 6 months the 
submittal of the information required for part two.
    (2) Combined license under part 52. An applicant for a combined 
license under part 52 of this chapter may submit the information 
required of applicants by the provisions of this chapter in three 
parts:
    (i) Part one shall include or be accompanied by any information 
required by Sec. Sec.  52.79(a)(1) and 50.30(f) of this chapter which 
relates to the issue(s) of site suitability for which an early review, 
hearing, and partial decision are sought, except that information with 
respect to operation of the facility at the projected initial power 
level need not be supplied, and shall include the information required 
by Sec. Sec.  50.33(a) through (e) and 50.37 of this chapter. The 
information submitted shall also include:
    (A) Proposed findings on the issues of site suitability on which 
the applicant has requested review and a statement of the bases or the 
reasons for those findings;
    (B) A range of postulated facility design and operation parameters 
that is sufficient to enable the Commission to perform the requested 
review of site suitability issues under the applicable provisions of 
parts 50, 51, 52, and 100; and
    (C) Information concerning the applicant's site selection process 
and long-range plans for ultimate development of the site required by 
Sec.  2.621(b)(1).
    (ii) Part two shall include or be accompanied by the remaining 
information required by Sec. Sec.  50.30(f), 50.33, and 52.79(a)(1) of 
this chapter.
    (iii) Part three shall include the remaining information required 
by Sec. Sec.  52.79 and 52.80 of this chapter.
    (iv) The information required for part two or part three shall be 
submitted during the period the partial decision on part one is 
effective. Submittal of the information required for part three may 
precede by no more than 6 months or follow by no more than 6 months the 
submittal of the information required for part two.
* * * * *

0
9. In Sec.  2.102, paragraph (a) is revised to read as follows:


Sec.  2.102  Administrative review of application.

    (a) During review of an application by the NRC staff, an applicant 
may be required to supply additional information. The staff may request 
any one party to the proceeding to confer with the staff informally. In 
the case of a docketed application for a construction permit, operating 
license, early site permit, standard design approval, combined license, 
or manufacturing license of this chapter, the staff shall establish a 
schedule for its review of the application, specifying the key 
intermediate steps from the time of docketing until the completion of 
its review.
* * * * *

0
10. Section 2.104 is revised to read as follows:


Sec.  2.104  Notice of hearing.

    (a) In the case of an application on which a hearing is required by 
the Act or this chapter, or in which the Commission finds that a 
hearing is required in the public interest, the Secretary will issue a 
notice of hearing to be published in the Federal Register. The notice 
must be published at least 15 days, and in the case of an application 
concerning a construction permit, early site permit, or combined 
license for a facility of the type described in Sec.  50.21(b) or Sec.  
50.22 of this chapter or a testing facility, at least 30 days before 
the date set for hearing in the notice.\1\ In addition, in the case of 
an application for a construction permit, early site permit, or 
combined license for a facility of the type described in Sec.  50.22 of 
this chapter, or a testing facility, the notice must be issued as soon 
as practicable after the NRC has docketed the application; provided, 
that if the NRC decides, under Sec.  2.101(a)(2), to determine the 
acceptability of the application based upon its technical adequacy as 
well as completeness, the notice shall be issued as soon as practicable 
after the application has been tendered.
---------------------------------------------------------------------------

    \1\ If the notice of hearing concerning an application for a 
construction permit, early site permit, or combined license for a 
facility of the type described in Sec.  50.22 of this chapter or a 
testing facility does not specify the time and place of initial 
hearing, a subsequent notice will be published in the Federal 
Register which will provide at least 30 days notice of the time and 
place of that hearing. After this notice is given, the presiding 
officer may reschedule the commencement of the initial hearing for a 
later date or reconvene a recessed hearing without again providing 
at least 30 days notice.
---------------------------------------------------------------------------

    (b) The notice of hearing must state:
    (1) The nature of the hearing;
    (2) The authority under which the hearing is to be held;
    (3) The matters of fact and law to be considered;
    (4) The date by which requests for hearing or petitions to 
intervene must be filed;
    (5) The presiding officer designated for the hearing, or the 
procedure that the Commission will use to designate a presiding officer 
for the hearing.
    (c)(1) The Secretary will transmit a notice of hearing on an 
application for a license for a production or utilization facility 
including an early site permit, combined license (but not for a 
manufacturing license), for a license for

[[Page 49473]]

receipt of waste radioactive material from other persons for the 
purpose of commercial disposal by the waste disposal licensee, for a 
license under part 61 of this chapter, for a construction authorization 
for an HLW repository at a geologic repository operations area under 
parts 60 or 63 of this chapter, for a license to receive and possess 
high-level radioactive waste at a geologic repository operations area 
under parts 60 or 63 of this chapter, and for a license under part 72 
of this chapter to acquire, receive or possess spent fuel for the 
purpose of storage in an independent spent fuel storage installation 
(ISFSI) to the governor or other appropriate official of the State and 
to the chief executive of the municipality in which the facility is to 
be located or the activity is to be conducted or, if the facility is 
not to be located or the activity conducted within a municipality, to 
the chief executive of the county (or to the Tribal organization, if it 
is to be located or conducted within an Indian reservation).
    (2) The Secretary will transmit a notice of hearing on an 
application for a license under part 72 of this chapter to acquire, 
receive or possess spent fuel, high-level radioactive waste or 
radioactive material associated with high-level radioactive waste for 
the purpose of storage in a monitored retrievable storage installation 
(MRS) to the same persons who received the notice of docketing under 
Sec.  72.16(e) of this chapter.

0
11. In Sec.  2.105, the introductory text of paragraphs (a) and (a)(4) 
are revised, and paragraphs (a)(12), (a)(13), and (b)(3) are added to 
read as follows:


Sec.  2.105  Notice of proposed action.

    (a) If a hearing is not required by the Act or this chapter, and if 
the Commission has not found that a hearing is in the public interest, 
it will, before acting thereon, publish in the Federal Register, as 
applicable, either a notice of intended operation under Sec.  52.103(a) 
of this chapter and a proposed finding that inspections, tests, 
analysis, and acceptance criteria for a combined license under subpart 
C of part 52 have been or will be met, or a notice of proposed action 
with respect to an application for:
* * * * *
    (4) An amendment to an operating license, combined license, or 
manufacturing license for a facility licensed under Sec. Sec.  50.21(b) 
or 50.22 of this chapter, or for a testing facility, as follows:
* * * * *
    (12) An amendment to an early site permit issued under subpart A of 
part 52 of this chapter, as follows:
    (i) If the early site permit does not provide authority to conduct 
the activities allowed under Sec.  50.10(e)(1) of this chapter, the 
amendment will involve no significant hazards consideration, and though 
the NRC will provide notice of opportunity for a hearing under this 
section, it may make the amendment immediately effective and grant a 
hearing thereafter; and
    (ii) If the early site permit provides authority to conduct the 
activities allowed under Sec.  50.10(e)(1) and the Commission 
determines under Sec. Sec.  50.58 and 50.91 of this chapter that an 
emergency situation exists or that exigent circumstances exist and that 
the amendment involves no significant hazards consideration, it will 
provide notice of opportunity for a hearing under Sec.  2.106 of this 
chapter (if a hearing is requested, which will be held after issuance 
of the amendment).
    (13) A manufacturing license under subpart F of part 52 of this 
chapter.
    (b) * * *
    (3) For a notice of intended operation under Sec.  52.103(a) of 
this chapter, the following information:
    (i) The identification of the NRC action as making the finding 
required under Sec.  52.103(g) of this chapter;
    (ii) The manner in which the licensee notifications under 10 CFR 
52.99(c) which are required to be made available by 10 CFR 52.99(e)(2) 
may be obtained and examined;
    (iii) The manner in which copies of the safety analysis may be 
obtained and examined; and
    (iv) Any conditions, limitations, or restrictions to be placed on 
the license in connection with the finding under Sec.  52.103(g) of 
this chapter, and the expiration date or circumstances (if any) under 
which the conditions, limitations or restrictions will no longer apply.
* * * * *

0
12. In Sec.  2.106, paragraphs (a) and (b) are revised to read as 
follows:


Sec.  2.106  Notice of issuance.

    (a) The Director of New Reactors, Director of Nuclear Reactor 
Regulation, or Director of Nuclear Material Safety and Safeguards, as 
appropriate, will inform the State and local officials specified in 
Sec.  2.104(e) and publish a document in the Federal Register 
announcing the issuance of:
    (1) A license or an amendment of a license for which a notice of 
proposed action has been previously published;
    (2) An amendment of a license for a facility of the type described 
in Sec.  50.21(b) or Sec.  50.22 of this chapter, or a testing 
facility, whether or not a notice of proposed action has been 
previously published; and
    (3) The finding under Sec.  52.103(g) of this chapter.
    (b) The notice of issuance will set forth:
    (1) In the case of a license or amendment:
    (i) The nature of the license or amendment;
    (ii) The manner in which copies of the safety analysis, if any, may 
be obtained and examined; and
    (iii) A finding that the application for the license or amendment 
complies with the requirements of the Act and this chapter.
    (2) In the case of a finding under Sec.  52.103(g) of this chapter:
    (i) The manner in which copies of the safety analysis, if any, may 
be obtained and examined; and
    (ii) A finding that the prescribed inspections, tests, and analyses 
have been performed, the prescribed acceptance criteria have been met, 
and that the license complies with the requirements of the Act and this 
chapter.
* * * * *

0
13. Section 2.109 is revised to read as follows:


Sec.  2.109  Effect of timely renewal application.

    (a) Except for the renewal of an operating license for a nuclear 
power plant under 10 CFR 50.21(b) or 50.22, an early site permit under 
subpart A of part 52 of this chapter, a manufacturing license under 
subpart F of part 52 of this chapter, or a combined license under 
subpart C of part 52 of this chapter, if at least 30 days before the 
expiration of an existing license authorizing any activity of a 
continuing nature, the licensee files an application for a renewal or 
for a new license for the activity so authorized, the existing license 
will not be deemed to have expired until the application has been 
finally determined.
    (b) If the licensee of a nuclear power plant licensed under 10 CFR 
50.21(b) or 50.22 files a sufficient application for renewal of either 
an operating license or a combined license at least 5 years before the 
expiration of the existing license, the existing license will not be 
deemed to have expired until the application has been finally 
determined.
    (c) If the holder of an early site permit licensed under subpart A 
of part 52 of this chapter files a sufficient application for renewal 
under Sec.  52.29 of this chapter at least 12 months before the 
expiration of the existing early site permit, the

[[Page 49474]]

existing permit will not be deemed to have expired until the 
application has been finally determined.
    (d) If the licensee of a manufacturing license under subpart F of 
part 52 of this chapter files a sufficient application for renewal 
under Sec.  52.177 of this chapter at least 12 months before the 
expiration of the existing license, the existing license will not be 
deemed to have expired until the application has been finally 
determined.

0
14. Section 2.110 is revised to read as follows:


Sec.  2.110  Filing and administrative action on submittals for 
standard design approval or early review of site suitability issues.

    (a)(1) A submittal for a standard design approval under subpart E 
of part 52 of this chapter shall be subject to Sec. Sec.  2.101(a) and 
2.390 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 
appendix Q to parts 50 of this chapter, a submittal for early review of 
site suitability issues under appendix Q to parts 50 of this chapter 
shall be subject to Sec. Sec.  2.101(a)(2) through (4) to the same 
extent as if it were an application for a permit or license.
    (b) Upon initiation of review by the NRC staff of a submittal for 
an early review of site suitability issues under appendix Q of parts 50 
of this chapter, or for a standard design approval under subpart E of 
part 52 of this chapter, the Director of New Reactors or the Director 
of Nuclear Reactor Regulation shall publish in the Federal Register a 
notice of receipt of the submittal, inviting comments from interested 
persons within 60 days of publication or other time as may be 
specified, for consideration by the NRC staff and ACRS in their review.
    (c)(1) Upon completion of review by the NRC staff and the ACRS of a 
submittal for a standard design approval, the Director of New Reactors 
or the Director of the Office of Nuclear Reactor Regulation shall 
publish in the Federal Register a determination as to whether or not 
the design is acceptable, subject to terms and conditions as may be 
appropriate, and shall make available at the NRC Web site, http://www.nrc.gov, a report that analyzes the design.
    (2) Upon completion of review by the NRC staff and, if appropriate 
by the ACRS, of a submittal for early review of site suitability 
issues, the NRC staff shall prepare a staff site report which shall 
identify the location of the site, state the site suitability issues 
reviewed, explain the nature and scope of the review, state the 
conclusions of the staff regarding the issues reviewed and state 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 the report available at 
the NRC Web site, http://www.nrc.gov. The NRC staff shall also send a 
copy of the report to the Governor or other appropriate official of the 
State in which the site is located, and to the chief executive of the 
municipality in which the site is located or, if the site is not 
located in a municipality, to the chief executive of the county.

0
15. Section 2.111 is revised to read as follows:


Sec.  2.111  Prohibition of sex discrimination.

    No person shall on the grounds of sex be excluded from 
participation in, be denied a license, standard design approval, or 
petition for rulemaking (including a design certification), be denied 
the benefits of, or be subjected to discrimination under any program or 
activity carried on or receiving Federal assistance under the Act or 
the Energy Reorganization Act of 1974.

0
16. In Sec.  2.202, paragraph (e) is revised to read as follows:


Sec.  2.202  Orders.

* * * * *
    (e)(1) If the order involves the modification of a part 50 license 
and is a backfit, the requirements of Sec.  50.109 of this chapter 
shall be followed, unless the licensee has consented to the action 
required.
    (2) If the order involves the modification of combined license 
under subpart C of part 52 of this chapter, the requirements of Sec.  
52.98 of this chapter shall be followed unless the licensee has 
consented to the action required.
    (3) If the order involves a change to an early site permit under 
subpart A of part 52 of this chapter, the requirements of Sec.  52.39 
of this chapter must be followed, unless the applicant or licensee has 
consented to the action required.
    (4) If the order involves a change to a standard design 
certification rule referenced by that plant's application, the 
requirements, if any, in the referenced design certification rule with 
respect to changes must be followed, or, in the absence of these 
requirements, the requirements of Sec.  52.63 of this chapter must be 
followed, unless the applicant or licensee has consented to follow the 
action required.
    (5) If the order involves a change to a standard design approval 
referenced by that plant's application, the requirements of Sec.  
52.145 of this chapter must be followed unless the applicant or 
licensee has consented to follow the action required.
    (6) If the order involves a modification of a manufacturing license 
under subpart F of part 52, the requirements of Sec.  52.171 of this 
chapter must be followed, unless the applicant or licensee has 
consented to the action required.

0
17. In Sec.  2.309, paragraphs (a), (f)(1)(i), (f)(1)(v), and 
(f)(1)(vi) are revised, a new paragraph (f)(1)(vii) is added, and 
paragraphs (g), (h)(2), and (i) are revised to read as follows:


Sec.  2.309  Hearing requests, petitions to intervene, requirements for 
standing, and contentions.

    (a) General requirements. Any person whose interest may be affected 
by a proceeding and who desires to participate as a party must file a 
written request for hearing and a specification of the contentions 
which the person seeks to have litigated in the hearing. In a 
proceeding under 10 CFR 52.103, the Commission, acting as the presiding 
officer, will grant the request if it determines that the requestor has 
standing under the provisions of paragraph (d) of this section and has 
proposed at least one admissible contention that meets the requirements 
of paragraph (f) of this section. For all other proceedings, except as 
provided in paragraph (e) of this section, the Commission, presiding 
officer, or the Atomic Safety and Licensing Board designated to rule on 
the request for hearing and/or petition for leave to intervene, will 
grant the request/petition if it determines that the requestor/
petitioner has standing under the provisions of paragraph (d) of this 
section and has proposed at least one admissible contention that meets 
the requirements of paragraph (f) of this section. In ruling on the 
request for hearing/petition to intervene submitted by petitioners 
seeking to intervene in the proceeding on the HLW repository, the 
Commission, the presiding officer, or the Atomic Safety and Licensing 
Board shall also consider any failure of the petitioner to participate 
as a potential party in the pre-license application phase under subpart 
J of this part in addition to the factors in paragraph (d) of this 
section. If a request for hearing or petition to intervene is filed in 
response to any notice of hearing or opportunity for hearing, the 
applicant/licensee shall be deemed to be a party.
* * * * *
    (f) * * *
    (1) * * *

[[Page 49475]]

    (i) Provide a specific statement of the issue of law or fact to be 
raised or controverted, provided further, that the issue of law or fact 
to be raised in a request for hearing under 10 CFR 52.103(b) must be 
directed at demonstrating that one or more of the acceptance criteria 
in the combined license have not been, or will not be met, and that the 
specific operational consequences of nonconformance would be contrary 
to providing reasonable assurance of adequate protection of the public 
health and safety;
* * * * *
    (v) Provide a concise statement of the alleged facts or expert 
opinions which support the requestor's/petitioner's position on the 
issue and on which the petitioner intends to rely at hearing, together 
with references to the specific sources and documents on which the 
requestor/petitioner intends to rely to support its position on the 
issue;
    (vi) In a proceeding other than one under 10 CFR 52.103, provide 
sufficient information to show that a genuine dispute exists with the 
applicant/licensee on a material issue of law or fact. This information 
must include references to specific portions of the application 
(including the applicant's environmental report and safety report) that 
the petitioner disputes and the supporting reasons for each dispute, 
or, if the petitioner believes that the application fails to contain 
information on a relevant matter as required by law, the identification 
of each failure and the supporting reasons for the petitioner's belief; 
and
    (vii) In a proceeding under 10 CFR 52.103(b), the information must 
be sufficient, and include supporting information showing, prima facie, 
that one or more of the acceptance criteria in the combined license 
have not been, or will not be met, and that the specific operational 
consequences of nonconformance would be contrary to providing 
reasonable assurance of adequate protection of the public health and 
safety. This information must include the specific portion of the 
report required by 10 CFR 52.99(c) which the requestor believes is 
inaccurate, incorrect, and/or incomplete (i.e., fails to contain the 
necessary information required by Sec.  52.99(c)). If the requestor 
identifies a specific portion of the Sec.  52.99(c) report as 
incomplete and the requestor contends that the incomplete portion 
prevents the requestor from making the necessary prima facie showing, 
then the requestor must explain why this deficiency prevents the 
requestor from making the prima facie showing.
* * * * *
    (g) Selection of hearing procedures. A request for hearing and/or 
petition for leave to intervene may, except in a proceeding under 10 
CFR 52.103, also address the selection of hearing procedures, taking 
into account the provisions of Sec.  2.310. If a request/petition 
relies upon Sec.  2.310(d), the request/petition must demonstrate, by 
reference to the contention and the bases provided and the specific 
procedures in subpart G of this part, that resolution of the contention 
necessitates resolution of material issues of fact which may be best 
determined through the use of the identified procedures.
    (h) * * *
    (2) Except in a proceeding under 10 CFR 52.103, the requestor/
petitioner may file a reply to any answer. The reply must be filed 
within 7 days after service of that answer.
* * * * *
    (i) Decision on request/petition. In all proceedings other than a 
proceeding under 10 CFR 52.103, the presiding officer shall, within 45 
days after the filing of answers and replies under paragraph (h) of 
this section, issue a decision on each request for hearing/petition to 
intervene, absent an extension from the Commission. The Commission, 
acting as the presiding officer, shall expeditiously grant or deny the 
request for hearing in a proceeding under 10 CFR 52.103. The 
Commission's decision may not be the subject of any appeal under 10 CFR 
2.311.

0
18. In Sec.  2.310, paragraph (j) is redesignated as paragraph (k), and 
a new paragraph (j) is added to read as follows:


Sec.  2.310  Selection of hearing procedures.

* * * * *
    (j) Proceedings on a Commission finding under 10 CFR 52.103(c) and 
(g) shall be conducted in accordance with the procedures designated by 
the Commission in each proceeding.
* * * * *

0
19. In Sec.  2.339, paragraph (d) is revised to read as follows:


Sec.  2.339  Expedited decisionmaking procedure.

* * * * *
    (d) The provisions of this section do not apply to an initial 
decision directing the issuance of a limited work authorization under 
10 CFR 50.10, an early site permit under subpart A of part 52 of this 
chapter, a construction permit or construction authorization, a 
combined license under subpart C of part 52 of this chapter, or a 
manufacturing license under subpart F of part 52.

0
20. Section 2.340 is revised to read as follows:


Sec.  2.340  Initial decision in certain contested proceedings; 
immediate effectiveness of initial decisions; issuance of 
authorizations, permits, and licenses.

    (a) Initial decision--production or utilization facility operating 
license. In any initial decision in a contested proceeding on an 
application for an operating license (including an amendment to or 
renewal of an operating license) for a production or utilization 
facility, the presiding officer shall make findings of fact and 
conclusions of law on the matters put into controversy by the parties 
to the proceeding, any matter designated by the Commission to be 
decided by the presiding officer, and any matter not put into 
controversy by the parties, but only to the extent that the presiding 
officer determines that a serious safety, environmental, or common 
defense and security matter exists, and the Commission approves of an 
examination of and decision on the matter upon its referral by the 
presiding officer. Depending on the resolution of those matters, the 
Commission, the Director of Nuclear Reactor Regulation, or the Director 
of New Reactors, as appropriate, after making the requisite findings, 
will issue, deny or appropriately condition the license.
    (b) Initial decision--combined license under 10 CFR part 52. In any 
initial decision in a contested proceeding on an application for a 
combined license (including an amendment to or renewal of a combined 
license) under subpart C of part 52 of this chapter, the presiding 
officer shall make findings of fact and conclusions of law on the 
matters put into controversy by the parties to the proceeding, and any 
matter designated by the Commission to be decided by the presiding 
officer. Depending on the resolution of those matters, the Commission, 
the Director of New Reactors, or the Director of Nuclear Reactor 
Regulation, as appropriate, after making the requisite findings, will 
issue, deny or appropriately condition the license.
    (c) Initial decision on finding under 10 CFR 52.103 with respect to 
acceptance criteria in nuclear power reactor combined licenses. In any 
initial decision under Sec.  52.103(g) of this chapter with respect to 
whether acceptance criteria have been or will be met, the presiding 
officer shall make findings of fact and conclusions of law

[[Page 49476]]

on the matters put into controversy by the parties to the proceeding, 
and on any matters designated by the Commission to be decided by the 
presiding officer. Matters not put into controversy by the parties 
shall be referred to the Commission for its determination. The 
Commission may, in its discretion, treat the matter as a request for 
action under 10 CFR 2.206 and process the matter in accordance with 
Sec.  52.103(f). Depending on the resolution of those matters, the 
Commission, the Director of New Reactors, or the Director of Nuclear 
Reactor Regulation, as appropriate, will make the finding under 10 CFR 
52.103, or appropriately condition that finding.
    (d) Initial decision--manufacturing license under 10 CFR part 52. 
In any initial decision in a contested proceeding on an application for 
a manufactured license (including an amendment to or renewal of a 
combined license) under subpart C of part 52 of this chapter, the 
presiding officer shall make findings of fact and conclusions of law on 
the matters put into controversy by the parties to the proceeding, and 
any matter designated by the Commission to be decided by the presiding 
officer. Depending on the resolution of those matters, the Commission, 
the Director of New Reactors, or the Director of Nuclear Reactor 
Regulation, as appropriate, after making the requisite findings, will 
issue, deny, or appropriately condition the manufacturing license.
    (e) Initial decision--other proceedings not involving production or 
utilization facilities. In proceedings not involving production or 
utilization facilities, the presiding officer shall make findings of 
fact and conclusions of law on the matters put into controversy by the 
parties to the proceeding, and on any matters designated by the 
Commission to be decided by the presiding officer. Matters not put into 
controversy by the parties must be referred to the Director of Nuclear 
Material Safety and Safeguards, or the Director of the Office of 
Federal and State Materials and Environmental Management Programs, as 
appropriate. Depending on the resolution of those matters, the Director 
of Nuclear Material Safety and Safeguards or the Director of the Office 
of Federal and State Materials and Environmental Management Programs, 
as appropriate, after making the requisite findings, will issue, deny, 
revoke or appropriately condition the license, or take other action as 
necessary or appropriate.
    (f) Immediate effectiveness of certain decisions. An initial 
decision directing the issuance or amendment of a limited work 
authorization under 10 CFR 50.10, an early site permit under subpart A 
of part 52 of this chapter, a construction permit or construction 
authorization under part 50 of this chapter, an operating license under 
part 50 of this chapter, a combined license under subpart C of part 52 
of this chapter, a manufacturing license under subpart F of part 52 of 
this chapter, or a license under 10 CFR part 72 to store spent fuel in 
an independent spent fuel storage facility (ISFSI) or a monitored 
retrievable storage installation (MRS), an initial decision directing 
issuance of a license under part 61 of this chapter, or an initial 
decision under 10 CFR 52.103(g) that acceptance criteria in a combined 
license have been met, is immediately effective upon issuance unless 
the presiding officer finds that good cause has been shown by a party 
why the initial decision should not become immediately effective.
    (g)-(h) [Reserved]
    (i) Issuance of authorizations, permits, and licenses--production 
and utilization facilities. The Commission, the Director of New 
Reactors, or the Director of Nuclear Reactor Regulation, as 
appropriate, shall issue a limited work authorization under 10 CFR 
50.10, an early site permit under subpart A of part 52 of this chapter, 
a construction permit or construction authorization under part 50 of 
this chapter, an operating license under part 50 of this chapter, a 
combined license under subpart C of part 52 of this chapter, or a 
manufacturing license under subpart F of part 52 of this chapter within 
10 days from the date of issuance of the initial decision:
    (1) If the Commission or the appropriate Director has made all 
findings necessary for issuance of the authorization, permit or 
license, not within the scope of the initial decision of the presiding 
officer; and
    (2) Notwithstanding the pendency of a petition for reconsideration 
under Sec.  2.345, a petition for review under Sec.  2.341, or a motion 
for stay under Sec.  2.342, or the filing of a petition under Sec.  
2.206.
    (j) Issuance of finding on acceptance criteria under 10 CFR 52.103. 
The Commission, the Director of New Reactors, or the Director of 
Nuclear Reactor Regulation, as appropriate, shall make the finding 
under 10 CFR 52.103(g) that acceptance criteria in a combined license 
have been, or will be met, within 10 days from the date of issuance of 
the initial decision:
    (1) If the Commission or the appropriate Director has made the 
finding under Sec.  52.103(g) that acceptance criteria have been, or 
will be met, for those acceptance criteria which are not within the 
scope of the initial decision of the presiding officer; and
    (2) Notwithstanding the pendency of a petition for reconsideration 
under Sec.  2.345, a petition for review under Sec.  2.341, or a motion 
for stay under Sec.  2.342, or the filing of a petition under Sec.  
2.206.
    (k) Issuance of other licenses. The Commission or the Director of 
Nuclear Material Safety and Safeguards, or the Director of the Office 
of Federal and State Materials and Environmental Management Programs, 
as appropriate, shall issue a license, including a license under 10 CFR 
part 72 to store spent fuel in either an independent spent fuel storage 
facility (ISFSI) located away from a reactor site or at a monitored 
retrievable storage installation (MRS), within 10 days from the date of 
issuance of the initial decision:
    (1) If the Commission or the appropriate Director has made all 
findings necessary for issuance of the license, not within the scope of 
the initial decision of the presiding officer; and
    (2) Notwithstanding the pendency of a petition for reconsideration 
under Sec.  2.345, a petition for review under Sec.  2.341, or a motion 
for stay under Sec.  2.342, or the filing of a petition under Sec.  
2.206.

0
21. In Sec.  2.341, paragraph (a)(1) is revised to read as follows:


Sec.  2.341  Review of decisions and actions of a presiding officer.

    (a)(1) Except for requests for review or appeals under Sec.  2.311 
or in a proceeding on the high-level radioactive waste repository 
(which are governed by Sec.  2.1015), review of decisions and actions 
of a presiding officer are treated under this section, provided, 
however, that no party may request a further Commission review of a 
Commission determination to allow a period of interim operation under 
10 CFR 52.103(c).
* * * * *

0
22. In Sec.  2.347, paragraph (a) is revised, and new paragraph (f)(5) 
is added to read as follows:


Sec.  2.347  Ex parte communications.

* * * * *
    (a)(1) Interested persons outside the agency may not make or 
knowingly cause to be made to any Commission adjudicatory employee, any 
ex parte communication relevant to the merits of the proceeding.
    (2) For purposes of this section, merits of the proceeding 
includes:
    (i) A disputed issue;

[[Page 49477]]

    (ii) A matter which a presiding officer seeks to be referred to the 
Commission under 10 CFR 2.340(a); and
    (iii) A matter for which the Commission has approved examination by 
the presiding officer under Sec.  2.340(a).
* * * * *
    (f) * * *
    (5) Communications, in contested proceedings and uncontested 
mandatory proceeding, regarding an undisputed issue.

0
23. In Sec.  2.348, the introductory text of paragraph (a) is revised, 
and new paragraphs (d)(1)(iii), (d)(1)(iv), and (d)(3) are added to 
read as follows:


Sec.  2.348  Separation of functions.

    (a) In any proceeding under this part, any NRC officer or employee 
engaged in the performance of any investigative or litigating function 
in the proceeding or in a factually related proceeding with respect to 
a disputed issue in that proceeding, may not participate in or advise a 
Commission adjudicatory employee about the initial or final decision 
with respect to that disputed issue, except--
* * * * *
    (d) * * *
    (1) * * *
    (iii) A matter which a presiding officer seeks to be referred to 
the Commission under 10 CFR 2.340(a); and
    (iv) A matter for which the Commission has approved examination by 
the presiding officer under Sec.  2.340(a).
* * * * *
    (3) Separation of functions does not apply to uncontested 
proceedings, or to an undisputed issue in contested initial licensing 
proceedings.
* * * * *

0
24. In Sec.  2.390, the introductory text of paragraph (a) is revised 
to read as follows:


Sec.  2.390  Public inspections, exemptions, requests for withholding.

    (a) Subject to the provisions of paragraphs (b), (d), (e), and (f) 
of this section, final NRC records and documents, including but not 
limited to correspondence to and from the NRC regarding the issuance, 
denial, amendment, transfer, renewal, modification, suspension, 
revocation, or violation of a license, permit, order, or standard 
design approval, or regarding a rulemaking proceeding subject to this 
part shall not, in the absence of an NRC determination of a compelling 
reason for nondisclosure after a balancing of the interests of the 
person or agency urging nondisclosure and the public interest in 
disclosure, be exempt from disclosure and will be made available for 
inspection and copying at the NRC Web site, http://www.nrc.gov, and/or 
at the NRC Public Document Room, except for matters that are:
* * * * *
0
25. Subpart D is revised to read as follows:

Subpart D--Additional Procedures Applicable to Proceedings for the 
Issuance of Licenses To Construct and/or Operate Nuclear Power 
Plants of Identical Design at Multiple Sites

Sec.
2.400 Scope of subpart.
2.401 Notice of hearing on construction permit or combined license 
applications pursuant to appendix N of 10 CFR parts 50 or 52.
2.402 Separate hearings on separate issues; consolidation of 
proceedings.
2.403 Notice of proposed action on applications for operating 
licenses pursuant to appendix N of 10 CFR part 50.
2.404 Hearings on applications for operating licenses pursuant to 
appendix N of 10 CFR part 50.
2.405 Initial decisions in consolidated hearings.
2.406 Finality of decisions on separate issues.
2.407 Applicability of other sections.


Sec.  2.400  Scope of subpart.

    This subpart describes procedures applicable to licensing 
proceedings which involve the consideration in hearings of a number of 
applications, filed by one or more applicants pursuant to appendix N of 
parts 50 or 52 of this chapter, for licenses to construct and/or 
operate nuclear power reactors of identical design to be located at 
multiple sites.


Sec.  2.401  Notice of hearing on construction permit or combined 
license applications pursuant to appendix N of 10 CFR parts 50 or 52.

    (a) In the case of applications pursuant to appendix N of part 50 
of this chapter for construction permits for nuclear power reactors of 
the type described in Sec.  50.22 of this chapter, or applications 
pursuant to appendix N of part 52 of this chapter for combined 
licenses, the Secretary will issue notices of hearing pursuant to 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.


Sec.  2.402  Separate hearings on separate issues; consolidation of 
proceedings.

    (a) In the case of applications under appendix N of part 50 of this 
chapter for construction permits for nuclear power reactors of a type 
described in 10 CFR 50.22, or applications pursuant to appendix N of 
part 52 of this chapter for combined licenses, 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 or environmental matters.
    (b) If a separate hearing is held on a particular phase of the 
proceeding, the Commission or presiding officers of each affected 
proceeding may, under 10 CFR 2.317, consolidate for hearing on that 
phase two or more proceedings to consider common issues relating to the 
applications involved in the proceedings, if it finds that this action 
will be conducive to the proper dispatch of its business and to the 
ends of justice. In specifying the place of this consolidated hearing, 
due regard will be given to the convenience and necessity of the 
parties, petitioners for leave to intervene, or the attorneys or 
representatives of such persons, and the public interest.


Sec.  2.403  Notice of proposed action on applications for operating 
licenses pursuant to appendix N of 10 CFR part 50.

    In the case of applications pursuant to appendix N of part 50 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 
Commission, the Director of New Reactors, or the Director of Nuclear 
Reactor Regulation will, prior to acting thereon, cause to be published 
in the Federal Register, pursuant to Sec.  2.105, a notice of proposed 
action with respect to each application as soon as practicable after 
the applications have been docketed.


Sec.  2.404  Hearings on applications for operating licenses pursuant 
to appendix N of 10 CFR part 50.

    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 pursuant to appendix N of part 
50 of this chapter with respect to a specific reactor(s) at a specific 
site, and the Commission, the Chief Administrative Judge, or a 
presiding officer has issued a notice of hearing or other appropriate 
order, then the Commission, the Chief Administrative Judge, or the 
presiding officer 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.

[[Page 49478]]

Sec.  2.405  Initial decisions in consolidated hearings.

    At the conclusion of a hearing held under this subpart, the 
presiding officer will render a partial initial decision on the common 
design. The partial initial decision on the common design may be 
appealed under Sec.  2.341. If the proceedings have also been 
consolidated with respect to matters other than the common design under 
Sec.  2.317(b), the presiding officer may issue a consolidated partial 
initial decision for those proceedings. No construction permit, full-
power operating license, or combined license under part 52 of this 
chapter will be issued until an initial decision has been issued on all 
phases of the hearing and all issues under the Act and the National 
Environmental Policy Act of 1969 appropriate to the proceeding have 
been resolved.


Sec.  2.406  Finality of decisions on separate issues.

    Notwithstanding any other provision of this chapter, in a 
proceeding conducted pursuant to this subpart and appendices N of parts 
50 or 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.


Sec.  2.407  Applicability of other sections.

    The provisions of subparts A, C, G, L, and N of this part relating 
to construction permits, operating licenses, and combined licenses 
apply, respectively, to construction permits, operating licenses, and 
combined licenses subject to this subpart, except as may be qualified 
by the provisions of this subpart.

0
26. 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 F of part 52 of this chapter.

0
27. In Sec.  2.501, the section heading, the introductory text of 
paragraph (a) and paragraph (b) are revised to read as follows:


Sec.  2.501  Notice of hearing on application under subpart F of 10 CFR 
part 52 for a license to manufacture nuclear power reactors.

    (a) In the case of an application under subpart F 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 will issue a 
notice of hearing to be published in the Federal Register at least 30 
days before the date set for hearing in the notice.\1\ The notice shall 
be issued as soon as practicable after the application has been 
docketed. The notice will state:
---------------------------------------------------------------------------

    \1\ The thirty-day (30) requirement of this paragraph is not 
applicable to a notice of the time and place of hearing published by 
the presiding officer after the notice of hearing described in this 
section has been published.
---------------------------------------------------------------------------

* * * * *
    (b) The notice of hearing shall comply with the requirements of 
Sec.  2.104(f) of this chapter.
* * * * *


Sec.  2.502  [Removed]

0
28. Remove and reserve Sec.  2.502.


Sec.  2.503  [Removed]

0
29. Remove and reserve Sec.  2.503.


Sec.  2.504  [Removed]

0
30. Remove and reserve Sec.  2.504.

0
31. Subpart F is revised to read as follows:

Subpart F--Additional Procedures Applicable to Early Partial 
Decisions on Site Suitability Issues in Connection With an 
Application for a Construction Permit or Combined License for 
Certain Utilization Facilities

Sec.
2.600 Scope of subpart.
2.601 Applicability of other sections.

Early Partial Decisions on Site Suitability--Construction Permit

2.602 Filing Fees.
2.603 Acceptance and docketing of application for early review of 
site suitability issues in a construction permit proceeding.
2.604 Notice of hearing on application for early review of site 
suitability issues in construction permit proceeding.
2.605 Additional considerations.
2.606 Partial decision on site suitability issues in construction 
permit proceeding.

Early Partial Decisions on Site Suitability--Combined License Under 10 
CFR Part 52

2.621 Acceptance and docketing of application for early review of 
site suitability issues in a combined license proceeding.
2.623 Notice of hearing on application for early review of site 
suitability issues in combined license proceeding.
2.625 Additional considerations.
2.627 Partial decision on site suitability issues in combined 
license proceeding.
2.629 Finality of partial decision on site suitability issues in 
combined license proceeding.


Sec.  2.600  Scope of subpart.

    This subpart prescribes procedures applicable to licensing 
proceedings which involve an early submittal of site suitability 
information in accordance with Sec.  2.101(a-1) and (a-2), and a 
hearing and early partial decision on issues of site suitability, in 
connection with an application for a permit to construct a utilization 
facility which is subject to Sec.  51.20(b) of this chapter and is of 
the type specified in Sec.  50.21(b)(2) or (3) or Sec.  50.22 of this 
chapter or is a testing facility; or an application for a combined 
license under part 52 of this chapter for a nuclear power facility.
    (a) The procedures in Sec. Sec.  2.601 through 2.609 apply to all 
applications under this subpart.
    (b) The procedures in Sec. Sec.  2.611 through 2.619 apply to 
applications for a permit to construct a utilization facility which is 
subject to Sec.  51.20(b) of this chapter and is of the type specified 
in Sec.  50.21(b)(2) or (3) or Sec.  50.22 of this chapter or is a 
testing facility.
    (c) The procedures in Sec. Sec.  2.621 through 2.629 apply to 
applications for combined license under part 52 of this chapter for a 
nuclear power facility.


Sec.  2.601  Applicability of other sections.

    The provisions of subparts A, C, G, L, and N relating to 
applications for construction permits and combined licenses, and 
proceedings thereon apply, respectively, to such applications and 
proceedings in accordance with this subpart, except as specifically 
provided otherwise by the provisions of this subpart.

Early Partial Decisions on Site Suitability--Construction Permit


Sec.  2.602  Filing fees.

    Each application which contains a request for early review of site 
suitability issues under the procedures of this subpart shall be 
accompanied by any fee required by Sec.  50.30(e) and part 170 of this 
chapter.


Sec.  2.603  Acceptance and docketing of application for early review 
of site suitability issues in a construction permit proceeding.

    (a) Each part of an application for a construction permit submitted 
in accordance with Sec.  2.101(a-1) of this part will be initially 
treated as a tendered

[[Page 49479]]

application. If it is determined that any one of the parts as described 
in Sec.  2.101(a-1) is incomplete and not acceptable for processing, 
the Director of the Office of New Reactors or the Director of the 
Office of Nuclear Reactor Regulation, as appropriate, will inform the 
applicant of this determination and the respects in which the document 
is deficient. Such a determination of completeness will generally be 
made within a period of 30 days.
    (b)(1) The Director of the Office of New Reactors or the Director 
of the Office of Nuclear Reactor Regulation, as appropriate, will 
accept for docketing part one of an application for a construction 
permit for a utilization facility which is subject to Sec.  51.20(b) of 
this chapter and is of the type specified in Sec.  50.21(b)(2) or (3) 
or Sec.  50.22 of this chapter, or is a testing facility where part one 
of the application as described in Sec.  2.101(a-1) is complete. Part 
one of any application will not be considered complete unless it 
contains proposed findings as required by Sec.  2.101(a-1)(1)(i) and 
unless it describes the applicant's site selection process, specifies 
the extent to which that process involves the consideration of 
alternative sites, explains the relationship between that process and 
the application for early review of site suitability issues, and 
briefly describes the applicant's long-range plans for ultimate 
development of the site. Upon assignment of a docket number, the 
procedures in Sec.  2.101(a)(3) and (4) relating to formal docketing 
and the submission and distribution of additional copies of the 
application shall be followed.
    (2) Additional parts of the application will be docketed upon a 
determination by the Director of the Office of New Reactors or the 
Director of the Office of Nuclear Reactor Regulation, as appropriate, 
that they are complete.
    (c) If part one of the application is docketed, the Director of the 
Office of New Reactors or the Director of the Office of Nuclear Reactor 
Regulation, as appropriate, will cause to be published in the Federal 
Register and send to the Governor or other appropriate official of the 
State in which the site is located, a notice of docketing of the 
application which states the purpose of the application, states the 
location of the proposed site, states that a notice of hearing will be 
published, requests comments within 120 days or such other time as may 
be specified on the initiation or outcome of an early site review from 
Federal, State, and local agencies and interested persons.


Sec.  2.604  Notice of hearing on application for early review of site 
suitability issues in construction permit proceeding.

    (a) Where an applicant for a construction permit requests an early 
review and hearing and an early partial decision on issues of site 
suitability pursuant to Sec.  2.101(a-1), the provisions in the notice 
of hearing setting forth the matters of fact and law to be considered, 
as required by Sec.  2.104, shall be modified so as to relate only to 
the site suitability issue or issues under review.
    (b) After docketing of part two of the application, as provided in 
Sec. Sec.  2.101(a-1) and 2.603, a supplementary notice of hearing will 
be published under Sec.  2.104 with respect to the remaining unresolved 
issues in the proceeding within the scope of Sec.  2.104. This 
supplementary notice of hearing will provide that any person whose 
interest may be affected by the proceeding and who desires to 
participate as a party in the resolution of the remaining issues shall 
file a petition for leave to intervene pursuant to Sec.  2.309 within 
the time prescribed in the notice. This supplementary notice will also 
provide appropriate opportunities for participation by a representative 
of an interested State under Sec.  2.315(c) and for limited appearances 
under Sec.  2.315(a).
    (c) Any person who was permitted to intervene as a party under the 
initial notice of hearing on site suitability issues and who was not 
dismissed or did not withdraw as a party may continue to participate as 
a party to the proceeding with respect to the remaining unresolved 
issues, provided that within the time prescribed for filing of 
petitions for leave to intervene in the supplementary notice of 
hearing, he or she files a notice of his intent to continue as a party, 
along with a supporting affidavit identifying the specific aspect or 
aspects of the subject matter of the proceeding as to which he or she 
wishes to continue to participate as a party and setting forth with 
particularity the basis for his contentions with regard to each aspect 
or aspects. A party who files a non-timely notice of intent to continue 
as a party may be dismissed from the proceeding, absent a determination 
that the party has made a substantial showing of good cause for failure 
to file on time, and with particular reference to the factors specified 
in Sec. Sec.  2.309(c)(1)(i) through (iv) and 2.309(d). The notice will 
be ruled upon by the Commission or presiding officer designated to rule 
on petitions for leave to intervene.
    (d) To the maximum extent practicable, the membership of any atomic 
safety and licensing board designated to preside in the proceeding on 
the remaining unresolved issues pursuant to the supplemental notice of 
hearing will be the same as the membership designated to preside in the 
initial notice of hearing on site suitability issues.


Sec.  2.605  Additional considerations.

    (a) The Commission will not conduct more than one review of site 
suitability issues with regard to a particular site prior to filing and 
review of part two of the application described in Sec.  2.101(a-1) of 
this part.
    (b) The Commission, upon its own initiative, or upon the motion of 
any party to the proceeding filed at least 60 days prior to the date of 
the commencement of the evidentiary hearing on site suitability issues, 
may decline to initiate an early hearing or render an early partial 
decision on any issue or issues of site suitability:
    (1) In cases where no partial decision on the relative merits of 
the proposed site and alternative sites under subpart A of part 51 of 
this chapter is requested, upon determination that there is a 
reasonable likelihood that further review would identify one or more 
preferable alternative sites and the partial decision on one or more 
site suitability issues would lead to an irreversible and irretrievable 
commitment of resources prior to the submittal of the remainder of the 
information required by Sec.  50.30(f) of this chapter that would 
prejudice the later review and decision on such alternative sites; or
    (2) In cases where it appears that an early partial decision on any 
issue or issues of site suitability would not be in the public interest 
considering:
    (i) The degree of likelihood that any early findings on those 
issues would retain their validity in later reviews;
    (ii) The objections, if any, of cognizant State or local government 
agencies to the conduct of an early review on those issues; and
    (iii) The possible effect on the public interest and the parties of 
having an early, if not necessarily conclusive, resolution of those 
issues.


Sec.  2.606  Partial decision on site suitability issues in 
construction permit proceeding.

    (a) The provisions of Sec. Sec.  2.331, 2.339, 2.340, 2.343, 2.712, 
and 2.713 shall apply to any partial initial decision rendered in 
accordance with this subpart. A limited work authorization may not be 
issued under 10 CFR 50.10(e) and no construction permit may be issued 
without completion of the full review required by Section 102(2) of the 
National Environmental Policy Act of 1969, as amended, and

[[Page 49480]]

subpart A of part 51 of this chapter. The authority of the Commission 
to review such a partial initial decision sua sponte, or to raise sua 
sponte an issue that has not been raised by the parties, will be 
exercised within the same time period as in the case of a full decision 
relating to the issuance of a construction permit.
    (b)(1) A partial decision on one or more site suitability issues 
pursuant to the applicable provisions of part 50, subpart A of part 51, 
and part 100 of this chapter issued in accordance with this subpart 
shall:
    (i) Clearly identify the site to which the partial decision 
applies; and
    (ii) Indicate to what extent additional information may be needed 
and additional review may be required to enable the Commission to 
determine in accordance with the provisions of the Act and the 
applicable provisions of the regulations in this chapter whether a 
construction permit for a facility to be located on the site identified 
in the partial decision should be issued or denied.
    (2) Following either the Commission (acting in the function of a 
presiding officer) issuance of a partial initial decision, or 
completion of Commission review of the partial initial decision of the 
Atomic Safety and Licensing Board, after hearing, on the site 
suitability issues, the partial decision shall remain in effect either 
for a period of 5 years or, where the applicant for the construction 
permit has made timely submittal of the information required to support 
the application as provided in Sec.  2.101(a-1), until the proceeding 
for a permit to construct a facility on the site identified in the 
partial decision has been concluded,\3\ unless the Commission or Atomic 
Safety and Licensing Board, upon its own initiative or upon motion by a 
party to the proceeding, finds that there exists significant new 
information that substantially affects the earlier conclusions and 
reopens the hearing record on site suitability issues. Upon good cause 
shown, the Commission may extend the 5-year period during which a 
partial decision shall remain in effect for a reasonable period of time 
not to exceed 1 year.
---------------------------------------------------------------------------

    \3\ The partial decision on site suitability issues shall be 
incorporated in the decision regarding issuance of the combined 
license to the extent that it serves as a basis for the decision on 
a specific site issue.
---------------------------------------------------------------------------

Early Partial Decisions on Site Suitability--Combined License Under 10 
CFR Part 52


Sec.  2.621  Acceptance and docketing of application for early review 
of site suitability issues in a combined license proceeding.

    (a) Each part of an application submitted in accordance with Sec.  
2.101(a-1) of this part will be initially treated as a tendered 
application. If it is determined that any one of the parts as described 
in Sec.  2.101(a-1) is incomplete and not acceptable for processing, 
the Director of the Office of New Reactors or the Director of the 
Office of Nuclear Reactor Regulation, as appropriate, will inform the 
applicant of this determination and the respects in which the document 
is deficient. Such a determination of completeness will generally be 
made within a period of 30 days.
    (b)(1) The Director of the Office of New Reactors or the Director 
of the Office of Nuclear Reactor Regulation, as appropriate, will 
accept for docketing an application for a combined license for a 
nuclear power facility where part one of the application as described 
in Sec.  2.101(a-1) is complete. Part one of any application will not 
be considered complete unless it contains proposed findings as required 
by Sec.  2.101(a-1)(1)(i) and unless it describes the applicant's site 
selection process, specifies the extent to which that process involves 
the consideration of alternative sites, explains the relationship 
between that process and the application for early review of site 
suitability issues, and briefly describes the applicant's long-range 
plans for ultimate development of the site. Upon assignment of a docket 
number, the procedures in Sec.  2.101(a)(3) and (4) relating to formal 
docketing and the submission and distribution of additional copies of 
the application shall be followed.
    (2) Additional parts of the application will be docketed upon a 
determination by the Director of the Office of New Reactors or the 
Director of the Office of Nuclear Reactor Regulation, as appropriate, 
that they are complete.
    (c) If part one of the application is docketed, the Director of the 
Office of New Reactors or the Director of the Office of Nuclear Reactor 
Regulation, as appropriate, will cause to be published in the Federal 
Register and send to the Governor or other appropriate official of the 
State in which the site is located, a notice of docketing of the 
application which states the purpose of the application, states the 
location of the proposed site, states that a notice of hearing will be 
published, requests comments within 120 days or such other time as may 
be specified on the initiation or outcome of an early site review from 
Federal, State, and local agencies and interested persons.


Sec.  2.623  Notice of hearing on application for early review of site 
suitability issues in combined license proceeding.

    (a) Where an applicant for a combined license under part 52 of this 
chapter requests an early review and hearing and an early partial 
decision on issues of site suitability pursuant to Sec.  2.101(a-2), 
the provisions in the notice of hearing setting forth the matters of 
fact and law to be considered, as required by Sec.  2.104, shall be 
modified so as to relate only to the site suitability issue or issues 
under review. The notice will provide appropriate opportunities for 
participation by a representative of an interested State under Sec.  
2.315(c) and for limited appearances under Sec.  2.315(a), limited 
however, to the issues of site suitability for which early review has 
been requested by the applicant.
    (b) After docketing of part two of the application, as provided in 
Sec. Sec.  2.101(a-1) and 2.603, a supplementary notice of hearing will 
be published under Sec.  2.104 with respect to the remaining unresolved 
issues in the proceeding within the scope of Sec.  2.104. This 
supplementary notice of hearing will provide that any person whose 
interest may be affected by the proceeding and who desires to 
participate as a party in the resolution of the remaining issues shall 
file a petition for leave to intervene pursuant to Sec.  2.309 within 
the time prescribed in the notice. This supplementary notice will also 
provide appropriate opportunities for participation by a representative 
of an interested State under Sec.  2.315(c) and for limited appearances 
under Sec.  2.315(a).
    (c) Any person who was permitted to intervene as a party under the 
initial notice of hearing on site suitability issues and who was not 
dismissed or did not withdraw as a party may continue to participate as 
a party to the proceeding without having to demonstrate standing under 
Sec.  2.309(d), provided, however, that within the time prescribed for 
filing of petitions for leave to intervene in the supplementary notice 
of hearing, the party files a notice of intent to continue as a party. 
The notice must include the information required by Sec.  2.309(f). A 
party who files a non-timely notice of intent to continue as a party 
may be dismissed from the proceeding, absent a determination that the 
party has made a substantial showing of good cause for failure to file 
on time, and with particular reference to the factors specified in 
Sec. Sec.  2.309(c)(1)(i)

[[Page 49481]]

through (iv) and 2.309(d). The notice will be ruled upon by the 
Commission or presiding officer designated to rule on petitions for 
leave to intervene.
    (d) To the maximum extent practicable, the presiding officer (as 
applicable, the membership of the licensing board) designated to 
preside in the proceeding on the remaining unresolved issues pursuant 
to the supplemental notice of hearing will be the same as the presiding 
officer (as applicable, the membership of the licensing board) 
designated to preside in the initial notice of hearing on site 
suitability issues.


Sec.  2.625  Additional considerations.

    (a) The Commission will not conduct more than one review of site 
suitability issues with regard to a particular site prior to filing and 
review of part two of the application described in Sec.  2.101(a-1) of 
this part.
    (b) The Commission, upon its own initiative, or upon the motion of 
any party to the proceeding filed at least 60 days prior to the date of 
the commencement of the evidentiary hearing on site suitability issues, 
may decline to initiate an early hearing or render an early partial 
decision on any issue or issues of site suitability:
    (1) In cases where no partial decision on the relative merits of 
the proposed site and alternative sites under subpart A of part 51 is 
requested, upon determination that there is a reasonable likelihood 
that further review would identify one or more preferable alternative 
sites and the partial decision on one or more site suitability issues 
would lead to an irreversible and irretrievable commitment of resources 
prior to the submittal of the remainder of the information required by 
Sec.  50.30(f) of this chapter that would prejudice the later review 
and decision on such alternative sites; or
    (2) In cases where it appears that an early partial decision on any 
issue or issues of site suitability would not be in the public interest 
considering:
    (i) The degree of likelihood that any early findings on those 
issues would retain their validity in later reviews;
    (ii) The objections, if any, of cognizant State or local government 
agencies to the conduct of an early review on those issues; and
    (iii) The possible effect on the public interest and the parties of 
having an early, if not necessarily conclusive, resolution of those 
issues.


Sec.  2.627  Partial decision on site suitability issues in combined 
license proceeding.

    (a) The provisions of Sec. Sec.  2.331, 2.339, 2.340(b), 2.343, 
2.712, and 2.713 shall apply to any partial initial decision rendered 
in accordance with this subpart. Section 2.340(c) shall not apply to 
any partial initial decision rendered in accordance with this subpart. 
A limited work authorization may not be issued under 10 CFR 50.10(e) 
and no construction permit may be issued without completion of the full 
review required by Section 102(2) of the National Environmental Policy 
Act of 1969, as amended, and subpart A of part 51 of this chapter. The 
authority of the Commission to review such a partial initial decision 
sua sponte, or to raise sua sponte an issue that has not been raised by 
the parties, will be exercised within the same time period as in the 
case of a full decision relating to the issuance of a construction 
permit.
    (b)(1) A partial decision on one or more site suitability issues 
pursuant to the applicable provisions of part 50, subpart A of part 51, 
and part 100 of this chapter issued in accordance with this subpart 
shall:
    (i) Clearly identify the site to which the partial decision 
applies; and
    (ii) Indicate to what extent additional information may be needed 
and additional review may be required to enable the Commission to 
determine in accordance with the provisions of the Act and the 
applicable provisions of the regulations in this chapter whether a 
construction permit for a facility to be located on the site identified 
in the partial decision should be issued or denied.
    (2) Following either the Commission (acting in the function of a 
presiding officer) issuance of a partial initial decision, or 
completion of Commission review of the partial initial decision of the 
presiding officer, after hearing, on the site suitability issues, the 
partial decision shall remain in effect either for a period of 5 years 
or, where the applicant for the combined license has made timely 
submittal of the information required to support the application as 
provided in Sec.  2.101(a-2), until the proceeding for a combined 
license on the site identified in the partial decision has been 
concluded, unless the Commission or presiding officer, upon its own 
initiative or upon motion by a party to the proceeding, finds that 
there exists significant new information that substantially affects the 
earlier conclusions and reopens the hearing record on site suitability 
issues. Upon good cause shown, the Commission may extend the 5-year 
period during which a partial decision shall remain in effect for a 
reasonable period of time not to exceed 1 year.


Sec.  2.629  Finality of partial decision on site suitability issues in 
a combined license proceeding.

    (a) The partial decision on site suitability issues in a combined 
license proceeding shall be incorporated in the decision regarding 
issuance of a combined license. Except as provided in 10 CFR 2.758, in 
making the findings required for issuance of a combined license, the 
Commission shall treat as resolved those matters resolved in connection 
with the issuance of the partial decision on site suitability issues. 
If the Commission reaches an adverse decision, the application shall be 
denied without prejudice for resubmission, provided, however, that in 
determining whether the resubmitted application is complete and 
acceptable for docketing under Sec.  2.101(a)(3), the Director of the 
Office of New Reactors or the Director of the Office of Nuclear Reactor 
Regulation, as appropriate, shall determine whether the resubmitted 
application addresses those matters identified as bases for denial of 
the original application.
    (b) Notwithstanding any provision in 10 CFR 50.109, while a partial 
decision on site suitability is in effect under Sec.  2.617(b)(2), the 
Commission may not modify, rescind, or impose new requirements with 
respect to matters within the scope of the site suitability decision, 
whether on its own motion, or in response to a request or petition from 
any person, unless the Commission determines that a modification to the 
original decision is necessary either for compliance with the 
Commission's regulations applicable and in effect at the time the 
partial decision was issued, or to assure adequate protection of the 
public health and safety or the common defense and security.

0
32. Section 2.800 is revised to read as follows:


Sec.  2.800  Scope and applicability.

    (a) This subpart governs the issuance, amendment, and repeal of 
regulations in which participation by interested persons is prescribed 
under Section 553 of title 5 of the U.S. Code.
    (b) The procedures in Sec. Sec.  2.804 through 2.810 apply to all 
rulemakings.
    (c) The procedures in Sec. Sec.  2.802 through 2.803 apply to all 
petitions for rulemaking except for initial applications for standard 
design certification rulemaking under subpart B of part 52 of this 
chapter, and subsequent petitions for amendment of an existing design 
certification rule filed by the original applicant for the design 
certification rule.
    (d) The procedures in Sec. Sec.  2.811 through 2.819, as 
supplemented by the

[[Page 49482]]

provisions of subpart B of part 52, apply to standard design 
certification rulemaking.
0
33. Section 2.801 is revised to read as follows:


Sec.  2.801  Initiation of rulemaking.

    Rulemaking may be initiated by the Commission at its own instance, 
on the recommendation of another agency of the United States, or on the 
petition of any other interested person, including an application for 
design certification under subpart B of part 52 of this chapter.
0
34. In subpart H, Sec. Sec.  2.811, 2.813, 2.815, 2.817 and 2.819 are 
added to read as follows:


Sec.  2.811  Filing of standard design certification application; 
required copies.

    (a) Serving of applications. The signed original of an application 
for a standard design certification, including all amendments to the 
applications, must be sent either by mail addressed: ATTN: Document 
Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; by facsimile; by hand delivery to the NRC's offices at 11555 
Rockville Pike, Rockville, Maryland, between the hours of 7:30 a.m. and 
4:15 p.m. eastern time; or, where practicable, by electronic 
submission, for example, via Electronic Information Exchange, e-mail, 
or CD-ROM. Electronic submissions must be made in a manner that enables 
the NRC to receive, read, authenticate, distribute, and archive the 
submission, and process and retrieve it a single page at a time. 
Detailed guidance on making electronic submissions can be obtained by 
visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by e-mail at <a href="/cdn-cgi/l/email-protection#3d7874787d534f5e135a524b"><span class="__cf_email__" data-cfemail="71343834311f03125f161e07">[email protected]</span></a>, 
or by writing the Office of Information Services, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001. The guidance 
discusses, among other topics, the formats the NRC can accept, the use 
of electronic signatures, and the treatment of nonpublic information. 
If the communication is on paper, the signed original must be sent.
    (b) Form of application. Each original of an application and an 
amendment of an application must meet the requirements in Sec.  2.813.
    (c) Capability to provide additional copies. The applicant shall 
maintain the capability to generate additional copies of the general 
information and the safety analysis report, or part thereof or 
amendment thereto, for subsequent distribution in accordance with the 
written instructions of the Director, Office of New Reactors, the 
Director, Office of Nuclear Reactor Regulation, or the Director, Office 
of Nuclear Material Safety and Safeguards, as appropriate.
    (d) Public hearing copy. In any hearing conducted under subpart O 
of this part for a design certification rulemaking, the applicant must 
make a copy of the updated application available at the public hearing 
for the use of any other parties to the proceeding, and shall certify 
that the updated copies of the application contain the current contents 
of the application submitted in accordance with the requirements of 
this part.
    (e) Pre-application consultation. A prospective applicant for a 
standard design certification may consult with the NRC before filing an 
application by writing to the Director, Division of New Reactor 
Licensing, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, with respect to the subject matters listed in Sec.  
2.802(a)(1)(i) through (iii) of this chapter. A prospective petitioner 
also may telephone the Rulemaking, Directives, and Editing Branch on 
(301) 415-7163, or toll free on (800) 368-5642, or send e-mail to 
<a href="/cdn-cgi/l/email-protection#561804150413061638243578313920"><span class="__cf_email__" data-cfemail="632d3120312633230d11004d040c15">[email protected]</span></a> on these subject matters. In addition, a prospective 
applicant may confer informally with the NRC staff BEFORE filing an 
application for a standard design certification, and the limitations in 
Sec.  2.802(a)(2) do not apply.


Sec.  2.813  Written communications.

    (a) General requirements. All correspondence, reports, and other 
written communications from the applicant to the Nuclear Regulatory 
Commission concerning the regulations in this subpart, and parts 50, 
52, and 100 of this chapter must be sent either by mail addressed: 
ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; by hand delivery to the NRC's offices at 
11555 Rockville Pike, Rockville, Maryland, between the hours of 7:30 
a.m. and 4:15 p.m. eastern time; or, where practicable, by electronic 
submission, for example, via Electronic Information Exchange, e-mail, 
or CD-ROM. Electronic submissions must be made in a manner that enables 
the NRC to receive, read, authenticate, distribute, and archive the 
submission, and process and retrieve it a single page at a time. 
Detailed guidance on making electronic submissions can be obtained by 
visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by e-mail at <a href="/cdn-cgi/l/email-protection#91d4d8d4d1ffe3f2bff6fee7"><span class="__cf_email__" data-cfemail="a0e5e9e5e0ced2c38ec7cfd6">[email protected]</span></a>, 
or by writing the Office of Information Services, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001. The guidance 
discusses, among other topics, the formats the NRC can accept, the use 
of electronic signatures, and the treatment of nonpublic information. 
If the communication is on paper, the signed original must be sent. If 
a submission due date falls on a Saturday, Sunday, or Federal holiday, 
the next Federal working day becomes the official due date.
    (b) Form of communications. All paper copies submitted to meet the 
requirements set forth in paragraph (a) of this section must be 
typewritten, printed or otherwise reproduced in permanent form on 
unglazed paper. Exceptions to these requirements imposed on paper 
submissions may be granted for the submission of micrographic, 
photographic, or similar forms.
    (c) Regulation governing submission. An applicant submitting 
correspondence, reports, and other written communications under the 
regulations of this chapter is requested but not required to cite 
whenever practical, in the upper right corner of the first page of the 
submission, the specific regulation or other basis requiring 
submission.


Sec.  2.815  Docketing and acceptance review.

    (a) Each application for a standard design certification will be 
assigned a docket number. However, to allow a determination as to 
whether an application is complete and acceptable for docketing, it 
will be initially treated as a tendered application. A copy of the 
tendered application will be available for public inspection at the NRC 
Web site, http://www.nrc.gov, and/or at the NRC Public Document Room. 
Generally, the determination on acceptability for docketing will be 
made within a period of 30 days. The Commission may decide to determine 
acceptability on the basis of the technical adequacy of the application 
as well as its completeness.
    (b) If the Commission determines that a tendered application is 
complete and acceptable for docketing, a docket number will be assigned 
to the application or part thereof, and the applicant will be notified 
of the determination.


Sec.  2.817  Withdrawal of application.

    (a) The Commission may permit an applicant to withdraw an 
application for a standard design certification before the issuance of 
a notice of proposed rulemaking on such terms and conditions as the 
Commission may prescribe, or may, on receiving a request for withdrawal 
of an application, deny the application or dismiss it without 
prejudice. The NRC will publish in the

[[Page 49483]]

Federal Register a document withdrawing the application, if the notice 
of receipt of the application, an advance notice of proposed 
rulemaking, or a notice of proposed rulemaking for the standard design 
certification has been previously published in the Federal Register. If 
the notice of receipt, advance notice of proposed rulemaking or notice 
of proposed rulemaking was published on the NRC Web site, then the 
notice of action on the withdrawal will also be published on the NRC 
Web site.
    (b) The withdrawal of an application does not authorize the removal 
of any document from the files of the Commission.


Sec.  2.819  Denial of application for failure to supply information.

    (a) The Commission may deny an application for a standard design 
certification if an applicant fails to respond to a request for 
additional information within 30 days from the date of the request, or 
within such other time as may be specified.
    (b) If the Commission denies an application because the applicant 
has failed to respond in a timely fashion to a request for additional 
information, the NRC will publish in the Federal Register a notice of 
denial and will notify the applicant with a simple statement of the 
grounds of denial. If a notice of receipt of application, advance 
notice of proposed rulemaking, or notice of proposed rulemaking for a 
standard design certification was published on the NRC Web site, then 
the notice of action on the denial will also be published on the NRC 
Web site.


0
35. In Sec.  2.1202, paragraph (a) is revised to read as follows:


Sec.  2.1202  Authority and role of NRC staff.

    (a) During the pendency of any hearing under this subpart, 
consistent with the NRC staff's findings in its review of the 
application or matter which is the subject of the hearing and as 
authorized by law, the NRC staff is expected to issue its approval or 
denial of the application promptly, or take other appropriate action on 
the underlying regulatory matter for which a hearing was provided. When 
the NRC staff takes its action, it shall notify the presiding officer 
and the parties to the proceeding of its action. That notice must 
include the NRC staff's position on the matters in controversy before 
the presiding officer with respect to the staff action. The NRC staff's 
action on the matter is effective upon issuance by the staff, except in 
matters involving:
    (1) An application to construct and/or operate a production or 
utilization facility (including an application for a limited work 
authorization under 10 CFR 50.12, or an application for a combined 
license under subpart C of 10 CFR part 52);
    (2) An application for an early site permit under subpart A of 10 
CFR part 52;
    (3) An application for a manufacturing license under subpart F of 
10 CFR part 52;
    (4) An application for an amendment to a construction authorization 
for a high-level radioactive waste repository at a geologic repository 
operations area falling under either 10 CFR 60.32(c)(1) or 10 CFR part 
63;
    (5) An application for the construction and operation of an 
independent spent fuel storage installation (ISFSI) located at a site 
other than a reactor site or a monitored retrievable storage 
installation (MRS) under 10 CFR part 72; and
    (6) Production or utilization facility licensing actions that 
involve significant hazards considerations as defined in 10 CFR 50.92.
* * * * *


Sec.  2.1211  [Removed]

0
36. Section 2.1211 is removed.

PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR 
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN 
EMPLOYMENT CLEARANCE

0
37. The authority citation for part 10 continues to read as follows:

    Authority: Secs. 145, 161, 68 Stat. 942, 948, as amended (42 
U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C. 
5841); E.O. 10450, 3 CFR parts 1949-1953 COMP., p. 936, as amended; 
E.O. 10865, 3 CFR 1959-1963 COMP., p. 398, as amended; 3 CFR Table 
4; E.O. 12968, 3 CFR 1995 COM., p. 396.


0
38. In Sec.  10.1, paragraphs (a)(1) and (a)(2) are revised and 
paragraph (a)(3) is added to read as follows:


Sec.  10.1  Purpose.

    (a) * * *
    (1) The eligibility of individuals who are employed by or 
applicants for employment with NRC contractors, agents, and other 
individuals who are NRC employees or applicants for NRC employment, and 
other persons designated by the Deputy Executive Director for 
Information Services and Administration and Chief Information Officer 
of the NRC, for access to Restricted Data under the Atomic Energy Act 
of 1954, as amended, and the Energy Reorganization Act of 1974, or for 
access to national security information;
    (2) The eligibility of NRC employees, or the eligibility of 
applicants for employment with the NRC, for employment clearance; and
    (3) The eligibility of individuals who are employed by or are 
applicants for employment with NRC licensees, certificate holders, 
holders of standard design approvals under part 52 of this chapter, 
applicants for licenses, certificates, and NRC approvals, and others 
who may require access related to a license, certificate, or NRC 
approval, or other activities as the Commission may determine, for 
access to Restricted Data under the Atomic Energy Act of 1954, as 
amended, and the Energy Reorganization Act of 1974, or for access to 
national security information.
* * * * *

0
39. In Sec.  10.2, paragraph (b) is revised to read as follows:


Sec.  10.2  Scope.

* * * * *
    (b) NRC licensees, certificate holders and holders of standard 
design approvals under part 52 of this chapter, applicants for 
licenses, certificates, and standard design approvals under part 52 of 
this chapter, and their employees (including consultants) and 
applicants for employment (including consulting);
* * * * *

PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS; INSPECTION 
AND INVESTIGATIONS

0
40. The authority citation for part 19 is revised to read as follows:

    Authority: Secs. 53, 63, 81, 103, 104, 161, 186, 68 Stat. 930, 
933, 935, 936, 937, 948, 955, as amended, sec. 234, 83 Stat. 444, as 
amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 
2093, 2111, 2133, 2134, 2201, 2236, 2282, 2297f); sec. 201, 88 Stat. 
1242, as amended (42 U.S.C. 5841); Pub. L. 95-601, sec. 10, 92 Stat. 
2951 (42 U.S.C. 5851); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 
note).
    Section 19.32 is also issued under sec. 401, 88 Stat. 1254 (42 
U.S.C. 2000d, 42 U.S.C. 5891).


0
41. Section 19.1 is revised to read as follows:


Sec.  19.1  Purpose.

    The regulations in this part establish requirements for notices, 
instructions, and reports by licensees and regulated entities to 
individuals participating in NRC-licensed and regulated activities and 
options available to these individuals in connection with Commission 
inspections of licensees and regulated entities, and to ascertain

[[Page 49484]]

compliance with the provisions of the Atomic Energy Act of 1954, as 
amended, titles II and IV of the Energy Reorganization Act of 1974, and 
regulations, orders, and licenses thereunder. The regulations in this 
part also establish the rights and responsibilities of the Commission 
and individuals during interviews compelled by subpoena as part of 
agency inspections or investigations under Section 161c of the Atomic 
Energy Act of 1954, as amended, on any matter within the Commission's 
jurisdiction.


0
42. Section 19.2 is revised to read as follows:


Sec.  19.2  Scope.

    (a) The regulations in this part apply to:
    (1) All persons who receive, possess, use, or transfer material 
licensed by the NRC under the regulations in parts 30 through 36, 39, 
40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed 
to operate a production or utilization facility under parts 50 or 52 of 
this chapter, persons licensed to possess power reactor spent fuel in 
an independent spent fuel storage installation (ISFSI) under part 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;
    (2) All applicants for and holders of licenses (including 
construction permits and early site permits) under parts 50, 52, and 54 
of this chapter;
    (3) All applicants for and holders of a standard design approval 
under subpart E of part 52 of this chapter; and
    (4) All applicants for a standard design certification under 
subpart B of part 52 of this chapter, and those (former) applicants 
whose designs have been certified under that subpart.
    (b) The regulations in this part regarding interviews of 
individuals under subpoena apply to all investigations and inspections 
within the jurisdiction of the NRC other than those involving NRC 
employees or NRC contractors. The regulations in this part do not apply 
to subpoenas issued under 10 CFR 2.702.


0
43. In Sec.  19.3 the definitions of License and Worker are revised, 
and the definitions of Regulated entities and Regulated activities are 
added to read as follows:


Sec.  19.3  Definitions.

* * * * *
    License means a license issued under the regulations in parts 30 
through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including 
licenses to manufacture, construct and/or operate a production or 
utilization facility under parts 50, 52, or 54 of this chapter.
* * * * *
    Regulated activities means any activity carried on which is under 
the jurisdiction of the NRC under the Atomic Energy Act of 1954, as 
amended, or any title of the Energy Reorganization Act of 1972, as 
amended.
    Regulated entities means any individual, person, organization, or 
corporation that is subject to the regulatory jurisdiction of the NRC, 
including (but not limited to) an applicant for or holder of a standard 
design approval under subpart E of part 52 of this chapter or a 
standard design certification under subpart B of part 52 of this 
chapter.
* * * * *
    Worker means an individual engaged in activities licensed or 
regulated by the Commission and controlled by a licensee or regulated 
entity, but does not include the licensee or regulated entity.


0
44. In Sec.  19.11, paragraph (c) is removed and reserved, and the 
introductory text of paragraph (a), paragraphs (b), (d), and (e) are 
revised, and paragraphs (f) and (g) are added to read as follows:


Sec.  19.11  Posting of notices to workers.

    (a) Each licensee (except for a holder of an early site permit 
under subpart A of part 52 of this chapter, or a holder of a 
manufacturing license under subpart F of part 52 of this chapter) shall 
post current copies of the following documents:
* * * * *
    (b) Each applicant for and holder of a standard design approval 
under subpart E of part 52 of this chapter, each applicant for an early 
site permit under subpart A of part 52 of this chapter, each applicant 
for a standard design certification under subpart B of part 52 of this 
chapter, and each applicant for and holder of a manufacturing license 
under subpart F of part 52 of this chapter shall post:
    (1) The regulations in this part;
    (2) The operating procedures applicable to the activities regulated 
by the NRC which are being conducted by the applicant or holder; and
    (3) Any notice of violation, proposed imposition of civil penalty, 
or order issued under subpart B of part 2 of this chapter, and any 
response from the applicant or holder.
    (c) [Reserved]
    (d) If posting of a document specified in paragraphs (a)(1), (2) or 
(3), or (b)(1) or (2) of this section is not practicable, the licensee 
or regulated entity may post a notice which describes the document and 
states where it may be examined.
    (e)(1) Each licensee, each applicant for a specific license, each 
applicant for or holder of a standard design approval under subpart E 
of part 52 of this chapter, each applicant for an early site permit 
under subpart A of part 52 of this chapter, and each applicant for a 
standard design certification under subpart B of part 52 of this 
chapter shall prominently post NRC Form 3, ``Notice to Employees,'' 
dated August 1997. Later versions of NRC Form 3 that supersede the 
August 1997 version shall replace the previously posted version within 
30 days of receiving the revised NRC Form 3 from the Commission.
    (2) Additional copies of NRC Form 3 may be obtained by writing to 
the Regional Administrator of the appropriate U.S. Nuclear Regulatory 
Commission Regional Office listed in appendix D to part 20 of this 
chapter, by calling (301) 415-7232, via e-mail to <a href="/cdn-cgi/l/email-protection#375158455a447759455419505841"><span class="__cf_email__" data-cfemail="11777e637c62517f63723f767e67">[email protected]</span></a>, or by 
visiting the NRC's Web site at http://www.nrc.gov and selecting forms 
from the index found on the home page.
    (f) Documents, notices, or forms posted under this section shall 
appear in a sufficient number of places to permit individuals engaged 
in NRC-licensed or regulated activities to observe them on the way to 
or from any particular licensed or regulated activity location to which 
the document applies, shall be conspicuous, and shall be replaced if 
defaced or altered.
    (g) Commission documents posted under paragraphs (a)(4) or (b)(3) 
of this section shall be posted within 2 working days after receipt of 
the documents from the Commission; the licensee's or regulated entity's 
response, if any, shall be posted within 2 working days after dispatch 
by the licensee or regulated entity. These documents shall remain 
posted for a minimum of 5 working days or until action correcting the 
violation has been completed, whichever is later.


0
45. Section 19.14 is revised to read as follows:


Sec.  19.14  Presence of representatives of licensees and regulated 
entities, and workers during inspections.

    (a) Each licensee, applicant for a license, applicant for or holder 
of a standard design approval under subpart E of part 52 of this 
chapter, applicant for an early site permit under subpart A of part 52 
of this chapter, and applicant for a standard design certification 
under subpart B of part 52 of this chapter shall afford to the 
Commission at all

[[Page 49485]]

reasonable times opportunity to inspect materials, activities, 
facilities, premises, and records under the regulations in this 
chapter.
    (b) During an inspection, Commission inspectors may consult 
privately with workers as specified in Sec.  19.15. The licensee, 
regulated entity, or the licensee's or regulated entity's 
representative may accompany Commission inspectors during other phrases 
of an inspection.
    (c) If, at the time of inspection, an individual has been 
authorized by the workers to represent them during Commission 
inspections, the licensee or regulated entity shall notify the 
inspectors of such authorization and shall give the workers' 
representative an opportunity to accompany the inspectors during the 
inspection of physical working conditions.
    (d) Each workers' representative shall be routinely engaged in NRC-
licensed or regulated activities under control of the licensee or 
regulated entity, and shall have received instructions as specified in 
Sec.  19.12.
    (e) Different representatives of licensees or regulated entities, 
and workers may accompany the inspectors during different phases of an 
inspection if there is no resulting interference with the conduct of 
the inspection. However, only one workers' representative at a time may 
accompany the inspectors.
    (f) With the approval of the licensee or regulated entity, and the 
workers' representative an individual who is not routinely engaged in 
licensed or regulated activities under control of the license or 
regulated entity (for example, a consultant to the licensee, the 
regulated entity, or the workers' representative), shall be afforded 
the opportunity to accompany Commission inspectors during the 
inspection of physical working conditions.
    (g) Notwithstanding the other provisions of this section, 
Commission inspectors are authorized to refuse to permit accompaniment 
by any individual who deliberately interferes with a fair and orderly 
inspection. With regard to areas containing information classified by 
an agency of the U.S. Government in the interest of national security, 
an individual who accompanies an inspector may have access to such 
information only if authorized to do so. With regard to any area 
containing proprietary information, the workers' representative for 
that area shall be an individual previously authorized by the licensee 
or regulated entity to enter that area.


0
46. Section 19.20 is revised to read as follows:


Sec.  19.20  Employee protection.

    Employment discrimination by a licensee, a holder of a certificate 
of compliance issued under part 76 of this chapter or regulated entity 
subject to the requirements in this part as delineated in Sec.  
19.2(a), or a contractor or subcontractor of a licensee, a holder of a 
certificate of compliance issued under part 76 of this chapter, or 
regulated entity subject to the requirements in this part as delineated 
in Sec.  19.2(a), against an employee for engaging in protected 
activities under this part or parts 30, 40, 50, 52, 54, 60, 61, 63, 70, 
72, 76, or 150 of this chapter is prohibited.


0
47. Section 19.31 is revised to read as follows:


Sec.  19.31  Application for exemptions.

    The Commission may, upon application by any interested person or 
upon its own initiative, grant such exemptions from the requirements of 
the regulations in this part as it determines are authorized by law, 
will not result in undue hazard to life and property.


0
48. Section 19.32 is revised to read as follows:


Sec.  19.32  Discrimination prohibited.

    No person shall on the grounds of sex be excluded from 
participation in, be denied a license, be denied the benefit of, or be 
subjected to discrimination under any program or activity carried on 
which is under the jurisdiction of the NRC under the Atomic Energy Act 
of 1954, as amended, or under any title of the Energy Reorganization 
Act of 1974, as amended. This provision will be enforced through agency 
provisions and regulations similar to those already established, with 
respect to racial and other discrimination, under Title VI of the Civil 
Rights Act of 1964. This remedy is not exclusive, however, and will not 
prejudice or cut off any other legal remedies available to a 
discriminatee.

PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION

0
49. 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); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).


0
50. 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, 
63, 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 under Sec.  35.75, or to exposure from voluntary participation 
in medical research programs.

0
51. In Sec.  20.1401 paragraph (a) is revised to read as follows:


Sec.  20.1401  General provisions and scope.

    (a) The criteria in this subpart apply to the decommissioning of 
facilities licensed under parts 30, 40, 50, 52, 60, 61, 63, 70, and 72 
of this chapter, and release of part of a facility or site for 
unrestricted use in accordance with Sec.  50.83 of this chapter, as 
well as other facilities subject to the Commission's jurisdiction under 
the Atomic Energy Act of 1954, as amended, and the Energy 
Reorganization Act of 1974, as amended. For high-level and low-level 
waste disposal facilities (10 CFR parts 60, 61, and 63), the criteria 
apply only to ancillary surface facilities that support radioactive 
waste disposal activities. The criteria do not apply to uranium and 
thorium recovery facilities already subject to appendix A to 10 CFR 
part 40 or the uranium solution extraction facilities.
* * * * *

0
52. Section 20.1406 is revised to read as follows:


Sec.  20.1406  Minimization of contamination.

    (a) Applicants for licenses, other than early site permits and 
manufacturing licenses under part 52 of this chapter and renewals, 
whose applications are submitted after August 20, 1997, shall describe 
in the application how facility design and procedures for operation 
will minimize, to the extent practicable, contamination of the facility 
and the environment, facilitate eventual decommissioning, and minimize, 
to the extent practicable, the generation of radioactive waste.
    (b) Applicants for standard design certifications, standard design

[[Page 49486]]

approvals, and manufacturing licenses under part 52 of this chapter, 
whose applications are submitted after August 20, 1997, shall describe 
in the application how facility design will minimize, to the extent 
practicable, contamination of the facility and the environment, 
facilitate eventual decommissioning, and minimize, to the extent 
practicable, the generation of radioactive waste.


0
53. In Sec.  20.2203, paragraphs (c) and (d) are revised to read as 
follows:


Sec.  20.2203  Reports of exposures, radiation levels, and 
concentrations of radioactive material exceeding the constraints or 
limits.

* * * * *
    (c) For holders of an operating license or a combined license for a 
nuclear power plant, the occurrences included in paragraph (a) of this 
section must be reported in accordance with the procedures described in 
Sec. Sec.  50.73(b), (c), (d), (e), and (g) of this chapter, and must 
include the information required by paragraph (b) of this section. 
Occurrences reported in accordance with Sec.  50.73 of this chapter 
need not be reported by a duplicate report under paragraph (a) of this 
section.
    (d) All licensees, other than those holding an operating license or 
a combined license for a nuclear power plant, who make reports under 
paragraph (a) of this section shall submit the report in writing either 
by mail addressed to the U.S. Nuclear Regulatory Commission, ATTN: 
Document Control Desk, Washington, DC 20555-0001; by hand delivery to 
the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, 
where practicable, by electronic submission, for example, Electronic 
Information Exchange, or CD-ROM. Electronic submissions must be made in 
a manner that enables the NRC to receive, read, authenticate, 
distribute, and archive the submission, and process and retrieve it a 
single page at a time. Detailed guidance on making electronic 
submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by 
e-mail to <a href="/cdn-cgi/l/email-protection#a9ece0ece9c7dbca87cec6df"><span class="__cf_email__" data-cfemail="bffaf6faffd1cddc91d8d0c9">[email protected]</span></a>, or by writing the Office of Information 
Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001. A copy should be sent to the appropriate NRC Regional Office 
listed in appendix D to this part.

PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE

0
54. 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); sec. 1704, 112 Stat. 2750 (44 
U.S.C. 3504 note).
    Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425, 
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).


0
55. 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 
applying for or holding a license or permit under 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;
    (2) Each individual, corporation, partnership, or other entity 
doing business within the United States, and each director and 
responsible officer of such an organization, that constructs a 
production or utilization facility licensed for manufacture, 
construction, or operation under parts 50 or 52 of this chapter, an 
ISFSI for the storage of spent fuel licensed under part 72 of this 
chapter, an MRS for the storage of spent fuel or high-level radioactive 
waste under 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;
    (3) Each individual, corporation, partnership, or other entity 
doing business within the United States, and each director and 
responsible officer of such an organization, applying for a design 
certification rule under part 52 of this chapter; or supplying basic 
components with respect to that design certification, and each 
individual, corporation, partnership, or other entity doing business 
within the United States, and each director and responsible officer of 
such an organization, whose application for design certification has 
been granted under part 52 of this chapter, or who has supplied or is 
supplying basic components with respect to that design certification;
    (4) Each individual, corporation, partnership, or other entity 
doing business within the United States, and each director and 
responsible officer of such an organization, applying for or holding a 
standard design approval under part 52 of this chapter; or supplying 
basic components with respect to a standard design approval under part 
52 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 under part 52 of this chapter (for the period of 
construction until the date that the Commission makes the finding under 
Sec.  52.103(g) of this chapter), or to manufacture a facility under 
part 52 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 these 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 these licensees to report defects under Section 206 of the 
Energy Reorganization Act of 1974.
* * * * *

0
56. In Sec.  21.3 the definitions of basic component, defect, 
deviation, and substantial safety hazard are revised to read as 
follows:


Sec.  21.3  Definitions.

* * * * *
    Basic component. (1)(i) When applied to nuclear power plants 
licensed under 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

[[Page 49487]]

    (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. Sec.  50.34(a)(1), 50.67(b)(2), or 100.11 
of this chapter, as applicable.
    (ii) Basic components are items designed and manufactured under a 
quality assurance program complying with appendix B to part 50 of this 
chapter, or commercial grade items which have successfully completed 
the dedication process.
    (2) When applied to standard design certifications under subpart C 
of part 52 of this chapter and standard design approvals under part 52 
of this chapter, basic component means the design or procurement 
information approved or to be approved within the scope of the design 
certification or approval for a structure, system, or component, or 
part thereof, that affects its safety function necessary to assure:
    (i) The integrity of the reactor coolant pressure boundary;
    (ii) The capability to shut down the reactor and maintain it in a 
safe-shutdown condition; or
    (iii) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. Sec.  50.34(a)(1), 50.67(b)(2), or 100.11 
of this chapter, as applicable.
    (3) When applied to other facilities and other activities licensed 
under 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.
    (4) 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, 
design certification, design approval, or information in support of an 
early site permit application under part 52 of this chapter, whether 
these services are performed by the component supplier or others.
* * * * *
    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;
    (2) The installation, use, or operation of a basic component 
containing a defect as defined in this section;
    (3) A deviation in a portion of a facility subject to the early 
site permit, standard design certification, standard design approval, 
construction permit, combined license 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;
    (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 under part 
50 or part 52 of this chapter; or
    (5) An error, omission or other circumstance in a design 
certification, or standard design approval that, on the basis of an 
evaluation, could create a substantial safety hazard.
    Deviation means a departure from the technical requirements 
included in a procurement document, or specified in early site permit 
information, a standard design certification or standard design 
approval.
* * * * *
    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 or otherwise approved or regulated by the NRC, other than for 
export, under parts 30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of this 
chapter.
* * * * *

0
57. Section 21.5 is revised to read as follows:


Sec.  21.5  Communications.

    Except where otherwise specified in this part, written 
communications and reports concerning the regulations in this part must 
be addressed to the NRC's Document Control Desk, and sent by mail to 
the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by 
hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, 
Maryland; or, where practicable, by electronic submission, for example, 
Electronic Information Exchange, or CD-ROM. Electronic submissions must 
be made in a manner that enables the NRC to receive, read, 
authenticate, distribute, and archive the submission, and process and 
retrieve it a single page at a time. Detailed guidance on making 
electronic submissions can be obtained by visiting the NRC's Web site 
at http://www.nrc.gov/site-help/eie.html, by calling (301) 415-6030, by 
e-mail to <a href="/cdn-cgi/l/email-protection#befbf7fbfed0ccdd90d9d1c8"><span class="__cf_email__" data-cfemail="97d2ded2d7f9e5f4b9f0f8e1">[email protected]</span></a>, or by writing the Office of Information 
Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001. The guidance discusses, among other topics, the formats the NRC 
can accept, the use of electronic signatures, and the treatment of 
nonpublic information. In the case of a licensee or permit holder, a 
copy of the communication must also be sent to the appropriate Regional 
Administrator at the address specified in appendix D to part 20 of this 
chapter.

0
58. In Sec.  21.21 the introductory text of paragraph (a)(3), paragraph 
(a)(3)(i), and paragraphs (d)(1)(i), (d)(1)(ii), and (d)(4)(vi) are 
revised and paragraph (d)(4)(ix) is added to read as follows:


Sec.  21.21  Notification of failure to comply or existence of a defect 
and its evaluation.

    (a) * * *
    (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 paragraphs (a)(1) or (a)(2) of this section if the 
manufacture, construction, or operation of a facility or activity, a 
basic component supplied for such facility or activity, or the design 
certification or design approval under part 52 of this chapter--
    (i) Fails to comply with the Atomic Energy Act of 1954, as amended, 
or any applicable rule, regulation, order, or license of the Commission 
or standard design approval under part 52 of this chapter, relating to 
a substantial safety hazard, or
* * * * *
    (d)(1) * * *
    (i) The manufacture, 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, design certification, 
or approval requirements under parts 30, 40, 50, 52, 60, 61, 63, 70, 
71, or 72 of this chapter.
* * * * *
    (4) * * *
    (vi) In the case of a basic component which contains a defect or 
fails to

[[Page 49488]]

comply, the number and location of these components in use at, supplied 
for, being supplied for, or may be supplied for, manufactured, or being 
manufactured for one or more facilities or activities subject to the 
regulations in this part.
* * * * *
    (ix) In the case of an early site permit, the entities to whom an 
early site permit was transferred.
* * * * *

0
59. In Sec.  21.51 paragraphs (a)(4) and (a)(5) are added and paragraph 
(b) is revised to read as follows:


Sec.  21.51  Maintenance and inspection of records.

    (a) * * *
    (4) Applicants for standard design certification under subpart B of 
part 52 of this chapter and others providing a design which is the 
subject of a design certification, during and following Commission 
adoption of a final design certification rule for that design, shall 
retain any notifications sent to purchasers and affected licensees for 
a minimum of 5 years after the date of the notification, and retain a 
record of the purchasers for 15 years after delivery of design which is 
the subject of the design certification rule or service associated with 
the design.
    (5) Applicants for or holders of a standard design approval under 
subpart E of part 52 of this chapter and others providing a design 
which is the subject of a design approval shall retain any 
notifications sent to purchasers and affected licensees for a minimum 
of 5 years after the date of the notification, and retain a record of 
the purchasers for 15 years after delivery of the design which is the 
subject of the design approval or service associated with the design.
    (b) Each individual, corporation, partnership, dedicating entity, 
or other entity subject to the regulations in this part shall permit 
the Commission the opportunity to inspect records pertaining to basic 
components that relate to the identification and evaluation of 
deviations, and the reporting of defects and failures to comply, 
including (but not limited to) any advice given to purchasers or 
licensees on the placement, erection, installation, operation, 
maintenance, modification, or inspection of a basic component.

0
60. In Sec.  21.61, paragraph (b) is revised to read as follows:


Sec.  21.61  Failure to notify.

* * * * *
    (b) Any NRC licensee or applicant for a license (including an 
applicant for, or holder of, a permit), applicant for a design 
certification under part 52 of this chapter during the pendency of its 
application, applicant for a design certification after Commission 
adoption of a final design certification rule for that design, or 
applicant for or holder of a standard design approval under part 52 of 
this chapter subject to the regulations in this part who fails to 
provide the notice required by Sec.  21.21, or otherwise fails to 
comply with the applicable requirements of this part shall be subject 
to a civil penalty as provided by Section 234 of the Atomic Energy Act 
of 1954, as amended.
* * * * *

PART 25--ACCESS AUTHORIZATION

0
61. The authority citation for part 25 continues to read as follows:

    Authority: Secs. 145, 161, 68 Stat. 942, 948, as amended (42 
U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C. 
5841); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); E.O. 10865, 
as amended, 3 CFR 1959-1963 Comp., p. 398 (50 U.S.C. 401, note); 
E.O. 12829, 3 CFR, 1993 Comp., p. 570; E.O. 12958, as amended, 3 
CFR, 1995 Comp., p. 333 as amended by E.O. 13292, 3 CFR 2004 Comp., 
p. 196; E.O. 12968, 3 CFR, 1995 Comp., p. 396.
    Appendix A also issued under 96 Stat. 1051 (31 U.S.C. 9701).


0
62. The heading of part 25 is revised to read as set forth above.

0
63. In Sec.  25.35, paragraph (a) is revised to read as follows:


Sec.  25.35  Classified visits.

    (a) The number of classified visits must be held to a minimum. The 
licensee, certificate holder, applicant for a standard design 
certification under part 52 of this chapter (including an applicant 
after the Commission has adopted a final standard design certification 
rule under part 52 of this chapter), or other facility, or an applicant 
for or holder of a standard design approval under part 52 of this 
chapter shall determine that the visit is necessary and that the 
purpose of the visit cannot be achieved without access to, or 
disclosure of, classified information. All classified visits require 
advance notification to, and approval of, the organization to be 
visited. In urgent cases, visit information may be furnished by 
telephone and confirmed in writing.
* * * * *

PART 26--FITNESS FOR DUTY PROGRAMS

0
64. The authority citation for part 26 continues to read as follows:

    Authority: Secs. 53, 81, 103, 104, 107, 161, 68 Stat. 930, 935, 
936, 937, 948, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 
U.S.C. 2073, 2111, 2112, 2133, 2134, 2137, 2201, 2297f); secs. 201, 
202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 
5842, 5846).


0
65. In Sec.  26.2, the introductory text of paragraph (a), and 
paragraph (c) are revised to read as follows:


Sec.  26.2  Scope.

    (a) The regulations in this part apply to licensees authorized to 
operate a nuclear power reactor, including a holder of a combined 
license after the Commission makes the finding under Sec.  52.103(g) of 
this chapter, and licensees who are authorized to possess or use 
formula quantities of SSNM, or to transport formula quantities of SSNM. 
Each licensee shall implement a fitness-for-duty program which complies 
with this part. The provisions of the fitness-for-duty program must 
apply to all persons granted unescorted access to nuclear power plant 
protected areas, to licensee, vendor, or contractor personnel required 
to physically report to a licensee's Technical Support Center (TSC) or 
Emergency Operations Facility (EOF) in accordance with licensee 
emergency plans and procedures, and to SSNM licensee and transporter 
personnel who:
* * * * *
    (c) Certain regulations in this part apply to licensees holding 
permits to construct a nuclear power plant, including a holder of a 
combined license before the date that the Commission makes the finding 
under Sec.  52.103(g) of this chapter, holders of manufacturing 
licenses under part 52, and persons authorized to conduct the 
activities under Sec.  50.10(e)(3) of this chapter. Each licensee with 
a construction permit, a combined license before the Commission makes 
the finding under Sec.  52.103(g) of this chapter, a manufacturing 
license, or person authorized to conduct the activities under Sec.  
50.10(e)(3) of this chapter, with a plant or reactor under active 
construction or manufacture, shall--
    (1) Comply with Sec. Sec.  26.10, 26.20, 26.23, 26.70, and 26.73;
    (2) Implement a chemical testing program, including random tests; 
and
    (3) Make provisions for employee assistance programs, imposition of 
sanctions, appeals procedures, the protection of information, and 
recordkeeping.
* * * * *

0
66. In Sec.  26.10, paragraph (a) is revised to read as follows:

[[Page 49489]]

Sec.  26.10  General performance objectives.

* * * * *
    (a) Provide reasonable assurance that nuclear power plant 
personnel, personnel of a holder of a manufacturing license, personnel 
of a person authorized to conduct activities under Sec.  50.10(e)(3) of 
this chapter, transporter personnel, and personnel of licensees 
authorized to possess or use formula quantities of SSNM, will perform 
their tasks in a reliable and trustworthy manner and are not under the 
influence of any substance, legal or illegal, or mentally or physically 
impaired from any cause, which in any way adversely affects their 
ability to safely and competently perform their duties;
* * * * *

0
67. In Appendix A of Part 26, paragraph (1) of Section 1.1 of Subpart A 
is revised to read as follows:

Appendix A to Part 26--Guidelines for Drug and Alcohol Testing Programs

    1.1 Applicability.
    (1) These guidelines apply to licensees authorized to operate 
nuclear power reactors, including a holder of a combined license 
after the Commission makes the finding under Sec.  52.103(g) of this 
chapter, and licensees who are authorized to possess, use, or 
transport formula quantities of strategic special nuclear material 
(SSNM).
* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
68. 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, 937, 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, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
    Section 50.7 also issued under Pub. L. 95--601, sec. 10, 92 
Stat. 2951 (42 U.S.C. 5841). Section 50.10 also issued under secs. 
101, 185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, 
Pub. L. 91--190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 
50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as 
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 
50.33a, 50.55a and appendix Q also issued under sec. 102, Pub. L. 
91--190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 
also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 
50.58, 50.91, and 50.92 also issued under Pub. L. 97--415, 96 Stat. 
2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 
Stat. 939 (42 U.S.C. 2152). Sections 50.80--50.81 also issued under 
sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also 
issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).


0
69. In Section 50.2, definitions of applicant, license, licensee, and 
prototype plant, are added to read as follows:


Sec.  50.2  Definitions.

* * * * *
    Applicant means a person or an entity applying for a license, 
permit, or other form of Commission permission or approval under this 
part or part 52 of this chapter.
* * * * *
    License means a license, including a construction permit or 
operating license under this part, an early site permit, combined 
license or manufacturing license under part 52 of this chapter, or a 
renewed license issued by the Commission under this part, part 52, or 
part 54 of this chapter.
    Licensee means a person who is authorized to conduct activities 
under a license issued by the Commission.
* * * * *
    Prototype plant means a nuclear reactor that is used to test design 
features, such as the testing required under Sec.  50.43(e). The 
prototype plant is similar to a first-of-a-kind or standard plant 
design in all features and size, but may include additional safety 
features to protect the public and the plant staff from the possible 
consequences of accidents during the testing period.
* * * * *

0
70. In Sec.  50.10 the introductory text of paragraphs (b) and (c), and 
paragraphs (e)(1), (e)(2), and (e)(3) are revised to read as follows:


Sec.  50.10  License required.

* * * * *
    (b) No person shall begin the construction of a production or 
utilization facility on a site on which the facility is to be operated 
until either a construction permit under this part, or a combined 
license under subpart C of part 52 of this chapter has been issued. As 
used in this paragraph, the term ``construction'' includes pouring the 
foundation for, or the installation of, any portion of the permanent 
facility on the site, but does not include:
* * * * *
    (c) Notwithstanding the provisions of paragraph (b) of this 
section, and subject to paragraphs (d) and (e) of this section, no 
person shall effect commencement of construction of a production or 
utilization facility subject to the provisions of Sec.  51.20(b) of 
this chapter on a site on which the facility is to be operated until an 
early site permit, construction permit, or combined license has been 
issued. As used in this paragraph, the term ``commencement of 
construction'' means any clearing of land, excavation or other 
substantial action that would adversely affect the environment of a 
site, but does not include:
* * * * *
    (e)(1) The Director of Nuclear Reactor Regulation may authorize an 
applicant for a construction permit for a utilization facility which is 
subject to Sec.  51.20(b) of this chapter, and is of the type specified 
in Sec. Sec.  50.21(b)(2) or (3), or Sec.  50.22 or is a testing 
facility, or an applicant for a combined license to conduct the 
following activities:
    (i) Preparation of the site for construction of the facility 
(including activities as clearing, grading, construction of temporary 
access roads and borrow areas);
    (ii) Installation of temporary construction support facilities 
(including items such as warehouse and shop facilities, utilities, 
concrete mixing plants, docking and unloading facilities, and 
construction support buildings);
    (iii) Excavation for facility structures;
    (iv) Construction of service facilities (including facilities such 
as roadways, paving, railroad spurs, fencing, exterior utility and 
lighting systems, transmission lines, and sanitary sewerage treatment 
facilities); and
    (v) The construction of structures, systems and components which do 
not prevent or mitigate the consequences of postulated accidents that 
could cause undue risk to the health and safety of the public.
    (2) No authorization shall be granted unless the staff has 
completed a final environmental impact statement on the issuance of the 
construction permit or combined license as required by subpart A of 
part 51 of this chapter. An authorization shall be granted only after 
the presiding officer in the proceeding on the construction permit or 
combined license application:
    (i) Has made all the findings required by Sec. Sec.  51.104(b), 
51.105, and 51.107 of this chapter to be made before issuance of the 
construction permit, or combined license for the facility; and
    (ii) Has determined that, based upon the available information and 
review to date, there is reasonable assurance that the proposed site is 
a suitable location for a reactor of the general size and type proposed 
from the standpoint of radiological health and safety considerations 
under the Act and regulations issued by the Commission.
    (3)(i) The Director of New Reactors or the Director of Nuclear 
Reactor

[[Page 49490]]

Regulation, as appropriate, may authorize an applicant for a 
construction permit for a utilization facility which is subject to 
Sec.  51.20(b) of this chapter, and is of the type specified in 
Sec. Sec.  50.21(b)(2) or (3), or Sec.  50.22 or is a testing facility, 
or an applicant for a combined license to conduct, in addition to the 
activities described in paragraph (e)(1) of this section, the 
installation of structural foundations, including any necessary 
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.
    (ii) Such an authorization, which may be combined with the 
authorization described in paragraph (e)(1) of this section, or may be 
granted at a later time, shall be granted only after the presiding 
officer in the proceeding on the construction permit or combined 
license application has, in addition to making the findings and 
determinations required by paragraph (e)(2) of this section, determined 
that there are no unresolved safety issues relating to the additional 
activities that may be authorized under this paragraph that would 
constitute good cause for withholding authorization.
* * * * *

0
71. Section 50.23 is revised to read as follows:


Sec.  50.23  Construction permits.

    A construction permit for the construction of a production or 
utilization facility will be issued before the issuance of a license if 
the application is otherwise acceptable, and will be converted upon 
completion of the facility and Commission action, into a license as 
provided in Sec.  50.56. However, if a combined license for a nuclear 
power reactor is issued under part 52 of this chapter, the construction 
permit and operating license are deemed to be combined in a single 
license. A construction permit for the alteration of a production or 
utilization facility will be issued before the issuance of an amendment 
of a license, if the application for amendment is otherwise acceptable, 
as provided in Sec.  50.91.

0
72. The undesignated center heading before Sec.  50.30 is revised to 
read as follows:
Applications for Licenses, Certifications, and Regulatory Approvals; 
Form; Contents; Ineligibility of Certain Applicants

0
73. In Sec.  50.30, the section heading and paragraphs (a)(1), (a)(3), 
(a)(5), (a)(6), (b), (e), and (f) are revised to read as follows:


Sec.  50.30  Filing of application; oath or affirmation.

    (a) * * *
    (1) Each filing of an application for a standard design approval or 
license to construct and/or operate, or manufacture, a production or 
utilization facility (including an early site permit, combined license, 
and manufacturing license under part 52 of this chapter), and any 
amendments to the applications, must be submitted to the U.S. Nuclear 
Regulatory Commission in accordance with Sec.  50.4 or Sec.  52.3 of 
this chapter, as applicable.
* * * * *
    (3) Each applicant for a construction permit under this part, or an 
early site permit, combined license, or manufacturing license under 
part 52 of this chapter, shall, upon notification by the Atomic Safety 
and Licensing Board appointed to conduct the public hearing required by 
the Atomic Energy Act, update the application and serve the updated 
copies of the application or parts of it, eliminating all superseded 
information, together with an index of the updated application, as 
directed by the Atomic Safety and Licensing Board. Any subsequent 
amendment to the application must be served on those served copies of 
the application and must be submitted to the U.S. Nuclear Regulatory 
Commission as specified in Sec.  50.4 or Sec.  52.3 of this chapter, as 
applicable.
* * * * *
    (5) At the time of filing an application, the Commission will make 
available at the NRC Web site, http://www.nrc.gov, a copy of the 
application, subsequent amendments, and other records pertinent to the 
matter which is the subject of the application for public inspection 
and copying.
    (6) The serving of copies required by this section must not occur 
until the application has been docketed under Sec.  2.101(a) of this 
chapter. Copies must be submitted to the Commission, as specified in 
Sec.  50.4 or Sec.  52.3 of this chapter, as applicable, to enable the 
Director, Office of New Reactors, or the Director, Office of Nuclear 
Reactor Regulation, or the Director, Office of Nuclear Material Safety 
and Safeguards, as appropriate, to determine whether the application is 
sufficiently complete to permit docketing.
    (b) Oath or affirmation. Each application for a standard design 
approval or license, including, whenever appropriate, a construction 
permit or early site permit, or amendment of it, and each amendment of 
each application must be executed in a signed original by the applicant 
or duly authorized officer thereof under oath or affirmation.
* * * * *
    (e) Filing Fees. Each application for a standard design approval or 
production or utilization facility license, including, whenever 
appropriate, a construction permit or early site permit, other than a 
license exempted from part 170 of this chapter, shall be accompanied by 
the fee prescribed in part 170 of this chapter. No fee will be required 
to accompany an application for renewal, amendment, or termination of a 
construction permit, operating license, combined license, or 
manufacturing license, except as provided in Sec.  170.21 of this 
chapter.
    (f) Environmental report. An application for a construction permit, 
operating license, early site permit, combined license, or 
manufacturing license for a nuclear power reactor, testing facility, 
fuel reprocessing plant, or other production or utilization facility 
whose construction or operation may be determined by the Commission to 
have a significant impact in the environment, shall be accompanied by 
an Environmental Report required under subpart A of part 51 of this 
chapter.

0
74. In Sec.  50.33, paragraphs (f)(3) and (f)(4) are redesignated as 
(f)(4)and (f)(5), respectively, and are revised, a new paragraph (f)(3) 
is added, and paragraphs (g), (h), and (k)(1) are revised to read as 
follows:


Sec.  50.33  Contents of applications; general information.

* * * * *
    (f) * * *
    (3) If the application is for a combined license under subpart C of 
part 52 of this chapter, the applicant shall submit the information 
described in paragraphs (f)(1) and (f)(2) of this section.
    (4) Each application for a construction permit, operating license, 
or combined license submitted by a newly-formed entity organized for 
the primary purpose of constructing and/or operating a facility must 
also include information showing:
    (i) The legal and financial relationships it has or proposes to 
have with its stockholders or owners;
    (ii) The stockholders' or owners' financial ability to meet any 
contractual obligation to the entity which they have incurred or 
proposed to incur; and

[[Page 49491]]

    (iii) Any other information considered necessary by the Commission 
to enable it to determine the applicant's financial qualification.
    (5) The Commission may request an established entity or newly-
formed entity to submit additional or more detailed information 
respecting its financial arrangements and status of funds if the 
Commission considers this information appropriate. This may include 
information regarding a licensee's ability to continue the conduct of 
the activities authorized by the license and to decommission the 
facility.
    (g) If the application is for an operating license or combined 
license for a nuclear power reactor, or if the application is for an 
early site permit and contains plans for coping with emergencies under 
Sec.  52.17(b)(2)(ii) of this chapter, the applicant shall submit 
radiological emergency response plans of State and local governmental 
entities in the United States that are wholly or partially within the 
plume exposure pathway emergency planning zone (EPZ),\4\ as well as the 
plans of State governments wholly or partially within the ingestion 
pathway EPZ.\5\ If the application is for an early site permit that, 
under 10 CFR 52.17(b)(2)(i), proposes major features of the emergency 
plans describing the EPZs, then the descriptions of the EPZs must meet 
the requirements of this paragraph. Generally, the plume exposure 
pathway EPZ for nuclear power reactors shall consist of an area about 
10 miles (16 km) in radius and the ingestion pathway EPZ shall consist 
of an area about 50 miles (80 km) in radius. The exact size and 
configuration of the EPZs surrounding a particular nuclear power 
reactor shall be determined in relation to the local emergency response 
needs and capabilities as they are affected by such conditions as 
demography, topography, land characteristics, access routes, and 
jurisdictional boundaries. The size of the EPZs also may be determined 
on a case-by-case basis for gas-cooled reactors and for reactors with 
an authorized power level less than 250 MW thermal. The plans for the 
ingestion pathway shall focus on such actions as are appropriate to 
protect the food ingestion pathway.
---------------------------------------------------------------------------

    \4\ Emergency planning zones (EPZs) are discussed in NUREG-0396, 
EPA 520/1-78-016, ``Planning Basis for the Development of State and 
Local Government Radiological Emergency Response Plans in Support of 
Light-Water Nuclear Power Plants,'' December 1978.
    \5\ If the State and local emergency response plans have been 
previously provided to the NRC for inclusion in the facility docket, 
the applicant need only provide the appropriate reference to meet 
this requirement.
---------------------------------------------------------------------------

    (h) If the applicant, other than an applicant for a combined 
license, proposes to construct or alter a production or utilization 
facility, the application shall state the earliest and latest dates for 
completion of the construction or alteration.
* * * * *
    (k)(1) For an application for an operating license or combined 
license for a production or utilization facility, information in the 
form of a report, as described in Sec.  50.75, indicating how 
reasonable assurance will be provided that funds will be available to 
decommission the facility.
* * * * *

0
75. In Sec.  50.34, the section heading, the introductory text of 
paragraph (a)(1), paragraphs (a)(1)(ii)(E) and (a)(12), the 
introductory text of paragraph (b), paragraphs (b)(10) and (b)(11), and 
paragraphs (c), (d), and (e), the introductory text of paragraphs (f) 
and(f)(1), and paragraphs (g), and (h)(1)(ii) are revised to read as 
follows:


Sec.  50.34  Contents of construction permit and operating license 
applications; technical information.

    (a) * * *
    (1) Stationary power reactor applicants for a construction permit 
who apply on or after January 10, 1997, shall comply with paragraph 
(a)(1)(ii) of this section. All other applicants for a construction 
permit shall comply with paragraph (a)(1)(i) of this section.
* * * * *
    (ii) * * *
    (E) With respect to operation at the projected initial power level, 
the applicant is required to submit information prescribed in 
paragraphs (a)(2) through (a)(8) of this section, as well as the 
information required by paragraph (a)(1)(i) of this section, in support 
of the application for a construction permit.
* * * * *
    (12) On or after January 10, 1997, stationary power reactor 
applicants who apply for a construction permit, as partial conformance 
to General Design Criterion 2 of appendix A to this part, shall comply 
with the earthquake engineering criteria in appendix S to this part.
    (b) Final safety analysis report. Each application for an operating 
license shall include a final safety analysis report. The final safety 
analysis report shall include information that describes the facility, 
presents the design bases and the limits on its operation, and presents 
a safety analysis of the structures, systems, and components and of the 
facility as a whole, and shall include the following:
* * * * *
    (10) On or after January 10, 1997, stationary power reactor 
applicants who apply for an operating license, as partial conformance 
to General Design Criterion 2 of appendix A to this part, shall comply 
with the earthquake engineering criteria of appendix S to this part. 
However, for those operating license applicants and holders whose 
construction permit was issued before January 10, 1997, the earthquake 
engineering criteria in Section VI of appendix A to part 100 of this 
chapter continues to apply.
    (11) On or after January 10, 1997, stationary power reactor 
applicants who apply for an operating license, shall provide a 
description and safety assessment of the site and of the facility as in 
Sec.  50.34(a)(1)(ii). However, for either an operating license 
applicant or holder whose construction permit was issued before January 
10, 1997, the reactor site criteria in part 100 of this chapter and the 
seismic and geologic siting criteria in appendix A to part 100 of this 
chapter continues to apply.
    (c) Physical Security Plan. Each application for an operating 
license for a production or utilization facility must include a 
physical security plan. The plan must describe how the applicant will 
meet the requirements of part 73 of this chapter (and part 11 of this 
chapter, if applicable, including the identification and description of 
jobs as required by Sec.  11.11(a) of this chapter, at the proposed 
facility). The plan must list tests, inspections, audits, and other 
means to be used to demonstrate compliance with the requirements of 10 
CFR parts 11 and 73, if applicable.
    (d) Safeguards contingency plan. Each application for an operating 
license for a production or utilization facility that will be subject 
to Sec. Sec.  73.50, 73.55, or Sec.  73.60 of this chapter, must 
include a licensee safeguards contingency plan in accordance with the 
criteria set forth in appendix C to 10 CFR part 73. The safeguards 
contingency plan shall include plans for dealing with threats, thefts, 
and radiological sabotage, as defined in part 73 of this chapter, 
relating to the special nuclear material and nuclear facilities 
licensed under this chapter and in the applicant's possession and 
control. Each application for such a license shall include the first 
four categories of information contained in the applicant's safeguards 
contingency plan. (The first four categories of information as set 
forth in appendix C to 10 CFR part 73

[[Page 49492]]

of this chapter are Background, Generic Planning Base, Licensee 
Planning Base, and Responsibility Matrix. The fifth category of 
information, Procedures, does not have to be submitted for approval.) 
\9\
---------------------------------------------------------------------------

    \9\ A physical security plan that contains all the information 
required in both Sec.  73.55 and appendix C to part 73 of this 
chapter satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------

    (e) Protection against unauthorized disclosure. Each applicant for 
an operating license for a production or utilization facility, who 
prepares a physical security plan, a safeguards contingency plan, or a 
guard qualification and training plan, shall protect the plans and 
other related safeguards information against unauthorized disclosure in 
accordance with the requirements of Sec.  73.21 of this chapter, as 
appropriate.
    (f) Additional TMI-related requirements. In addition to the 
requirements of paragraph (a) of this section, each applicant for a 
light-water-reactor construction permit or manufacturing license whose 
application was pending as of February 16, 1982, shall meet the 
requirements in paragraphs (f)(1) through (3) of this section. This 
regulation applies to the pending applications by Duke Power Company 
(Perkins Nuclear Station, Units 1, 2, and 3), Houston Lighting & Power 
Company (Allens Creek Nuclear Generating Station, Unit 1), Portland 
General Electric Company (Pebble Springs Nuclear Plant, Units 1 and 2), 
Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2), 
Puget Sound Power & Light Company (Skagit/Hanford Nuclear Power 
Project, Units 1 and 2), and Offshore Power Systems (License to 
Manufacture Floating Nuclear Plants). The number of units that will be 
specified in the manufacturing license above, if issued, will be that 
number whose start of manufacture, as defined in the license 
application, can practically begin within a 10-year period commencing 
on the date of issuance of the manufacturing license, but in no event 
will that number be in excess of ten. The manufacturing license will 
require the plant design to be updated no later than 5 years after its 
approval. Paragraphs (f)(1)(xii), (2)(ix), and (3)(v) of this section, 
pertaining to hydrogen control measures, must be met by all applicants 
covered by this regulation. However, the Commission may decide to 
impose additional requirements and the issue of whether compliance with 
these provisions, together with 10 CFR 50.44 and criterion 50 of 
appendix A to 10 CFR part 50, is sufficient for issuance of that 
manufacturing license which may be considered in the manufacturing 
license proceeding. In addition, each applicant for a design 
certification, design approval, combined license, or manufacturing 
license under part 52 of this chapter shall demonstrate compliance with 
the technically relevant portions of the requirements in paragraphs 
(f)(1) through (3) of this section, except for paragraphs (f)(1)(xii), 
(f)(2)(ix), and (f)(3)(v).
    (1) To satisfy the following requirements, the application shall 
provide sufficient information to describe the nature of the studies, 
how they are to be conducted, estimated submittal dates, and a program 
to ensure that the results of these studies are factored into the final 
design of the facility. For licensees identified in the introduction to 
paragraph (f) of this section, all studies must be completed no later 
than 2 years following the issuance of the construction permit or 
manufacturing license.\10\ For all other applicants, the studies must 
be submitted as part of the final safety analysis report.
---------------------------------------------------------------------------

    \10\ Alphanumeric designations correspond to the related action 
plan items in NUREG 0718 and NUREG-0660, ``NRC Action Plan Developed 
as a Result of the TMI-2 Accident.'' They are provided herein for 
information only.
---------------------------------------------------------------------------

* * * * *
    (g) Combustible gas control. All applicants for a reactor 
construction permit or operating license whose application is submitted 
after October 16, 2003, shall include the analyses, and the 
descriptions of the equipment and systems required by Sec.  50.44 as a 
part of their application.
    (h) * * *
    (1) * * *
    (ii) Applications for light-water-cooled nuclear power plant 
construction permits docketed after May 17, 1982, shall include an 
evaluation of the facility against the SRP in effect on May 17, 1982, 
or the SRP revision in effect six months before the docket date of the 
application, whichever is later.
* * * * *

0
76. Section 50.34a is revised to read as follows:


Sec.  50.34a  Design objectives for equipment to control releases of 
radioactive material in effluents--nuclear power reactors.

    (a) An application for a construction permit shall include a 
description of the preliminary design of equipment to be installed to 
maintain control over radioactive materials in gaseous and liquid 
effluents produced during normal reactor operations, including expected 
operational occurrences. In the case of an application filed on or 
after January 2, 1971, the application shall also identify the design 
objectives, and the means to be employed, for keeping levels of 
radioactive material in effluents to unrestricted areas as low as is 
reasonably achievable. The term ``as low as is reasonably achievable'' 
as used in this part means as low as is reasonably achievable taking 
into account the state of technology, and the economics of improvements 
in relation to benefits to the public health and safety and other 
societal and socioeconomic considerations, and in relation to the use 
of atomic energy in the public interest. The guides set out in appendix 
I to this part provide numerical guidance on design objectives for 
light-water-cooled nuclear power reactors to meet the requirements that 
radioactive material in effluents released to unrestricted areas be 
kept as low as is reasonably achievable. These numerical guides for 
design objectives and limiting conditions for operation are not to be 
construed as radiation protection standards.
    (b) Each application for a construction permit shall include:
    (1) A description of the preliminary design of equipment to be 
installed under paragraph (a) of this section;
    (2) An estimate of:
    (i) The quantity of each of the principal radionuclides expected to 
be released annually to unrestricted areas in liquid effluents produced 
during normal reactor operations; and
    (ii) The quantity of each of the principal radionuclides of the 
gases, halides, and particulates expected to be released annually to 
unrestricted areas in gaseous effluents produced during normal reactor 
operations.
    (3) A general description of the provisions for packaging, storage, 
and shipment offsite of solid waste containing radioactive materials 
resulting from treatment of gaseous and liquid effluents and from other 
sources.
    (c) Each application for an operating license shall include:
    (1) A description of the equipment and procedures for the control 
of gaseous and liquid effluents and for the maintenance and use of 
equipment installed in radioactive waste systems, under paragraph (a) 
of this section; and
    (2) A revised estimate of the information required in paragraph 
(b)(2) of this section if the expected releases and exposures differ 
significantly from the estimates submitted in the application for a 
construction permit.
    (d) Each application for a combined license under part 52 of this 
chapter shall include:

[[Page 49493]]

    (1) A description of the equipment and procedures for the control 
of gaseous and liquid effluents and for the maintenance and use of 
equipment installed in radioactive waste systems, under paragraph (a) 
of this section; and
    (2) The information required in paragraph (b)(2) of this section.
    (e) Each application for a design approval, a design certification, 
or a manufacturing license under part 52 of this chapter shall include:
    (1) A description of the equipment for the control of gaseous and 
liquid effluents and for the maintenance and use of equipment installed 
in radioactive waste systems, under paragraph (a) of this section; and
    (2) The information required in paragraph (b)(2) of this section.

0
77. In Sec.  50.36, paragraphs (c), (d), and (e) are redesignated as 
paragraphs (d), (e), and (f), respectively, and a new paragraph (c) is 
added to read as follows:


Sec.  50.36  Technical specifications.

* * * * *
    (c) Each applicant for a design certification or manufacturing 
license under part 52 of this chapter shall include in its application 
proposed generic technical specifications in accordance with the 
requirements of this section for the portion of the plant that is 
within the scope of the design certification or manufacturing license 
application.
* * * * *

0
78. In Sec.  50.36a, paragraph (a) is revised to read as follows:


Sec.  50.36a  Technical specifications on effluents from nuclear power 
reactors.

    (a) To keep releases of radioactive materials to unrestricted areas 
during normal conditions, including expected occurrences, as low as is 
reasonably achievable, each licensee of a nuclear power reactor and 
each applicant for a design certification or a manufacturing license 
will include technical specifications that, in addition to requiring 
compliance with applicable provisions of Sec.  20.1301 of this chapter, 
require that:
    (1) Operating procedures developed pursuant to Sec.  50.34a(c) for 
the control of effluents be established and followed and that the 
radioactive waste system, pursuant to Sec.  50.34a, be maintained and 
used. The licensee shall retain the operating procedures in effect as a 
record until the Commission terminates the license and shall retain 
each superseded revision of the procedures for 3 years from the date it 
was superseded.
    (2) Each holder of an operating license, and each holder of a 
combined license after the Commission has made the finding under Sec.  
52.103(g) of this chapter, shall submit a report to the Commission 
annually that specifies the quantity of each of the principal 
radionuclides released to unrestricted areas in liquid and in gaseous 
effluents during the previous 12 months, including any other 
information as may be required by the Commission to estimate maximum 
potential annual radiation doses to the public resulting from effluent 
releases. The report must be submitted as specified in Sec.  50.4, and 
the time between submission of the reports must be no longer than 12 
months. If quantities of radioactive materials released during the 
reporting period are significantly above design objectives, the report 
must cover this specifically. On the basis of these reports and any 
additional information the Commission may obtain from the licensee or 
others, the Commission may require the licensee to take action as the 
Commission deems appropriate.
* * * * *

0
79. Section 50.36b is revised to read as follows:


Sec.  50.36b  Environmental conditions.

    (a) Each construction permit under this part, each early site 
permit under part 52 of this chapter, and each combined license under 
part 52 of this chapter may include conditions to protect the 
environment during construction. These conditions are to be set out in 
an attachment to the permit or license, which is incorporated in and 
made a part of the permit or license. These conditions will be derived 
from information contained in the environmental report submitted 
pursuant to Sec.  51.50 of this chapter as analyzed and evaluated in 
the NRC record of decision, and will identify the obligations of the 
licensee in the environmental area, including, as appropriate, 
requirements for reporting and keeping records of environmental data, 
and any conditions and monitoring requirement for the protection of the 
nonaquatic environment.
    (b) Each license authorizing operation of a production or 
utilization facility, including a combined license under part 52 of 
this chapter, and each license for a nuclear power reactor facility for 
which the certification of permanent cessation of operations required 
under Sec.  50.82(a)(1) or Sec.  52.110(a) of this chapter has been 
submitted, which is of a type described in Sec.  50.21(b)(2) or (3) or 
Sec.  50.22 or is a testing facility, may include conditions to protect 
the environment during operation and decommissioning. These conditions 
are to be set out in an attachment to the license which is incorporated 
in and made a part of the license. These conditions will be derived 
from information contained in the environmental report or the 
supplement to the environmental report submitted pursuant to Sec. Sec.  
51.50 and 51.53 of this chapter as analyzed and evaluated in the NRC 
record of decision, and will identify the obligations of the licensee 
in the environmental area, including, as appropriate, requirements for 
reporting and keeping records of environmental data, and any conditions 
and monitoring requirement for the protection of the nonaquatic 
environment.

0
80. Section 50.37 is revised to read as follows:


Sec.  50.37  Agreement limiting access to Classified Information.

    As part of its application and in any event before the receipt of 
Restricted Data or classified National Security Information or the 
issuance of a license, construction permit, early site permit, or 
standard design approval, or before the Commission has adopted a final 
standard design certification rule under part 52 of this chapter, the 
applicant shall agree in writing that it will not permit any individual 
to have access to any facility to possess Restricted Data or classified 
National Security Information until the individual and/or facility has 
been approved for access under the provisions of 10 CFR parts 25 and/or 
95. The agreement of the applicant becomes part of the license, or 
construction permit, or standard design approval.

0
81. The undesignated center heading before Sec.  50.40 is revised to 
read as follows:

Standards for Licenses, Certifications, and Regulatory Approvals

0
82. Section 50.40 is revised to read as follows:


Sec.  50.40  Common standards.

    In determining that a construction permit or operating license in 
this part, or early site permit, combined license, or manufacturing 
license in part 52 of this chapter will be issued to an applicant, the 
Commission will be guided by the following considerations:
    (a) Except for an early site permit or manufacturing license, the 
processes to be performed, the operating procedures, the facility and 
equipment, the use of the facility, and other technical specifications, 
or the proposals, in regard to any of the foregoing

[[Page 49494]]

collectively provide reasonable assurance that the applicant will 
comply with the regulations in this chapter, including the regulations 
in part 20 of this chapter, and that the health and safety of the 
public will not be endangered.
    (b) The applicant for a construction permit, operating license, 
combined license, or manufacturing license is technically and 
financially qualified to engage in the proposed activities in 
accordance with the regulations in this chapter. However, no 
consideration of financial qualification is necessary for an electric 
utility applicant for an operating license for a utilization facility 
of the type described in Sec.  50.21(b) or Sec.  50.22 or for an 
applicant for a manufacturing license.
    (c) The issuance of a construction permit, operating license, early 
site permit, combined license, or manufacturing license to the 
applicant will not, in the opinion of the Commission, be inimical to 
the common defense and security or to the health and safety of the 
public.
    (d) Any applicable requirements of subpart A of 10 CFR part 51 have 
been satisfied.

0
83. In Sec.  50.43, the section heading, the introductory paragraph, 
and paragraph (d) are revised, and paragraph (e) is added to read as 
follows:


Sec.  50.43  Additional standards and provisions affecting class 103 
licenses and certifications for commercial power.

    In addition to applying the standards set forth in Sec. Sec.  50.40 
and 50.42, paragraphs (a) through (e) of this section apply in the case 
of a class 103 license for a facility for the generation of commercial 
power. For a design certification under part 52 of this chapter, only 
paragraph (e) of this section applies.
* * * * *
    (d) Nothing shall preclude any government agency, now or hereafter 
authorized by law to engage in the production, marketing, or 
distribution of electric energy, if otherwise qualified, from obtaining 
a construction permit or operating license under this part, or a 
combined license under part 52 of this chapter for a utilization 
facility for the primary purpose of producing electric energy for 
disposition for ultimate public consumption.
    (e) Applications for a design certification, combined license, 
manufacturing license, or operating license that propose nuclear 
reactor designs which differ significantly from light-water reactor 
designs that were licensed before 1997, or use simplified, inherent, 
passive, or other innovative means to accomplish their safety 
functions, will be approved only if:
    (1)(i) The performance of each safety feature of the design has 
been demonstrated through either analysis, appropriate test programs, 
experience, or a combination thereof;
    (ii) Interdependent effects among the safety features of the design 
are acceptable, as demonstrated by analysis, appropriate test programs, 
experience, or a combination thereof; and
    (iii) 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; or
    (2) 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 a prototype plant is used to comply with the testing 
requirements, then the NRC may impose additional requirements on 
siting, safety features, or operational conditions for the prototype 
plant to protect the public and the plant staff from the possible 
consequences of accidents during the testing period.

0
84. Section 50.45 is revised to read as follows:


Sec.  50.45  Standards for construction permits, operating licenses, 
and combined licenses.

    (a) An applicant for an operating license or an amendment of an 
operating license who proposes to construct or alter a production or 
utilization facility will be initially granted a construction permit if 
the application is in conformity with and acceptable under the criteria 
of Sec. Sec.  50.31 through 50.38, and the standards of Sec. Sec.  
50.40 through 50.43, as applicable.
    (b) A holder of a combined license who proposes, after the 
Commission makes the finding under Sec.  52.103(g) of this chapter, to 
alter the licensed facility will be initially granted a construction 
permit if the application is in conformity with and acceptable under 
the criteria of Sec. Sec.  50.30 through 50.33, Sec.  50.34(f), 
Sec. Sec.  50.34a through 50.38, the standards of Sec. Sec.  50.40 
through 50.43, as applicable, and Sec. Sec.  52.79 and 52.80 of this 
chapter.

0
85. In Sec.  50.46, paragraph (a)(3) is revised to read as follows:


Sec.  50.46  Acceptance criteria for emergency core cooling systems for 
light-water nuclear power reactors.

    (a) * * *
    (3)(i) Each applicant for or holder of an operating license or 
construction permit issued under this part, applicant for a standard 
design certification under part 52 of this chapter (including an 
applicant after the Commission has adopted a final design certification 
regulation), or an applicant for or holder of a standard design 
approval, a combined license or a manufacturing license issued under 
part 52 of this chapter, shall estimate the effect of any change to or 
error in an acceptable evaluation model or in the application of such a 
model to determine if the change or error is significant. For this 
purpose, a significant change or error is one which results in a 
calculated peak fuel cladding temperature different by more than 50 
[deg]F from the temperature calculated for the limiting transient using 
the last acceptable model, or is a cumulation of changes and errors 
such that the sum of the absolute magnitudes of the respective 
temperature changes is greater than 50 [deg]F.
    (ii) For each change to or error discovered in an acceptable 
evaluation model or in the application of such a model that affects the 
temperature calculation, the applicant or holder of a construction 
permit, operating license, combined license, or manufacturing license 
shall report the nature of the change or error and its estimated effect 
on the limiting ECCS analysis to the Commission at least annually as 
specified in Sec.  50.4 or Sec.  52.3 of this chapter, as applicable. 
If the change or error is significant, the applicant or licensee shall 
provide this report within 30 days and include with the report a 
proposed schedule for providing a reanalysis or taking other action as 
may be needed to show compliance with Sec.  50.46 requirements. This 
schedule may be developed using an integrated scheduling system 
previously approved for the facility by the NRC. For those facilities 
not using an NRC approved integrated scheduling system, a schedule will 
be established by the NRC staff within 60 days of receipt of the 
proposed schedule. Any change or error correction that results in a 
calculated ECCS performance that does not conform to the criteria set 
forth in paragraph (b) of this section is a reportable event as 
described in Sec. Sec.  50.55(e), 50.72, and 50.73. The affected 
applicant or licensee shall propose immediate steps to demonstrate 
compliance or bring plant design or operation into compliance with 
Sec.  50.46 requirements.
    (iii) For each change to or error discovered in an acceptable 
evaluation model or in the application of such a model that affects the 
temperature

[[Page 49495]]

calculation, the applicant or holder of a standard design approval or 
the applicant for a standard design certification (including an 
applicant after the Commission has adopted a final design certification 
rule) shall report the nature of the change or error and its estimated 
effect on the limiting ECCS analysis to the Commission and to any 
applicant or licensee referencing the design approval or design 
certification at least annually as specified in Sec.  52.3 of this 
chapter. If the change or error is significant, the applicant or holder 
of the design approval or the applicant for the design certification 
shall provide this report within 30 days and include with the report a 
proposed schedule for providing a reanalysis or taking other action as 
may be needed to show compliance with Sec.  50.46 requirements. The 
affected applicant or holder shall propose immediate steps to 
demonstrate compliance or bring plant design into compliance with Sec.  
50.46 requirements.
* * * * *

0
86. In Sec.  50.47, paragraph (a)(1) is revised and paragraph (e) is 
added to read as follows:


Sec.  50.47  Emergency plans.

    (a)(1)(i) Except as provided in paragraph (d) of this section, no 
initial operating license for a nuclear power reactor will be issued 
unless a finding is made by the NRC that there is reasonable assurance 
that adequate protective measures can and will be taken in the event of 
a radiological emergency. No finding under this section is necessary 
for issuance of a renewed nuclear power reactor operating license.
    (ii) No initial combined license under part 52 of this chapter will 
be issued unless a finding is made by the NRC that there is reasonable 
assurance that adequate protective measures can and will be taken in 
the event of a radiological emergency. No finding under this section is 
necessary for issuance of a renewed combined license.
    (iii) If an application for an early site permit under subpart A of 
part 52 of this chapter includes complete and integrated emergency 
plans under 10 CFR 52.17(b)(2)(ii), no early site permit will be issued 
unless a finding is made by the NRC that the emergency plans provide 
reasonable assurance that adequate protective measures can and will be 
taken in the event of a radiological emergency.
    (iv) If an application for an early site permit proposes major 
features of the emergency plans under 10 CFR 52.17(b)(2)(i), no early 
site permit will be issued unless a finding is made by the NRC that the 
major features are acceptable in accordance with the applicable 
standards of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the 
scope of emergency preparedness matters addressed in the major 
features.
* * * * *
    (e) Notwithstanding the requirements of paragraph (b) of this 
section and the provisions of Sec.  52.103 of this chapter, a holder of 
a combined license under part 52 of this chapter may not load fuel or 
operate except as provided in accordance with appendix E to part 50 and 
Sec.  50.54(gg).

0
87. In Sec.  50.48, the introductory text of paragraph (a)(1) is 
revised and paragraph (a)(4) is added to read as follows:


Sec.  50.48  Fire protection.

    (a)(1) Each holder of an operating license issued under this part 
or a combined license issued under part 52 of this chapter must have a 
fire protection plan that satisfies Criterion 3 of appendix A to this 
part. This fire protection plan must:
* * * * *
    (a)(4) Each applicant for a design approval, design certification, 
or manufacturing license under part 52 of this chapter must have a 
description and analysis of the fire protection design features for the 
standard plant necessary to demonstrate compliance with Criterion 3 of 
appendix A to this part.
* * * * *

0
88. In Sec.  50.49, paragraph (a) is revised to read as follows:


Sec.  50.49  Environmental qualification of electric equipment 
important to safety for nuclear power plants.

    (a) Each holder of or an applicant for an operating license issued 
under this part, or a combined license or manufacturing license issued 
under part 52 of this chapter, other than a nuclear power plant for 
which the certifications required under Sec.  50.82(a)(1) or Sec.  
52.110(a)(1) of this chapter have been submitted, shall establish a 
program for qualifying the electric equipment defined in paragraph (b) 
of this section. For a manufacturing license, only electric equipment 
defined in paragraph (b) which is within the scope of the manufactured 
reactor must be included in the program.
* * * * *

0
89. In Sec.  50.54, the introductory text, and paragraphs (a)(1), (i-
1), (o), (p), and (q) are revised and paragraph (gg) is added to read 
as follows:


Sec.  50.54  Conditions of licenses.

    The following paragraphs with the exception of paragraphs (r) and 
(gg) of this section are conditions in every nuclear power reactor 
operating license issued under this part. The following paragraphs with 
the exception of paragraph (r), (s), and (u) of this section are 
conditions in every combined license issued under part 52 of this 
chapter, provided, however, that paragraphs (i), (i-1), (j), (k), (l), 
(m), (n), (w), (x), (y), and (z) of this section are only applicable 
after the Commission makes the finding under Sec.  52.103(g) of this 
chapter.
    (a)(1) Each nuclear power plant or fuel reprocessing plant licensee 
subject to the quality assurance criteria in appendix B of this part 
shall implement, under Sec.  50.34(b)(6)(ii) or Sec.  52.79 of this 
chapter, the quality assurance program described or referenced in the 
safety analysis report, including changes to that report. However, a 
holder of a combined license under part 52 of this chapter shall 
implement the quality assurance program described or referenced in the 
safety analysis report applicable to operation 30 days prior to the 
scheduled date for the initial loading of fuel.
* * * * *
    (i-1) Within 3 months after either the issuance of an operating 
license or the date that the Commission makes the finding under Sec.  
52.103(g) of this chapter for a combined license, as applicable, the 
licensee shall have in effect an operator requalification program. The 
operator requalification program must, as a minimum, meet the 
requirements of Sec.  55.59(c) of this chapter. Notwithstanding the 
provisions of Sec.  50.59, the licensee may not, except as specifically 
authorized by the Commission decrease the scope of an approved operator 
requalification program.
* * * * *
    (o) Primary reactor containments for water cooled power reactors, 
other than facilities for which the certifications required under 
Sec. Sec.  50.82(a)(1) or 52.110(a)(1) of this chapter have been 
submitted, shall be subject to the requirements set forth in appendix J 
to this part.
    (p)(1) The licensee shall prepare and maintain safeguards 
contingency plan procedures in accordance with appendix C of part 73 of 
this chapter for effecting the actions and decisions contained in the 
Responsibility Matrix of the safeguards contingency plan. The licensee 
may make no change which would decrease the effectiveness of a security 
plan, or guard training and

[[Page 49496]]

qualification plan, prepared pursuant to Sec.  50.34(c) or Sec.  
52.79(a), or part 73 of this chapter, or of the first four categories 
of information (Background, Generic Planning Base, Licensee Planning 
Base, Responsibility Matrix) contained in a licensee safeguards 
contingency plan prepared pursuant to Sec.  50.34(d) or Sec.  52.79(a) 
or part 73 of this chapter, as applicable, without prior approval of 
the Commission. A licensee desiring to make such a change shall submit 
an application for an amendment to the licensee's license pursuant to 
Sec.  50.90.
    (2) The licensee may make changes to the plans referenced in 
paragraph (p)(1) of this section, without prior Commission approval if 
the changes do not decrease the safeguards effectiveness of the plan. 
The licensee shall maintain records of changes to the plans made 
without prior Commission approval for a period of 3 years from the date 
of the change, and shall submit, as specified in Sec.  50.4 or Sec.  
52.3 of this chapter, a report containing a description of each change 
within 2 months after the change is made. Prior to the safeguards 
contingency plan being put into effect, the licensee shall have:
    (i) All safeguards capabilities specified in the safeguards 
contingency plan available and functional;
    (ii) Detailed procedures developed according to appendix C to part 
73 of this chapter available at the licensee's site; and
    (iii) All appropriate personnel trained to respond to safeguards 
incidents as outlined in the plan and specified in the detailed 
procedures.
    (3) The licensee shall provide for the development, revision, 
implementation, and maintenance of its safeguards contingency plan. The 
licensee shall ensure that all program elements are reviewed by 
individuals independent of both security program management and 
personnel who have direct responsibility for implementation of the 
security program either:
    (i) At intervals not to exceed 12 months; or
    (ii) As necessary, based on an assessment by the licensee against 
performance indicators, and as soon as reasonably practicable after a 
change occurs in personnel, procedures, equipment, or facilities that 
potentially could adversely affect security, but no longer than 12 
months after the change. In any case, all elements of the safeguards 
contingency plan must be reviewed at least once every 24 months.
    (4) The review must include a review and audit of safeguards 
contingency procedures and practices, an audit of the security system 
testing and maintenance program, and a test of the safeguards systems 
along with commitments established for response by local law 
enforcement authorities. The results of the review and audit, along 
with recommendations for improvements, must be documented, reported to 
the licensee's corporate and plant management, and kept available at 
the plant for inspection for a period of 3 years.
    (q) A holder of a nuclear power reactor operating license under 
this part, or a combined license under part 52 of this chapter after 
the Commission makes the finding under Sec.  52.103(g) of this chapter, 
shall follow and maintain in effect emergency plans which meet the 
standards in Sec.  50.47(b) and the requirements in appendix E of this 
part. A licensee authorized to possess and/or operate a research 
reactor or a fuel facility shall follow and maintain in effect 
emergency plans which meet the requirements in appendix E to this part. 
The licensee shall retain the emergency plan and each change that 
decreases the effectiveness of the plan as a record until the 
Commission terminates the license for the nuclear power reactor. The 
nuclear power reactor licensee may make changes to these plans without 
Commission approval only if the changes do not decrease the 
effectiveness of the plans and the plans, as changed, continue to meet 
the standards of Sec.  50.47(b) and the requirements of appendix E to 
this part. The research reactor and/or the fuel facility licensee may 
make changes to these plans without Commission approval only if these 
changes do not decrease the effectiveness of the plans and the plans, 
as changed, continue to meet the requirements of appendix E to this 
part. This nuclear power reactor, research reactor, or fuel facility 
licensee shall retain a record of each change to the emergency plan 
made without prior Commission approval for a period of 3 years from the 
date of the change. Proposed changes that decrease the effectiveness of 
the approved emergency plans may not be implemented without application 
to and approval by the Commission. The licensee shall submit, as 
specified in Sec.  50.4, a report of each proposed change for approval. 
If a change is made without approval, the licensee shall submit, as 
specified in Sec.  50.4, a report of each change within 30 days after 
the change is made.
* * * * *
    (gg)(1) Notwithstanding 10 CFR 52.103, if following the conduct of 
the exercise required by paragraph IV.f.2.a of appendix E to part 50 of 
this chapter, DHS identifies one or more deficiencies in the state of 
offsite emergency preparedness, the holder of a combined license under 
10 CFR part 52 may operate at up to 5 percent of rated thermal power 
only if the Commission finds that the state of onsite emergency 
preparedness provides reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological 
emergency. The NRC will base this finding on its assessment of the 
applicant's onsite emergency plans against the pertinent standards in 
Sec.  50.47 and appendix E to this part. Review of the applicant's 
emergency plans will include the following standards with offsite 
aspects:
    (i) Arrangements for requesting and effectively using offsite 
assistance onsite have been made, arrangements to accommodate State and 
local staff at the licensee's near-site Emergency Operations Facility 
have been made, and other organizations capable of augmenting the 
planned onsite response have been identified.
    (ii) Procedures have been established for licensee communications 
with State and local response organizations, including initial 
notification of the declaration of emergency and periodic provision of 
plant and response status reports.
    (iii) Provisions exist for prompt communications among principal 
response organizations to offsite emergency personnel who would be 
responding onsite.
    (iv) Adequate emergency facilities and equipment to support the 
emergency response onsite are provided and maintained.
    (v) Adequate methods, systems, and equipment for assessing and 
monitoring actual or potential offsite consequences of a radiological 
emergency condition are in use onsite.
    (vi) Arrangements are made for medical services for contaminated 
and injured onsite individuals.
    (vii) Radiological emergency response training has been made 
available to those offsite who may be called to assist in an emergency 
onsite.
    (2) The condition in this paragraph, regarding operation at up to 5 
percent power, ceases to apply 30 days after DHS informs the NRC that 
the offsite deficiencies have been corrected, unless the NRC notifies 
the combined license holder before the expiration of the 30-day period 
that the Commission finds under paragraphs (s)(2) and (3) of this 
section that the state of emergency preparedness does not provide 
reasonable assurance that adequate protective measures can and will be 
taken in the event of a radiological emergency.

[[Page 49497]]


0
90. In Sec.  50.55, the heading, the introductory text and paragraphs 
(a), (b), and (e) are revised, and a new paragraph (f)(4) is added to 
read as follows:


Sec.  50.55  Conditions of construction permits, early site permits, 
combined licenses, and manufacturing licenses.

    Each construction permit is subject to the following terms and 
conditions; each early site permit is subject to the terms and 
conditions in paragraph (f) of this section; each manufacturing license 
is subject to the terms and conditions in paragraphs (e) and (f) of 
this section; and each combined license is subject to the terms and 
conditions in paragraphs (e) and (f) of this section until the date 
that the Commission makes the finding under Sec.  52.103(g) of this 
chapter:
    (a) The construction permit shall state the earliest and latest 
dates for completion of the construction or modification.
    (b) If the proposed construction or modification of the facility is 
not completed by the latest completion date, the construction permit 
shall expire and all rights are forfeited. However, upon good cause 
shown, the Commission will extend the completion date for a reasonable 
period of time. The Commission will recognize, among other things, 
developmental problems attributable to the experimental nature of the 
facility or fire, flood, explosion, strike, sabotage, domestic 
violence, enemy action, an act of the elements, and other acts beyond 
the control of the permit holder, as a basis for extending the 
completion date.
* * * * *
    (e)(1) Definitions. For purposes of this paragraph, the definitions 
in Sec.  21.3 of this chapter apply.
    (2) Posting requirements. (i) Each individual, partnership, 
corporation, dedicating entity, or other entity subject to the 
regulations in this part shall post current copies of the regulations 
in this part; Section 206 of the Energy Reorganization Act of 1974 
(ERA); and procedures adopted under the regulations in this part. These 
documents must be posted in a conspicuous position on any premises 
within the United States where the activities subject to this part are 
conducted.
    (ii) If posting of the regulations in this part or the procedures 
adopted under the regulations in this part is not practicable, the 
licensee or firm subject to the regulations in this part may, in 
addition to posting Section 206 of the ERA, post a notice which 
describes the regulations/procedures, including the name of the 
individual to whom reports may be made, and states where the 
regulation, procedures, and reports may be examined.
    (3) Procedures. Each individual, corporation, partnership, or other 
entity holding a facility construction permit subject to this part, 
combined license (until the Commission makes the finding under 10 CFR 
52.103(g)), and manufacturing license under 10 CFR part 52 must adopt 
appropriate procedures to--
    (i) 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 (e)(3)(ii) of 
this section, in all cases within 60 days of discovery, to identify a 
reportable defect or failure to comply that could create a substantial 
safety hazard, were it to remain uncorrected.
    (ii) 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 paragraph (e)(4)(v) of this 
section. 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.
    (iii) Ensure that a director or responsible officer of the holder 
of a facility construction permit subject to this part, combined 
license (until the Commission makes the finding under 10 CFR 
52.103(g)), and manufacturing license under 10 CFR part 52 is informed 
as soon as practicable, and, in all cases, within the 5 working days 
after completion of the evaluation described in paragraph (e)(3)(i) or 
(e)(3)(ii) of this section, if the construction or manufacture of a 
facility or activity, or a basic component supplied for such facility 
or activity--
    (A) Fails to comply with the AEA, as amended, or any applicable 
regulation, order, or license of the Commission, relating to a 
substantial safety hazard;
    (B) Contains a defect; or
    (C) Undergoes any significant breakdown in any portion of the 
quality assurance program conducted under the requirements of appendix 
B to 10 CFR part 50 which could have produced a defect in a basic 
component. These breakdowns in the quality assurance program are 
reportable whether or not the breakdown actually resulted in a defect 
in a design approved and released for construction, installation, or 
manufacture.
    (4) Notification. (i) The holder of a facility construction permit 
subject to this part, combined license (until the Commission makes the 
finding under 10 CFR 52.103(g)), and manufacturing license who obtains 
information reasonably indicating that the facility fails to comply 
with the AEA, as amended, or any applicable regulation, order, or 
license of the Commission relating to a substantial safety hazard must 
notify the Commission of the failure to comply through a director or 
responsible officer or designated person as discussed in paragraph 
(e)(10) of this section.
    (ii) The holder of a facility construction permit subject to this 
part, combined license, or manufacturing license, who obtains 
information reasonably indicating the existence of any defect found in 
the construction or manufacture, or any defect found in the final 
design of a facility as approved and released for construction or 
manufacture, must notify the Commission of the defect through a 
director or responsible officer or designated person as discussed in 
paragraph (e)(4)(v) of this section.
    (iii) The holder of a facility construction permit subject to this 
part, combined license, or manufacturing license, who obtains 
information reasonably indicating that the quality assurance program 
has undergone any significant breakdown discussed in paragraph 
(e)(3)(ii)(C) of this section must notify the Commission of the 
breakdown in the quality assurance program through a director or 
responsible officer or designated person as discussed in paragraph 
(4)(v) of this section.
    (iv) A dedicating entity is responsible for identifying and 
evaluating deviations and reporting defects and failures to comply 
associated with substantial safety hazards for dedicated items; and 
maintaining auditable records for the dedication process.
    (v) The notification requirements of this paragraph apply to all 
defects and failures to comply associated with a substantial safety 
hazard regardless of whether extensive evaluation, redesign, or repair 
is required to conform to the criteria and bases stated in the safety 
analysis report, construction permit, combined license, or 
manufacturing license. Evaluation of potential defects and failures to 
comply and reporting of defects and failures to comply under this 
section satisfies the construction permit holder's, combined license 
holder's, and manufacturing license holder's evaluation and 
notification

[[Page 49498]]

obligations under part 21 of this chapter, and satisfies the 
responsibility of individual directors or responsible officers of 
holders of construction permits issued under Sec.  50.23, holders of 
combined licenses (until the Commission makes the finding under Sec.  
52.103 of this chapter), and holders of manufacturing licenses to 
report defects, and failures to comply associated with substantial 
safety hazards under Section 206 of the ERA. The director or 
responsible officer may authorize an individual to provide the 
notification required by this section, provided that this must not 
relieve the director or responsible officer of his or her 
responsibility under this section.
    (5) Notification--timing and where sent. The notification required 
by paragraph (e)(4) of this section must consist of--
    (i) Initial notification by facsimile, which is the preferred 
method of notification, to the NRC Operations Center at (301) 816-5151 
or by telephone at (301) 816-5100 within 2 days following receipt of 
information by the director or responsible corporate officer under 
paragraph (e)(3)(iii) 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 paragraph 
(e)(3)(ii) of this section.
    (ii) Written notification submitted to the Document Control Desk, 
U.S. Nuclear Regulatory Commission, by an appropriate method listed in 
Sec.  50.4, with a copy to the appropriate Regional Administrator at 
the address specified in appendix D to part 20 of this chapter and a 
copy to the appropriate NRC resident inspector within 30 days following 
receipt of information by the director or responsible corporate officer 
under paragraph (e)(3)(iii) of this section, on the identification of a 
defect or failure to comply.
    (6) Content of notification. The written notification required by 
paragraph (e)(9)(ii) of this section must clearly indicate that the 
written notification is being submitted under Sec.  50.55(e) and 
include 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 the facility or the activity within the United 
States which contains a defect or fails to comply.
    (iii) Identification of the firm constructing or manufacturing 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 the defect or failure to 
comply.
    (v) The date on which the information of a 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 the basic components in 
use at the facility subject to the regulations in this part.
    (vii) In the case of a completed reactor manufactured under part 52 
of this chapter, the entities to which the reactor was supplied.
    (viii) 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.
    (ix) 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 other entities.
    (7) Procurement documents. Each individual, corporation, 
partnership, dedicating entity, or other entity subject to the 
regulations in this part shall ensure that each procurement document 
for a facility, or a basic component specifies or is issued by the 
entity subject to the regulations, when applicable, that the provisions 
of 10 CFR part 21 or 10 CFR 50.55(e) applies, as applicable.
    (8) Coordination with 10 CFR part 21. The requirements of Sec.  
50.55(e) are satisfied when the defect or failure to comply associated 
with a substantial safety hazard has been previously reported under 
part 21 of this chapter, under Sec.  73.71 of this chapter, or under 
Sec. Sec.  50.55(e) or 50.73. For holders of construction permits 
issued before October 29, 1991, evaluation, reporting and recordkeeping 
requirements of Sec.  50.55(e) may be met by complying with the 
comparable requirements of part 21 of this chapter.
    (9) Records retention. The holder of a construction permit, 
combined license, and manufacturing license must prepare and maintain 
records necessary to accomplish the purposes of this section, 
specifically--
    (i) Retain procurement documents, which define the requirements 
that facilities or basic components must meet in order to be considered 
acceptable, for the lifetime of the facility or basic component.
    (ii) Retain records of evaluations of all deviations and failures 
to comply under paragraph (e)(3)(i) of this section for the longest of:
    (A) Ten (10) years from the date of the evaluation;
    (B) Five (5) years from the date that an early site permit is 
referenced in an application for a combined license; or
    (C) Five (5) years from the date of delivery of a manufactured 
reactor.
    (iii) Retain records of all interim reports to the Commission made 
under paragraph (e)(3)(ii) of this section, or notifications to the 
Commission made under paragraph (e)(4) of this section for the minimum 
time periods stated in paragraph (e)(9)(ii) of this section;
    (iv) Suppliers of basic components must retain records of:
    (A) All notifications sent to affected licensees or purchasers 
under paragraph (e)(4)(iv) of this section for a minimum of ten (10) 
years following the date of the notification;
    (B) The facilities or other purchasers to whom basic components or 
associated services were supplied for a minimum of fifteen (15) years 
from the delivery of the basic component or associated services.
    (v) Maintaining records in accordance with this section satisfies 
the recordkeeping obligations under part 21 of this chapter of the 
entities, including directors or responsible officers thereof, subject 
to this section.
    (f) * * *
    (4) Each holder of an early site permit or a manufacturing license 
under part 52 of this chapter shall implement the quality assurance 
program described or referenced in the safety analysis report, 
including changes to that report. Each holder of a combined license 
shall implement the quality assurance program for design and 
construction described or referenced in the safety analysis report, 
including changes to that report, provided, however, that the holder of 
a combined license is not subject to the terms and conditions in this 
paragraph after the Commission makes the finding under Sec.  52.103(g) 
of this chapter.
    (i) Each holder described in paragraph (f)(4) of this section may 
make a change to a previously accepted quality assurance program 
description included or referenced in the safety analysis report, if 
the change does not reduce the commitments in the program description 
previously accepted by the NRC. Changes to the quality assurance 
program description that do not reduce the commitments must be 
submitted to NRC within 90 days. Changes to the quality assurance 
program description that reduce the commitments must be submitted to 
NRC and receive NRC

[[Page 49499]]

approval before implementation, as follows:
    (A) Changes to the safety analysis report must be submitted for 
review as specified in Sec.  50.4. Changes made to NRC-accepted quality 
assurance topical report descriptions must be submitted as specified in 
Sec.  50.4.
    (B) The submittal of a change to the safety analysis report quality 
assurance program description must include all pages affected by that 
change and must be accompanied by a forwarding letter identifying the 
change, the reason for the change, and the basis for concluding that 
the revised program incorporating the change continues to satisfy the 
criteria of appendix B of this part and the safety analysis report 
quality assurance program description commitments previously accepted 
by the NRC (the letter need not provide the basis for changes that 
correct spelling, punctuation, or editorial items).
    (C) A copy of the forwarding letter identifying the changes must be 
maintained as a facility record for three (3) years.
    (D) Changes to the quality assurance program description included 
or referenced in the safety analysis report shall be regarded as 
accepted by the Commission upon receipt of a letter to this effect from 
the appropriate reviewing office of the Commission or 60 days after 
submittal to the Commission, whichever occurs first.
    (ii) [Reserved]

0
91. In Section 50.55a, the introductory paragraphs (b)(1)(i), 
(b)(1)(ii), (b)(1)(iii), (b)(1)(v), the introductory text of paragraphs 
(b)(4) and (d)(1), paragraph (e)(1), the introductory text of paragraph 
(f)(3), paragraphs (f)(3)(iii), (f)(3)(iv)(B), (f)(4)(i), the 
introductory text of paragraph (g)(3), paragraphs (g)(4)(i), the 
introductory text of paragraph (g)(4)(v), and paragraph (h)(3) are 
revised to read as follows:


Sec.  50.55a  Codes and standards.

    Each construction permit for a utilization facility is subject to 
the following conditions in addition to those specified in Sec.  50.55. 
Each combined license for a utilization facility is subject to the 
following conditions in addition to those specified in Sec.  50.55, 
except that each combined license for a boiling or pressurized water-
cooled nuclear power facility is subject to the conditions in 
paragraphs (f) and (g) of this section, but only after the Commission 
makes the finding under Sec.  52.103(g) of this chapter. Each operating 
license for a boiling or pressurized water-cooled nuclear power 
facility is subject to the conditions in paragraphs (f) and (g) of this 
section in addition to those specified in Sec.  50.55. Each 
manufacturing license, standard design approval, and standard design 
certification application under part 52 of this chapter is subject to 
the conditions in paragraphs (a), (b)(1), (b)(4), (c), (d), (e), 
(f)(3), and (g)(3) of this section.
* * * * *
    (b) * * *
    (1) * * *
    (i) Section III Materials. When applying the 1992 Edition of 
Section III, applicants or licensees must apply the 1992 Edition with 
the 1992 Addenda of Section II of the ASME Boiler and Pressure Vessel 
Code.
    (ii) Weld leg dimensions. When applying the 1989 Addenda through 
the latest edition, and addenda incorporated by reference in paragraph 
(b)(1) of this section, applicants or licensees may not apply paragraph 
NB-3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-
3673.2(b)-1.
    (iii) Seismic design. Applicants or licensees may use Articles NB-
3200, NB-3600, NC-3600, and ND-3600 up to and including the 1993 
Addenda, subject to the limitation specified in paragraph (b)(1)(ii) of 
this section. Applicants or licensees may not use these articles in the 
1994 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (b)(1) of this section.
* * * * *
    (v) Independence of inspection. Applicants or licensees may not 
apply NCA-4134.10(a) of Section III, 1995 Edition, through the latest 
edition and addenda incorporated by reference in paragraph (b)(1) of 
this section.
* * * * *
    (4) Design, Fabrication, and Materials Code Cases. Applicants or 
licensees may apply the ASME Boiler and Pressure Vessel Code cases 
listed in NRC Regulatory Guide 1.84, Revision 33, without prior NRC 
approval subject to the following:
* * * * *
    (d) * * *
    (1) For a nuclear power plant whose application for a construction 
permit under this part, or a combined license or manufacturing license 
under part 52 of this chapter is docketed after May 14, 1984, or for an 
application for a standard design approval or a standard design 
certification docketed after May 14, 1984, components classified 
Quality Group B \9\ must meet the requirements for Class 2 Components 
in Section III of the ASME Boiler and Pressure Vessel Code.
---------------------------------------------------------------------------

    \9\ See footnotes at end of section.
---------------------------------------------------------------------------

* * * * *
    (e) * * *
    (1) For a nuclear power plant whose application for a construction 
permit under this part, or a combined license or manufacturing license 
under part 52 of this chapter is docketed after May 14, 1984, or for an 
application for a standard design approval or a standard design 
certification docketed after May 14, 1984, components classified 
Quality Group C <SUP>9</SUP> must meet the requirements for Class 3 
components in Section III of the ASME Boiler and Pressure Vessel Code.
* * * * *
    (f) * * *
    (3) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit under this part or design approval, 
design certification, combined license, or manufacturing license under 
part 52 of this chapter, was issued on or after July 1, 1974:
* * * * *
    (iii)(A) Pumps and valves, in facilities whose construction permit 
under this part, or design certification or design approval under part 
52 of this chapter was issued before November 22, 1999, which are 
classified as ASME Code Class 1 must be designed and be provided with 
access to enable the performance of inservice testing of the pumps and 
valves for assessing operational readiness set forth in the editions 
and addenda of Section XI of the ASME Boiler and Pressure Vessel Code 
incorporated by reference in paragraph (b) of this section (or the 
optional ASME Code cases that are listed in NRC Regulatory Guide 1.147, 
through Revision 14 or Regulatory Guide 1.192, that are incorporated by 
reference in paragraph (b) of this section) applied to the construction 
of the particular pump or valve or the summer 1973 Addenda, whichever 
is later.
    (B) Pumps and valves, in facilities whose construction permit under 
this part, or design certification, design approval, combined license, 
or manufacturing license under part 52 of this chapter, is issued on or 
after November 22, 1999, which are classified as ASME Code Class 1 must 
be designed and be provided with access to enable the performance of 
inservice testing of the pumps and valves for assessing operational 
readiness set forth in editions and addenda of the ASME OM Code (or the 
optional ASME Code cases listed in the NRC Regulatory Guide 1.192 that 
is incorporated by reference in paragraph (b) of this section)

[[Page 49500]]

referenced in paragraph (b)(3) of this section at the time the 
construction permit, combined license, manufacturing license, design 
certification, or design approval is issued.
    (iv) * * *
    (B) Pumps and valves, in facilities whose construction permit under 
this part or design certification or combined license under part 52 of 
this chapter is issued on or after November 22, 1999, which are 
classified as ASME Code Class 2 and 3 must be designed and be provided 
with access to enable the performance of inservice testing of the pumps 
and valves for assessing operational readiness set forth in editions 
and addenda of the ASME OM Code (or the optional ASME Code cases listed 
in the NRC Regulatory Guide 1.192 that is incorporated by reference in 
paragraph (b) of this section) referenced in paragraph (b)(3) of this 
section at the time the construction permit, combined license, or 
design certification is issued.
* * * * *
    (4) * * *
    (i) Inservice tests to verify operational readiness of pumps and 
valves, whose function is required for safety, conducted during the 
initial 120-month interval must comply with the requirements in the 
latest edition and addenda of the Code incorporated by reference in 
paragraph (b) of this section on the date 12 months before the date of 
issuance of the operating license under this part, or 12 months before 
the date scheduled for initial loading fuel under a combined license 
under part 52 of this chapter (or the optional ASME Code cases listed 
in NRC Regulatory Guide 1.192, that is incorporated by reference in 
paragraph (b) of this section), subject to the limitations and 
modifications listed in paragraph (b) of this section.
* * * * *
    (g) * * *
    (3) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit under this part, or design 
certification, design approval, combined license, or manufacturing 
license under part 52 of this chapter, was issued on or after July 1, 
1974:
* * * * *
    (4) * * *
    (i) Inservice examinations of components and system pressure tests 
conducted during the initial 120-month inspection interval must comply 
with the requirements in the latest edition and addenda of the Code 
incorporated by reference in paragraph (b) of this section on the date 
12 months before the date of issuance of the operating license under 
this part, or 12 months before the date scheduled for initial loading 
of fuel under a combined license under part 52 of this chapter (or the 
optional ASME Code cases listed in NRC Regulatory Guide 1.147, through 
Revision 14, that are incorporated by reference in paragraph (b) of 
this section), subject to the limitations and modifications listed in 
paragraph (b) of this section.
* * * * *
    (v) For a boiling or pressurized water-cooled nuclear power 
facility whose construction permit under this part or combined license 
under part 52 of this chapter was issued after January 1, 1956:
* * * * *
    (h) * * *
    (3) Safety systems. Applications filed on or after May 13, 1999, 
for construction permits and operating licenses under this part, and 
for design approvals, design certifications, and combined licenses 
under part 52 of this chapter, must meet the requirements for safety 
systems in IEEE Std. 603-1991 and the correction sheet dated January 
30, 1995.

0
92. In Sec.  50.59, paragraphs (b), (d)(2), and (d)(3) are revised to 
read as follows:


Sec.  50.59  Changes, tests, and experiments.

* * * * *
    (b) This section applies to each holder of an operating license 
issued under this part or a combined license issued under part 52 of 
this chapter, including the holder of a license authorizing operation 
of a nuclear power reactor that has submitted the certification of 
permanent cessation of operations required under Sec.  50.82(a)(1) or 
Sec.  50.110 or a reactor licensee whose license has been amended to 
allow possession of nuclear fuel but not operation of the facility.
* * * * *
    (d) * * *
    (2) The licensee shall submit, as specified in Sec.  50.4 or Sec.  
52.3 of this chapter, as applicable, a report containing a brief 
description of any changes, tests, and experiments, including a summary 
of the evaluation of each. A report must be submitted at intervals not 
to exceed 24 months. For combined licenses, the report must be 
submitted at intervals not to exceed 6 months during the period from 
the date of application for a combined license to the date the 
Commission makes its findings under 10 CFR 52.103(g).
    (3) The records of changes in the facility must be maintained until 
the termination of an operating license issued under this part, a 
combined license issued under part 52 of this chapter, or the 
termination of a license issued under 10 CFR part 54, whichever is 
later. Records of changes in procedures and records of tests and 
experiments must be maintained for a period of 5 years.

0
93. In Sec.  50.61, paragraph (b)(1) is revised to read as follows:


Sec.  50.61  Fracture toughness requirements for protection against 
pressurized thermal shock events.

* * * * *
    (b) * * *
    (1) For each pressurized water nuclear power reactor for which an 
operating license has been issued under this part or a combined license 
has been issued under part 52 of this chapter, other than a nuclear 
power reactor facility for which the certifications required under 
Sec.  50.82(a)(1) have been submitted, the licensee shall have 
projected values of RT<INF>PTS</INF>, accepted by the NRC, for each 
reactor vessel beltline material for the EOL fluence of the material. 
The assessment of RT<INF>PTS</INF> must use the calculation procedures 
given in paragraph (c)(1) of this section, except as provided in 
paragraphs (c)(2) and (c)(3) of this section. The assessment must 
specify the bases for the projected value of RT<INF>PTS</INF> for each 
vessel beltline material, including the assumptions regarding core 
loading patterns, and must specify the copper and nickel contents and 
the fluence value used in the calculation for each beltline material. 
This assessment must be updated whenever there is a significant \2\ 
change in projected values of RT<INF>PTS</INF>, or upon request for a 
change in the expiration date for operation of the facility.
---------------------------------------------------------------------------

    \2\ Changes to RT<INF>PTS</INF> values are considered 
significant if either the previous value or the current value, or 
both values, exceed the screening criterion before the expiration of 
the operating license or the combined license under part 52 of this 
chapter, including any renewed term, if applicable for the plant.
---------------------------------------------------------------------------

* * * * *

0
94. In Sec.  50.62, paragraph (d) is revised to read as follows:


Sec.  50.62  Requirements for reduction of risk from anticipated 
transients without scram (ATWS) events for light-water-cooled nuclear 
power plants.

* * * * *
    (d) Implementation. For each light-water-cooled nuclear power plant 
operating license issued before September 27, 2007, by 180 days after 
the issuance of the QA guidance for non-safety related components, each 
licensee shall develop and submit to the Commission, as specified in 
Sec.  50.4, a proposed schedule for meeting the

[[Page 49501]]

requirements of paragraphs (c)(1) through (c)(5) of this section. Each 
shall include an explanation of the schedule along with a justification 
if the schedule calls for final implementation later than the second 
refueling outage after July 26, 1984, or the date of issuance of a 
license authorizing operation above 5 percent of full power. A final 
schedule shall then be mutually agreed upon by the Commission and 
licensee. For each light-water-cooled nuclear power plant operating 
license application submitted after September 27, 2007, the applicant 
shall submit information in its final safety analysis report 
demonstrating how it will comply with paragraphs (c)(1) through (c)(5) 
of this section.

0
95. In Sec.  50.63, the introductory text of paragraphs (a)(1) and 
(c)(1) are revised to read as follows:


Sec.  50.63  Loss of all alternating current power.

    (a) * * *
    (1) Each light-water-cooled nuclear power plant licensed to operate 
under this part, each light-water-cooled nuclear power plant licensed 
under subpart C of 10 CFR part 52 after the Commission makes the 
finding under Sec.  52.103(g) of this chapter, and each design for a 
light-water-cooled nuclear power plant approved under a standard design 
approval, standard design certification, and manufacturing license 
under part 52 of this chapter must be able to withstand for a specified 
duration and recover from a station blackout as defined in Sec.  50.2. 
The specified station blackout duration shall be based on the following 
factors:
* * * * *
    (c) * * *
    (1) Information Submittal. For each light-water-cooled nuclear 
power plant licensed to operate on or before July 21, 1988, the 
licensee shall submit the information defined below to the Director of 
the Office of Nuclear Reactor Regulation by April 17, 1989. For each 
light-water-cooled nuclear power plant licensed to operate after July 
21, 1988, but before September 27, 2007, the licensee shall submit the 
information defined in this section to the Director of the Office of 
Nuclear Reactor Regulation, by 270 days after the date of license 
issuance. For each light-water-cooled nuclear power plant operating 
license application submitted after September 27, 2007, the applicant 
shall submit the information defined below in its final safety analysis 
report.
* * * * *

0
96. In Sec.  50.65, paragraph (a)(1) is revised to read as follows:


Sec.  50.65  Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants.

* * * * *
    (a)(1) Each holder of an operating license for a nuclear power 
plant under this part and each holder of a combined license under part 
52 of this chapter after the Commission makes the finding under Sec.  
52.103(g) of this chapter, shall monitor the performance or condition 
of structures, systems, or components, against licensee-established 
goals, in a manner sufficient to provide reasonable assurance that 
these structures, systems, and components, as defined in paragraph (b) 
of this section, are capable of fulfilling their intended functions. 
These goals shall be established commensurate with safety and, where 
practical, take into account industry-wide operating experience. When 
the performance or condition of a structure, system, or component does 
not meet established goals, appropriate corrective action shall be 
taken. For a nuclear power plant for which the licensee has submitted 
the certifications specified in Sec.  50.82(a)(1) or 52.110(a)(1) of 
this chapter, as applicable, this section shall only apply to the 
extent that the licensee shall monitor the performance or condition of 
all structures, systems, or components associated with the storage, 
control, and maintenance of spent fuel in a safe condition, in a manner 
sufficient to provide reasonable assurance that these structures, 
systems, and components are capable of fulfilling their intended 
functions.
* * * * *

0
97. In Sec.  50.70 paragraphs (a) and (b)(2) are revised to read as 
follows:


Sec.  50.70  Inspections.

    (a) Each applicant for or holder of a license, including a 
construction permit or an early site permit, shall permit inspection, 
by duly authorized representatives of the Commission, of his records, 
premises, activities, and of licensed materials in possession or use, 
related to the license or construction permit or early site permit as 
may be necessary to effectuate the purposes of the Act, as amended, 
including Section 105 of the Act, and the Energy Reorganization Act of 
1974, as amended.
    (b) * * *
    (2) For a site with a single power reactor or fuel facility 
licensed under part 50 or part 52 of this chapter, or a facility issued 
a manufacturing license under part 52, the space provided shall be 
adequate to accommodate a full-time inspector, a part-time secretary 
and transient NRC personnel and will be generally commensurate with 
other office facilities at the site. A space of 250 square feet either 
within the site's office complex or in an office trailer or other 
onsite space is suggested as a guide. For sites containing multiple 
power reactor units or fuel facilities, additional space may be 
requested to accommodate additional full-time inspector(s). The office 
space that is provided shall be subject to the approval of the 
Director, Office of New Reactors, or the Director, Office of Nuclear 
Reactor Regulation. All furniture, supplies and communication equipment 
will be furnished by the Commission.
* * * * *

0
98. In Sec.  50.71, paragraphs (a), (c), (d)(1), and the introductory 
text of paragraph (e) are revised, paragraph (e)(3)(iii) is added, 
paragraph (f) is redesignated as paragraph (g) and revised, and new 
paragraphs (f) and (h) are added to read as follows:


Sec.  50.71  Maintenance of records, making of reports.

    (a) Each licensee, including each holder of a construction permit 
or early site permit, shall maintain all records and make all reports, 
in connection with the activity, as may be required by the conditions 
of the license or permit or by the regulations, and orders of the 
Commission in effectuating the purposes of the Act, including Section 
105 of the Act, and the Energy Reorganization Act of 1974, as amended. 
Reports must be submitted in accordance with Sec.  50.4 or 10 CFR 52.3, 
as applicable.
* * * * *
    (c) Records that are required by the regulations in this part or 
part 52 of this chapter, by license condition, or by technical 
specifications must be retained for the period specified by the 
appropriate regulation, license condition, or technical specification. 
If a retention period is not otherwise specified, these records must be 
retained until the Commission terminates the facility license or, in 
the case of an early site permit, until the permit expires.
    (d)(1) Records which must be maintained under this part or part 52 
of this chapter may be the original or a reproduced copy or microform 
if the reproduced copy or microform is duly authenticated by authorized 
personnel and the microform is capable of producing a clear and legible 
copy after storage for the period specified by Commission regulations. 
The record may also be stored in electronic media with the capability 
of producing legible, accurate, and complete records during

[[Page 49502]]

the required retention period. Records such as letters, drawings, and 
specifications, must include all pertinent information such as stamps, 
initials, and signatures. The licensee shall maintain adequate 
safeguards against tampering with, and loss of records.
* * * * *
    (e) Each person licensed to operate a nuclear power reactor under 
the provisions of Sec.  50.21 or Sec.  50.22, and each applicant for a 
combined license under part 52 of this chapter, shall update 
periodically, as provided in paragraphs (e) (3) and (4) of this 
section, the final safety analysis report (FSAR) originally submitted 
as part of the application for the license, to assure that the 
information included in the report contains the latest information 
developed. This submittal shall contain all the changes necessary to 
reflect information and analyses submitted to the Commission by the 
applicant or licensee or prepared by the applicant or licensee pursuant 
to Commission requirement since the submittal of the original FSAR, or 
as appropriate, the last update to the FSAR under this section. The 
submittal shall include the effects \1\ of all changes made in the 
facility or procedures as described in the FSAR; all safety analyses 
and evaluations performed by the applicant or licensee either in 
support of approved license amendments or in support of conclusions 
that changes did not require a license amendment in accordance with 
Sec.  50.59(c)(2) or, in the case of a license that references a 
certified design, in accordance with Sec.  52.98(c) of this chapter; 
and all analyses of new safety issues performed by or on behalf of the 
applicant or licensee at Commission request. The updated information 
shall be appropriately located within the update to the FSAR.
---------------------------------------------------------------------------

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

* * * * *
    (3) * * *
    (iii) During the period from the docketing of an application for a 
combined license under subpart C of part 52 of this chapter until the 
Commission makes the finding under Sec.  52.103(g) of this chapter, the 
update to the FSAR must be submitted annually.
* * * * *
    (f) Each person licensed to manufacture a nuclear power reactor 
under subpart F of 10 CFR part 52 shall update the FSAR originally 
submitted as part of the application to reflect any modification to the 
design that is approved by the Commission under Sec.  52.171 of this 
chapter, and any new analyses of the design performed by or on behalf 
of the licensee at the NRC's request. This submittal shall contain all 
the changes necessary to reflect information and analyses submitted to 
the Commission by the licensee or prepared by the licensee with respect 
to the modification approved under Sec.  52.171 of this chapter or the 
analyses requested by the Commission under Sec.  52.171 of this 
chapter. The updated information shall be appropriately located within 
the update to the FSAR.
    (g) The provisions of this section apply to nuclear power reactor 
licensees that have submitted the certification of permanent cessation 
of operations required under Sec. Sec.  50.82(a)(1)(i) or 52.110(a)(1) 
of this chapter. The provisions of paragraphs (a), (c), and (d) of this 
section also apply to non-power reactor licensees that are no longer 
authorized to operate.
    (h)(1) No later than the scheduled date for initial loading of 
fuel, each holder of a combined license under subpart C of 10 CFR part 
52 shall develop a level 1 and a level 2 probabilistic risk assessment 
(PRA). The PRA must cover those initiating events and modes for which 
NRC-endorsed consensus standards on PRA exist one year prior to the 
scheduled date for initial loading of fuel.
    (2) Each holder of a combined license shall maintain and upgrade 
the PRA required by paragraph (h)(1) of this section. The upgraded PRA 
must cover initiating events and modes of operation contained in NRC-
endorsed consensus standards on PRA in effect one year prior to each 
required upgrade. The PRA must be upgraded every four years until the 
permanent cessation of operations under Sec.  52.110(a) of this 
chapter.
    (3) Each holder of a combined license shall, no later than the date 
on which the licensee submits an application for a renewed license, 
upgrade the PRA required by paragraph (h)(1) of this section to cover 
all modes and all initiating events.

0
99. In Sec.  50.72, the introductory text of paragraph (a)(1) is 
revised to read as follows:


Sec.  50.72  Immediate notification requirements for operating nuclear 
power reactors.

    (a) * * *
    (1) Each nuclear power reactor licensee licensed under Sec. Sec.  
50.21(b) or 50.22 holding an operating license under this part or a 
combined license under part 52 of this chapter after the Commission 
makes the finding under Sec.  52.103(g), shall notify the NRC 
Operations Center via the Emergency Notification System of:
* * * * *

0
100. In Sec.  50.73, paragraph (a)(1) is revised to read as follows:


Sec.  50.73  Licensee event report system.

    (a) * * *
    (1) The holder of an operating license under this part or a 
combined license under part 52 of this chapter (after the Commission 
has made the finding under Sec.  52.103(g) of this chapter) for a 
nuclear power plant (licensee) shall submit a Licensee Event Report 
(LER) for any event of the type described in this paragraph within 60 
days after the discovery of the event. In the case of an invalid 
actuation reported under Sec.  50.73(a)(2)(iv), other than actuation of 
the reactor protection system (RPS) when the reactor is critical, the 
licensee may, at its option, provide a telephone notification to the 
NRC Operations Center within 60 days after discovery of the event 
instead of submitting a written LER. Unless otherwise specified in this 
section, the licensee shall report an event if it occurred within 3 
years of the date of discovery regardless of the plant mode or power 
level, and regardless of the significance of the structure, system, or 
component that initiated the event.
* * * * *

0
101. In Sec.  50.75, paragraphs (a) and (b) are revised, paragraph 
(e)(3) is added, paragraphs (f)(1), (f)(2), (f)(3), and (f)(4) are 
redesignated as paragraphs (f)(2), (f)(3), (f)(4), and (f)(5), 
respectively, and paragraph (f)(1) is added to read as follows:


Sec.  50.75  Reporting and recordkeeping for decommissioning planning.

    (a) This section establishes requirements for indicating to NRC how 
a licensee will provide reasonable assurance that funds will be 
available for the decommissioning process. For power reactor licensees 
(except a holder of a manufacturing license under part 52 of this 
chapter), reasonable assurance consists of a series of steps as 
provided in paragraphs (b), (c), (e), and (f) of this section. Funding 
for the decommissioning of power reactors may also be subject to the 
regulation of Federal or State Government agencies (e.g., Federal 
Energy Regulatory Commission (FERC) and State Public Utility 
Commissions) that have jurisdiction over rate regulation. The 
requirements of this section, in particular paragraph (c) of this 
section, are in addition to, and not substitution for, other 
requirements, and are not intended to be used by themselves or by other 
agencies to establish rates.

[[Page 49503]]

    (b) Each power reactor applicant for or holder of an operating 
license, and each applicant for a combined license under subpart C of 
10 CFR part 52 for a production or utilization facility of the type and 
power level specified in paragraph (c) of this section shall submit a 
decommissioning report, as required by Sec.  50.33(k).
    (1) For an applicant for or holder of an operating license under 
part 50, the report must contain a certification that financial 
assurance for decommissioning will be (for a license applicant), or has 
been (for a license holder), provided in an amount which may be more, 
but not less, than the amount stated in the table in paragraph (c)(1) 
of this section adjusted using a rate at least equal to that stated in 
paragraph (c)(2) of this section. For an applicant for a combined 
license under subpart C of 10 CFR part 52, the report must contain a 
certification that financial assurance for decommissioning will be 
provided no later than 30 days after the Commission publishes notice in 
the Federal Register under Sec.  52.103(a) in an amount which may be 
more, but not less, than the amount stated in the table in paragraph 
(c)(1) of this section, adjusted using a rate at least equal to that 
stated in paragraph (c)(2) of this section.
    (2) The amount to be provided must be adjusted annually using a 
rate at least equal to that stated in paragraph (c)(2) of this section.
    (3) The amount must be covered by one or more of the methods 
described in paragraph (e) of this section as acceptable to the NRC.
    (4) The amount stated in the applicant's or licensee's 
certification may be based on a cost estimate for decommissioning the 
facility. As part of the certification, a copy of the financial 
instrument obtained to satisfy the requirements of paragraph (e) of 
this section must be submitted to NRC; provided, however, that an 
applicant for or holder of a combined license need not obtain such 
financial instrument or submit a copy to the Commission except as 
provided in paragraph (e)(3) of this section.
* * * * *
    (e) * * *
    (3) Each holder of a combined license under subpart C of 10 CFR 
part 52 shall, 2 years before and 1 year before the scheduled date for 
initial loading of fuel, consistent with the schedule required by Sec.  
52.99(a), submit a report to the NRC containing a certification 
updating the information described under paragraph (b)(1) of this 
section, including a copy of the financial instrument to be used. No 
later than 30 days after the Commission publishes notice in the Federal 
Register under 10 CFR 52.103(a), the licensee shall submit a report 
containing a certification that financial assurance for decommissioning 
is being provided in an amount specified in the licensee's most recent 
updated certification, including a copy of the financial instrument 
obtained to satisfy the requirements of paragraph (e) of this section.
    (f)(1) Each power reactor licensee shall report, on a calendar-year 
basis, to the NRC by March 31, 1999, and at least once every 2 years on 
the status of its decommissioning funding for each reactor or part of a 
reactor that it owns. However, each holder of a combined license under 
part 52 of this chapter need not begin reporting until the date that 
the Commission has made the finding under Sec.  52.103(g) of this 
chapter. The information in this report must include, at a minimum the 
amount of decommissioning funds estimated to be required under 10 CFR 
50.75(b) and (c); the amount accumulated to the end of the calendar 
year preceding the date of the report; a schedule of the annual amounts 
remaining to be collected; the assumptions used regarding rates of 
escalation in decommissioning costs, rates of earnings on 
decommissioning funds, and rates of other factors used in funding 
projections; any contracts upon which the licensee is relying under 
paragraph (e)(1)(v) of this section; any modifications occurring to a 
licensee's current method of providing financial assurance since the 
last submitted report; and any material changes to trust agreements. 
Any licensee for a plant that is within 5 years of the projected end of 
its operation, or where conditions have changed so that it will close 
within 5 years (before the end of its licensed life), or has already 
closed (before the end of its licensed life), or for plants involved in 
mergers or acquisitions shall submit this report annually.
* * * * *

0
102. Section 50.78 is revised to read as follows:


Sec.  50.78  Installation information and verification.

    Each holder of a construction permit and each holder of a combined 
license shall, if requested by the Commission, submit installation 
information on Form-71, permit verification thereof by the 
International Atomic Energy Agency, and take other action as may be 
necessary to implement the US/IAEA Safeguards Agreement, in the manner 
set forth in Sec.  75.6 and Sec. Sec.  75.11 through 75.14 of this 
chapter.


0
103. In Sec.  50.80, paragraphs (a) and (b) are revised to read as 
follows:


Sec.  50.80  Transfer of licenses.

    (a) No license for a production or utilization facility (including, 
but not limited to, permits under this part and part 52 of this 
chapter, and licenses under parts 50 and 52 of this chapter), or any 
right thereunder, shall be transferred, assigned, or in any manner 
disposed of, either voluntarily or involuntarily, directly or 
indirectly, through transfer of control of the license to any person, 
unless the Commission gives its consent in writing.
    (b)(1) An application for transfer of a license shall include:
    (i) For a construction permit or operating license under this part, 
as much of the information described in Sec. Sec.  50.33 and 50.34 of 
this part with respect to the identity and technical and financial 
qualifications of the proposed transferee as would be required by those 
sections if the application were for an initial license. The Commission 
may require additional information such as data respecting proposed 
safeguards against hazards from radioactive materials and the 
applicant's qualifications to protect against such hazards.
    (ii) For an early site permit under part 52 of this chapter, as 
much of the information described in Sec. Sec.  52.16 and 52.17 of this 
chapter with respect to the identity and technical qualifications of 
the proposed transferee as would be required by those sections if the 
application were for an initial license.
    (iii) For a combined license under part 52 of this chapter, as much 
of the information described in Sec. Sec.  52.77 and 52.79 of this 
chapter with respect to the identity and technical and financial 
qualifications of the proposed transferee as would be required by those 
sections if the application were for an initial license. The Commission 
may require additional information such as data respecting proposed 
safeguards against hazards from radioactive materials and the 
applicant's qualifications to protect against such hazards.
    (iv) For a manufacturing license under part 52 of this chapter, as 
much of the information described in Sec. Sec.  52.156 and 52.157 of 
this chapter with respect to the identity and technical qualifications 
of the proposed transferee as would be required by those sections if 
the application were for an initial license.
    (2) The application shall include also a statement of the purposes 
for which the transfer of the license is requested,

[[Page 49504]]

the nature of the transaction necessitating or making desirable the 
transfer of the license, and an agreement to limit access to Restricted 
Data pursuant to Sec.  50.37. The Commission may require any person who 
submits an application for license pursuant to the provisions of this 
section to file a written consent from the existing licensee or a 
certified copy of an order or judgment of a court of competent 
jurisdiction attesting to the person's right (subject to the licensing 
requirements of the Act and these regulations) to possession of the 
facility or site involved.
* * * * *

0
104. In Sec.  50.81, paragraph (d)(1) is revised, and a new paragraph 
(d)(3) is added to read as follows:


Sec.  50.81  Creditor regulations.

* * * * *
    (d) * * *
    (1) License includes any license under this chapter, any 
construction permit under this part, and any early site permit under 
part 52 of this chapter, which may be issued by the Commission with 
regard to a facility;
* * * * *
    (3) Facility includes but is not limited to, a site which is the 
subject of an early site permit under subpart A of part 52 of this 
chapter, and a reactor manufactured under a manufacturing license under 
subpart F of part 52 of this chapter.


0
105. Section 50.90 is revised to read as follows:


Sec.  50.90  Application for amendment of license, construction permit, 
or early site permit.

    Whenever a holder of a license, including a construction permit and 
operating license under this part, and an early site permit, combined 
license, and manufacturing license under part 52 of this chapter, 
desires to amend the license or permit, application for an amendment 
must be filed with the Commission, as specified in Sec. Sec.  50.4 or 
52.3 of this chapter, as applicable, fully describing the changes 
desired, and following as far as applicable, the form prescribed for 
original applications.


0
106. In Sec.  50.91, the introductory text is revised to read as 
follows:


Sec.  50.91  Notice for public comment; State consultation.

    The Commission will use the following procedures for an application 
requesting an amendment to an operating license under this part or a 
combined license under part 52 of this chapter for a facility licensed 
under Sec. Sec.  50.21(b) or 50.22, or for a testing facility, except 
for amendments subject to hearings governed by 10 CFR part 2, subpart 
L. For amendments subject to 10 CFR part 2, subpart L, the following 
procedures will apply only to the extent specifically referenced in 
Sec.  2.309(b) of this chapter, except that notice of opportunity for 
hearing must be published in the Federal Register at least 30 days 
before the requested amendment is issued by the Commission:
* * * * *

0
107. Section 50.92 paragraph (a), and the introductory text of 
paragraph (c) are revised to read as follows:


Sec.  50.92  Issuance of amendment.

    (a) In determining whether an amendment to a license, construction 
permit, or early site permit will be issued to the applicant, the 
Commission will be guided by the considerations which govern the 
issuance of initial licenses, construction permits, or early site 
permits to the extent applicable and appropriate. If the application 
involves the material alteration of a licensed facility, a construction 
permit will be issued before the issuance of the amendment to the 
license, provided however, that if the application involves a material 
alteration to a nuclear power reactor manufactured under part 52 of 
this chapter before its installation at a site, or a combined license 
before the date that the Commission makes the finding under Sec.  
52.103(g) of this chapter, no application for a construction permit is 
required. If the amendment involves a significant hazards 
consideration, the Commission will give notice of its proposed action:
    (1) Under Sec.  2.105 of this chapter before acting thereon; and
    (2) As soon as practicable after the application has been docketed.
* * * * *
    (c) The Commission may make a final determination, under the 
procedures in Sec.  50.91, that a proposed amendment to an operating 
license or a combined license for a facility or reactor licensed under 
Sec. Sec.  50.21(b) or 50.22, or for a testing facility involves no 
significant hazards consideration, if operation of the facility in 
accordance with the proposed amendment would not:
* * * * *

0
108. Section 50.100 is revised to read as follows:


Sec.  50.100  Revocation, suspension, modification of licenses, 
permits, and approvals for cause.

    A license, permit, or standard design approval under parts 50 or 52 
of this chapter may be revoked, suspended, or modified, in whole or in 
part, for any material false statement in the application or in the 
supplemental or other statement of fact required of the applicant; or 
because of conditions revealed by the application or statement of fact 
of any report, record, inspection, or other means which would warrant 
the Commission to refuse to grant a license, permit, or approval on an 
original application (other than those relating to Sec. Sec.  50.51, 
50.42(a), and 50.43(b)); or for failure to manufacture a reactor, or 
construct or operate a facility in accordance with the terms of the 
permit or license, provided, however, that failure to make timely 
completion of the proposed construction or alteration of a facility 
under a construction permit under part 50 of this chapter or a combined 
license under part 52 of this chapter shall be governed by the 
provisions of Sec.  50.55(b); or for violation of, or failure to 
observe, any of the terms and provisions of the act, regulations, 
license, permit, approval, or order of the Commission.


0
109. 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's 
regulations or the imposition of a regulatory staff position 
interpreting the Commission's regulations 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;
    (ii) Six (6) months before the date of docketing of the operating 
license application for the facility for facilities having construction 
permits issued before October 21, 1985;
    (iii) The date of issuance of the operating license for the 
facility for facilities having operating licenses;
    (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 F 
of part 52 of this chapter;
    (vi) The date of issuance of the first construction permit issued 
for a duplicate design under appendix N of this part; or

[[Page 49505]]

    (vii) The date of issuance of a combined license under subpart C of 
part 52 of this chapter, provided that if the combined license 
references an early site permit, the provisions in Sec.  52.39 of this 
chapter apply with respect to the site characteristics, design 
parameters, and terms and conditions specified in the early site 
permit. If the combined license references a standard design 
certification rule under subpart B of 10 CFR part 52, the provisions in 
Sec.  52.63 of this chapter apply with respect to the design matters 
resolved in the standard design certification rule, provided however, 
that if any specific backfitting limitations are included in a 
referenced design certification rule, those limitations shall govern. 
If the combined license references a standard design approval under 
subpart E of 10 CFR part 52, the provisions in Sec.  52.145 of this 
chapter apply with respect to the design matters resolved in the 
standard design approval. If the combined license uses a reactor 
manufactured under a manufacturing license under subpart F of 10 CFR 
part 52, the provisions of Sec.  52.171 of this chapter apply with 
respect to matters resolved in the manufacturing license proceeding.
* * * * *

0
110. Section 50.120 is revised to read as follows:


Sec.  50.120  Training and qualification of nuclear power plant 
personnel.

    (a) Applicability. The requirements of this section apply to each 
applicant for and each holder of an operating license issued under this 
part and each holder of a combined license issued under part 52 of this 
chapter for a nuclear power plant of the type specified in Sec.  
50.21(b) or Sec.  50.22.
    (b) Requirements. (1)(i) Each nuclear power plant operating license 
applicant, by 18 months prior to fuel load, and each holder of an 
operating license shall establish, implement, and maintain a training 
program that meets the requirements of paragraphs (b)(2) and (b)(3) of 
this section.
    (ii) Each holder of a combined license shall establish, implement, 
and maintain the training program that meets the requirements of 
paragraphs (b)(2) and (b)(3) of this section, as described in the final 
safety analysis report no later than 18 months before the scheduled 
date for initial loading of fuel.
    (2) The training program must be derived from a systems approach to 
training as defined in 10 CFR 55.4, and must provide for the training 
and qualification of the following categories of nuclear power plant 
personnel:
    (i) Non-licensed operator.
    (ii) Shift supervisor.
    (iii) Shift technical advisor.
    (iv) Instrument and control technician.
    (v) Electrical maintenance personnel.
    (vi) Mechanical maintenance personnel.
    (vii) Radiological protection technician.
    (viii) Chemistry technician.
    (ix) Engineering support personnel.
    (3) The training program must incorporate the instructional 
requirements necessary to provide qualified personnel to operate and 
maintain the facility in a safe manner in all modes of operation. The 
training program must be developed to be in compliance with the 
facility license, including all technical specifications and applicable 
regulations. The training program must be periodically evaluated and 
revised as appropriate to reflect industry experience as well as 
changes to the facility, procedures, regulations, and quality assurance 
requirements. The training program must be periodically reviewed by 
licensee management for effectiveness. Sufficient records must be 
maintained by the licensee to maintain program integrity and kept 
available for NRC inspection to verify the adequacy of the program.


0
111. In Appendix A to Part 50, the first paragraph under the 
introduction and the second paragraph under Criterion 19 are revised to 
read as follows:

Appendix A to Part 50--General Design Criteria for Nuclear Power Plants

* * * * *

Introduction

    Under the provisions of Sec.  50.34, an application for a 
construction permit must include the principal design criteria for a 
proposed facility. Under the provisions of 10 CFR 52.47, 52.79, 
52.137, and 52.157, an application for a design certification, 
combined license, design approval, or manufacturing license, 
respectively, must include the principal design criteria for a 
proposed facility. The principal design criteria establish the 
necessary design, fabrication, construction, testing, and 
performance requirements for structures, systems, and components 
important to safety; that is, structures, systems, and components 
that provide reasonable assurance that the facility can be operated 
without undue risk to the health and safety of the public.
* * * * *
    Criterion 19--Control Room.
* * * * *
    Applicants for and holders of construction permits and operating 
licenses under this part who apply on or after January 10, 1997, 
applicants for design approvals or certifications under part 52 of 
this chapter who apply on or after January 10, 1997, applicants for 
and holders of combined licenses or manufacturing licenses under 
part 52 of this chapter who do not reference a standard design 
approval or certification, or holders of operating licenses using an 
alternative source term under Sec.  50.67, shall meet the 
requirements of this criterion, except that with regard to control 
room access and occupancy, adequate radiation protection shall be 
provided to ensure that radiation exposures shall not exceed 0.05 Sv 
(5 rem) total effective dose equivalent (TEDE) as defined in Sec.  
50.2 for the duration of the accident.
* * * * *

0
112. In Appendix B to Part 50, the Introduction and Section I are 
revised to read as follows:

Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power 
Plants and Fuel Reprocessing Plants

    Introduction. Every applicant for a construction permit is 
required by the provisions of Sec.  50.34 to include in its 
preliminary safety analysis report a description of the quality 
assurance program to be applied to the design, fabrication, 
construction, and testing of the structures, systems, and components 
of the facility. Every applicant for an operating license is 
required to include, in its final safety analysis report, 
information pertaining to the managerial and administrative controls 
to be used to assure safe operation. Every applicant for a combined 
license under part 52 of this chapter is required by the provisions 
of Sec.  52.79 of this chapter to include in its final safety 
analysis report a description of the quality assurance applied to 
the design, and to be applied to the fabrication, construction, and 
testing of the structures, systems, and components of the facility 
and to the managerial and administrative controls to be used to 
assure safe operation. For applications submitted after September 
27, 2007, every applicant for an early site permit under part 52 of 
this chapter is required by the provisions of Sec.  52.17 of this 
chapter to include in its site safety analysis report a description 
of the quality assurance program applied to site activities related 
to the design, fabrication, construction, and testing of the 
structures, systems, and components of a facility or facilities that 
may be constructed on the site. Every applicant for a design 
approval or design certification under part 52 of this chapter is 
required by the provisions of 10 CFR 52.137 and 52.47, respectively, 
to include in its final safety analysis report a description of the 
quality assurance program applied to the design of the structures, 
systems, and components of the facility. Every applicant for a 
manufacturing license under part 52 of this chapter is required by 
the provisions of 10 CFR 52.157 to include in its final safety 
analysis report a description of the quality assurance program 
applied to the design, and to be applied to the manufacture of, the 
structures, systems, and components of the reactor. Nuclear power 
plants and fuel reprocessing plants include structures, systems, and 
components that prevent or mitigate the consequences of

[[Page 49506]]

postulated accidents that could cause undue risk to the health and 
safety of the public. This appendix establishes quality assurance 
requirements for the design, manufacture, construction, and 
operation of those structures, systems, and components. The 
pertinent requirements of this appendix apply to all activities 
affecting the safety-related functions of those structures, systems, 
and components; these activities include designing, purchasing, 
fabricating, handling, shipping, storing, cleaning, erecting, 
installing, inspecting, testing, operating, maintaining, repairing, 
refueling, and modifying.
    As used in this appendix, ``quality assurance'' comprises all 
those planned and systematic actions necessary to provide adequate 
confidence that a structure, system, or component will perform 
satisfactorily in service. Quality assurance includes quality 
control, which comprises those quality assurance actions related to 
the physical characteristics of a material, structure, component, or 
system which provide a means to control the quality of the material, 
structure, component, or system to predetermined requirements.

I. Organization

    The applicant \1\ shall be responsible for the establishment and 
execution of the quality assurance program. The applicant may 
delegate to others, such as contractors, agents, or consultants, the 
work of establishing and executing the quality assurance program, or 
any part thereof, but shall retain responsibility for the quality 
assurance program. The authority and duties of persons and 
organizations performing activities affecting the safety-related 
functions of structures, systems, and components shall be clearly 
established and delineated in writing. These activities include both 
the performing functions of attaining quality objectives and the 
quality assurance functions. The quality assurance functions are 
those of (1) assuring that an appropriate quality assurance program 
is established and effectively executed; and (2) verifying, such as 
by checking, auditing, and inspecting, that activities affecting the 
safety-related functions have been correctly performed. The persons 
and organizations performing quality assurance functions shall have 
sufficient authority and organizational freedom to identify quality 
problems; to initiate, recommend, or provide solutions; and to 
verify implementation of solutions. There persons and organizations 
performing quality assurance functions shall report to a management 
level so that the required authority and organizational freedom, 
including sufficient independence from cost and schedule when 
opposed to safety considerations, are provided. Because of the many 
variables involved, such as the number of personnel, the type of 
activity being performed, and the location or locations where 
activities are performed, the organizational structure for executing 
the quality assurance program may take various forms, provided that 
the persons and organizations assigned the quality assurance 
functions have the required authority and organizational freedom. 
Irrespective of the organizational structure, the individual(s) 
assigned the responsibility for assuring effective execution of any 
portion of the quality assurance program at any location where 
activities subject to this appendix are being performed, shall have 
direct access to the levels of management necessary to perform this 
function.
---------------------------------------------------------------------------

    \1\ While the term ``applicant'' is used in these criteria, the 
requirements are, of course, applicable after such a person has 
received a license to construct and operate a nuclear power plant or 
a fuel reprocessing plant or has received an early site permit, 
design approval, design certification, or manufacturing license, as 
applicable. These criteria will also be used for guidance in 
evaluating the adequacy of quality assurance programs in use by 
holders of construction permits, operating licenses, early site 
permits, design approvals, combined licenses, and manufacturing 
licenses.
---------------------------------------------------------------------------

* * * * *


0
113. In Appendix C to Part 50, the heading, the first paragraph of 
General Information, and the headings of Sections I.A and II.A, and 
Section III are revised to read as follows:

Appendix C to Part 50--A Guide for the Financial Data and Related 
Information Required To Establish Financial Qualifications for 
Construction Permits and Combined Licenses

General Information

    This appendix is intended to appraise applicants for 
construction permits and combined licenses for production or 
utilization facilities of the types described in Sec.  50.21(b) or 
Sec.  50.22, or testing facilities, of the general kinds of 
financial data and other related information that will demonstrate 
the financial qualification of the applicant to carry out the 
activities for which the permit or license is sought. The kind and 
depth of information described in this guide is not intended to be a 
rigid and absolute requirement. In some instances, additional 
pertinent material may be needed. In any case, the applicant should 
include information other than that specified, if the information is 
pertinent to establishing the applicant's financial ability to carry 
out the activities for which the permit or license is sought.
* * * * *

I. * * *

A. Applications for Construction Permits or Combined Licenses

* * * * *

II. * * *

A. Applications for Construction Permits or Combined Licenses

* * * * *

III. Annual Financial Statement

    Each holder of a construction permit for a production or 
utilization facility of a type described in Sec.  50.21(b) or Sec.  
50.22 or a testing facility, and each holder of a combined license 
issued under part 52 of this chapter, is required by Sec.  50.71(b) 
to file its annual financial report with the Commission at the time 
of issuance. This requirement does not apply to licensees or holders 
of construction permits for medical and research reactors.
* * * * *

0
114. In Appendix E to Part 50, Sections I, III, IV.F.2.a, IV.F.2.c, and 
V are revised, and footnotes 6, 7, 8, 9, 10, and 11 are redesignated as 
7, 8, 9, 10, 11, and 12, respectively, and a new footnote 6 is added to 
read as follows:

Appendix E to Part 50--Emergency Planning and Preparedness for 
Production and Utilization Facilities

* * * * *

I. Introduction

    Each applicant for a construction permit is required by Sec.  
50.34(a) to include in the preliminary safety analysis report a 
discussion of preliminary plans for coping with emergencies. Each 
applicant for an operating license is required by Sec.  50.34(b) to 
include in the final safety analysis report plans for coping with 
emergencies. Each applicant for a combined license under subpart C 
of part 52 of this chapter is required by Sec.  52.79 of this 
chapter to include in the application plans for coping with 
emergencies. Each applicant for an early site permit under subpart A 
of part 52 of this chapter may submit plans for coping with 
emergencies under Sec.  52.17 of this chapter.
    This appendix establishes minimum requirements for emergency 
plans for use in attaining an acceptable state of emergency 
preparedness. These plans shall be described generally in the 
preliminary safety analysis report for a construction permit and 
submitted as part of the final safety analysis report for an 
operating license. These plans, or major features thereof, may be 
submitted as part of the site safety analysis report for an early 
site permit.
* * * * *

III. The Final Safety Analysis Report; Site Safety Analysis Report

    The final safety analysis report or the site safety analysis 
report for an early site permit that includes complete and 
integrated emergency plans under Sec.  52.17(b)(2)(ii) of this 
chapter shall contain the plans for coping with emergencies. The 
plans shall be an expression of the overall concept of operation; 
they shall describe the essential elements of advance planning that 
have been considered and the provisions that have been made to cope 
with emergency situations. The plans shall incorporate information 
about the emergency response roles of supporting organizations and 
offsite agencies. That information shall be sufficient to provide 
assurance of coordination among the supporting groups and with the 
licensee. The site safety analysis report for an early site permit 
which proposes major features must address the relevant provisions 
of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the scope of 
emergency preparedness matters addressed in the major features.
    The plans submitted must include a description of the elements 
set out in Section IV for the emergency planning zones (EPZs)

[[Page 49507]]

to an extent sufficient to demonstrate that the plans provide 
reasonable assurance that adequate protective measures can and will 
be taken in the event of an emergency.

IV. Content of Emergency Plans

* * * * *
    F. * * *
    2. * * *
    a. A full participation \4\ exercise which tests as much of the 
licensee, State, and local emergency plans as is reasonably 
achievable without mandatory public participation shall be conducted 
for each site at which a power reactor is located.
---------------------------------------------------------------------------

    \4\ Full participation when used in conjunction with emergency 
preparedness exercises for a particular site means appropriate 
offsite local and State authorities and licensee personnel 
physically and actively take part in testing their integrated 
capability to adequately assess and respond to an accident at a 
commercial nuclear power plant. Full participation includes testing 
major observable portions of the onsite and offsite emergency plans 
and mobilization of State, local and licensee personnel and other 
resources in sufficient numbers to verify the capability to respond 
to the accident scenario.
---------------------------------------------------------------------------

    (i) For an operating license issued under this part, this 
exercise must be conducted within two years before the issuance of 
the first operating license for full power (one authorizing 
operation above 5 percent of rated power) of the first reactor and 
shall include participation by each State and local government 
within the plume exposure pathway EPZ and each state within the 
ingestion exposure pathway EPZ. If the full participation exercise 
is conducted more than 1 year prior to issuance of an operating 
licensee for full power, an exercise which tests the licensee's 
onsite emergency plans must be conducted within one year before 
issuance of an operating license for full power. This exercise need 
not have State or local government participation.
    (ii) For a combined license issued under part 52 of this 
chapter, this exercise must be conducted within two years of the 
scheduled date for initial loading of fuel. If the first full 
participation exercise is conducted more than one year before the 
scheduled date for initial loading of fuel, an exercise which tests 
the licensee's onsite emergency plans must be conducted within one 
year before the scheduled date for initial loading of fuel. This 
exercise need not have State or local government participation. If 
DHS identifies one or more deficiencies in the state of offsite 
emergency preparedness as the result of the first full participation 
exercise, or if the Commission finds that the state of emergency 
preparedness does not provide reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency, the provisions of Sec.  50.54(gg) apply.
    (iii) For a combined licensee issued under part 52 of this 
chapter, if the applicant currently has an operating reactor at the 
site, an exercise, either full or partial participation,\5\ shall be 
conducted for each subsequent reactor constructed on the site. This 
exercise may be incorporated in the exercise requirements of 
Sections IV.F.2.b. and c. in this appendix. If DHS identifies one or 
more deficiencies in the state of offsite emergency preparedness as 
the result of this exercise for the new reactor, or if the 
Commission finds that the state of emergency preparedness does not 
provide reasonable assurance that adequate protective measures can 
and will be taken in the event of a radiological emergency, the 
provisions of Sec.  50.54(gg) apply.
---------------------------------------------------------------------------

    \5\ Partial participation when used in conjunction with 
emergency preparedness exercises for a particular site means 
appropriate offsite authorities shall actively take part in the 
exercise sufficient to test direction and control functions; i.e., 
(a) protective action decision making related to emergency action 
levels, and (b) communication capabilities among affected State and 
local authorities and the licensee.
---------------------------------------------------------------------------

* * * * *
    c. Offsite plans for each site shall be exercised biennially 
with full participation by each offsite authority having a role 
under the radiological response plan. Where the offsite authority 
has a role under a radiological response plan for more than one 
site, it shall fully participate in one exercise every two years and 
shall, at least, partially participate in other offsite plan 
exercises in this period. If two different licensees whose licensed 
facilities are located either on the same site or on adjacent, 
contiguous sites, and that share most of the elements defining co-
located licensees,\6\ each licensee shall:
---------------------------------------------------------------------------

    \6\ Co-located licensees are two different licensees whose 
licensed facilities are located either on the same site or on 
adjacent, contiguous sites, and that share most of the following 
emergency planning and siting elements:
    a. Plume exposure and ingestion emergency planning zones;
    b. Offsite governmental authorities;
    c. Offsite emergency response organizations;
    d. Public notification system; and/or
    e. Emergency facilities.
---------------------------------------------------------------------------

    (1) Conduct an exercise biennially of its onsite emergency plan; 
and
    (2) Participate quadrennially in an offsite biennial full or 
partial participation exercise; and
    (3) Conduct emergency preparedness activities and interactions 
in the years between its participation in the offsite full or 
partial participation exercise with offsite authorities, to test and 
maintain interface among the affected State and local authorities 
and the licensee. Co-located licensees shall also participate in 
emergency preparedness activities and interaction with offsite 
authorities for the period between exercises.
* * * * *

V. Implementing Procedures

    No less than 180 days before the scheduled issuance of an 
operating license for a nuclear power reactor or a license to 
possess nuclear material, or the scheduled date for initial loading 
of fuel for a combined license under part 52 of this chapter, the 
applicant's or licensee's detailed implementing procedures for its 
emergency plan shall be submitted to the Commission as specified in 
Sec.  50.4. Licensees who are authorized to operate a nuclear power 
facility shall submit any changes to the emergency plan or 
procedures to the Commission, as specified in Sec.  50.4, within 30 
days of such changes.
* * * * *

0
115. In Appendix I to Part 50, the first paragraphs of Sections I, II, 
IV, and V are revised to read as follows:

Appendix I to Part 50--Numerical Guides for Design Objectives and 
Limiting Conditions for Operation To Meet the Criterion ``as Low as Is 
Reasonably Achievable'' for Radioactive Material in Light-Water-Cooled 
Nuclear Power Reactor Effluents

    SECTION I. Introduction. Section 50.34a provides that an 
application for a construction permit shall include a description of 
the preliminary design of equipment to be installed to maintain 
control over radioactive materials in gaseous and liquid effluents 
produced during normal conditions, including expected occurrences. 
In the case of an application filed on or after January 2, 1971, the 
application must also identify the design objectives, and the means 
to be employed, for keeping levels of radioactive material in 
effluents to unrestricted areas as low as practicable. Sections 
52.47, 52.79, 52.137, and 52.157 of this chapter provide that 
applications for design certification, combined license, design 
approval, or manufacturing license, respectively, shall include a 
description of the equipment and procedures for the control of 
gaseous and liquid effluents and for the maintenance and use of 
equipment installed in radioactive waste systems.
* * * * *
    SECTION II. Guides on design objectives for light-water-cooled 
nuclear power reactors licensed under 10 CFR part 50 or part 52 of 
this chapter. The guides on design objectives set forth in this 
section may be used by an applicant for a construction permit as 
guidance in meeting the requirements of Sec.  50.34a(a), or by an 
applicant for a combined license under part 52 of this chapter as 
guidance in meeting the requirements of Sec.  50.34a(d), or by an 
applicant for a design approval, a design certification, or a 
manufacturing license as guidance in meeting the requirements of 
Sec.  50.34a(e). The applicant shall provide reasonable assurance 
that the following design objectives will be met.
* * * * *
    SECTION IV. Guides on technical specifications for limiting 
conditions for operation for light-water-cooled nuclear power 
reactors licensed under 10 CFR part 50 or part 52 of this chapter. 
The guides on limiting conditions for operation for light-water-
cooled nuclear power reactors set forth below may be used by an 
applicant for an operating license under this part or a design 
certification or combined license under part 52 of this chapter, or 
a licensee who has submitted a certification of permanent cessation 
of operations under Sec.  50.82(a)(1) or Sec.  52.110 of this 
chapter as guidance in developing technical specifications under 
Sec.  50.36a(a) to keep levels of radioactive materials in effluents 
to unrestricted areas as low as is reasonably achievable.
* * * * *

[[Page 49508]]

    SECTION V. Effective dates. A. The guides for limiting 
conditions for operation set forth in this appendix shall be 
applicable in any case in which an application was filed on or after 
January 2, 1971, for a construction permit for a light-water-cooled 
nuclear power reactor under this part, or a design certification, a 
combined license, or a manufacturing license for a light-water-
cooled nuclear power reactor under part 52 of this chapter.
* * * * *

0
116. In Appendix J to Part 50 in Option A, Section I, and paragraph 
II.K are revised and in Option B, Section I, and paragraphs V.B.2 and 3 
are revised to read as follows:

Appendix J to Part 50--Primary Reactor Containment Leakage Testing for 
Water-Cooled Reactors

* * * * *

Option A--Prescriptive Requirements

* * * * *

I. Introduction

    One of the conditions of all operating licenses under this part 
and combined licenses under part 52 of this chapter for water-cooled 
power reactors as specified in Sec.  50.54(o) is that primary 
reactor containments shall meet the containment leakage test 
requirements set forth in this appendix. These test requirements 
provide for preoperational and periodic verification by tests of the 
leak-tight integrity of the primary reactor containment, and systems 
and components which penetrate containment of water-cooled power 
reactors, and establish the acceptance criteria for these tests. The 
purposes of the tests are to assure that (a) leakage through the 
primary reactor containment and systems and components penetrating 
primary containment shall not exceed allowable leakage rate values 
as specified in the technical specifications or associated bases; 
and (b) periodic surveillance of reactor containment penetrations 
and isolation valves is performed so that proper maintenance and 
repairs are made during the service life of the containment, and 
systems and components penetrating primary containment. These test 
requirements may also be used for guidance in establishing 
appropriate containment leakage test requirements in technical 
specifications or associated bases for other types of nuclear power 
reactors.

II. * * *

    K. La (percent/24 hours) means the maximum allowable leakage 
rate at pressure Pa as specified for preoperational tests in the 
technical specifications or associated bases, and as specified for 
periodic tests in the operating license or combined license, 
including the technical specifications in any referenced design 
certification or manufactured reactor used at the facility.
* * * * *

Option B--Performance-Based Requirements

* * * * *

I. Introduction

    One of the conditions required of all operating licenses and 
combined licenses for light-water-cooled power reactors as specified 
in Sec.  50.54(o) is that primary reactor containments meet the 
leakage-rate test requirements in either Option A or B of this 
appendix. These test requirements ensure that (a) leakage through 
these containments or systems and components penetrating these 
containments does not exceed allowable leakage rates specified in 
the technical specifications; and (b) integrity of the containment 
structure is maintained during its service life. Option B of this 
appendix identifies the performance-based requirements and criteria 
for preoperational and subsequent periodic leakage-rate testing.\3\
---------------------------------------------------------------------------

    \3\ Specific guidance concerning a performance-based leakage-
test program, acceptable leakage-rate test methods, procedures, and 
analyses that may be used to implement these requirements and 
criteria are provided in Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program.''
---------------------------------------------------------------------------

* * * * *

V. * * *

    B. * * *
    2. A licensee or applicant for an operating license under this 
part or a combined license under part 52 of this chapter may adopt 
Option B, or parts thereof, as specified in Section V.A of this 
appendix, by submitting its implementation plan and request for 
revision to technical specifications (see paragraph B.3 of this 
section) to the Director of the Office of Nuclear Reactor Regulation 
or the Director of the Office of New Reactors, as appropriate.
    3. The regulatory guide or other implementation document used by 
a licensee or applicant for an operating license under this part or 
a combined license under part 52 of this chapter to develop a 
performance-based leakage-testing program must be included, by 
general reference, in the plant technical specifications. The 
submittal for technical specification revisions must contain 
justification, including supporting analyses, if the licensee 
chooses to deviate from methods approved by the Commission and 
endorsed in a regulatory guide.
* * * * *

Appendix M to Part 50 [Removed]

0
117. Appendix M to Part 50 is removed and reserved.

0
118. The heading for appendix N to part 50 is revised to read as 
follows:

Appendix N to Part 50--Standardization of Nuclear Power Plant Designs: 
Permits To Construct and Licenses To Operate Nuclear Power Reactors of 
Identical Design at Multiple Sites

Appendix O to Part 50 [Removed]

0
119. Appendix O to Part 50 is removed and reserved.

0
120. In Appendix S to Part 50, the first paragraph titled ``General 
Information,'' Section I(a), and Section III are revised to read as 
follows:

Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear 
Power Plants

General Information

    This appendix applies to applicants for a construction permit or 
operating license under part 50, or a design certification, combined 
license, design approval, or manufacturing license under part 52 of 
this chapter, on or after January 10, 1997. However, for either an 
operating license applicant or holder whose construction permit was 
issued before January 10, 1997, the earthquake engineering criteria 
in Section VI of appendix A to 10 CFR part 100 continue to apply. 
Paragraphs IV.a.1.i, IV.a.1.ii, IV.4.b, and IV.4.c of this appendix 
apply to applicants for an early site permit under part 52.

I. Introduction

    (a) Each applicant for a construction permit, operating license, 
design certification, combined license, design approval, or 
manufacturing license is required by Sec. Sec.  50.34(a)(12), 
50.34(b)(10), or 10 CFR 52.47, 52.79, 52.137, or 52.157, and General 
Design Criterion 2 of appendix A to this part, to design nuclear 
power plant structures, systems, and components important to safety 
to withstand the effects of natural phenomena, such as earthquakes, 
without loss of capability to perform their safety functions. Also, 
as specified in Sec.  50.54(ff), nuclear power plants that have 
implemented the earthquake engineering criteria described herein 
must shut down if the criteria in paragraph IV(a)(3) of this 
appendix are exceeded.
* * * * *

III. Definitions

    As used in these criteria:
    Combined license means a combined construction permit and 
operating license with conditions for a nuclear power facility 
issued under subpart C of part 52 of this chapter.
    Design Approval means an NRC staff approval, issued under 
subpart E of part 52 of this chapter, of a final standard design for 
a nuclear power reactor of the type described in 10 CFR 50.22.
    Design Certification means a Commission approval, issued under 
subpart B of part 52 of this chapter, of a standard design for a 
nuclear power facility.
    Manufacturing license means a license, issued under subpart F of 
part 52 of this chapter, authorizing the manufacture of nuclear 
power reactors but not their installation into facilities located at 
the sites on which the facilities are to be operated.
    Operating basis earthquake ground motion (OBE) is the vibratory 
ground motion for which those features of the nuclear power plant 
necessary for continued operation without undue risk to the health 
and safety of the public will remain functional. The operating basis 
earthquake ground motion is only associated with plant shutdown and 
inspection unless specifically selected by the applicant as a design 
input.

[[Page 49509]]

    Response spectrum is a plot of the maximum responses 
(acceleration, velocity, or displacement) of idealized single-
degree-of-freedom oscillators as a function of the natural 
frequencies of the oscillators for a given damping value. The 
response spectrum is calculated for a specified vibratory motion 
input at the oscillators' supports.
    Safe-shutdown earthquake ground motion (SSE) is the vibratory 
ground motion for which certain structures, systems, and components 
must be designed to remain functional.
    Structures, systems, and components required to withstand the 
effects of the safe-shutdown earthquake ground motion or surface 
deformation are those necessary to assure:
    (1) The integrity of the reactor coolant pressure boundary;
    (2) The capability to shut down the reactor and maintain it in a 
safe-shutdown condition; or
    (3) The capability to prevent or mitigate the consequences of 
accidents that could result in potential offsite exposures 
comparable to the guideline exposures of Sec.  50.34(a)(1).
    Surface deformation is distortion of geologic strata at or near 
the ground surface by the processes of folding or faulting as a 
result of various earth forces. Tectonic surface deformation is 
associated with earthquake processes.
* * * * *

PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC 
LICENSING AND RELATED REGULATORY FUNCTIONS

0
121. 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, 1244 (42 U.S.C. 5841, 
5842); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note). Subpart A 
also issued under National Environmental Policy Act of 1969, secs. 
102, 104, 105, 83 Stat. 853-854, as amended (42 U.S.C. 4332, 4334, 
4335); and Pub. L. 95-604, Title II, 92 Stat. 3033-3041; and sec. 
193, Pub. L. 101-575, 104 Stat. 2835 (42 U.S.C. 2243). Sections 
51.20, 51.30, 51.60, 51.80, and 51.97 also issued under secs. 135, 
141, Pub. L. 97-425, 96 Stat. 2232, 2241, and sec. 148, Pub. L. 100-
203, 101 Stat. 1330-223 (42 U.S.C. 10155, 10161, 10168). Section 
51.22 also issued under sec. 274, 73 Stat. 688, as amended by 92 
Stat. 3036-3038 (42 U.S.C. 2021) and under Nuclear Waste Policy Act 
of 1982, sec. 121, 96 Stat. 2228 (42 U.S.C. 10141). Sections 51.43, 
51.67, and 51.109 also issued under Nuclear Waste Policy Act of 
1982, sec. 114(f), 96 Stat. 2216, as amended (42 U.S.C. 10134(f)).


0
122. In Sec.  51.17, paragraph (b) is revised to read as follows:


Sec.  51.17  Information collection requirements; OMB approval.

* * * * *
    (b) The approved information collection requirements in this part 
appear in Sec. Sec.  51.6, 51.16, 51.41, 51.45, 51.50, 51.51, 51.52, 
51.53, 51.54, 51.55, 51.58, 51.60, 51.61, 51.62, 51.66, 51.68, and 
51.69.

0
123. In Sec.  51.20, paragraphs (b)(1) and (b)(2) are revised, and 
paragraph (b)(6) is removed and reserved.
    The revisions read as follows:


Sec.  51.20  Criteria for and identification of licensing and 
regulatory actions requiring environmental impact statements.

* * * * *
    (b)* * *
    (1) Issuance of a limited work authorization or a permit to 
construct a nuclear power reactor, testing facility, or fuel 
reprocessing plant under part 50 of this chapter, or issuance of an 
early site permit under part 52 of this chapter.
    (2) Issuance or renewal of a full power or design capacity license 
to operate a nuclear power reactor, testing facility, or fuel 
reprocessing plant under part 50 of this chapter, or a combined license 
under part 52 of this chapter.
* * * * *
    (6) [Reserved]
* * * * *

0
124. In Sec.  51.22, the introductory text of paragraph (c)(3), 
paragraphs (c)(3)(i) and (c)(9), the introductory text of paragraphs 
(c)(10) and (c)(12), and paragraph (c)(17) are revised, and paragraphs 
(c)(22) and (c)(23) are added to read as follows:


Sec.  51.22  Criterion for categorical exclusion; identification of 
licensing and regulatory actions eligible for categorical exclusion or 
otherwise not requiring environmental review.

* * * * *
    (c) * * *
    (3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 39, 40, 50, 51, 
52, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this chapter 
which relate to--
    (i) Procedures for filing and reviewing applications for licenses 
or construction permits or early site permits or other forms of 
permission or for amendments to or renewals of licenses or construction 
permits or early site permits or other forms of permission;
* * * * *
    (9) Issuance of an amendment to a permit or license for a reactor 
under part 50 or part 52 of this chapter, which changes a requirement 
with respect to installation or use of a facility component located 
within the restricted area, as defined in part 20 of this chapter, or 
which changes an inspection or a surveillance requirement, provided 
that--
    (i) The amendment involves no significant hazards consideration;
    (ii) There is no significant change in the types or significant 
increase in the amounts of any effluents that may be released offsite; 
and
    (iii) There is no significant increase in individual or cumulative 
occupational radiation exposure.
    (10) Issuance of an amendment to a permit or license under parts 
30, 31, 32, 33, 34, 35, 36, 39, 40, 50, 52, 60, 61, 63, 70, or part 72 
of this chapter which--
* * * * *
    (12) Issuance of an amendment to a license under parts 50, 52, 60, 
61, 63, 70, 72, or 75 of this chapter relating solely to safeguards 
matters (i.e., protection against sabotage or loss or diversion of 
special nuclear material) or issuance of an approval of a safeguards 
plan submitted under parts 50, 52, 70, 72, and 73 of this chapter, 
provided that the amendment or approval does not involve any 
significant construction impacts. These amendments and approvals are 
confined to--
* * * * *
    (17) Issuance of an amendment to a permit or license under parts 
30, 40, 50, 52, or part 70 of this chapter which deletes any limiting 
condition of operation or monitoring requirement based on or applicable 
to any matter subject to the provisions of the Federal Water Pollution 
Control Act.
* * * * *
    (22) Issuance of a standard design approval under part 52 of this 
chapter.
    (23) The Commission finding for a combined license under Sec.  
52.103(g) of this chapter.
* * * * *

0
125. In Sec.  51.23 paragraphs (b) and (c) are revised to read as 
follows:


Sec.  51.23  Temporary storage of spent fuel after cessation of reactor 
operation--generic determination of no significant environmental 
impact.

* * * * *
    (b) Accordingly, as provided in Sec. Sec.  51.30(b), 51.53, 51.61, 
51.80(b), 51.95, and 51.97(a), and within the scope of the generic 
determination in paragraph (a) of this section, no discussion of any 
environmental impact of spent fuel storage in reactor facility storage 
pools or independent spent fuel storage installations (ISFSI) for the 
period following the term of the reactor operating license or 
amendment, reactor combined license or amendment, or initial ISFSI 
license or amendment for which application is made, is required in any 
environmental report, environmental impact statement, environmental 
assessment, or other analysis prepared in connection with the issuance 
or amendment of an

[[Page 49510]]

operating license for a nuclear power reactor under parts 50 and 54 of 
this chapter, or issuance or amendment of a combined license for a 
nuclear power reactor under parts 52 and 54 of this chapter, or the 
issuance of an initial license for storage of spent fuel at an ISFSI, 
or any amendment thereto.
    (c) This section does not alter any requirements to consider the 
environmental impacts of spent fuel storage during the term of a 
reactor operating license or combined license, or a license for an 
ISFSI in a licensing proceeding.


0
126. In Sec.  51.26, a new paragraph (d) is added to read as follows:


Sec.  51.26  Requirement to publish notice of intent and conduct 
scoping process.

* * * * *
    (d) Whenever the appropriate NRC staff director determines that a 
supplement to an environmental impact statement will be prepared by the 
NRC, a notice of intent will be prepared as provided in Sec.  51.27, 
and will be published in the Federal Register as provided in Sec.  
51.116. The NRC staff need not conduct a scoping process (see 
Sec. Sec.  51.27, 51.28, and 51.29), provided, however, that if scoping 
is conducted, then the scoping must be directed at matters to be 
addressed in the supplement. If scoping is conducted in a proceeding 
for a combined license referencing an early site permit under part 52, 
then the scoping must be directed at matters to be addressed in the 
supplement as described in Sec.  51.92(e).


0
127. In Sec.  51.27, the introductory text of paragraph (a) is revised, 
and a new paragraph (b) is added to read as follows:


Sec.  51.27  Notice of intent.

    (a) The notice of intent required by Sec.  51.26(a) shall:
* * * * *
    (b) The notice of intent required by Sec.  51.26(d) shall:
    (1) State that a supplement to a final environmental impact 
statement will be prepared in accordance with Sec.  51.72 or Sec.  
51.92. For a combined license application that references an early site 
permit, the supplement to the early site permit environmental impact 
statement will be prepared in accordance with Sec.  51.92(e);
    (2) Describe the proposed action and, to the extent required, 
possible alternatives. For the case of a combined license referencing 
an early site permit, identify the proposed action as the issuance of a 
combined license for the construction and operation of a nuclear power 
plant as described in the combined license application at the site 
described in the early site permit referenced in the combined license 
application;
    (3) Identify the environmental report prepared by the applicant and 
information on where copies are available for public inspection;
    (4) Describe the matters to be addressed in the supplement to the 
final environmental impact statement;
    (5) Describe any proposed scoping process that the NRC staff may 
conduct, including the role of participants, whether written comments 
will be accepted, the last date for submitting comments and where 
comments should be sent, whether a public scoping meeting will be held, 
the time and place of any scoping meeting or when the time and place of 
the meeting will be announced; and
    (6) State the name, address, and telephone number of an individual 
in NRC who can provide information about the proposed action, the 
scoping process, and the supplement to the environmental impact 
statement.

0
128. In Sec.  51.29, the section heading and paragraph (a)(1) are 
revised to read as follows:


Sec.  51.29  Scoping-environmental impact statement and supplement to 
environmental impact statement.

    (a) * * *
    (1) Define the proposed action which is to be the subject of the 
statement or supplement. For environmental impact statements other than 
a supplement to an early site permit final environmental impact 
statement prepared for a combined license application, the provisions 
of 40 CFR 1502.4 will be used for this purpose. For a supplement to an 
early site permit final environmental impact statement prepared for a 
combined license application, the proposed action shall be as set forth 
in the relevant provisions of Sec.  51.92(e).
* * * * *

0
129. In Sec.  51.30, the introductory text of paragraph (a) is revised, 
and paragraphs (d) and (e) are added to read as follows:


Sec.  51.30  Environmental assessment.

    (a) An environmental assessment for proposed actions, other than 
those for a standard design certification under 10 CFR part 52 or a 
manufacturing license under part 52, shall identify the proposed action 
and include:
* * * * *
    (d) An environmental assessment for a standard design certification 
under subpart B of part 52 of this chapter must identify the proposed 
action, and will be limited to the consideration of the costs and 
benefits of severe accident mitigation design alternatives and the 
bases for not incorporating severe accident mitigation design 
alternatives in the design certification. An environmental assessment 
for an amendment to a design certification will be limited to the 
consideration of whether the design change which is the subject of the 
proposed amendment renders a severe accident mitigation design 
alternative previously rejected in the earlier environmental assessment 
to become cost beneficial, or results in the identification of new 
severe accident mitigation design alternatives, in which case the costs 
and benefits of new severe accident mitigation design alternatives and 
the bases for not incorporating new severe accident mitigation design 
alternatives in the design certification must be addressed.
    (e) An environmental assessment for a manufacturing license under 
subpart F of part 52 of this chapter must identify the proposed action, 
and will be limited to the consideration of the costs and benefits of 
severe accident mitigation design alternatives and the bases for not 
incorporating severe accident mitigation design alternatives in the 
manufacturing license. An environmental assessment for an amendment to 
a manufacturing license will be limited to consideration of whether the 
design change which is the subject of the proposed amendment either 
renders a severe accident mitigation design alternative previously 
rejected in an environmental assessment to become cost beneficial, or 
results in the identification of new severe accident mitigation design 
alternatives, in which case the costs and benefits of new severe 
accident mitigation design alternatives and the bases for not 
incorporating new severe accident mitigation design alternatives in the 
manufacturing license must be addressed. In either case, the 
environmental assessment will not address the environmental impacts 
associated with manufacturing the reactor under the manufacturing 
license.

0
130. Section 51.31 is revised to read as follows:


Sec.  51.31  Determinations based on environmental assessment.

    (a) General. Upon completion of an environmental assessment for 
proposed actions other than those involving a standard design 
certification or a manufacturing license under part 52 of this chapter, 
the appropriate NRC staff director will determine whether to prepare an 
environmental impact statement or a finding of no significant

[[Page 49511]]

impact on the proposed action. As provided in Sec.  51.33, a 
determination to prepare a draft finding of no significant impact may 
be made.
    (b) Standard design certification. (1) For actions involving the 
issuance or amendment of a standard design certification, the 
Commission shall prepare a draft environmental assessment for public 
comment as part of the proposed rule. The proposed rule must state 
that:
    (i) The Commission has determined in Sec.  51.32 that there is no 
significant environmental impact associated with the issuance of the 
standard design certification or its amendment, as applicable; and
    (ii) Comments on the environmental assessment will be limited to 
the consideration of SAMDAs as required by Sec.  51.30(d).
    (2) The Commission will prepare a final environmental assessment 
following the close of the public comment period for the proposed 
standard design certification.
    (c) Manufacturing license. (1) Upon completion of the environmental 
assessment for actions involving issuance or amendment of a 
manufacturing license (manufacturing license environmental assessment), 
the appropriate NRC staff director will determine the costs and 
benefits of severe accident mitigation design alternatives and the 
bases for not incorporating severe accident mitigation design 
alternatives in the design of the reactor to be manufactured under the 
manufacturing license. The NRC staff director may determine to prepare 
a draft environmental assessment.
    (2) The manufacturing license environmental assessment must state 
that:
    (i) The Commission has determined in Sec.  51.32 that there is no 
significant environmental impact associated with the issuance of a 
manufacturing license or an amendment to a manufacturing license, as 
applicable;
    (ii) The environmental assessment will not address the 
environmental impacts associated with manufacturing the reactor under 
the manufacturing license; and
    (iii) Comments on the environmental assessment will be limited to 
the consideration of severe accident mitigation design alternatives as 
required by Sec.  51.30(e).
    (3) If the NRC staff director makes a determination to prepare and 
issue a draft environmental assessment for public review and comment 
before making a final determination on the manufacturing license 
application, the assessment will be marked, ``Draft.'' The NRC notice 
of availability on the draft environmental assessment will include a 
request for comments which specifies where comments should be submitted 
and when the comment period expires. The notice will state that copies 
of the environmental assessment and any related environmental documents 
are available for public inspection and where inspections can be made. 
A copy of the final environmental assessment will be sent to the U.S. 
Environmental Protection Agency, the applicant, any party to a 
proceeding, each commenter, and any other Federal, State, and local 
agencies, and Indian tribes, State, regional, and metropolitan 
clearinghouses expressing an interest in the action. Additional copies 
will be made available in accordance with Sec.  51.123.
    (4) When a hearing is held under the regulations in part 2 of this 
chapter on the proposed issuance of the manufacturing license or 
amendment, the NRC staff director will prepare a final environmental 
assessment which may be subject to modification as a result of review 
and decision as appropriate to the nature and scope of the proceeding.
    (5) Only a party admitted into the proceeding with respect to a 
contention on the environmental assessment, or an entity participating 
in the proceeding pursuant to Sec.  2.315(c) of this chapter, may take 
a position and offer evidence on the matters within the scope of the 
environmental assessment.

0
131. In Sec.  51.32, paragraph (b) is added to read as follows:


Sec.  51.32  Finding of no significant impact.

* * * * *
    (b) The Commission finds that there is no significant environmental 
impact associated with the issuance of:
    (1) A standard design certification under subpart B of part 52 of 
this chapter;
    (2) An amendment to a design certification;
    (3) A manufacturing license under subpart F of part 52 of this 
chapter; or
    (4) An amendment to a manufacturing license.

0
132. In Sec.  51.45, paragraphs (a) and (c) are revised to read as 
follows:


Sec.  51.45  Environmental report.

    (a) General. As required by Sec. Sec.  51.50, 51.53, 51.54, 51.55, 
51.60, 51.61, 51.62, or 51.68, as appropriate, each applicant or 
petitioner for rulemaking shall submit with its application or petition 
for rulemaking one signed original of a separate document entitled 
``Applicant's'' or ``Petitioner's Environmental Report,'' as 
appropriate. An applicant or petitioner for rulemaking may submit a 
supplement to an environmental report at any time.
* * * * *
    (c) Analysis. The environmental report shall include an analysis 
that considers and balances the environmental effects of the proposed 
action, the environmental impacts of alternatives to the proposed 
action, and alternatives available for reducing or avoiding adverse 
environmental effects. Except for environmental reports prepared at the 
early site permit stage under Sec.  51.50(b), or environmental reports 
prepared at the license renewal stage under Sec.  51.53(c), the 
analysis in the environmental report should also include consideration 
of the economic, technical, and other benefits and costs of the 
proposed action and of alternatives. Environmental reports prepared at 
the license renewal stage under Sec.  51.53(c) need not discuss the 
economic or technical benefits and costs of either the proposed action 
or alternatives except insofar as these benefits and costs are either 
essential for a determination regarding the inclusion of an alternative 
in the range of alternatives considered or relevant to mitigation. In 
addition, environmental reports prepared under Sec.  51.53(c) need not 
discuss issues not related to the environmental effects of the proposed 
action and its alternatives. The analyses for environmental reports 
shall, to the fullest extent practicable, quantify the various factors 
considered. To the extent that there are important qualitative 
considerations or factors that cannot be quantified, those 
considerations or factors shall be discussed in qualitative terms. The 
environmental report should contain sufficient data to aid the 
Commission in its development of an independent analysis.
* * * * *

0
133. Section 51.50 is revised to read as follows:


Sec.  51.50  Environmental report--construction permit, early site 
permit, or combined license stage.

    (a) Construction permit stage. Each applicant for a permit to 
construct a production or utilization facility covered by Sec.  51.20 
shall submit with its application a separate document, entitled 
``Applicant's Environmental Report--Construction Permit Stage,'' which 
shall contain the information specified in Sec. Sec.  51.45, 51.51, and 
51.52. Each environmental report shall identify procedures for 
reporting and keeping records of environmental data, and any conditions 
and monitoring requirements for protecting the non-aquatic

[[Page 49512]]

environment, proposed for possible inclusion in the license as 
environmental conditions in accordance with Sec.  50.36b of this 
chapter.
    (b) Early site permit stage. Each applicant for an early site 
permit shall submit with its application a separate document, entitled 
``Applicant's Environmental Report--Early Site Permit Stage,'' which 
shall contain the information specified in Sec. Sec.  51.45, 51.51, and 
51.52, as modified in this paragraph.
    (1) The environmental report must include an evaluation of 
alternative sites to determine whether there is any obviously superior 
alternative to the site proposed.
    (2) The environmental report may address one or more of the 
environmental effects of construction and operation of a reactor, or 
reactors, which have design characteristics that fall within the site 
characteristics and design parameters for the early site permit 
application, provided however, that the environmental report must 
address all environmental effects of construction and operation 
necessary to determine whether there is any obviously superior 
alternative to the site proposed. The environmental report need not 
include an assessment of the economic, technical, or other benefits 
(for example, need for power) and costs of the proposed action or an 
evaluation of alternative energy sources.
    (3) For other than light-water-cooled nuclear power reactors, the 
environmental report must contain the basis for evaluating the 
contribution of the environmental effects of fuel cycle activities for 
the nuclear power reactor.
    (4) Each environmental report must identify the procedures for 
reporting and keeping records of environmental data, and any conditions 
and monitoring requirements for protecting the non-aquatic environment, 
proposed for possible inclusion in the license as environmental 
conditions in accordance with Sec.  50.36b of this chapter.
    (c) Combined license stage. Each applicant for a combined license 
shall submit with its application a separate document, entitled 
``Applicant's Environmental Report--Combined License Stage.'' Each 
environmental report shall contain the information specified in 
Sec. Sec.  51.45, 51.51, and 51.52, as modified in this paragraph. For 
other than light-water-cooled nuclear power reactors, the environmental 
report shall contain the basis for evaluating the contribution of the 
environmental effects of fuel cycle activities for the nuclear power 
reactor. Each environmental report shall identify procedures for 
reporting and keeping records of environmental data, and any conditions 
and monitoring requirements for protecting the non-aquatic environment, 
proposed for possible inclusion in the license as environmental 
conditions in accordance with Sec.  50.36b of this chapter. The 
combined license environmental report may reference information 
contained in a final environmental document previously prepared by the 
NRC staff.
    (1) Application referencing an early site permit. If the combined 
license application references an early site permit, then the 
``Applicant's Environmental Report--Combined License Stage'' need not 
contain information or analyses submitted to the Commission in 
``Applicant's Environmental Report--Early Site Permit Stage,'' or 
resolved in the Commission's early site permit environmental impact 
statement, but must contain, in addition to the environmental 
information and analyses otherwise required:
    (i) Information to demonstrate that the design of the facility 
falls within the site characteristics and design parameters specified 
in the early site permit;
    (ii) Information to resolve any significant environmental issue 
that was not resolved in the early site permit proceeding;
    (iii) Any new and significant information for issues related to the 
impacts of construction and operation of the facility that were 
resolved in the early site permit proceeding;
    (iv) A description of the process used to identify new and 
significant information regarding the NRC's conclusions in the early 
site permit environmental impact statement. The process must use a 
reasonable methodology for identifying such new and significant 
information; and
    (v) A demonstration that all environmental terms and conditions 
that have been included in the early site permit will be satisfied by 
the date of issuance of the combined license. Any terms or conditions 
of the early site permit that could not be met by the time of issuance 
of the combined license, must be set forth as terms or conditions of 
the combined license.
    (2) Application referencing standard design certification. If the 
combined license references a standard design certification, then the 
combined license environmental report may incorporate by reference the 
environmental assessment previously prepared by the NRC for the 
referenced design certification. If the design certification 
environmental assessment is referenced, then the combined license 
environmental report must contain information to demonstrate that the 
site characteristics for the combined license site fall within the site 
parameters in the design certification environmental assessment.
    (3) Application referencing a manufactured reactor. If the combined 
license application proposes to use a manufactured reactor, then the 
combined license environmental report may incorporate by reference the 
environmental assessment previously prepared by the NRC for the 
underlying manufacturing license. If the manufacturing license 
environmental assessment is referenced, then the combined license 
environmental report must contain information to demonstrate that the 
site characteristics for the combined license site fall within the site 
parameters in the manufacturing license environmental assessment. The 
environmental report need not address the environmental impacts 
associated with manufacturing the reactor under the manufacturing 
license.

0
134. In Sec.  51.51 paragraph (a) is revised to read as follows:


Sec.  51.51  Uranium fuel cycle environmental data--Table S-3.

    (a) Under Sec.  51.50, every environmental report prepared for the 
construction permit stage or early site permit stage or combined 
license stage of a light-water-cooled nuclear power reactor, and 
submitted on or after September 4, 1979, shall take Table S-3, Table of 
Uranium Fuel Cycle Environmental Data, as the basis for evaluating the 
contribution of the environmental effects of uranium mining and 
milling, the production of uranium hexafluoride, isotopic enrichment, 
fuel fabrication, reprocessing of irradiated fuel, transportation of 
radioactive materials and management of low-level wastes and high-level 
wastes related to uranium fuel cycle activities to the environmental 
costs of licensing the nuclear power reactor. Table S-3 shall be 
included in the environmental report and may be supplemented by a 
discussion of the environmental significance of the data set forth in 
the table as weighed in the analysis for the proposed facility.
* * * * *

0
135. In Sec.  51.52, the introductory paragraph is revised to read as 
follows:


Sec.  51.52  Environmental effects of transportation of fuel and 
waste--Table S-4.

    Under Sec.  51.50, every environmental report prepared for the 
construction permit stage or early site permit stage or combined 
license stage of a light-water-

[[Page 49513]]

cooled nuclear power reactor, and submitted after February 4, 1975, 
shall contain a statement concerning transportation of fuel and 
radioactive wastes to and from the reactor. That statement shall 
indicate that the reactor and this transportation either meet all of 
the conditions in paragraph (a) of this section or all of the 
conditions of paragraph (b) of this section.
* * * * *

0
136. In Sec.  51.53, paragraph (a) and the introductory text of 
paragraph (c)(3) are revised to read as follows:


Sec.  51.53  Postconstruction environmental reports.

    (a) General. Any environmental report prepared under the provisions 
of this section may incorporate by reference any information contained 
in a prior environmental report or supplement thereto that relates to 
the production or utilization facility or site, or any information 
contained in a final environmental document previously prepared by the 
NRC staff that relates to the production or utilization facility or 
site. Documents that may be referenced include, but are not limited to, 
the final environmental impact statement; supplements to the final 
environmental impact statement, including supplements prepared at the 
license renewal stage; NRC staff-prepared final generic environmental 
impact statements; and environmental assessments and records of 
decisions prepared in connection with the construction permit, 
operating license, early site permit, combined license and any license 
amendment for that facility.
* * * * *
    (c) * * *
    (3) For those applicants seeking an initial renewed license and 
holding an operating license, construction permit, or combined license 
as of June 30, 1995, the environmental report shall include the 
information required in paragraph (c)(2) of this section subject to the 
following conditions and considerations:
* * * * *

0
137. Section 51.54 is revised to read as follows:


Sec.  51.54  Environmental report--manufacturing license.

    (a) Each applicant for a manufacturing license under subpart F of 
part 52 of this chapter shall submit with its application a separate 
document entitled, ``Applicant's Environmental Report--Manufacturing 
License.'' The environmental report must address the costs and benefits 
of severe accident mitigation design alternatives, and the bases for 
not incorporating severe accident mitigation design alternatives into 
the design of the reactor to be manufactured. The environmental report 
need not address the environmental impacts associated with 
manufacturing the reactor under the manufacturing license, the benefits 
and impacts of utilizing the reactor in a nuclear power plant, or an 
evaluation of alternative energy sources.
    (b) Each applicant for an amendment to a manufacturing license 
shall submit with its application a separate document entitled, 
``Applicant's Supplemental Environmental Report--Amendment to 
Manufacturing License.'' The environmental report must address whether 
the design change which is the subject of the proposed amendment either 
renders a severe accident mitigation design alternative previously 
rejected in an environmental assessment to become cost beneficial, or 
results in the identification of new severe accident mitigation design 
alternatives that may be reasonably incorporated into the design of the 
manufactured reactor. The environmental report need not address the 
environmental impacts associated with manufacturing the reactor under 
the manufacturing license.

0
138. Section 51.55 is redesignated as Sec.  51.58, and is revised to 
read as follows:


Sec.  51.58  Environmental report-number of copies; distribution.

    (a) Each applicant for a license or permit to site, construct, 
manufacture, or operate a production or utilization facility covered by 
Sec. Sec.  51.20(b)(1), (b)(2), (b)(3), or (b)(4), each applicant for 
renewal of an operating or combined license for a nuclear power plant, 
each applicant for a license amendment authorizing the decommissioning 
of a production or utilization facility covered by Sec.  51.20, and 
each applicant for a license or license amendment to store spent fuel 
at a nuclear power plant after expiration of the operating license or 
combined license for the nuclear power plant shall submit a copy to the 
Director of the Office of Nuclear Reactor Regulation, the Director of 
the Office of New Reactors, the Director of the Office of Nuclear 
Material Safety and Safeguards, as appropriate, of an environmental 
report or any supplement to an environmental report. These reports must 
be sent either by mail addressed: ATTN: Document Control Desk; U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand 
delivery to the NRC's offices at 11555 Rockville Pike, Rockville, 
Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time; 
or, where practicable, by electronic submission, for example, via 
Electronic Information Exchange, or CD-ROM. Electronic submissions must 
be made in a manner that enables the NRC to receive, read, 
authenticate, distribute, and archive the submission, and process and 
retrieve it a single page at a time. Detailed guidance on making 
electronic submissions can be obtained by visiting the NRC's Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 
415-0439, by e-mail to <a href="/cdn-cgi/l/email-protection#97d2ded2d7f9e5f4b9f0f8e1"><span class="__cf_email__" data-cfemail="9bded2dedbf5e9f8b5fcf4ed">[email protected]</span></a>, or by writing the Office of 
Information Services, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001. The guidance discusses, among other topics, the formats 
the NRC can accept, the use of electronic signatures, and the treatment 
of nonpublic information. If the communication is on paper, the signed 
original must be sent. If a submission due date falls on a Saturday, 
Sunday, or Federal holiday, the next Federal working day becomes the 
official due date. The applicant shall maintain the capability to 
generate additional copies of the environmental report or any 
supplement to the environmental report for subsequent distribution to 
parties and Boards in the NRC proceedings; Federal, State, and local 
officials; and any affected Indian tribes, in accordance with written 
instructions issued by the Director of the Office of New Reactors, the 
Director of the Office of Nuclear Reactor Regulation, or the Director 
of the Office of Nuclear Material Safety and Safeguards, as 
appropriate.
    (b) Each applicant for a license to manufacture a nuclear power 
reactor, or for an amendment to a license to manufacture, seeking 
approval of the final design of the nuclear power reactor under subpart 
F of part 52 of this chapter, shall submit to the Commission an 
environmental report or any supplement to an environmental report in 
the manner specified in Sec.  50.3 of this chapter. The applicant shall 
maintain the capability to generate additional copies of the 
environmental report or any supplement to the environmental report for 
subsequent distribution to parties and Boards in the NRC proceeding; 
Federal, State, and local officials; and any affected Indian tribes, in 
accordance with written instructions issued by the Director of the 
Office of New Reactors or the Director of the Office of Nuclear Reactor 
Regulation.

0
139. Section 51.55 is added to read as follows:

[[Page 49514]]

Sec.  51.55  Environmental report--standard design certification.

    (a) Each applicant for a standard design certification under 
subpart B of part 52 of this chapter shall submit with its application 
a separate document entitled, ``Applicant's Environmental Report--
Standard Design Certification.'' The environmental report must address 
the costs and benefits of severe accident mitigation design 
alternatives, and the bases for not incorporating severe accident 
mitigation design alternatives in the design to be certified.
    (b) Each applicant for an amendment to a design certification shall 
submit with its application a separate document entitled, ``Applicant's 
Supplemental Environmental Report--Amendment to Standard Design 
Certification.'' The environmental report must address whether the 
design change which is the subject of the proposed amendment either 
renders a severe accident mitigation design alternative previously 
rejected in an environmental assessment to become cost beneficial, or 
results in the identification of new severe accident mitigation design 
alternatives that may be reasonably incorporated into the design 
certification.

0
140. Section 51.66 is revised to read as follows:


Sec.  51.66  Environmental report--number of copies; distribution.

    Each applicant for a license or other form of permission, or an 
amendment to or renewal of a license or other form of permission issued 
under parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70, and/or 72 of this 
chapter, and covered by Sec. Sec.  51.60(b)(1) through (6); or by 
Sec. Sec.  51.61 or 51.62 shall submit to the Director of Nuclear 
Material Safety and Safeguards an environmental report or any 
supplement to an environmental report in the manner specified in Sec.  
51.58(a). The applicant shall maintain the capability to generate 
additional copies of the environmental report or any supplement to the 
environmental report for subsequent distribution to Federal, State, and 
local officials, and any affected Indian tribes in accordance with 
written instructions issued by the Director of Nuclear Material Safety 
and Safeguards.

0
141. In Sec.  51.71 paragraph (d) and Footnote 3 are revised to read as 
follows:


Sec.  51.71  Draft environmental impact statement--contents.

* * * * *
    (d) Analysis. Unless excepted in this paragraph or Sec.  51.75, the 
draft environmental impact statement will include a preliminary 
analysis that considers and weighs the environmental effects of the 
proposed action; the environmental impacts of alternatives to the 
proposed action; and alternatives available for reducing or avoiding 
adverse environmental effects and consideration of the economic, 
technical, and other benefits and costs of the proposed action and 
alternatives and indicate what other interests and considerations of 
Federal policy, including factors not related to environmental quality 
if applicable, are relevant to the consideration of environmental 
effects of the proposed action identified under paragraph (a) of this 
section. The draft supplemental environmental impact statement prepared 
at the license renewal stage under Sec.  51.95(c) need not discuss the 
economic or technical benefits and costs of either the proposed action 
or alternatives except if benefits and costs are either essential for a 
determination regarding the inclusion of an alternative in the range of 
alternatives considered or relevant to mitigation. In addition, the 
supplemental environmental impact statement prepared at the license 
renewal stage need not discuss other issues not related to the 
environmental effects of the proposed action and associated 
alternatives. The draft supplemental environmental impact statement for 
license renewal prepared under Sec.  51.95(c) will rely on conclusions 
as amplified by the supporting information in the GEIS for issues 
designated as Category 1 in appendix B to subpart A of this part. The 
draft supplemental environmental impact statement must contain an 
analysis of those issues identified as Category 2 in appendix B to 
subpart A of this part that are open for the proposed action. The 
analysis for all draft environmental impact statements will, to the 
fullest extent practicable, quantify the various factors considered. To 
the extent that there are important qualitative considerations or 
factors that cannot be quantified, these considerations or factors will 
be discussed in qualitative terms. Consideration will be given to 
compliance with environmental quality standards and requirements that 
have been imposed by Federal, State, regional, and local agencies 
having responsibility for environmental protection, including 
applicable zoning and land-use regulations and water pollution 
limitations or requirements issued or imposed under the Federal Water 
Pollution Control Act. The environmental impact of the proposed action 
will be considered in the analysis with respect to matters covered by 
environmental quality standards and requirements irrespective of 
whether a certification or license from the appropriate authority has 
been obtained.\3\ While satisfaction of Commission standards and 
criteria pertaining to radiological effects will be necessary to meet 
the licensing requirements of the Atomic Energy Act, the analysis will, 
for the purposes of NEPA, consider the radiological effects of the 
proposed action and alternatives.
---------------------------------------------------------------------------

    \3\ Compliance with the environmental quality standards and 
requirements of the Federal Water Pollution Control Act (imposed by 
EPA or designated permitting states) is not a substitute for, and 
does not negate the requirement for NRC to weigh all environmental 
effects of the proposed action, including the degradation, if any, 
of water quality, and to consider alternatives to the proposed 
action that are available for reducing adverse effects. Where an 
environmental assessment of aquatic impact from plant discharges is 
available from the permitting authority, the NRC will consider the 
assessment in its determination of the magnitude of environmental 
impacts for striking an overall cost-benefit balance at the 
construction permit and operating license and early site permit and 
combined license stages, and in its determination of whether the 
adverse environmental impacts of license renewal are so great that 
preserving the option of license renewal for energy planning 
decision-makers would be unreasonable at the license renewal stage. 
When no such assessment of aquatic impacts is available from the 
permitting authority, NRC will establish on its own, or in 
conjunction with the permitting authority and other agencies having 
relevant expertise, the magnitude of potential impacts for striking 
an overall cost-benefit balance for the facility at the construction 
permit and operating license and early site permit and combined 
license stages, and in its determination of whether the adverse 
environmental impacts of license renewal are so great that 
preserving the option of license renewal for energy planning 
decision-makers would be unreasonable at the license renewal stage.
---------------------------------------------------------------------------

* * * * *

0
142. Section 51.75 is revised to read as follows:


Sec.  51.75  Draft environmental impact statement--construction permit, 
early site permit, or combined license.

    (a) Construction permit stage. A draft environmental impact 
statement relating to issuance of a construction permit for a 
production or utilization facility will be prepared in accordance with 
the procedures and measures described in Sec. Sec.  51.70, 51.71, 
51.72, and 51.73. The contribution of the environmental effects of the 
uranium fuel cycle activities specified in Sec.  51.51 shall be 
evaluated on the basis of impact values set forth in Table S-3, Table 
of Uranium Fuel Cycle Environmental Data, which shall be set out in the 
draft environmental impact statement. With the exception of radon-222 
and technetium-99 releases, no further discussion of fuel cycle release 
values and other numerical data that appear explicitly in the table 
shall be required.\5\

[[Page 49515]]

The impact statement shall take account of dose commitments and health 
effects from fuel cycle effluents set forth in Table S-3 and shall in 
addition take account of economic, socioeconomic, and possible 
cumulative impacts and other fuel cycle impacts as may reasonably 
appear significant.
    (b) Early site permit stage. A draft environmental impact statement 
relating to issuance of an early site permit for a production or 
utilization facility will be prepared in accordance with the procedures 
and measures described in Sec. Sec.  51.70, 51.71, 51.72, 51.73, and 
this section. The contribution of the environmental effects of the 
uranium fuel cycle activities specified in Sec.  51.51 shall be 
evaluated on the basis of impact values set forth in Table S-3, Table 
of Uranium Fuel Cycle Environmental Data, which shall be set out in the 
draft environmental impact statement. With the exception of radon-222 
and technetium-99 releases, no further discussion of fuel cycle release 
values and other numerical data that appear explicitly in the table 
shall be required.\5\ The impact statement shall take account of dose 
commitments and health effects from fuel cycle effluents set forth in 
Table S-3 and shall in addition take account of economic, 
socioeconomic, and possible cumulative impacts and other fuel cycle 
impacts as may reasonably appear significant. The draft environmental 
impact statement must include an evaluation of alternative sites to 
determine whether there is any obviously superior alternative to the 
site proposed. The draft environmental impact statement must also 
include an evaluation of the environmental effects of construction and 
operation of a reactor, or reactors, which have design characteristics 
that fall within the site characteristics and design parameters for the 
early site permit application, but only to the extent addressed in the 
early site permit environmental report or otherwise necessary to 
determine whether there is any obviously superior alternative to the 
site proposed. The draft environmental impact statement must not 
include an assessment of the economic, technical, or other benefits 
(for example, need for power) and costs of the proposed action or an 
evaluation of alternative energy sources, unless these matters are 
addressed in the early site permit environmental report.
---------------------------------------------------------------------------

    \5\ Values for releases of Rn-222 and Tc-99 are not given in the 
table. The amount and significance of Rn-222 releases from the fuel 
cycle and Tc-99 releases from waste management or reprocessing 
activities shall be considered in the draft environmental impact 
statement and may be the subject of litigation in individual 
licensing proceedings.
---------------------------------------------------------------------------

    (c) Combined license stage. A draft environmental impact statement 
relating to issuance of a combined license that does not reference an 
early site permit will be prepared in accordance with the procedures 
and measures described in Sec. Sec.  51.70, 51.71, 51.72, and 51.73. 
The contribution of the environmental effects of the uranium fuel cycle 
activities specified in Sec.  51.51 shall be evaluated on the basis of 
impact values set forth in Table S-3, Table of Uranium Fuel Cycle 
Environmental Data, which shall be set out in the draft environmental 
impact statement. With the exception of radon-222 and technetium-99 
releases, no further discussion of fuel cycle release values and other 
numerical data that appear explicitly in the table shall be 
required.\5\ The impact statement shall take account of dose 
commitments and health effects from fuel cycle effluents set forth in 
Table S-3 and shall in addition take account of economic, 
socioeconomic, and possible cumulative impacts and other fuel cycle 
impacts as may reasonably appear significant. The impact statement will 
include a discussion of the storage of spent fuel for the nuclear power 
plant within the scope of the generic determination in Sec.  51.23(a) 
and in accordance with Sec.  51.23(b).
    (1) Combined license application referencing an early site permit. 
If the combined license application references an early site permit, 
then the NRC staff shall prepare a draft supplement to the early site 
permit environmental impact statement. The supplement must be prepared 
in accordance with Sec.  51.92(e).
    (2) Combined license application referencing a standard design 
certification. If the combined license application references a 
standard design certification and the site characteristics of the 
combined license's site fall within the site parameters specified in 
the design certification environmental assessment, then the draft 
combined license environmental impact statement shall incorporate by 
reference the design certification environmental assessment, and 
summarize the findings and conclusions of the environmental assessment 
with respect to severe accident mitigation design alternatives.
    (3) Combined license application referencing a manufactured 
reactor. If the combined license application proposes to use a 
manufactured reactor and the site characteristics of the combined 
license's site fall within the site parameters specified in the 
manufacturing license environmental assessment, then the draft combined 
license environmental impact statement shall incorporate by reference 
the manufacturing license environmental assessment, and summarize the 
findings and conclusions of the environmental assessment with respect 
to severe accident mitigation design alternatives. The combined license 
environmental impact statement report will not address the 
environmental impacts associated with manufacturing the reactor under 
the manufacturing license.


Sec.  51.76  [Removed]

0
143. Section 51.76 is removed and reserved.

0
144. Section 51.92 is revised to read as follows:


Sec.  51.92   Supplement to the final environmental impact statement.

    (a) If the proposed action has not been taken, the NRC staff will 
prepare a supplement to a final environmental impact statement for 
which a notice of availability has been published in the Federal 
Register as provided in Sec.  51.118, if:
    (1) There are substantial changes in the proposed action that are 
relevant to environmental concerns; or
    (2) There are new and significant circumstances or information 
relevant to environmental concerns and bearing on the proposed action 
or its impacts.
    (b) In a proceeding for a combined license application under 10 CFR 
part 52 referencing an early site permit under part 52, the NRC staff 
shall prepare a supplement to the final environmental impact statement 
for the referenced early site permit in accordance with paragraph (e) 
of this section.
    (c) The NRC staff may prepare a supplement to a final environmental 
impact statement when, in its opinion, preparation of a supplement will 
further the purposes of NEPA.
    (d) The supplement to a final environmental impact statement will 
be prepared in the same manner as the final environmental impact 
statement except that a scoping process need not be used.
    (e) The supplement to an early site permit final environmental 
impact statement which is prepared for a combined license application 
in accordance with Sec.  51.75(c)(1) and paragraph (b) of this section 
must:
    (1) Identify the proposed action as the issuance of a combined 
license for the construction and operation of a nuclear power plant as 
described in the combined license application at the site described in 
the early site permit referenced in the combined license application;

[[Page 49516]]

    (2) Incorporate by reference the final environmental impact 
statement prepared for the early site permit;
    (3) Contain no separate discussion of alternative sites;
    (4) Include an analysis of the economic, technical, and other 
benefits and costs of the proposed action, to the extent that the final 
environmental impact statement prepared for the early site permit did 
not include an assessment of these benefits and costs;
    (5) Include an analysis of other energy alternatives, to the extent 
that the final environmental impact statement prepared for the early 
site permit did not include an assessment of energy alternatives;
    (6) Include an analysis of any environmental issue related to the 
impacts of construction or operation of the facility that was not 
resolved in the proceeding on the early site permit; and
    (7) Include an analysis of the issues related to the impacts of 
construction and operation of the facility that were resolved in the 
early site permit proceeding for which new and significant information 
has been identified, including, but not limited to, new and significant 
information demonstrating that the design of the facility falls outside 
the site characteristics and design parameters specified in the early 
site permit.
    (f)(1) A supplement to a final environmental impact statement will 
be accompanied by or will include a request for comments as provided in 
Sec.  51.73 and a notice of availability will be published in the 
Federal Register as provided in Sec.  51.117 if paragraphs (a) or (b) 
of this section applies.
    (2) If comments are not requested, a notice of availability of a 
supplement to a final environmental impact statement will be published 
in the Federal Register as provided in Sec.  51.118.

0
145. In Sec.  51.95, paragraph (a), the introductory text of paragraph 
(c), and paragraph (d) are revised to read as follows:


Sec.  51.95  Postconstruction environmental impact statements.

    (a) General. Any supplement to a final environmental impact 
statement or any environmental assessment prepared under the provisions 
of this section may incorporate by reference any information contained 
in a final environmental document previously prepared by the NRC staff 
that relates to the same production or utilization facility. Documents 
that may be referenced include, but are not limited to, the final 
environmental impact statement; supplements to the final environmental 
impact statement, including supplements prepared at the operating 
license stage; NRC staff-prepared final generic environmental impact 
statements; environmental assessments and records of decisions prepared 
in connection with the construction permit, the operating license, the 
early site permit, or the combined license and any license amendment 
for that facility. A supplement to a final environmental impact 
statement will include a request for comments as provided in Sec.  
51.73.
* * * * *
    (c) Operating license renewal stage. In connection with the renewal 
of an operating license or combined license for a nuclear power plant 
under parts 52 or 54 of this chapter, the Commission shall prepare an 
environmental impact statement, which is a supplement to the 
Commission's NUREG-1437, ``Generic Environmental Impact Statement for 
License Renewal of Nuclear Plants'' (May 1996), which is available in 
the NRC Public Document Room, 11555 Rockville Pike, Rockville, 
Maryland.
* * * * *
    (d) Postoperating license stage. In connection with the amendment 
of an operating or combined license authorizing decommissioning 
activities at a production or utilization facility covered by Sec.  
51.20, either for unrestricted use or based on continuing use 
restrictions applicable to the site, or with the issuance, amendment or 
renewal of a license to store spent fuel at a nuclear power reactor 
after expiration of the operating or combined license for the nuclear 
power reactor, the NRC staff will prepare a supplemental environmental 
impact statement for the post operating or post combined license stage 
or an environmental assessment, as appropriate, which will update the 
prior environmental documentation prepared by the NRC for compliance 
with NEPA under the provisions of this part. The supplement or 
assessment may incorporate by reference any information contained in 
the final environmental impact statement--for the operating or combined 
license stage, as appropriate, or in the records of decision prepared 
in connection with the early site permit, construction permit, 
operating license, or combined license for that facility. The 
supplement will include a request for comments as provided in Sec.  
51.73. Unless otherwise required by the Commission in accordance with 
the generic determination in Sec.  51.23(a) and the provisions of Sec.  
51.23(b), a supplemental environmental impact statement for the 
postoperating or post combined license stage or an environmental 
assessment, as appropriate, will address the environmental impacts of 
spent fuel storage only for the term of the license, license amendment 
or license renewal applied for.

0
146. Section 51.105 is revised to read as follows:


Sec.  51.105  Public hearings in proceedings for issuance of 
construction permits or early site permits.

    (a) In addition to complying with applicable requirements of Sec.  
51.104, in a proceeding for the issuance of a construction permit or 
early site permit for a nuclear power reactor, testing facility, fuel 
reprocessing plant or isotopic enrichment plant, the presiding officer 
will:
    (1) Determine whether the requirements of Sections 102(2) (A), (C), 
and (E) of NEPA and the regulations in this subpart have been met;
    (2) Independently consider the final balance among conflicting 
factors contained in the record of the proceeding with a view to 
determining the appropriate action to be taken;
    (3) Determine, after weighing the environmental, economic, 
technical, and other benefits against environmental and other costs, 
and considering reasonable alternatives, whether the construction 
permit or early site permit should be issued, denied, or appropriately 
conditioned to protect environmental values;
    (4) Determine, in an uncontested proceeding, whether the NEPA 
review conducted by the NRC staff has been adequate; and
    (5) Determine, in a contested proceeding, whether in accordance 
with the regulations in this subpart, the construction permit or early 
site permit should be issued as proposed by the NRC's Director of New 
Reactors or Director of Nuclear Reactor Regulation.
    (b) The presiding officer in an early site permit hearing shall not 
admit contentions proffered by any party concerning the benefits 
assessment (e.g., need for power) or alternative energy sources if 
those issues were not addressed by the applicant in the early site 
permit application.

0
147. Section 51.105a is added to read as follows:


Sec.  51.105a  Public hearings in proceedings for issuance of 
manufacturing licenses.

    In addition to complying with applicable requirements of Sec.  
51.31(c), in a proceeding for the issuance of a manufacturing license, 
the presiding officer will determine whether, in accordance with the 
regulations in this subpart, the manufacturing license

[[Page 49517]]

should be issued as proposed by the NRC's Director of New Reactors or 
Director of Nuclear Reactor Regulation.

0
148. Section 51.107 is added under the undesignated center heading 
``Production and Utilization Facilities'' to read as follows:


Sec.  51.107  Public hearings in proceedings for issuance of combined 
licenses.

    (a) In addition to complying with the applicable requirements of 
Sec.  51.104, in a proceeding for the issuance of a combined license 
for a nuclear power reactor under part 52 of this chapter, the 
presiding officer will:
    (1) Determine whether the requirements of Sections 102(2) (A), (C), 
and (E) of NEPA and the regulations in this subpart have been met;
    (2) Independently consider the final balance among conflicting 
factors contained in the record of the proceeding with a view to 
determining the appropriate action to be taken;
    (3) Determine, after weighing the environmental, economic, 
technical, and other benefits against environmental and other costs, 
and considering reasonable alternatives, whether the combined license 
should be issued, denied, or appropriately conditioned to protect 
environmental values;
    (4) Determine, in an uncontested proceeding, whether the NEPA 
review conducted by the NRC staff has been adequate; and
    (5) Determine, in a contested proceeding, whether in accordance 
with the regulations in this subpart, the combined license should be 
issued as proposed by the NRC's Director of New Reactors or Director of 
Nuclear Reactor Regulation.
    (b) If a combined license application references an early site 
permit, then the presiding officer in the combined license hearing 
shall not admit any contention proffered by any party on environmental 
issues which have been accorded finality under Sec.  52.39 of this 
chapter, unless the contention:
    (1) Demonstrates that the nuclear power reactor proposed to be 
built does not fit within one or more of the site characteristics or 
design parameters included in the early site permit;
    (2) Raises any significant environmental issue that was not 
resolved in the early site permit proceeding; or
    (3) Raises any issue involving the impacts of construction and 
operation of the facility that was resolved in the early site permit 
proceeding for which new and significant information has been 
identified.
    (c) If the combined license application references a standard 
design certification, or proposes to use a manufactured reactor, then 
the presiding officer in a combined license hearing shall not admit 
contentions proffered by any party concerning severe accident 
mitigation design alternatives unless the contention demonstrates that 
the site characteristics fall outside of the site parameters in the 
standard design certification or underlying manufacturing license for 
the manufactured reactor.

0
149. Section 51.108 is added under the undesignated center heading 
``Production and Utilization Facilities,'' to read as follows:


Sec.  51.108  Public hearings on Commission findings that inspections, 
tests, analyses, and acceptance criteria of combined licenses are met.

    In any public hearing requested under 10 CFR 52.103(b), the 
Commission will not admit any contentions on environmental issues, the 
adequacy of the environmental impact statement for the combined license 
issued under subpart C of part 52, or the adequacy of any other 
environmental impact statement or environmental assessment referenced 
in the combined license application. The Commission will not make any 
environmental findings in connection with the finding under 10 CFR 
52.103(g).

0
150. Part 52 is revised to read as follows:

PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER 
PLANTS

General Provisions

Sec.
52.0 Scope; applicability of 10 CFR Chapter I provisions.
52.1 Definitions.
52.2 Interpretations.
52.3 Written communications.
52.4 Deliberate misconduct.
52.5 Employee protection.
52.6 Completeness and accuracy of information.
52.7 Specific exemptions.
52.8 Combining licenses; elimination of repetition.
52.9 Jurisdictional limits.
52.10 Attacks and destructive acts.
52.11 Information collection requirements: OMB approval.
Subpart A--Early Site Permits
52.12 Scope of subpart.
52.13 Relationship to other subparts.
52.15 Filing of applications.
52.16 Contents of applications; general information.
52.17 Contents of applications; technical information.
52.18 Standards for review of applications.
52.21 Administrative review of applications; hearings.
52.23 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).
52.24 Issuance of early site permit.
52.25 Extent of activities permitted.
52.27 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.
52.39 Finality of early site permit determinations.
Subpart B--Standard Design Certifications
52.41 Scope of subpart.
52.43 Relationship to other subparts.
52.45 Filing of applications.
52.46 Contents of applications; general information.
52.47 Contents of applications; technical information.
52.48 Standards for review of applications.
52.51 Administrative review of applications.
52.53 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).
52.54 Issuance of standard design certification.
52.55 Duration of certification.
52.57 Application for renewal.
52.59 Criteria for renewal.
52.61 Duration of renewal.
52.63 Finality of standard design certifications.
Subpart C--Combined Licenses
52.71 Scope of subpart.
52.73 Relationship to other subparts.
52.75 Filing of applications.
52.77 Contents of applications; general information.
52.79 Contents of applications; technical information in final 
safety analysis report.
52.80 Contents of applications; additional technical information.
52.81 Standards for review of applications.
52.83 Finality of referenced NRC approvals; partial initial decision 
on site suitability.
52.85 Administrative review of applications; hearings.
52.87 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).
52.89 Reserved.
52.91 Authorization to conduct site activities.
52.93 Exemptions and variances.
52.97 Issuance of combined licenses.
52.98 Finality of combined licenses; information requests.
52.99 Inspection during construction.
52.103 Operation under a combined license.
52.104 Duration of combined license.
52.105 Transfer of combined license.
52.107 Application for renewal.
52.109 Continuation of combined license.
52.110 Termination of license.
Subpart D--Reserved
Subpart E--Standard Design Approvals
52.131 Scope of subpart.
52.133 Relationship to other subparts.
52.135 Filing of applications.

[[Page 49518]]

52.136 Contents of applications; general information.
52.137 Contents of applications; technical information.
52.139 Standards for review of applications.
52.141 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).
52.143 Staff approval of design.
52.145 Finality of standard design approvals; information requests.
52.147 Duration of design approval.
Subpart F--Manufacturing Licenses
52.151 Scope of subpart.
52.153 Relationship to other subparts.
52.155 Filing of applications.
52.156 Contents of applications; general information.
52.157 Contents of applications; technical information in final 
safety analysis report.
52.158 Contents of application; additional technical information.
52.159 Standards for review of application.
52.161 Reserved.
52.163 Administrative review of applications; hearings.
52.165 Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).
52.167 Issuance of manufacturing license.
52.169 Reserved.
52.171 Finality of manufacturing licenses; information requests.
52.173 Duration of manufacturing license.
52.175 Transfer of manufacturing license.
52.177 Application for renewal.
52.179 Criteria for renewal.
52.181 Duration of renewal.
Subpart G--Reserved
Subpart H--Enforcement
52.301 Violations.
52.303 Criminal penalties.
Appendix A to Part 52--Design Certification Rule for the U.S. 
Advanced Boiling Water Reactor
Appendix B to Part 52--Design Certification Rule for the System 80+ 
Design
Appendix C to Part 52--Design Certification Rule for the AP600 
Design
Appendix D to Part 52--Design Certification Rule for the AP1000 
Design
Appendixes E Through M to Part 52 [Reserved]
Appendix N to Part 52--Standardization of Nuclear Power Plant 
Designs: Combined Licenses to Construct and Operate Nuclear Power 
Reactors of Identical Design at Multiple Sites

    Authority:  Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as 
amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282); secs. 
201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 
5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).

General Provisions


Sec.  52.0  Scope; applicability of 10 CFR Chapter I provisions.

    (a) This part governs the issuance of early site permits, standard 
design certifications, combined licenses, standard design approvals, 
and manufacturing licenses for nuclear power facilities licensed under 
Section 103 of the Atomic Energy Act of 1954, as amended (68 Stat. 
919), and Title II of the Energy Reorganization Act of 1974 (88 Stat. 
1242). This part also gives notice to all persons who knowingly provide 
to any holder of or applicant for an approval, certification, permit, 
or license, or to a contractor, subcontractor, or consultant of any of 
them, components, equipment, materials, or other goods or services that 
relate to the activities of a holder of or applicant for an approval, 
certification, permit, or license, subject to this part, that they may 
be individually subject to NRC enforcement action for violation of the 
provisions in 10 CFR 52.4.
    (b) Unless otherwise specifically provided for in this part, the 
regulations in 10 CFR Chapter I apply to a holder of or applicant for 
an approval, certification, permit, or license. A holder of or 
applicant for an approval, certification, permit, or license issued 
under this part shall comply with all requirements in 10 CFR Chapter I 
that are applicable. A license, approval, certification, or permit 
issued under this part is subject to all requirements in 10 CFR Chapter 
I which, by their terms, are applicable to early site permits, design 
certifications, combined licenses, design approvals, or manufacturing 
licenses.


Sec.  52.1  Definitions.

    (a) As used in this part--
    Combined license means a combined construction permit and operating 
license with conditions for a nuclear power facility issued under 
subpart C of this part.
    Decommission means to remove a facility or site safely from service 
and reduce residual radioactivity to a level that permits--
    (i) Release of the property for unrestricted use and termination of 
the license; or
    (ii) Release of the property under restricted conditions and 
termination of the license.
    Design characteristics are the actual features of a reactor or 
reactors. Design characteristics are specified in a standard design 
approval, a standard design certification, a combined license 
application, or a manufacturing license.
    Design parameters are the postulated features of a reactor or 
reactors that could be built at a proposed site. Design parameters are 
specified in an early site permit.
    Early site permit means a Commission approval, issued under subpart 
A of this part, for a site or sites for one or more nuclear power 
facilities. An early site permit is a partial construction permit.
    License means a license, including an early site permit, combined 
license or manufacturing license under this part or a renewed license 
issued by the Commission under this part or part 54 of this chapter.
    Licensee means a person who is authorized to conduct activities 
under a license issued by the Commission.
    Major feature of the emergency plans means an aspect of those plans 
necessary to:
    (i) Address in whole or part one or more of the 16 standards in 10 
CFR 50.47(b); or
    (ii) Describe the emergency planning zones as required in 10 CFR 
50.33(g).
    Manufacturing license means a license, issued under subpart F of 
this part, authorizing the manufacture of nuclear power reactors but 
not their construction, installation, or operation at the sites on 
which the reactors are to be operated.
    Modular design means a nuclear power station that consists of two 
or more essentially identical nuclear reactors (modules) and each 
module is a separate nuclear reactor capable of being operated 
independent of the state of completion or operating condition of any 
other module co-located on the same site, even though the nuclear power 
station may have some shared or common systems.
    Prototype plant means a nuclear power plant that is used to test 
new safety features, such as the testing required under 10 CFR 
50.43(e). The prototype plant is similar to a first-of-a-kind or 
standard plant design in all features and size, but may include 
additional safety features to protect the public and the plant staff 
from the possible consequences of accidents during the testing period.
    Site characteristics are the actual physical, environmental and 
demographic features of a site. Site characteristics are specified in 
an early site permit or in a final safety analysis report for a 
combined license.
    Site parameters are the postulated physical, environmental and 
demographic features of an assumed site. Site parameters are specified 
in a standard design approval, standard design certification, or 
manufacturing license.
    Standard design means a design which is sufficiently detailed and 
complete to support certification or approval in accordance with 
subpart B or E of this part, and which is usable for a multiple number 
of units or at a multiple number of sites without reopening or 
repeating the review.

[[Page 49519]]

    Standard design approval or design approval means an NRC staff 
approval, issued under subpart E of this part, of a final standard 
design for a nuclear power reactor of the type described in 10 CFR 
50.22. The approval may be for either the final design for the entire 
reactor facility or the final design of major portions thereof.
    Standard design certification or design certification means a 
Commission approval, issued under subpart B of this part, of a final 
standard design for a nuclear power facility. This design may be 
referred to as a certified standard design.
    (b) All other terms in this part have the meaning set out in 10 CFR 
50.2, or Section 11 of the Atomic Energy Act, as applicable.


Sec.  52.2  Interpretations.

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


Sec.  52.3  Written communications.

    (a) General requirements. All correspondence, reports, 
applications, and other written communications from an applicant, 
licensee, or holder of a standard design approval to the Nuclear 
Regulatory Commission concerning the regulations in this part, 
individual license conditions, or the terms and conditions of an early 
site permit or standard design approval, must be sent either by mail 
addressed: ATTN: Document Control Desk, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001; by hand delivery to the NRC's 
offices at 11555 Rockville Pike, Rockville, Maryland, between the hours 
of 7:30 a.m. and 4:15 p.m. eastern time; or, where practicable, by 
electronic submission, for example, via Electronic Information 
Exchange, e-mail, or CD-ROM. Electronic submissions must be made in a 
manner that enables the NRC to receive, read, authenticate, distribute, 
and archive the submission, and process and retrieve it a single page 
at a time. Detailed guidance on making electronic submissions can be 
obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/eie.html, by calling (301) 415-6030, by e-mail at <a href="/cdn-cgi/l/email-protection#16535f535678647538717960"><span class="__cf_email__" data-cfemail="e3a6aaa6a38d9180cd848c95">[email protected]</span></a>, or 
by writing the Office of Information Services, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001. The guidance discusses, among 
other topics, the formats the NRC can accept, the use of electronic 
signatures, and the treatment of nonpublic information. If the 
communication is on paper, the signed original must be sent. If a 
submission due date falls on a Saturday, Sunday, or Federal holiday, 
the next Federal working day becomes the official due date.
    (b) Distribution requirements. Copies of all correspondence, 
reports, and other written communications concerning the regulations in 
this part or individual license conditions, or the terms and conditions 
of an early site permit or standard design approval, must be submitted 
to the persons listed in paragraph (b)(1) of this section (addresses 
for the NRC Regional Offices are listed in appendix D to part 20 of 
this chapter).
    (1) Applications for amendment of permits and licenses; reports; 
and other communications. All written communications (including 
responses to: generic letters, bulletins, information notices, 
regulatory information summaries, inspection reports, and miscellaneous 
requests for additional information) that are required of holders of 
early site permits, standard design approvals, combined licenses, or 
manufacturing licenses issued under this part must be submitted as 
follows, except as otherwise specified in paragraphs (b)(2) through 
(b)(7) of this section: to the NRC's Document Control Desk (if on 
paper, the signed original), with a copy to the appropriate Regional 
Office, and a copy to the appropriate NRC Resident Inspector, if one 
has been assigned to the site of the facility or the place of 
manufacture of a reactor licensed under subpart F of this part.
    (2) Applications and amendments to applications. Applications for 
early site permits, standard design approvals, combined licenses, 
manufacturing licenses and amendments to any of these types of 
applications must be submitted to the NRC's Document Control Desk, with 
a copy to the appropriate Regional Office, and a copy to the 
appropriate NRC Resident Inspector, if one has been assigned to the 
site of the facility or the place of manufacture of a reactor licensed 
under subpart F of this part, except as otherwise specified in 
paragraphs (b)(3) through (b)(7) of this section. If the application or 
amendment is on paper, the submission to the Document Control Desk must 
be the signed original.
    (3) Acceptance review application. Written communications required 
for an application for determination of suitability for docketing must 
be submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office. If the communication is on paper, the 
submission to the Document Control Desk must be the signed original.
    (4) Security plan and related submissions. Written communications, 
as defined in paragraphs (b)(4)(i) through (iv) of this section, must 
be submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office. If the communication is on paper, the 
submission to the Document Control Desk must be the signed original.
    (i) Physical security plan under Sec.  52.79 of this chapter;
    (ii) Safeguards contingency plan under Sec.  52.79 of this chapter;
    (iii) Change to security plan, guard training and qualification 
plan, or safeguards contingency plan made without prior Commission 
approval under Sec.  50.54(p) of this chapter;
    (iv) Application for amendment of physical security plan, guard 
training and qualification plan, or safeguards contingency plan under 
Sec.  50.90 of this chapter.
    (5) Emergency plan and related submissions. Written communications 
as defined in paragraphs (b)(5)(i) through (iii) of this section must 
be submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office, and a copy to the appropriate NRC Resident 
Inspector if one has been assigned to the site of the facility. If the 
communication is on paper, the submission to the Document Control Desk 
must be the signed original.
    (i) Emergency plan under Sec.  52.17(b) or Sec.  52.79(a);
    (ii) Change to an emergency plan under Sec.  50.54(q) of this 
chapter;
    (iii) Emergency implementing procedures under appendix E, Section V 
of part 50 of this chapter.
    (6) Updated FSAR. An updated final safety analysis report (FSAR) or 
replacement pages under Sec.  50.71(e) of this chapter, or the 
regulations in this part must be submitted to the NRC's Document 
Control Desk, with a copy to the appropriate Regional Office, and a 
copy to the appropriate NRC Resident Inspector if one has been assigned 
to the site of the facility or the place of manufacture of a reactor 
licensed under subpart F of this part. Paper copy submissions may be 
made using replacement pages; however, if a licensee chooses to use 
electronic submission, all subsequent updates or submissions must be 
performed electronically on a total replacement basis. If the 
communication is on paper, the submission to the Document Control Desk 
must be the signed original. If the communications are submitted 
electronically, see Guidance for

[[Page 49520]]

Electronic Submissions to the Commission.
    (7) Quality assurance related submissions. (i) A change to the 
safety analysis report quality assurance program description under 
Sec.  50.54(a)(3) or Sec.  50.55(f)(4) of this chapter, or a change to 
a licensee's NRC-accepted quality assurance topical report under Sec.  
50.54(a)(3) or Sec.  50.55(f)(4) of this chapter, must be submitted to 
the NRC's Document Control Desk, with a copy to the appropriate 
Regional Office, and a copy to the appropriate NRC Resident Inspector 
if one has been assigned to the site of the facility. If the 
communication is on paper, the submission to the Document Control Desk 
must be the signed original.
    (ii) A change to an NRC-accepted quality assurance topical report 
from nonlicensees (i.e., architect/engineers, NSSS suppliers, fuel 
suppliers, constructors, etc.) must be submitted to the NRC's Document 
Control Desk. If the communication is on paper, the signed original 
must be sent.
    (8) Certification of permanent cessation of operations. The 
licensee's certification of permanent cessation of operations under 
Sec.  52.110(a)(1), must state the date on which operations have ceased 
or will cease, and must be submitted to the NRC's Document Control 
Desk. This submission must be under oath or affirmation.
    (9) Certification of permanent fuel removal. The licensee's 
certification of permanent fuel removal under Sec.  52.110(a)(1), must 
state the date on which the fuel was removed from the reactor vessel 
and the disposition of the fuel, and must be submitted to the NRC's 
Document Control Desk. This submission must be under oath or 
affirmation.
    (c) Form of communications. All paper copies submitted to meet the 
requirements set forth in paragraph (b) of this section must be 
typewritten, printed or otherwise reproduced in permanent form on 
unglazed paper. Exceptions to these requirements imposed on paper 
submissions may be granted for the submission of micrographic, 
photographic, or similar forms.
    (d) Regulation governing submission. Applicants, licensees, and 
holders of standard design approvals submitting correspondence, 
reports, and other written communications under the regulations of this 
part are requested but not required to cite whenever practical, in the 
upper right corner of the first page of the submission, the specific 
regulation or other basis requiring submission.


Sec.  52.4  Deliberate misconduct.

    (a) Applicability. This section applies to any:
    (1) Licensee;
    (2) Holder of a standard design approval;
    (3) Applicant for a standard design certification;
    (4) Applicant for a license or permit;
    (5) Applicant for a standard design approval;
    (6) Employee of a licensee;
    (7) Employee of an applicant for a license, a standard design 
certification, or a standard design approval;
    (8) Any contractor (including a supplier or consultant), 
subcontractor, or employee of a contractor or subcontractor of any 
licensee; or
    (9) Any contractor (including a supplier or consultant), 
subcontractor, or employee of a contractor or subcontractor of any 
applicant for a license, a standard design certification, or a standard 
design approval.
    (b) Definitions. For purposes of this section:
    Deliberate misconduct means an intentional act or omission that a 
person or entity knows:
    (i) Would cause a licensee or an applicant for a license, standard 
design certification, or standard design approval to be in violation of 
any rule, regulation, or order; or any term, condition, or limitation, 
of any license, standard design certification, or standard design 
approval; or
    (ii) Constitutes a violation of a requirement, procedure, 
instruction, contract, purchase order, or policy of a licensee, holder 
of a standard design approval, applicant for a license, standard design 
certification, or standard design approval, or contractor, or 
subcontractor.
    (c) Prohibition against deliberate misconduct. Any person or entity 
subject to this section, who knowingly provides to any licensee, any 
applicant for a license, standard design certification or standard 
design approval, or a contractor, or subcontractor of a person or 
entity subject to this section, any components, equipment, materials, 
or other goods or services that relate to a licensee's or applicant's 
activities under this part, may not:
    (1) Engage in deliberate misconduct that causes or would have 
caused, if not detected, a licensee, holder of a standard design 
approval, or applicant to be in violation of any rule, regulation, or 
order; or any term, condition, or limitation of any license issued by 
the Commission, any standard design approval, or standard design 
certification; or
    (2) Deliberately submit to the NRC; a licensee, an applicant for a 
license, standard design certification or standard design approval; or 
a licensee's, standard design approval holder's, or applicant's 
contractor or subcontractor, information that the person submitting the 
information knows to be incomplete or inaccurate in some respect 
material to the NRC.
    (d) A person or entity who violates paragraph (c)(1) or (c)(2) of 
this section may be subject to enforcement action in accordance with 
the procedures in 10 CFR part 2, subpart B.


Sec.  52.5  Employee protection.

    (a) Discrimination by a Commission licensee, holder of a standard 
design approval, an applicant for a license, standard design 
certification, or standard design approval, a contractor or 
subcontractor of a Commission licensee, holder of a standard design 
approval, applicant for a license, standard design certification, or 
standard design approval, against an employee for engaging in certain 
protected activities is prohibited. Discrimination includes discharge 
and other actions that relate to compensation, terms, conditions, or 
privileges of employment. The protected activities are established in 
Section 211 of the Energy Reorganization Act of 1974, as amended, and 
in general are related to the administration or enforcement of a 
requirement imposed under the Atomic Energy Act or the Energy 
Reorganization Act.
    (1) The protected activities include but are not limited to:
    (i) Providing the Commission or his or her employer information 
about alleged violations of either of the statutes named in the 
introductory text of paragraph (a) of this section or possible 
violations of requirements imposed under either of those statutes;
    (ii) Refusing to engage in any practice made unlawful under either 
of the statutes named in the introductory text of paragraph (a) of this 
section or under these requirements if the employee has identified the 
alleged illegality to the employer;
    (iii) Requesting the Commission to institute action against his or 
her employer for the administration or enforcement of these 
requirements;
    (iv) Testifying in any Commission proceeding, or before Congress, 
or at any Federal or State proceeding regarding any provision (or 
proposed provision) of either of the statutes named in the introductory 
text of paragraph (a) of this section; and

[[Page 49521]]

    (v) Assisting or participating in, or is about to assist or 
participate in, these activities.
    (2) These activities are protected even if no formal proceeding is 
actually initiated as a result of the employee assistance or 
participation.
    (3) This section has no application to any employee alleging 
discrimination prohibited by this section who, acting without direction 
from his or her employer (or the employer's agent), deliberately causes 
a violation of any requirement of the Energy Reorganization Act of 
1974, as amended, or the Atomic Energy Act of 1954, as amended.
    (b) Any employee who believes that he or she has been discharged or 
otherwise discriminated against by any person for engaging in protected 
activities specified in paragraph (a)(1) of this section may seek a 
remedy for the discharge or discrimination through an administrative 
proceeding in the Department of Labor. The administrative proceeding 
must be initiated within 180 days after an alleged violation occurs. 
The employee may do this by filing a complaint alleging the violation 
with the Department of Labor, Employment Standards Administration, Wage 
and Hour Division. The Department of Labor may order reinstatement, 
back pay, and compensatory damages.
    (c) A violation of paragraph (a), (e), or (f) of this section by a 
Commission licensee, a holder of a standard design approval, an 
applicant for a Commission license, standard design certification, or a 
standard design approval, or a contractor or subcontractor of a 
Commission licensee, holder of a standard design approval, or any 
applicant may be grounds for--
    (1) Denial, revocation, or suspension of the license or standard 
design approval;
    (2) Withdrawal or revocation of a proposed or final standard design 
certification;
    (3) Imposition of a civil penalty on the licensee, holder of a 
standard design approval, or applicant (including an applicant for a 
standard design certification under this part following Commission 
adoption of final design certification rule).
    (4) Other enforcement action.
    (d) Actions taken by an employer, or others, which adversely affect 
an employee may be predicated upon nondiscriminatory grounds. The 
prohibition applies when the adverse action occurs because the employee 
has engaged in protected activities. An employee's engagement in 
protected activities does not automatically render him or her immune 
from discharge or discipline for legitimate reasons or from adverse 
action dictated by nonprohibited considerations.
    (e)(1) Each licensee, each holder of a standard design approval, 
and each applicant for a license, standard design certification, or 
standard design approval, shall prominently post the revision of NRC 
Form 3, ``Notice to Employees,'' referenced in 10 CFR 19.11(e). This 
form must be posted at locations sufficient to permit employees 
protected by this section to observe a copy on the way to or from their 
place of work. Premises must be posted not later than thirty (30) days 
after an application is docketed and remain posted while the 
application is pending before the Commission, during the term of the 
license, standard design certification, or standard design approval 
under 10 CFR part 52, and for 30 days following license termination or 
the expiration or termination of the standard design certification or 
standard design approval under 10 CFR part 52.
    (2) Copies of NRC Form 3 may be obtained by writing to the Regional 
Administrator of the appropriate U.S. Nuclear Regulatory Commission 
Regional Office listed in appendix D to part 20 of this chapter, by 
calling (301) 415-7232, via e-mail to <a href="/cdn-cgi/l/email-protection#a2c4cdd0cfd1e2ccd0c18cc5cdd4"><span class="__cf_email__" data-cfemail="d4b2bba6b9a794baa6b7fab3bba2">[email protected]</span></a>, or by visiting the 
NRC's Web site at http://www.nrc.gov and selecting forms from the index 
found on the NRC's home page.
    (f) No agreement affecting the compensation, terms, conditions, or 
privileges of employment, including an agreement to settle a complaint 
filed by an employee with the Department of Labor under Section 211 of 
the Energy Reorganization Act of 1974, as amended, may contain any 
provision which would prohibit, restrict, or otherwise discourage an 
employee from participating in protected activity as defined in 
paragraph (a)(1) of this section including, but not limited to, 
providing information to the NRC or to his or her employer on potential 
violations or other matters within NRC's regulatory responsibilities.
    (g) Part 19 of this chapter sets forth requirements and regulatory 
provisions applicable to licensees, holders of a standard design 
approval, applicants for a license, standard design certification, or 
standard design approval, and contractors or subcontractors of a 
Commission licensee, or holder of a standard design approval, and are 
in addition to the requirements in this section.


Sec.  52.6  Completeness and accuracy of information.

    (a) Information provided to the Commission by a licensee (including 
an early site permit holder, a combined license holder, and a 
manufacturing license holder), a holder of a standard design approval 
under this part, and an applicant for a license or an applicant for a 
standard design certification or a standard design approval under this 
part, and information required by statute or by the Commission's 
regulations, orders, license conditions, or terms and conditions of a 
standard design approval to be maintained by the licensee, the holder 
of a standard design approval under this part, the applicant for a 
standard design certification under this part following Commission 
adoption of a final design certification rule, and an applicant for a 
license, a standard design certification, or a standard design approval 
under this part shall be complete and accurate in all material 
respects.
    (b) Each applicant or licensee, each holder of a standard design 
approval under this part, and each applicant for a standard design 
certification under this part following Commission adoption of a final 
design certification regulation, shall notify the Commission of 
information identified by the applicant or the licensee as having for 
the regulated activity a significant implication for public health and 
safety or common defense and security. An applicant, licensee, or 
holder violates this paragraph only if the applicant, licensee, or 
holder fails to notify the Commission of information that the 
applicant, licensee, or holder has been identified as having a 
significant implication for public health and safety or common defense 
and security. Notification shall be provided to the Administrator of 
the appropriate Regional Office within 2 working days of identifying 
the information. This requirement is not applicable to information 
which is already required to be provided to the Commission by other 
reporting or updating requirements.


Sec.  52.7  Specific exemptions.

    The Commission may, upon application by any interested person or 
upon its own initiative, grant exemptions from the requirements of the 
regulations of this part. The Commission's consideration will be 
governed by Sec.  50.12 of this chapter, unless other criteria are 
provided for in this part, in which case the Commission's consideration 
will be governed by the criteria in this part. Only if those criteria 
are not met will the Commission's consideration be

[[Page 49522]]

governed by Sec.  50.12 of this chapter. The Commission's consideration 
of requests for exemptions from requirements of the regulations of 
other parts in this chapter, which are applicable by virtue of this 
part, shall be governed by the exemption requirements of those parts.


Sec.  52.8  Combining licenses; elimination of repetition.

    (a) An applicant for a license under this part may combine in its 
application several applications for different kinds of licenses under 
the regulations of this chapter.
    (b) An applicant may incorporate by reference in its application 
information contained in previous applications, statements or reports 
filed with the Commission, provided, however, that such references are 
clear and specific.
    (c) The Commission may combine in a single license the activities 
of an applicant which would otherwise be licensed separately.


Sec.  52.9  Jurisdictional limits.

    No permit, license, standard design approval, or standard design 
certification under this part shall be deemed to have been issued for 
activities which are not under or within the jurisdiction of the United 
States.


Sec.  52.10  Attacks and destructive acts.

    Neither an applicant for a license to manufacture, construct, and 
operate a utilization facility under this part, nor for an amendment to 
this license, or an applicant for an early site permit, a standard 
design certification, or standard design approval under this part, or 
for an amendment to the early site permit, standard design 
certification, or standard design approval, is required to provide for 
design features or other measures for the specific purpose of 
protection against the effects of--
    (a) Attacks and destructive acts, including sabotage, directed 
against the facility by an enemy of the United States, whether a 
foreign government or other person; or
    (b) Use or deployment of weapons incident to U.S. defense 
activities.


Sec.  52.11  Information collection requirements: OMB approval.

    (a) The Nuclear Regulatory Commission has submitted the information 
collection requirements contained in this part to the Office of 
Management and Budget (OMB) for approval as required by the Paperwork 
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or 
sponsor, and a person is not required to respond to, a collection of 
information unless it displays a currently valid OMB control number. 
OMB has approved the information collection requirements contained in 
this part under Control Number 3150-0151.
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  52.7, 52.15, 52.16, 52.17, 52.29, 52.35, 
52.39, 52.45, 52.46, 52.47, 52.57, 52.63, 52.75, 52.77, 52.79, 52.80, 
52.93, 52.99, 52.110, 52.135, 52.136, 52.137, 52.155, 52.156, 52.157, 
52.158, 52.171, 52.177, and appendices A, B, C, D, and N of part 52.

Subpart A--Early Site Permits


Sec.  52.12  Scope of subpart.

    This subpart sets out the requirements and procedures applicable to 
Commission issuance of an early site permit for approval of a site for 
one or more nuclear power facilities separate from the filing of an 
application for a construction permit or combined license for the 
facility.


Sec.  52.13  Relationship to other subparts.

    This subpart applies when any person who may apply for a 
construction permit under 10 CFR part 50, or for a combined license 
under this part seeks an early site permit from the Commission 
separately from an application for a construction permit or a combined 
license.


Sec.  52.15  Filing of applications.

    (a) Any person who may apply for a construction permit under 10 CFR 
part 50, or for a combined license under this part, may file an 
application for an early site permit with the Director, Office of New 
Reactors, or the Director, Office of Nuclear Reactor Regulation, as 
appropriate. An application for an early site permit may be filed 
notwithstanding the fact that an application for a construction permit 
or a combined license has not been filed in connection with the site 
for which a permit is sought.
    (b) The application must comply with the applicable filing 
requirements of Sec. Sec.  52.3 and 50.30 of this chapter.
    (c) The fees associated with the filing and review of an 
application for the initial issuance or renewal of an early site permit 
are set forth in 10 CFR part 170.


Sec.  52.16  Contents of applications; general information.

    The application must contain all of the information required by 10 
CFR 50.33(a) through (d) and (j) of this chapter.


Sec.  52.17  Contents of applications; technical information.

    (a) For applications submitted before September 27, 2007, the rule 
provisions in effect at the date of docketing apply unless otherwise 
requested by the applicant in writing. The application must contain:
    (1) A site safety analysis report. The site safety analysis report 
shall include the following:
    (i) The specific number, type, and thermal power level of the 
facilities, or range of possible facilities, for which the site may be 
used;
    (ii) The anticipated maximum levels of radiological and thermal 
effluents each facility will produce;
    (iii) The type of cooling systems, intakes, and outflows that may 
be associated with each facility;
    (iv) The boundaries of the site;
    (v) The proposed general location of each facility on the site;
    (vi) The seismic, meteorological, hydrologic, and geologic 
characteristics of the proposed site with appropriate consideration of 
the most severe of the natural phenomena that have been historically 
reported for the site and surrounding area and with sufficient margin 
for the limited accuracy, quantity, and period of time in which the 
historical data have been accumulated;
    (vii) The location and description of any nearby industrial, 
military, or transportation facilities and routes;
    (viii) The existing and projected future population profile of the 
area surrounding the site;
    (ix) A description and safety assessment of the site on which a 
facility is to be located. The assessment must contain an analysis and 
evaluation of the major structures, systems, and components of the 
facility that bear significantly on the acceptability of the site under 
the radiological consequence evaluation factors identified in 
paragraphs (a)(1)(ix)(A) and (a)(1)(ix)(B) of this section. In 
performing this assessment, an applicant shall assume a fission product 
release \1\ from the core into the containment assuming that the 
facility is operated at the ultimate power level contemplated. The 
applicant shall perform an evaluation and analysis of the postulated 
fission product release, using the expected demonstrable

[[Page 49523]]

containment leak rate and any fission product cleanup systems intended 
to mitigate the consequences of the accidents, together with applicable 
site characteristics, including site meteorology, to evaluate the 
offsite radiological consequences. Site characteristics must comply 
with part 100 of this chapter. The evaluation must determine that:
---------------------------------------------------------------------------

    \1\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. Such accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (A) An individual located at any point on the boundary of the 
exclusion area for any 2 hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 25 rem \2\ total effective dose equivalent (TEDE).
---------------------------------------------------------------------------

    \2\ A whole body dose of 25 rem has been stated to correspond 
numerically to the once in a lifetime accidental or emergency dose 
for radiation workers which, according to NCRP recommendations at 
the time could be disregarded in the determination of their 
radiation exposure status (see NBS Handbook 69 dated June 5, 1959). 
However, its use is not intended to imply that this number 
constitutes an acceptable limit for an emergency dose to the public 
under accident conditions. Rather, this dose value has been set 
forth in this section as a reference value, which can be used in the 
evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accidents.
---------------------------------------------------------------------------

    (B) An individual located at any point on the outer boundary of the 
low population zone, who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
TEDE;
    (x) Information demonstrating that site characteristics are such 
that adequate security plans and measures can be developed;
    (xi) For applications submitted after September 27, 2007, a 
description of the quality assurance program applied to site-related 
activities for the future design, fabrication, construction, and 
testing of the structures, systems, and components of a facility or 
facilities that may be constructed on the site. Appendix B to 10 CFR 
part 50 sets forth the requirements for quality assurance programs for 
nuclear power plants. The description of the quality assurance program 
for a nuclear power plant site shall include a discussion of how the 
applicable requirements of appendix B to part 50 of this chapter will 
be satisfied; and
    (xii) An evaluation of the site against applicable sections of the 
Standard Review Plan (SRP) revision in effect 6 months before the 
docket date of the application. The evaluation required by this section 
shall include an identification and description of all differences in 
analytical techniques and procedural measures proposed for a site and 
those corresponding techniques and measures given in the SRP acceptance 
criteria. Where such a difference exists, the evaluation shall discuss 
how the proposed alternative provides an acceptable method of complying 
with the Commission's regulations, or portions thereof, that underlie 
the corresponding SRP acceptance criteria. The SRP is not a substitute 
for the regulations, and compliance is not a requirement.
    (2) A complete environmental report as required by 10 CFR 51.50(b).
    (b)(1) The site safety analysis report must identify physical 
characteristics of the proposed site, such as egress limitations from 
the area surrounding the site, that could pose a significant impediment 
to the development of emergency plans. If physical characteristics are 
identified that could pose a significant impediment to the development 
of emergency plans, the application must identify measures that would, 
when implemented, mitigate or eliminate the significant impediment.
    (2) The site safety analysis report may also:
    (i) Propose major features of the emergency plans, in accordance 
with the pertinent standards of 10 CFR 50.47, and the requirements of 
appendix E to 10 CFR part 50, such as the exact size and configuration 
of the emergency planning zones, for review and approval by NRC, in 
consultation with the Department of Homeland Security (DHS) in the 
absence of complete and integrated emergency plans; or
    (ii) Propose complete and integrated emergency plans for review and 
approval by the NRC, in consultation with DHS, in accordance with the 
applicable standards of 10 CFR 50.47, and the requirements of appendix 
E to 10 CFR part 50. To the extent approval of emergency plans is 
sought, the application must contain the information required by 
Sec. Sec.  50.33(g) and (j) of this chapter.
    (3) Emergency plans submitted under paragraph (b)(2)(ii) of this 
section must include the proposed inspections, tests, and analyses that 
the holder of a combined license referencing the early site permit 
shall perform, and the acceptance criteria that are necessary and 
sufficient to provide reasonable assurance that, if the inspections, 
tests, and analyses are performed and the acceptance criteria met, the 
facility has been constructed and will be operated in conformity with 
the emergency plans, the provisions of the Act, and the Commission's 
rules and regulations. Major features of an emergency plan submitted 
under paragraph (b)(2)(i) of this section may include proposed 
inspections, tests, analyses, and acceptance criteria.
    (4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the site 
safety analysis report must include a description of contacts and 
arrangements made with Federal, State, and local governmental agencies 
with emergency planning responsibilities. The site safety analysis 
report must contain any certifications that have been obtained. If 
these certifications cannot be obtained, the site safety analysis 
report must contain information, including a utility plan, sufficient 
to show that the proposed plans provide reasonable assurance that 
adequate protective measures can and will be taken in the event of a 
radiological emergency at the site. Under the option set forth in 
paragraph (b)(2)(ii) of this section, the applicant shall make good 
faith efforts to obtain from the same governmental agencies 
certifications that:
    (i) The proposed emergency plans are practicable;
    (ii) These agencies are committed to participating in any further 
development of the plans, including any required field demonstrations, 
and
    (iii) That these agencies are committed to executing their 
responsibilities under the plans in the event of an emergency.
    (c) If the applicant requests authorization to perform activities 
at the site, which are identified in 10 CFR 50.10(e)(1), after issuance 
of the early site permit and without a separate authorization under 10 
CFR 50.10(e)(1), the applicant must identify the activities that are 
requested, and propose a plan for redress of the site in the event that 
the activities are performed and the early site permit expires before 
it is referenced in an application for a construction permit or a 
combined license. 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. In addition, the Commission shall prepare an 
environmental impact statement during review of the application, in 
accordance with the applicable provisions of 10 CFR part 51. The 
Commission shall determine, after

[[Page 49524]]

consultation with DHS, whether the information required of the 
applicant by Sec.  52.17(b)(1) shows that there is no significant 
impediment to the development of emergency plans that cannot be 
mitigated or eliminated by measures proposed by the applicant, whether 
any major features of emergency plans submitted by the applicant under 
Sec.  52.17(b)(2)(i) are acceptable in accordance with the applicable 
standards of 10 CFR 50.47 and the requirements of appendix E to 10 CFR 
part 50, and whether any emergency plans submitted by the applicant 
under Sec.  52.17(b)(2)(ii) provide reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency.


Sec.  52.21  Administrative review of applications; hearings.

    An early site permit is subject to all procedural requirements in 
10 CFR part 2, including the requirements for docketing in Sec.  
2.101(a)(1) through (4) of this chapter, and the requirements for 
issuance of a notice of hearing in Sec. Sec.  2.104(a) and (d) of this 
chapter, provided that the designated sections may not be construed to 
require that the environmental report, or draft or final environmental 
impact statement include an assessment of the benefits of construction 
and operation of the reactor or reactors, or an analysis of alternative 
energy sources. The presiding officer in an early site permit hearing 
shall not admit contentions proffered by any party concerning an 
assessment of the benefits of construction and operation of the reactor 
or reactors, or an analysis of alternative energy sources if those 
issues were not addressed by the applicant in the early site permit 
application. All hearings conducted on applications for early site 
permits filed under this part are governed by the procedures contained 
in subparts C, G, L, and N of 10 CFR part 2, as applicable.


Sec.  52.23  Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).

    The Commission shall refer a copy of the application for an early 
site permit to the ACRS. The ACRS shall report on those portions of the 
application which concern safety.


Sec.  52.24  Issuance of early site permit.

    (a) After conducting a hearing under Sec.  52.21 and receiving the 
report to be submitted by the ACRS under Sec.  52.23, the Commission 
may issue an early site permit, in the form the Commission deems 
appropriate, if the Commission finds that:
    (1) An application for an early site permit meets the applicable 
standards and requirements of the Act and the Commission's regulations;
    (2) Notifications, if any, to other agencies or bodies have been 
duly made;
    (3) There is reasonable assurance that the site is in conformity 
with the provisions of the Act, and the Commission's regulations;
    (4) The applicant is technically qualified to engage in any 
activities authorized;
    (5) The proposed inspections, tests, analyses and acceptance 
criteria, including any on emergency planning, are necessary and 
sufficient, within the scope of the early site permit, to provide 
reasonable assurance that the facility has been constructed and will be 
operated in conformity with the license, the provisions of the Act, and 
the Commission's regulations;
    (6) Issuance of the permit will not be inimical to the common 
defense and security or to the health and safety of the public;
    (7) Any significant adverse environmental impact resulting from 
activities requested under Sec.  52.17(c) can be redressed; and
    (8) The findings required by subpart A of 10 CFR part 51 have been 
made.
    (b) The early site permit must specify the site characteristics, 
design parameters, and terms and conditions of the early site permit 
the Commission deems appropriate. Before issuance of either a 
construction permit or combined license referencing an early site 
permit, the Commission shall find that any relevant terms and 
conditions of the early site permit have been met. Any terms or 
conditions of the early site permit that could not be met by the time 
of issuance of the construction permit or combined license, must be set 
forth as terms or conditions of the construction permit or combined 
license.
    (c) The early site permit shall specify the activities under Sec.  
52.17(c) that the permit holder is authorized to perform.


Sec.  52.25  Extent of activities permitted.

    If the activities authorized by Sec.  52.24(c) are performed and 
the site is not referenced in an application for a construction permit 
or a combined license issued under subpart C of this part while the 
permit remains valid, then the early site permit remains in effect 
solely for the purpose of site redress, and the holder of the permit 
shall redress the site in accordance with the terms of the site redress 
plan required by Sec.  52.17(c). If, before redress is complete, a use 
not envisaged in the redress plan is found for the site or parts 
thereof, the holder of the permit shall carry out the redress plan to 
the greatest extent possible consistent with the alternate use.


Sec.  52.27  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 
10, nor more than 20 years from the date of issuance.
    (b) An early site permit continues to be valid beyond the date of 
expiration in any proceeding on a construction permit application or a 
combined license application that references the early site permit and 
is docketed before the date of expiration of the early site permit, or, 
if a timely application for renewal of the permit has been docketed, 
before the Commission has determined whether to renew the permit.
    (c) An applicant for a construction permit or combined license may, 
at its own risk, reference in its application a site for which an early 
site permit application has been docketed but not granted.
    (d) Upon issuance of a construction permit or combined license, a 
referenced early site permit is subsumed, to the extent referenced, 
into the construction permit or combined license.


Sec.  52.28  Transfer of early site permit.

    An application to transfer an early site permit will be processed 
under 10 CFR 50.80.


Sec.  52.29  Application for renewal.

    (a) Not less than 12, nor more than 36 months before the expiration 
date stated in the early site permit, or any later renewal period, the 
permit holder may apply for a renewal of the permit. An application for 
renewal must contain all information necessary to bring up to date the 
information and data contained in the previous application.
    (b) Any person whose interests may be affected by renewal of the 
permit may request a hearing on the application for renewal. The 
request for a hearing must comply with 10 CFR 2.309. If a hearing is 
granted, notice of the hearing will be published in accordance with 10 
CFR 2.309.
    (c) An early site permit, either original or renewed, for which a 
timely application for renewal has been filed, remains in effect until 
the Commission 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

[[Page 49525]]

ACRS. The ACRS shall report on those portions of the application which 
concern safety and shall apply the criteria set forth in Sec.  52.31.


Sec.  52.31  Criteria for renewal.

    (a) The Commission shall grant the renewal if it determines that:
    (1) The site complies with the Act, the Commission's regulations, 
and orders applicable and in effect at the time the site permit was 
originally issued; and
    (2) Any new requirements the Commission may wish to impose are:
    (i) Necessary for adequate protection to public health and safety 
or common defense and security;
    (ii) Necessary for compliance with the Commission's regulations, 
and orders applicable and in effect at the time the site permit was 
originally issued; or
    (iii) A substantial increase in overall protection of the public 
health and safety or the common defense and security to be derived from 
the new requirements, and the direct and indirect costs of 
implementation of those requirements are justified in view of this 
increased protection.
    (b) A denial of renewal for failure to comply with the provisions 
of Sec.  52.31(a) does not bar the permit holder or another applicant 
from filing a new application for the site which proposes changes to 
the site or the way that it is used to correct the deficiencies cited 
in the denial of the renewal.


Sec.  52.33  Duration of renewal.

    Each renewal of an early site permit may be for not less than 10, 
nor more than 20 years, plus any remaining years on the early site 
permit then in effect before renewal.


Sec.  52.35  Use of site for other purposes.

    A site for which an early site permit has been issued under this 
subpart may be used for purposes other than those described in the 
permit, including the location of other types of energy facilities. The 
permit holder shall inform the Director of New Reactors or the Director 
of Nuclear Reactor Regulation, as appropriate, (Director) of any 
significant uses for the site which have not been approved in the early 
site permit. The information about the activities must be given to the 
Director at least 30 days in advance of any actual construction or site 
modification for the activities. The information provided could be the 
basis for imposing new requirements on the permit, in accordance with 
the provisions of Sec.  52.39. If the permit holder informs the 
Director that the holder no longer intends to use the site for a 
nuclear power plant, the Director may terminate the permit.


Sec.  52.39  Finality of early site permit determinations.

    (a) Commission finality. (1) Notwithstanding any provision in 10 
CFR 50.109, while an early site permit is in effect under Sec. Sec.  
52.27 or 52.33, the Commission may not change or impose new site 
characteristics, design parameters, or terms and conditions, including 
emergency planning requirements, on the early site permit unless the 
Commission:
    (i) Determines that a modification is necessary to bring the permit 
or the site into compliance with the Commission's regulations and 
orders applicable and in effect at the time the permit was issued;
    (ii) Determines the modification is necessary to assure adequate 
protection of the public health and safety or the common defense and 
security;
    (iii) Determines that a modification is necessary based on an 
update under paragraph (b) of this section; or
    (iv) Issues a variance requested under paragraph (d) of this 
section.
    (2) In making the findings required for issuance of a construction 
permit or combined license, or the findings required by Sec.  52.103, 
or in any enforcement hearing other than one initiated by the 
Commission under paragraph (a)(1) of this section, if the application 
for the construction permit or combined license references an early 
site permit, the Commission shall treat as resolved those matters 
resolved in the proceeding on the application for issuance or renewal 
of the early site permit, except as provided for in paragraphs (b), 
(c), and (d) of this section.
    (i) If the early site permit approved an emergency plan (or major 
features thereof) that is in use by a licensee of a nuclear power 
plant, the Commission shall treat as resolved changes to the early site 
permit emergency plan (or major features thereof) that are identical to 
changes made to the licensee's emergency plans in compliance with Sec.  
50.54(q) of this chapter occurring after issuance of the early site 
permit.
    (ii) If the early site permit approved an emergency plan (or major 
features thereof) that is not in use by a licensee of a nuclear power 
plant, the Commission shall treat as resolved changes that are 
equivalent to those that could be made under Sec.  50.54(q) of this 
chapter without prior NRC approval had the emergency plan been in use 
by a licensee.
    (b) Updating of early site permit-emergency preparedness. An 
applicant for a construction permit, operating license, or combined 
license who has filed an application referencing an early site permit 
issued under this subpart shall update the emergency preparedness 
information that was provided under Sec.  52.17(b), and discuss whether 
the updated information materially changes the bases for compliance 
with applicable NRC requirements.
    (c) Hearings and petitions. (1) In any proceeding for the issuance 
of a construction permit, operating license, or combined license 
referencing an early site permit, contentions on the following matters 
may be litigated in the same manner as other issues material to the 
proceeding:
    (i) The nuclear power reactor proposed to be built does not fit 
within one or more of the site characteristics or design parameters 
included in the early site permit;
    (ii) One or more of the terms and conditions of the early site 
permit have not been met;
    (iii) A variance requested under paragraph (d) of this section is 
unwarranted or should be modified;
    (iv) New or additional information is provided in the application 
that substantially alters the bases for a previous NRC conclusion or 
constitutes a sufficient basis for the Commission to modify or impose 
new terms and conditions related to emergency preparedness; or
    (v) Any significant environmental issue that was not resolved in 
the early site permit proceeding, or any issue involving the impacts of 
construction and operation of the facility that was resolved in the 
early site permit proceeding for which significant new information has 
been identified.
    (2) Any person may file a petition requesting that the site 
characteristics, design parameters, or terms and conditions of the 
early site permit should be modified, or that the permit should be 
suspended or revoked. The petition will be considered in accordance 
with Sec.  2.206 of this chapter. Before construction commences, the 
Commission shall consider the petition and determine whether any 
immediate action is required. If the petition is granted, an 
appropriate order will be issued. Construction under the construction 
permit or combined license will not be affected by the granting of the 
petition unless the order is made immediately effective. Any change 
required by the Commission in response to the petition must meet the 
requirements of paragraph (a)(1) of this section.
    (d) Variances. An applicant for a construction permit, operating 
license, or combined license referencing an early site permit may 
include in its application a request for a variance from

[[Page 49526]]

one or more site characteristics, design parameters, or terms and 
conditions of the early site permit, or from the site safety analysis 
report. In determining whether to grant the variance, the Commission 
shall apply the same technically relevant criteria applicable to the 
application for the original or renewed early site permit. Once a 
construction permit or combined license referencing an early site 
permit is issued, variances from the early site permit will not be 
granted for that construction permit or combined license.
    (e) Early site permit amendment. The holder of an early site permit 
may not make changes to the early site permit, including the site 
safety analysis report, without prior Commission approval. The request 
for a change to the early site permit must be in the form of an 
application for a license amendment, and must meet the requirements of 
10 CFR 50.90 and 50.92.
    (f) Information requests. Except for information requests seeking 
to verify compliance with the current licensing basis of the early site 
permit, information requests to the holder of an early site permit must 
be evaluated before issuance to ensure that the burden to be imposed on 
respondents is justified in view of the potential safety significance 
of the issue to be addressed in the requested information. Each 
evaluation performed by the NRC staff must be in accordance with 10 CFR 
50.54(f), and must be approved by the Executive Director for Operations 
or his or her designee before issuance of the request.

Subpart B--Standard Design Certifications


Sec.  52.41  Scope of subpart.

    (a) This subpart sets forth the requirements and procedures 
applicable to Commission issuance of rules granting standard design 
certifications for nuclear power facilities separate from the filing of 
an application for a construction permit or combined license for such a 
facility.
    (b)(1) Any person may seek a standard design certification for an 
essentially complete nuclear power plant design which is an 
evolutionary change from light water reactor designs of plants which 
have been licensed and in commercial operation before April 18, 1989.
    (2) Any person may also seek a standard design certification for a 
nuclear power plant design which differs significantly from the light 
water reactor designs described in paragraph (b)(1) of this section or 
uses simplified, inherent, passive, or other innovative means to 
accomplish its safety functions.


Sec.  52.43  Relationship to other subparts.

    (a) This subpart applies to a person that requests a standard 
design certification from the NRC separately from an application for a 
combined license filed under subpart C of this part for a nuclear power 
facility. An applicant for a combined license may reference a standard 
design certification.
    (b) Subpart E of this part governs the NRC staff review and 
approval of a final standard design. Subpart E may be used 
independently of the provisions in this subpart.
    (c) Subpart F of this part governs the issuance of licenses to 
manufacture nuclear power reactors to be installed and operated at 
sites not identified in the manufacturing license application. Subpart 
F may be used independently of the provisions in this subpart. However, 
an applicant for a manufacturing license under subpart F may reference 
a design certification.


Sec.  52.45  Filing of applications.

    (a) An application for design certification may be filed 
notwithstanding the fact that an application for a construction permit, 
combined license, or manufacturing license for such a facility has not 
been filed.
    (b) The application must comply with the applicable filing 
requirements of Sec. Sec.  52.3 and Sec. Sec.  2.811 through 2.819 of 
this chapter.
    (c) The fees associated with the review of an application for the 
initial issuance or renewal of a standard design certification are set 
forth in 10 CFR part 170.


Sec.  52.46  Contents of applications; general information.

    The application must contain all of the information required by 10 
CFR 50.33(a) through (c) and (j).


Sec.  52.47  Contents of applications; technical information.

    The application must contain a level of design information 
sufficient to enable the Commission to judge the applicant's proposed 
means of assuring that construction conforms to the design and to reach 
a final conclusion on all safety questions associated with the design 
before the certification is granted. The information submitted for a 
design certification must include performance requirements and design 
information sufficiently detailed to permit the preparation of 
acceptance and inspection requirements by the NRC, and procurement 
specifications and construction and installation specifications by an 
applicant. The Commission will require, before design certification, 
that information normally contained in certain procurement 
specifications and construction and installation specifications be 
completed and available for audit if the information is necessary for 
the Commission to make its safety determination.
    (a) The application must contain a final safety analysis report 
(FSAR) that describes the facility, presents the design bases and the 
limits on its operation, and presents a safety analysis of the 
structures, systems, and components and of the facility as a whole, and 
must include the following information:
    (1) The site parameters postulated for the design, and an analysis 
and evaluation of the design in terms of those site parameters;
    (2) A description and analysis of the structures, systems, and 
components (SSCs) of the facility, with emphasis upon performance 
requirements, the bases, with technical justification therefor, upon 
which these requirements have been established, and the evaluations 
required to show that safety functions will be accomplished. It is 
expected that the standard plant will reflect through its design, 
construction, and operation an extremely low probability for accidents 
that could result in the release of significant quantities of 
radioactive fission products. The description shall be sufficient to 
permit understanding of the system designs and their relationship to 
the safety evaluations. Such items as the reactor core, reactor coolant 
system, instrumentation and control systems, electrical systems, 
containment system, other engineered safety features, auxiliary and 
emergency systems, power conversion systems, radioactive waste handling 
systems, and fuel handling systems shall be discussed insofar as they 
are pertinent. The following power reactor design characteristics will 
be taken into consideration by the Commission:
    (i) Intended use of the reactor including the proposed maximum 
power level and the nature and inventory of contained radioactive 
materials;
    (ii) The extent to which generally accepted engineering standards 
are applied to the design of the reactor;
    (iii) The extent to which the reactor incorporates unique, unusual 
or enhanced safety features having a significant bearing on the 
probability or consequences of accidental release of radioactive 
materials; and

[[Page 49527]]

    (iv) The safety features that are to be engineered into the 
facility and those barriers that must be breached as a result of an 
accident before a release of radioactive material to the environment 
can occur. Special attention must be directed to plant design features 
intended to mitigate the radiological consequences of accidents. In 
performing this assessment, an applicant shall assume a fission product 
release \3\ from the core into the containment assuming that the 
facility is operated at the ultimate power level contemplated. The 
applicant shall perform an evaluation and analysis of the postulated 
fission product release, using the expected demonstrable containment 
leak rate and any fission product cleanup systems intended to mitigate 
the consequences of the accidents, together with applicable postulated 
site parameters, including site meteorology, to evaluate the offsite 
radiological consequences. The evaluation must determine that:
---------------------------------------------------------------------------

    \3\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. These accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (A) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 25 rem \4\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------

    \4\ A whole body dose of 25 rem has been stated to correspond 
numerically to the once in a lifetime accidental or emergency dose 
for radiation workers which, according to NCRP recommendations at 
the time could be disregarded in the determination of their 
radiation exposure status (see NBS Handbook 69 dated June 5, 1959). 
However, its use is not intended to imply that this number 
constitutes an acceptable limit for an emergency dose to the public 
under accident conditions. This dose value has been set forth in 
this section as a reference value, which can be used in the 
evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accident.
---------------------------------------------------------------------------

    (B) An individual located at any point on the outer boundary of the 
low population zone, who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
TEDE;
    (3) The design of the facility including:
    (i) The principal design criteria for the facility. Appendix A to 
10 CFR part 50, general design criteria (GDC), establishes minimum 
requirements for the principal design criteria for water-cooled nuclear 
power plants similar in design and location to plants for which 
construction permits have previously been issued by the Commission and 
provides guidance to applicants in establishing principal design 
criteria for other types of nuclear power units;
    (ii) The design bases and the relation of the design bases to the 
principal design criteria;
    (iii) Information relative to materials of construction, general 
arrangement, and approximate dimensions, sufficient to provide 
reasonable assurance that the design will conform to the design bases 
with an adequate margin for safety;
    (4) An analysis and evaluation of the design and performance of 
structures, systems, and components with the objective of assessing the 
risk to public health and safety resulting from operation of the 
facility and including determination of the margins of safety during 
normal operations and transient conditions anticipated during the life 
of the facility, and the adequacy of structures, systems, and 
components provided for the prevention of accidents and the mitigation 
of the consequences of accidents. Analysis and evaluation of emergency 
core cooling system (ECCS) cooling performance and the need for high-
point vents following postulated loss-of-coolant accidents shall be 
performed in accordance with the requirements of Sec. Sec.  50.46 and 
50.46a of this chapter;
    (5) The kinds and quantities of radioactive materials expected to 
be produced in the operation and the means for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in part 20 of this chapter;
    (6) The information required by Sec.  20.1406 of this chapter;
    (7) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (8) The information necessary to demonstrate compliance with any 
technically relevant portions of the Three Mile Island requirements set 
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), 
and (f)(3)(v);
    (9) For applications for light-water-cooled nuclear power plants, 
an evaluation of the standard plant design against the Standard Review 
Plan (SRP) revision in effect 6 months before the docket date of the 
application. The evaluation required by this section shall include an 
identification and description of all differences in design features, 
analytical techniques, and procedural measures proposed for the design 
and those corresponding features, techniques, and measures given in the 
SRP acceptance criteria. Where a difference exists, the evaluation 
shall discuss how the proposed alternative provides an acceptable 
method of complying with the Commission's regulations, or portions 
thereof, that underlie the corresponding SRP acceptance criteria. The 
SRP is not a substitute for the regulations, and compliance is not a 
requirement.
    (10) The information with respect to the design of equipment to 
maintain control over radioactive materials in gaseous and liquid 
effluents produced during normal reactor operations described in 10 CFR 
50.34a(e);
    (11) Proposed technical specifications prepared in accordance with 
the requirements of Sec. Sec.  50.36 and 50.36a of this chapter;
    (12) An analysis and description of the equipment and systems for 
combustible gas control as required by 10 CFR 50.44;
    (13) The list of electric equipment important to safety that is 
required by 10 CFR 50.49(d);
    (14) A description of protection provided against pressurized 
thermal shock events, including projected values of the reference 
temperature for reactor vessel beltline materials as defined in 10 CFR 
50.60 and 50.61;
    (15) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram events in Sec.  50.62;
    (16) A coping analysis, and any design features necessary to 
address station blackout, as required by 10 CFR 50.63;
    (17) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68(b)(2)-(b)(4);
    (18) A description and analysis of the fire protection design 
features for the standard plant necessary to comply with 10 CFR part 
50, appendix A, GDC 3, and Sec.  50.48 of this chapter;
    (19) A description of the quality assurance program applied to the 
design of the structures, systems, and components of the facility. 
Appendix B to 10 CFR part 50, ``Quality Assurance Criteria for Nuclear 
Power Plants and Fuel Reprocessing Plants,'' sets forth the 
requirements for quality assurance programs for nuclear power plants. 
The description of the quality assurance program for a nuclear power 
plant shall include a discussion of how the applicable requirements of 
appendix B to 10 CFR part 50 were satisfied;
    (20) The information necessary to demonstrate that the standard 
plant complies with the earthquake

[[Page 49528]]

engineering criteria in 10 CFR part 50, appendix S;
    (21) Proposed technical resolutions of those Unresolved Safety 
Issues and medium- and high-priority generic safety issues which are 
identified in the version of NUREG-0933 current on the date up to 6 
months before the docket date of the application and which are 
technically relevant to the design;
    (22) The information necessary to demonstrate how operating 
experience insights have been incorporated into the plant design;
    (23) For light-water reactor designs, a description and analysis of 
design features for the prevention and mitigation of severe accidents, 
e.g., challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen combustion, and containment bypass;
    (24) A representative conceptual design for those portions of the 
plant for which the application does not seek certification, to aid the 
NRC in its review of the FSAR and to permit assessment of the adequacy 
of the interface requirements in paragraph (a)(25) of this section;
    (25) The interface requirements to be met by those portions of the 
plant for which the application does not seek certification. These 
requirements must be sufficiently detailed to allow completion of the 
FSAR;
    (26) Justification that compliance with the interface requirements 
of paragraph (a)(25) of this section is verifiable through inspections, 
tests, or analyses. The method to be used for verification of interface 
requirements must be included as part of the proposed ITAAC required by 
paragraph (b)(1) of this section; and
    (27) A description of the design-specific probabilistic risk 
assessment (PRA) and its results.
    (b) The application must also contain:
    (1) The proposed inspections, tests, analyses, and acceptance 
criteria that are necessary and sufficient to provide reasonable 
assurance that, if the inspections, tests, and analyses are performed 
and the acceptance criteria met, a facility that incorporates the 
design certification has been constructed and will be operated in 
conformity with the design certification, the provisions of the Act, 
and the Commission's rules and regulations; and
    (2) An environmental report as required by 10 CFR 51.55.
    (c) This paragraph applies, according to its provisions, to 
particular applications:
    (1) An application for certification of a nuclear power reactor 
design that is an evolutionary change from light-water reactor designs 
of plants that have been licensed and in commercial operation before 
April 18, 1989, must provide an essentially complete nuclear power 
plant design except for site-specific elements such as the service 
water intake structure and the ultimate heat sink;
    (2) An application for certification of a nuclear power reactor 
design that differs significantly from the light-water reactor designs 
described in paragraph (c)(1) of this section or uses simplified, 
inherent, passive, or other innovative means to accomplish its safety 
functions must provide an essentially complete nuclear power reactor 
design except for site-specific elements such as the service water 
intake structure and the ultimate heat sink, and must meet the 
requirements of 10 CFR 50.43(e); and
    (3) An application for certification of a modular nuclear power 
reactor design must describe and analyze the possible operating 
configurations of the reactor modules with common systems, interface 
requirements, and system interactions. The final safety analysis must 
also account for differences among the configurations, including any 
restrictions that will be necessary during the construction and startup 
of a given module to ensure the safe operation of any module already 
operating.


Sec.  52.48  Standards for review of applications.

    Applications filed under this subpart will be reviewed for 
compliance with the standards set out in 10 CFR parts 20, 50 and its 
appendices, 51, 73, and 100.


Sec.  52.51  Administrative review of applications.

    (a) A standard design certification is a rule that will be issued 
in accordance with the provisions of subpart H of 10 CFR part 2, as 
supplemented by the provisions of this section. The Commission shall 
initiate the rulemaking after an application has been filed under Sec.  
52.45 and shall specify the procedures to be used for the rulemaking. 
The notice of proposed rulemaking published in the Federal Register 
must provide an opportunity for the submission of comments on the 
proposed design certification rule. If, at the time a proposed design 
certification rule is published in the Federal Register under this 
paragraph (a), the Commission decides that a legislative hearing should 
be held, the information required by 10 CFR 2.1502(c) must be included 
in the Federal Register document for the proposed design certification.
    (b) Following the submission of comments on the proposed design 
certification rule, the Commission may, at its discretion, hold a 
legislative hearing under the procedures in subpart O of part 2 of this 
chapter. The Commission shall publish a document in the Federal 
Register of its decision to hold a legislative hearing. The document 
shall contain the information specified in paragraph (c) of this 
section, and specify whether the Commission or a presiding officer will 
conduct the legislative hearing.
    (c) Notwithstanding anything in 10 CFR 2.390 to the contrary, 
proprietary information will be protected in the same manner and to the 
same extent as proprietary information submitted in connection with 
applications for licenses, provided that the design certification shall 
be published in Chapter I of this title.


Sec.  52.53  Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).

    The Commission shall refer a copy of the application to the ACRS. 
The ACRS shall report on those portions of the application which 
concern safety.


Sec.  52.54  Issuance of standard design certification.

    (a) After conducting a rulemaking proceeding under Sec.  52.51 on 
an application for a standard design certification and receiving the 
report to be submitted by the Advisory Committee on Reactor Safeguards 
under Sec.  52.53, the Commission may issue a standard design 
certification in the form of a rule for the design which is the subject 
of the application, if the Commission determines that:
    (1) The application meets the applicable standards and requirements 
of the Atomic Energy Act and the Commission's regulations;
    (2) Notifications, if any, to other agencies or bodies have been 
duly made;
    (3) There is reasonable assurance that the standard design conforms 
with the provisions of the Act, and the Commission's regulations;
    (4) The applicant is technically qualified;
    (5) The proposed inspections, tests, analyses, and acceptance 
criteria are necessary and sufficient, within the scope of the standard 
design, to provide reasonable assurance that, if the inspections, 
tests, and analyses are performed and the acceptance criteria met, the 
facility has been constructed and will be operated in accordance with 
the design certification, the provisions of the Act, and the 
Commission's regulations;

[[Page 49529]]

    (6) Issuance of the standard design certification will not be 
inimical to the common defense and security or to the health and safety 
of the public;
    (7) The findings required by subpart A of part 51 of this chapter 
have been made; and
    (8) The applicant has implemented the quality assurance program 
described or referenced in the safety analysis report.
    (b) The design certification rule must specify the site parameters, 
design characteristics, and any additional requirements and 
restrictions of the design certification rule.
    (c) After the Commission has adopted a final design certification 
rule, the applicant shall not permit any individual to have access to 
or any facility to possess restricted data or classified National 
Security Information until the individual and/or facility has been 
approved for access under the provisions of 10 CFR parts 25 and/or 95, 
as applicable.


Sec.  52.55  Duration of certification.

    (a) Except as provided in paragraph (b) of this section, a standard 
design certification issued under this subpart is valid for 15 years 
from the date of issuance.
    (b) A standard design certification continues to be valid beyond 
the date of expiration in any proceeding on an application for a 
combined license or an operating license that references the standard 
design certification and is docketed either before the date of 
expiration of the certification, or, if a timely application for 
renewal of the certification has been filed, before the Commission has 
determined whether to renew the certification. A design certification 
also continues to be valid beyond the date of expiration in any hearing 
held under Sec.  52.103 before operation begins under a combined 
license that references the design certification.
    (c) An applicant for a construction permit or a combined license 
may, at its own risk, reference in its application a design for which a 
design certification application has been docketed but not granted.


Sec.  52.57  Application for renewal.

    (a) Not less than 12 nor more than 36 months before the expiration 
of the initial 15-year period, or any later renewal period, any person 
may apply for renewal of the certification. An application for renewal 
must contain all information necessary to bring up to date the 
information and data contained in the previous application. The 
Commission will require, before renewal of certification, that 
information normally contained in certain procurement specifications 
and construction and installation specifications be completed and 
available for audit if this information is necessary for the Commission 
to make its safety determination. Notice and comment procedures must be 
used for a rulemaking proceeding on the application for renewal. The 
Commission, in its discretion, may require the use of additional 
procedures in individual renewal proceedings.
    (b) A design certification, either original or renewed, for which a 
timely application for renewal has been filed remains in effect until 
the Commission has determined whether to renew the certification. If 
the certification is not renewed, it continues to be valid in certain 
proceedings, in accordance with the provisions of Sec.  52.55.
    (c) The Commission shall refer a copy of the application for 
renewal to the Advisory Committee on Reactor Safeguards (ACRS). The 
ACRS shall report on those portions of the application which concern 
safety and shall apply the criteria set forth in Sec.  52.59.


Sec.  52.59  Criteria for renewal.

    (a) The Commission shall issue a rule granting the renewal if the 
design, either as originally certified or as modified during the 
rulemaking on the renewal, complies with the Atomic Energy Act and the 
Commission's regulations applicable and in effect at the time the 
certification was issued.
    (b) The Commission may impose other requirements if it determines 
that:
    (1) They are necessary for adequate protection to public health and 
safety or common defense and security;
    (2) They are necessary for compliance with the Commission's 
regulations and orders applicable and in effect at the time the design 
certification was issued; or
    (3) There is a substantial increase in overall protection of the 
public health and safety or the common defense and security to be 
derived from the new requirements, and the direct and indirect costs of 
implementing those requirements are justified in view of this increased 
protection.
    (c) In addition, the applicant for renewal may request an amendment 
to the design certification. The Commission shall grant the amendment 
request if it determines that the amendment will comply with the Atomic 
Energy Act and the Commission's regulations in effect at the time of 
renewal. If the amendment request entails such an extensive change to 
the design certification that an essentially new standard design is 
being proposed, an application for a design certification must be filed 
in accordance with this subpart.
    (d) Denial of renewal does not bar the applicant, or another 
applicant, from filing a new application for certification of the 
design, which proposes design changes that correct the deficiencies 
cited in the denial of the renewal.


Sec.  52.61  Duration of renewal.

    Each renewal of certification for a standard design will be for not 
less than 10, nor more than 15 years.


Sec.  52.63  Finality of standard design certifications.

    (a)(1) Notwithstanding any provision in 10 CFR 50.109, while a 
standard design certification rule is in effect under Sec. Sec.  52.55 
or 52.61, the Commission may not modify, rescind, or impose new 
requirements on the certification information, whether on its own 
motion, or in response to a petition from any person, unless the 
Commission determines in a rulemaking that the change:
    (i) Is necessary either to bring the certification information or 
the referencing plants into compliance with the Commission's 
regulations applicable and in effect at the time the certification was 
issued;
    (ii) Is necessary to provide adequate protection of the public 
health and safety or the common defense and security;
    (iii) Reduces unnecessary regulatory burden and maintains 
protection to public health and safety and the common defense and 
security;
    (iv) Provides the detailed design information to be verified under 
those inspections, tests, analyses, and acceptance criteria (ITAAC) 
which are directed at certification information (i.e., design 
acceptance criteria);
    (v) Is necessary to correct material errors in the certification 
information;
    (vi) Substantially increases overall safety, reliability, or 
security of facility design, construction, or operation, and the direct 
and indirect costs of implementation of the rule change are justified 
in view of this increased safety, reliability, or security; or
    (vii) Contributes to increased standardization of the certification 
information.
    (2)(i) In a rulemaking under Sec.  52.63(a)(1), except for Sec.  
52.63(a)(1)(ii), the Commission will give consideration to whether the 
benefits justify the costs for plants that are already licensed or for 
which an application for a permit or license is under consideration.

[[Page 49530]]

    (ii) The rulemaking procedures for changes under Sec.  52.63(a)(1) 
must provide for notice and opportunity for public comment.
    (3) Any modification the NRC imposes on a design certification rule 
under paragraph (a)(1) of this section will be applied to all plants 
referencing the certified design, except those to which the 
modification has been rendered technically irrelevant by action taken 
under paragraphs (a)(4) or (b)(1) of this section.
    (4) The Commission may not impose new requirements by plant-
specific order on any part of the design of a specific plant 
referencing the design certification rule if that part was approved in 
the design certification while a design certification rule is in effect 
under Sec.  52.55 or Sec.  52.61, unless:
    (i) A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time the 
certification was issued, or to assure adequate protection of the 
public health and safety or the common defense and security; and
    (ii) Special circumstances as defined in 10 CFR 52.7 are present. 
In addition to the factors listed in Sec.  52.7, the Commission shall 
consider whether the special circumstances which Sec.  52.7 requires to 
be present outweigh any decrease in safety that may result from the 
reduction in standardization caused by the plant-specific order.
    (5) Except as provided in 10 CFR 2.335, in making the findings 
required for issuance of a combined license, construction permit, 
operating license, or manufacturing license, or for any hearing under 
Sec.  52.103, the Commission shall treat as resolved those matters 
resolved in connection with the issuance or renewal of a design 
certification rule.
    (b)(1) An applicant or licensee who references a design 
certification rule may request an exemption from one or more elements 
of the certification information. The Commission may grant such a 
request only if it determines that the exemption will comply with the 
requirements of Sec.  52.7. In addition to the factors listed in Sec.  
52.7, the Commission shall consider whether the special circumstances 
that Sec.  52.7 requires to be present outweigh any decrease in safety 
that may result from the reduction in standardization caused by the 
exemption. The granting of an exemption on request of an applicant is 
subject to litigation in the same manner as other issues in the 
operating license or combined license hearing.
    (2) Subject to Sec.  50.59 of this chapter, a licensee who 
references a design certification rule may make departures from the 
design of the nuclear power facility, without prior Commission 
approval, unless the proposed departure involves a change to the design 
as described in the rule certifying the design. The licensee shall 
maintain records of all departures from the facility and these records 
must be maintained and available for audit until the date of 
termination of the license.
    (c) The Commission will require, before granting a construction 
permit, combined license, operating license, or manufacturing license 
which references a design certification rule, that information normally 
contained in certain procurement specifications and construction and 
installation specifications be completed and available for audit if the 
information is necessary for the Commission to make its safety 
determinations, including the determination that the application is 
consistent with the certification information. This information may be 
acquired by appropriate arrangements with the design certification 
applicant.

Subpart C--Combined Licenses


Sec.  52.71  Scope of subpart.

    This subpart sets out the requirements and procedures applicable to 
Commission issuance of combined licenses for nuclear power facilities.


Sec.  52.73  Relationship to other subparts.

    (a) An application for a combined license under this subpart may, 
but need not, reference a standard design certification, standard 
design approval, or manufacturing license issued under subparts B, E, 
or F of this part, respectively, or an early site permit issued under 
subpart A of this part. In the absence of a demonstration that an 
entity other than the one originally sponsoring and obtaining a design 
certification is qualified to supply a design, the Commission will 
entertain an application for a combined license that references a 
standard design certification issued under subpart B of this part only 
if the entity that sponsored and obtained the certification supplies 
the design for the applicant's use.
    (b) The Commission will require, before granting a combined license 
that references a standard design certification, that information 
normally contained in certain procurement specifications and 
construction and installation specifications be completed and available 
for audit if the information is necessary for the Commission to make 
its safety determinations, including the determination that the 
application is consistent with the certification information.


Sec.  52.75  Filing of applications.

    (a) Any person except one excluded by 10 CFR 50.38 may file an 
application for a combined license for a nuclear power facility with 
the Director of New Reactors or the Director of Nuclear Reactor 
Regulation, as appropriate.
    (b) The application must comply with the applicable filing 
requirements of Sec. Sec.  52.3 and 50.30 of this chapter.
    (c) The fees associated with the filing and review of the 
application are set forth in 10 CFR part 170.


Sec.  52.77  Contents of applications; general information.

    The application must contain all of the information required by 10 
CFR 50.33.


Sec.  52.79  Contents of applications; technical information in final 
safety analysis report.

    (a) The application must contain a final safety analysis report 
that describes the facility, presents the design bases and the limits 
on its operation, and presents a safety analysis of the structures, 
systems, and components of the facility as a whole. The final safety 
analysis report shall include the following information, at a level of 
information sufficient to enable the Commission to reach a final 
conclusion on all safety matters that must be resolved by the 
Commission before issuance of a combined license:
    (1)(i) The boundaries of the site;
    (ii) The proposed general location of each facility on the site;
    (iii) The seismic, meteorological, hydrologic, and geologic 
characteristics of the proposed site with appropriate consideration of 
the most severe of the natural phenomena that have been historically 
reported for the site and surrounding area and with sufficient margin 
for the limited accuracy, quantity, and time in which the historical 
data have been accumulated;
    (iv) The location and description of any nearby industrial, 
military, or transportation facilities and routes;
    (v) The existing and projected future population profile of the 
area surrounding the site;
    (vi) A description and safety assessment of the site on which the 
facility is to be located. The assessment must contain an analysis and 
evaluation of the major structures, systems, and components of the 
facility that bear significantly on the acceptability of the site under 
the radiological consequence evaluation factors identified in 
paragraphs (a)(1)(vi)(A) and (a)(1)(vi)(B)

[[Page 49531]]

of this section. In performing this assessment, an applicant shall 
assume a fission product release \5\ from the core into the containment 
assuming that the facility is operated at the ultimate power level 
contemplated. The applicant shall perform an evaluation and analysis of 
the postulated fission product release, using the expected demonstrable 
containment leak rate and any fission product cleanup systems intended 
to mitigate the consequences of the accidents, together with applicable 
site characteristics, including site meteorology, to evaluate the 
offsite radiological consequences. Site characteristics must comply 
with part 100 of this chapter. The evaluation must determine that:
---------------------------------------------------------------------------

    \5\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. These accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (A) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 25 rem \6\ total effective dose equivalent (TEDE).
---------------------------------------------------------------------------

    \6\ A whole body dose of 25 rem has been stated to correspond 
numerically to the once in a lifetime accidental or emergency dose 
for radiation workers which, according to NCRP recommendations at 
the time could be disregarded in the determination of their 
radiation exposure status (see NBS Handbook 69 dated June 5, 1959). 
However, its use is not intended to imply that this number 
constitutes an acceptable limit for an emergency dose to the public 
under accident conditions. Rather, this dose value has been set 
forth in this section as a reference value, which can be used in the 
evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accident.
---------------------------------------------------------------------------

    (B) An individual located at any point on the outer boundary of the 
low population zone, who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
TEDE; and
    (2) A description and analysis of the structures, systems, and 
components of the facility with emphasis upon performance requirements, 
the bases, with technical justification therefor, upon which these 
requirements have been established, and the evaluations required to 
show that safety functions will be accomplished. It is expected that 
reactors will reflect through their design, construction, and operation 
an extremely low probability for accidents that could result in the 
release of significant quantities of radioactive fission products. The 
descriptions shall be sufficient to permit understanding of the system 
designs and their relationship to safety evaluations. Items such as the 
reactor core, reactor coolant system, instrumentation and control 
systems, electrical systems, containment system, other engineered 
safety features, auxiliary and emergency systems, power conversion 
systems, radioactive waste handling systems, and fuel handling systems 
shall be discussed insofar as they are pertinent. The following power 
reactor design characteristics and proposed operation will be taken 
into consideration by the Commission:
    (i) Intended use of the reactor including the proposed maximum 
power level and the nature and inventory of contained radioactive 
materials;
    (ii) The extent to which generally accepted engineering standards 
are applied to the design of the reactor;
    (iii) The extent to which the reactor incorporates unique, unusual 
or enhanced safety features having a significant bearing on the 
probability or consequences of accidental release of radioactive 
materials;
    (iv) The safety features that are to be engineered into the 
facility and those barriers that must be breached as a result of an 
accident before a release of radioactive material to the environment 
can occur. Special attention must be directed to plant design features 
intended to mitigate the radiological consequences of accidents. In 
performing this assessment, an applicant shall assume a fission product 
release \7\ from the core into the containment assuming that the 
facility is operated at the ultimate power level contemplated;
---------------------------------------------------------------------------

    \7\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. These accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (3) The kinds and quantities of radioactive materials expected to 
be produced in the operation and the means for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in part 20 of this chapter;
    (4) The design of the facility including:
    (i) The principal design criteria for the facility. Appendix A to 
part 50 of this chapter, ``General Design Criteria for Nuclear Power 
Plants,'' establishes minimum requirements for the principal design 
criteria for water-cooled nuclear power plants similar in design and 
location to plants for which construction permits have previously been 
issued by the Commission and provides guidance to applicants in 
establishing principal design criteria for other types of nuclear power 
units;
    (ii) The design bases and the relation of the design bases to the 
principal design criteria;
    (iii) Information relative to materials of construction, 
arrangement, and dimensions, sufficient to provide reasonable assurance 
that the design will conform to the design bases with adequate margin 
for safety.
    (5) An analysis and evaluation of the design and performance of 
structures, systems, and components with the objective of assessing the 
risk to public health and safety resulting from operation of the 
facility and including determination of the margins of safety during 
normal operations and transient conditions anticipated during the life 
of the facility, and the adequacy of structures, systems, and 
components provided for the prevention of accidents and the mitigation 
of the consequences of accidents. Analysis and evaluation of ECCS 
cooling performance and the need for high-point vents following 
postulated loss-of-coolant accidents shall be performed in accordance 
with the requirements of Sec. Sec.  50.46 and 50.46a of this chapter;
    (6) A description and analysis of the fire protection design 
features for the reactor necessary to comply with 10 CFR part 50, 
appendix A, GDC 3, and Sec.  50.48 of this chapter;
    (7) A description of protection provided against pressurized 
thermal shock events, including projected values of the reference 
temperature for reactor vessel beltline materials as defined in 
Sec. Sec.  50.60 and 50.61(b)(1) and (b)(2) of this chapter;
    (8) An analysis and description of the equipment and systems for 
combustible gas control as required by Sec.  50.44 of this chapter;
    (9) The coping analyses, and any design features necessary to 
address station blackout, as described in Sec.  50.63 of this chapter;
    (10) A description of the program, and its implementation, required 
by Sec.  50.49(a) of this chapter for the environmental qualification 
of electric equipment important to safety and the list of electric 
equipment important to safety that is required by 10 CFR 50.49(d);
    (11) A description of the program(s), and their implementation, 
necessary to ensure that the systems and components meet the 
requirements of the ASME

[[Page 49532]]

Boiler and Pressure Vessel Code and the ASME Code for Operation and 
Maintenance of Nuclear Power Plants in accordance with 50.55a of this 
chapter;
    (12) A description of the primary containment leakage rate testing 
program, and its implementation, necessary to ensure that the 
containment meets the requirements of appendix J to 10 CFR part 50;
    (13) A description of the reactor vessel material surveillance 
program required by appendix H to 10 CFR part 50 and its 
implementation;
    (14) A description of the operator training program, and its 
implementation, necessary to meet the requirements of 10 CFR part 55;
    (15) A description of the program, and its implementation, for 
monitoring the effectiveness of maintenance necessary to meet the 
requirements of Sec.  50.65 of this chapter;
    (16)(i) The information with respect to the design of equipment to 
maintain control over radioactive materials in gaseous and liquid 
effluents produced during normal reactor operations, as described in 
Sec.  50.34a(d) of this chapter;
    (ii) A description of the process and effluent monitoring and 
sampling program required by appendix I to 10 CFR part 50 and its 
implementation.
    (17) The information with respect to compliance with technically 
relevant positions of the Three Mile Island requirements in Sec.  
50.34(f) of this chapter, with the exception of Sec. Sec.  
50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v);
    (18) If the applicant seeks to use risk-informed treatment of SSCs 
in accordance with Sec.  50.69 of this chapter, the information 
required by Sec.  50.69(b)(2) of this chapter;
    (19) Information necessary to demonstrate that the plant complies 
with the earthquake engineering criteria in 10 CFR part 50, appendix S;
    (20) Proposed technical resolutions of those Unresolved Safety 
Issues and medium- and high-priority generic safety issues which are 
identified in the version of NUREG-0933 current on the date up to 6 
months before the docket date of the application and which are 
technically relevant to the design;
    (21) Emergency plans complying with the requirements of Sec.  50.47 
of this chapter, and 10 CFR part 50, appendix E;
    (22)(i) All emergency plan certifications that have been obtained 
from the State and local governmental agencies with emergency planning 
responsibilities must state that:
    (A) The proposed emergency plans are practicable;
    (B) These agencies are committed to participating in any further 
development of the plans, including any required field demonstrations; 
and
    (C) These agencies are committed to executing their 
responsibilities under the plans in the event of an emergency;
    (ii) If certifications cannot be obtained after sustained, good 
faith efforts by the applicant, then the application must contain 
information, including a utility plan, sufficient to show that the 
proposed plans provide reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological emergency 
at the site.
    (23) [Reserved]
    (24) If the application is for a nuclear power reactor design which 
differs significantly from light-water reactor designs that were 
licensed before 1997 or use simplified, inherent, passive, or other 
innovative means to accomplish their safety functions, the application 
must describe how the design meets the requirements in Sec.  50.43(e) 
of this chapter;
    (25) A description of the quality assurance program, applied to the 
design, and to be applied to the fabrication, construction, and 
testing, of the structures, systems, and components of the facility. 
Appendix B to 10 CFR part 50 sets forth the requirements for quality 
assurance programs for nuclear power plants. The description of the 
quality assurance program for a nuclear power plant must include a 
discussion of how the applicable requirements of appendix B to 10 CFR 
part 50 have been and will be satisfied, including a discussion of how 
the quality assurance program will be implemented;
    (26) The applicant's organizational structure, allocations or 
responsibilities and authorities, and personnel qualifications 
requirements for operation;
    (27) Managerial and administrative controls to be used to assure 
safe operation. Appendix B to 10 CFR part 50 sets forth the 
requirements for these controls for nuclear power plants. The 
information on the controls to be used for a nuclear power plant shall 
include a discussion of how the applicable requirements of appendix B 
to 10 CFR part 50 will be satisfied;
    (28) Plans for preoperational testing and initial operations;
    (29)(i) Plans for conduct of normal operations, including 
maintenance, surveillance, and periodic testing of structures, systems, 
and components;
    (ii) Plans for coping with emergencies, other than the plans 
required by Sec.  52.79(a)(21);
    (30) Proposed technical specifications prepared in accordance with 
the requirements of Sec. Sec.  50.36 and 50.36a of this chapter;
    (31) For nuclear power plants to be operated on multi-unit sites, 
an evaluation of the potential hazards to the structures, systems, and 
components important to safety of operating units resulting from 
construction activities, as well as a description of the managerial and 
administrative controls to be used to provide assurance that the 
limiting conditions for operation are not exceeded as a result of 
construction activities at the multi-unit sites;
    (32) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (33) A description of the training program required by Sec.  50.120 
of this chapter and its implementation;
    (34) A description and plans for implementation of an operator 
requalification program. The operator requalification program must as a 
minimum, meet the requirements for those programs contained in Sec.  
55.59 of this chapter;
    (35)(i) A physical security plan, describing how the applicant will 
meet the requirements of 10 CFR part 73 (and 10 CFR part 11, if 
applicable, including the identification and description of jobs as 
required by Sec.  11.11(a) of this chapter, at the proposed facility). 
The plan must list tests, inspections, audits, and other means to be 
used to demonstrate compliance with the requirements of 10 CFR parts 11 
and 73, if applicable;
    (ii) A description of the implementation of the physical security 
plan;
    (36)(i) A safeguards contingency plan in accordance with the 
criteria set forth in appendix C to 10 CFR part 73. The safeguards 
contingency plan shall include plans for dealing with threats, thefts, 
and radiological sabotage, as defined in part 73 of this chapter, 
relating to the special nuclear material and nuclear facilities 
licensed under this chapter and in the applicant's possession and 
control. Each application for this type of license shall include the 
information contained in the applicant's safeguards contingency 
plan.\8\ (Implementing procedures required for this plan need not be 
submitted for approval.)
---------------------------------------------------------------------------

    \8\ A physical security plan that contains all the information 
required in both Sec.  73.55 of this chapter and appendix C to 10 
CFR part 73 satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------

    (ii) A training and qualification plan in accordance with the 
criteria set forth in appendix B to 10 CFR part 73.
    (iii) A description of the implementation of the safeguards

[[Page 49533]]

contingency plan and the training and qualification plan;
    (iv) Each applicant who prepares a physical security plan, a 
safeguards contingency plan, or a guard qualification and training 
plan, shall protect the plans and other related Safeguards Information 
against unauthorized disclosure in accordance with the requirements of 
Sec.  73.21 of this chapter, as appropriate.
    (37) The information necessary to demonstrate how operating 
experience insights have been incorporated into the plant design;
    (38) For light-water reactor designs, a description and analysis of 
design features for the prevention and mitigation of severe accidents, 
e.g., challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen combustion, and containment bypass;
    (39) A description of the radiation protection program required by 
Sec.  20.1101 of this chapter and its implementation.
    (40) A description of the fire protection program required by Sec.  
50.48 of this chapter and its implementation.
    (41) For applications for light-water-cooled nuclear power plant 
combined licenses, an evaluation of the facility against the Standard 
Review Plan (SRP) revision in effect 6 months before the docket date of 
the application. The evaluation required by this section shall include 
an identification and description of all differences in design 
features, analytical techniques, and procedural measures proposed for a 
facility and those corresponding features, techniques, and measures 
given in the SRP acceptance criteria. Where a difference exists, the 
evaluation shall discuss how the proposed alternative provides an 
acceptable method of complying with the Commission's regulations, or 
portions thereof, that underlie the corresponding SRP acceptance 
criteria. The SRP is not a substitute for the regulations, and 
compliance is not a requirement;
    (42) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram (ATWS) events in Sec.  50.62 of this chapter;
    (43) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68 of this chapter;
    (44) A description of the fitness-for-duty program required by 10 
CFR part 26 and its implementation.
    (45) The information required by Sec.  20.1406 of this chapter.
    (46) A description of the plant-specific probabilistic risk 
assessment (PRA) and its results.
    (b) If the combined license application references an early site 
permit, then the following requirements apply:
    (1) The final safety analysis report need not contain information 
or analyses submitted to the Commission in connection with the early 
site permit, provided, however, that the final safety analysis report 
must either include or incorporate by reference the early site permit 
site safety analysis report and must contain, in addition to the 
information and analyses otherwise required, information sufficient to 
demonstrate that the design of the facility falls within the site 
characteristics and design parameters specified in the early site 
permit.
    (2) If the final safety analysis report does not demonstrate that 
design of the facility falls within the site characteristics and design 
parameters, the application shall include a request for a variance that 
complies with the requirements of Sec. Sec.  52.39 and 52.93.
    (3) The final safety analysis report must demonstrate that all 
terms and conditions that have been included in the early site permit, 
other than those imposed under Sec.  50.36b, will be satisfied by the 
date of issuance of the combined license. Any terms or conditions of 
the early site permit that could not be met by the time of issuance of 
the combined license, must be set forth as terms or conditions of the 
combined license.
    (4) If the early site permit approves complete and integrated 
emergency plans, or major features of emergency plans, then the final 
safety analysis report must include any new or additional information 
that updates and corrects the information that was provided under Sec.  
52.17(b), and discuss whether the new or additional information 
materially changes the bases for compliance with the applicable 
requirements. The application must identify changes to the emergency 
plans or major features of emergency plans that have been incorporated 
into the proposed facility emergency plans and that constitute or would 
constitute a decrease in effectiveness under Sec.  50.54(q) of this 
chapter.
    (5) If complete and integrated emergency plans are approved as part 
of the early site permit, new certifications meeting the requirements 
of paragraph (a)(22) of this section are not required.
    (c) If the combined license application references a standard 
design approval, then the following requirements apply:
    (1) The final safety analysis report need not contain information 
or analyses submitted to the Commission in connection with the design 
approval, provided, however, that the final safety analysis report must 
either include or incorporate by reference the standard design approval 
final safety analysis report and must contain, in addition to the 
information and analyses otherwise required, information sufficient to 
demonstrate that the characteristics of the site fall within the site 
parameters specified in the design approval. In addition, the plant-
specific PRA information must use the PRA information for the design 
approval and must be updated to account for site-specific design 
information and any design changes or departures.
    (2) The final safety analysis report must demonstrate that all 
terms and conditions that have been included in the final design 
approval will be satisfied by the date of issuance of the combined 
license.
    (d) If the combined license application references a standard 
design certification, then the following requirements apply:
    (1) The final safety analysis report need not contain information 
or analyses submitted to the Commission in connection with the design 
certification, provided, however, that the final safety analysis report 
must either include or incorporate by reference the standard design 
certification final safety analysis report and must contain, in 
addition to the information and analyses otherwise required, 
information sufficient to demonstrate that the site characteristics 
fall within the site parameters specified in the design certification. 
In addition, the plant-specific PRA information must use the PRA 
information for the design certification and must be updated to account 
for site-specific design information and any design changes or 
departures.
    (2) The final safety analysis report must demonstrate that the 
interface requirements established for the design under Sec.  52.47 
have been met.
    (3) The final safety analysis report must demonstrate that all 
requirements and restrictions set forth in the referenced design 
certification rule, other than those imposed under Sec.  50.36b, must 
be satisfied by the date of issuance of the combined license. Any 
requirements and restrictions set forth in the referenced design 
certification rule that could not be satisfied by the time of issuance 
of the combined license, must be set forth as terms or conditions of 
the combined license.

[[Page 49534]]

    (e) If the combined license application references the use of one 
or more manufactured nuclear power reactors licensed under subpart F of 
this part, then the following requirements apply:
    (1) The final safety analysis report need not contain information 
or analyses submitted to the Commission in connection with the 
manufacturing license, provided, however, that the final safety 
analysis report must either include or incorporate by reference the 
manufacturing license final safety analysis report and must contain, in 
addition to the information and analyses otherwise required, 
information sufficient to demonstrate that the site characteristics 
fall within the site parameters specified in the manufacturing license. 
In addition, the plant-specific PRA information must use the PRA 
information for the manufactured reactor and must be updated to account 
for site-specific design information and any design changes or 
departures.
    (2) The final safety analysis report must demonstrate that the 
interface requirements established for the design have been met.
    (3) The final safety analysis report must demonstrate that all 
terms and conditions that have been included in the manufacturing 
license, other than those imposed under Sec.  50.36b, will be satisfied 
by the date of issuance of the combined license. Any terms or 
conditions of the manufacturing license that could not be met by the 
time of issuance of the combined license, must be set forth as terms or 
conditions of the combined license.


Sec.  52.80  Contents of applications; additional technical 
information.

    The application must contain:
    (a) The proposed inspections, tests, and analyses, including those 
applicable to emergency planning, that the licensee shall perform, and 
the acceptance criteria that are necessary and sufficient to provide 
reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met, the facility has been 
constructed and will be operated in conformity with the combined 
license, the provisions of the Act, and the Commission's rules and 
regulations.
    (1) If the application references an early site permit with ITAAC, 
the early site permit ITAAC must apply to those aspects of the combined 
license which are approved in the early site permit.
    (2) If the application references a standard design certification, 
the ITAAC contained in the certified design must apply to those 
portions of the facility design which are approved in the design 
certification.
    (3) If the application references an early site permit with ITAAC 
or a standard design certification or both, the application may include 
a notification that a required inspection, test, or analysis in the 
ITAAC has been successfully completed and that the corresponding 
acceptance criterion has been met. The Federal Register notification 
required by Sec.  52.85 must indicate that the application includes 
this notification.
    (b) A complete environmental report as required by 10 CFR 51.50(c).
    (c) If the applicant wishes to be able to perform the activities at 
the site allowed by 10 CFR 50.10(e) before issuance of the combined 
license, the applicant must identify and describe the activities that 
are requested and propose a plan for redress of the site in the event 
that the activities are performed and either construction is abandoned 
or the combined license is revoked. 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.  52.81  Standards for review of applications.

    Applications filed under this subpart will be reviewed according to 
the standards set out in 10 CFR parts 20, 50, 51, 54, 55, 73, 100, and 
140.


Sec.  52.83  Finality of referenced NRC approvals; partial initial 
decision on site suitability.

    (a) If the application for a combined license under this subpart 
references an early site permit, design certification rule, standard 
design approval, or manufacturing license, the scope and nature of 
matters resolved for the application and any combined license issued 
are governed by the relevant provisions addressing finality, including 
Sec. Sec.  52.39, 52.63, 52.98, 52.145, and 52.171.
    (b) While a partial decision on site suitability is in effect under 
10 CFR 2.617(b)(2), the scope and nature of matters resolved in the 
proceeding are governed by the finality provisions in 10 CFR 2.629.


Sec.  52.85  Administrative review of applications; hearings.

    A proceeding on a combined license is subject to all applicable 
procedural requirements contained in 10 CFR part 2, including the 
requirements for docketing (Sec.  2.101 of this chapter) and issuance 
of a notice of hearing (Sec.  2.104 of this chapter). If an applicant 
requests a Commission finding on certain ITAAC with the issuance of the 
combined license, then those ITAAC will be identified in the notice of 
hearing. All hearings on combined licenses are governed by the 
procedures contained in 10 CFR part 2.


Sec.  52.87  Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).

    The Commission shall refer a copy of the application to the ACRS. 
The ACRS shall report on those portions of the application that concern 
safety and shall apply the standards referenced in Sec.  52.81, in 
accordance with the finality provisions in Sec.  52.83.


Sec.  52.89  [Reserved].


Sec.  52.91  Authorization to conduct site activities.

    (a) If the application does not reference an early site permit 
which authorizes the applicant to perform site preparation activities, 
the applicant may not perform the site preparation activities allowed 
by 10 CFR 50.10(e)(1) without 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 there is reasonable assurance that redress carried out 
under the site redress plan will achieve an environmentally stable and 
aesthetically acceptable site suitable for whatever non-nuclear use may 
conform with local zoning laws.
    (b) 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).
    (c) If, after an applicant for a combined license has performed the 
activities permitted by paragraph (a) or (b) of this section, and the 
application for the license is withdrawn or denied, then the applicant 
shall redress the site in accord with the terms of the site redress 
plan. If a use not envisaged in the redress plan is found for the site 
or parts before redress is complete, the applicant shall carry out the 
redress plan to the greatest extent possible consistent with the 
alternate use.


Sec.  52.93  Exemptions and variances.

    (a) Applicants for a combined license under this subpart, or any 
amendment to a combined license, may include in the application a 
request for an

[[Page 49535]]

exemption from one or more of the Commission's regulations.
    (1) If the request is for an exemption from any part of a 
referenced design certification rule, the Commission may grant the 
request if it determines that the exemption complies with any exemption 
provisions of the referenced design certification rule, or with Sec.  
52.63 if there are no applicable exemption provisions in the referenced 
design certification rule.
    (2) For all other requests for exemptions, the Commission may grant 
a request if it determines that the exemption complies with Sec.  52.7.
    (b) An applicant for a combined license who has filed an 
application referencing an early site permit issued under subpart A of 
this part may include in the application a request for a variance from 
one or more site characteristics, design parameters, or terms and 
conditions of the permit, or from the site safety analysis report. In 
determining whether to grant the variance, the Commission shall apply 
the same technically relevant criteria as were applicable to the 
application for the original or renewed site permit. Once a 
construction permit or combined license referencing an early site 
permit is issued, variances from the early site permit will not be 
granted for that construction permit or combined license.
    (c) An applicant for a combined license who has filed an 
application referencing a nuclear power reactor manufactured under a 
manufacturing license issued under subpart F of this part may include 
in the application a request for a departure from one or more design 
characteristics, site parameters, terms and conditions, or approved 
design of the manufactured reactor. The Commission may grant a request 
only if it determines that the departure will comply with the 
requirements of 10 CFR 52.7, and that the special circumstances 
outweigh any decrease in safety that may result from the reduction in 
standardization caused by the departure.
    (d) Issuance of a variance under paragraph (b) or a departure under 
paragraph (c) of this section is subject to litigation during the 
combined license proceeding in the same manner as other issues material 
to that proceeding.


Sec.  52.97  Issuance of combined licenses.

    (a)(1) After conducting a hearing in accordance with Sec.  52.85 
and receiving the report submitted by the ACRS, the Commission may 
issue a combined license if the Commission finds that:
    (i) The applicable standards and requirements of the Act and the 
Commission's regulations have been met;
    (ii) Any required notifications to other agencies or bodies have 
been duly made;
    (iii) There is reasonable assurance that the facility will be 
constructed and will operate in conformity with the license, the 
provisions of the Act, and the Commission's regulations.
    (iv) The applicant is technically and financially qualified to 
engage in the activities authorized; and
    (v) Issuance of the license will not be inimical to the common 
defense and security or to the health and safety of the public; and
    (vi) The findings required by subpart A of part 51 of this chapter 
have been made.
    (2) The Commission may also find, at the time it issues the 
combined license, that certain acceptance criteria in one or more of 
the inspections, tests, analyses, and acceptance criteria (ITAAC) in a 
referenced early site permit or standard design certification have been 
met. This finding will finally resolve that those acceptance criteria 
have been met, those acceptance criteria will be deemed to be excluded 
from the combined license, and findings under Sec.  52.103(g) with 
respect to those acceptance criteria are unnecessary.
    (b) The Commission shall identify within the combined license the 
inspections, tests, and analyses, including those applicable to 
emergency planning, that the licensee shall perform, and the acceptance 
criteria that, if met, are necessary and sufficient to provide 
reasonable assurance that the facility has been constructed and will be 
operated in conformity with the license, the provisions of the Act, and 
the Commission's rules and regulations.
    (c) A combined license shall contain the terms and conditions, 
including technical specifications, as the Commission deems necessary 
and appropriate.


Sec.  52.98  Finality of combined licenses; information requests.

    (a) After issuance of a combined license, the Commission may not 
modify, add, or delete any term or condition of the combined license, 
the design of the facility, the inspections, tests, analyses, and 
acceptance criteria contained in the license which are not derived from 
a referenced standard design certification or manufacturing license, 
except in accordance with the provisions of Sec.  52.103 or Sec.  
50.109 of this chapter, as applicable.
    (b) If the combined license does not reference a design 
certification or a reactor manufactured under a subpart F of this part 
manufacturing license, then a licensee may make changes in the facility 
as described in the final safety analysis report (as updated), make 
changes in the procedures as described in the final safety analysis 
report (as updated), and conduct tests or experiments not described in 
the final safety analysis report (as updated) under the applicable 
change processes in 10 CFR part 50 (e.g., Sec. Sec.  50.54, 50.59, or 
50.90 of this chapter).
    (c) If the combined license references a certified design, then--
    (1) Changes to or departures from information within the scope of 
the referenced design certification rule are subject to the applicable 
change processes in that rule; and
    (2) Changes that are not within the scope of the referenced design 
certification rule are subject to the applicable change processes in 10 
CFR part 50, unless they also involve changes to or noncompliance with 
information within the scope of the referenced design certification 
rule. In these cases, the applicable provisions of this section and the 
design certification rule apply.
    (d) If the combined license references a reactor manufactured under 
a subpart F of this part manufacturing license, then--
    (1) Changes to or departures from information within the scope of 
the manufactured reactor's design are subject to the change processes 
in Sec.  52.171; and
    (2) Changes that are not within the scope of the manufactured 
reactor's design are subject to the applicable change processes in 10 
CFR part 50.
    (e) The Commission may issue and make immediately effective any 
amendment to a combined license upon a determination by the Commission 
that the amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person. The amendment may be issued and made 
immediately effective in advance of the holding and completion of any 
required hearing. The amendment will be processed in accordance with 
the procedures specified in 10 CFR 50.91.
    (f) Any modification to, addition to, or deletion from the terms 
and conditions of a combined license, including any modification to, 
addition to, or deletion from the inspections, tests, analyses, or 
related acceptance criteria contained in the license is a proposed 
amendment to the license. There must be an opportunity for a hearing on 
the amendment.
    (g) Except for information sought to verify licensee compliance 
with the

[[Page 49536]]

current licensing basis for that facility, information requests to the 
holder of a combined license must be evaluated before issuance to 
ensure that the burden to be imposed on the licensee is justified in 
view of the potential safety significance of the issue to be addressed 
in the requested information. Each evaluation performed by the NRC 
staff must be in accordance with 10 CFR 50.54(f) and must be approved 
by the Executive Director for Operations or his or her designee before 
issuance of the request.


Sec.  52.99  Inspection during construction.

    (a) The licensee shall submit to the NRC, no later than 1 year 
after issuance of the combined license or at the start of construction 
as defined in 10 CFR 50.10(b), whichever is later, its schedule for 
completing the inspections, tests, or analyses in the ITAAC. The 
licensee shall submit updates to the ITAAC schedule every 6 months 
thereafter and, within 1 year of its scheduled date for initial loading 
of fuel, the licensee shall submit updates to the ITAAC schedule every 
30 days until the final notification is provided to the NRC under 
paragraph (c)(1) of this section.
    (b) With respect to activities subject to an ITAAC, an applicant 
for a combined license may proceed at its own risk with design and 
procurement activities, and a licensee may proceed at its own risk with 
design, procurement, construction, and pre-operational activities, even 
though the NRC may not have found that any one of the prescribed 
acceptance criteria have been met.
    (c)(1) The licensee shall notify the NRC that the prescribed 
inspections, tests, and analyses have been performed and that the 
prescribed acceptance criteria have been met. The notification must 
contain sufficient information to demonstrate that the prescribed 
inspections, tests, and analyses have been performed and that the 
prescribed acceptance criteria have been met.
    (2) If the licensee has not provided, by the date 225 days before 
the scheduled date for initial loading of fuel, the notification 
required by paragraph (c)(1) of this section for all ITAAC, then the 
licensee shall notify the NRC that the prescribed inspections, tests, 
or analyses for all uncompleted ITAAC will be performed and that the 
prescribed acceptance criteria will be met prior to operation. The 
notification must be provided no later than the date 225 days before 
the scheduled date for initial loading of fuel, and must provide 
sufficient information to demonstrate that the prescribed inspections, 
tests, or analyses will be performed and the prescribed acceptance 
criteria for the uncompleted ITAAC will be met, including, but not 
limited to, a description of the specific procedures and analytical 
methods to be used for performing the prescribed inspections, tests, 
and analyses and determining that the prescribed acceptance criteria 
have been met.
    (d)(1) In the event that an activity is subject to an ITAAC derived 
from a referenced standard design certification and the licensee has 
not demonstrated that the ITAAC has been met, the licensee may take 
corrective actions to successfully complete that ITAAC or request an 
exemption from the standard design certification ITAAC, as applicable. 
A request for an exemption must also be accompanied by a request for a 
license amendment under Sec.  52.98(f).
    (2) In the event that an activity is subject to an ITAAC not 
derived from a referenced standard design certification and the 
licensee has not demonstrated that the ITAAC has been met, the licensee 
may take corrective actions to successfully complete that ITAAC or 
request a license amendment under Sec.  52.98(f).
    (e) The NRC shall ensure that the prescribed inspections, tests, 
and analyses in the ITAAC are performed.
    (1) At appropriate intervals until the last date for submission of 
requests for hearing under Sec.  52.103(a), the NRC shall publish 
notices in the Federal Register of the NRC staff's determination of the 
successful completion of inspections, tests, and analyses.
    (2) The NRC shall make publicly available the licensee 
notifications under paragraph (c)(1), and, no later than the date of 
publication of the notice of intended operation required by Sec.  
52.103(a), make available all licensee notifications under paragraphs 
(c)(1) and (c)(2) of this section.


Sec.  52.103  Operation under a combined license.

    (a) The licensee shall notify the NRC of its scheduled date for 
initial loading of fuel no later than 270 days before the scheduled 
date and shall notify the NRC of updates to its schedule every 30 days 
thereafter. Not less than 180 days before the date scheduled for 
initial loading of fuel into a plant by a licensee that has been issued 
a combined license under this part, the Commission shall publish notice 
of intended operation in the Federal Register. The notice must provide 
that any person whose interest may be affected by operation of the 
plant may, within 60 days, request that the Commission hold a hearing 
on whether the facility as constructed complies, or on completion will 
comply, with the acceptance criteria in the combined license, except 
that a hearing shall not be granted for those ITAAC which the 
Commission found were met under Sec.  52.97(a)(2).
    (b) A request for hearing under paragraph (a) of this section must 
show, prima facie, that--
    (1) One or more of the acceptance criteria of the ITAAC in the 
combined license have not been, or will not be, met; and
    (2) The specific operational consequences of nonconformance that 
would be contrary to providing reasonable assurance of adequate 
protection of the public health and safety.
    (c) The Commission, acting as the presiding officer, shall 
determine whether to grant or deny the request for hearing in 
accordance with the applicable requirements of 10 CFR 2.309. If the 
Commission grants the request, the Commission, acting as the presiding 
officer, shall determine whether during a period of interim operation 
there will be reasonable assurance of adequate protection to the public 
health and safety. The Commission's determination must consider the 
petitioner's prima facie showing and any answers thereto. If the 
Commission determines there is such reasonable assurance, it shall 
allow operation during an interim period under the combined license.
    (d) The Commission, in its discretion, shall determine appropriate 
hearing procedures, whether informal or formal adjudicatory, for any 
hearing under paragraph (a) of this section, and shall state its 
reasons therefore.
    (e) The Commission shall, to the maximum possible extent, render a 
decision on issues raised by the hearing request within 180 days of the 
publication of the notice provided by paragraph (a) of this section or 
by the anticipated date for initial loading of fuel into the reactor, 
whichever is later.
    (f) A petition to modify the terms and conditions of the combined 
license will be processed as a request for action in accordance with 10 
CFR 2.206. The petitioner shall file the petition with the Secretary of 
the Commission. Before the licensed activity allegedly affected by the 
petition (fuel loading, low power testing, etc.) commences, the 
Commission shall determine whether any immediate action is required. If 
the petition is granted, then an appropriate order will be issued. Fuel 
loading and operation under the combined license will not be affected 
by the granting of the petition unless the order is made immediately 
effective.

[[Page 49537]]

    (g) The licensee shall not operate the facility until the 
Commission makes a finding that the acceptance criteria in the combined 
license are met, except for those acceptance criteria that the 
Commission found were met under Sec.  52.97(a)(2). If the combined 
license is for a modular design, each reactor module may require a 
separate finding as construction proceeds.
    (h) After the Commission has made the finding in paragraph (g) of 
this section, the ITAAC do not, by virtue of their inclusion in the 
combined license, constitute regulatory requirements either for 
licensees or for renewal of the license; except for the specific ITAAC 
for which the Commission has granted a hearing under paragraph (a) of 
this section, all ITAAC expire upon final Commission action in the 
proceeding. However, subsequent changes to the facility or procedures 
described in the final safety analysis report (as updated) must comply 
with the requirements in Sec. Sec.  52.98(e) or (f), as applicable.


Sec.  52.104  Duration of combined license.

    A combined license is issued for a specified period not to exceed 
40 years from the date on which the Commission makes a finding that 
acceptance criteria are met under Sec.  52.103(g) or allowing operation 
during an interim period under the combined license under Sec.  
52.103(c).


Sec.  52.105  Transfer of combined license.

    A combined license may be transferred in accordance with Sec.  
50.80 of this chapter.


Sec.  52.107  Application for renewal.

    The filing of an application for a renewed license must be in 
accordance with 10 CFR part 54.


Sec.  52.109  Continuation of combined license.

    Each combined license for a facility that has permanently ceased 
operations, continues in effect beyond the expiration date to authorize 
ownership and possession of the production or utilization facility, 
until the Commission notifies the licensee in writing that the license 
is terminated. During this period of continued effectiveness the 
licensee shall--
    (1) Take actions necessary to decommission and decontaminate the 
facility and continue to maintain the facility, including, where 
applicable, the storage, control and maintenance of the spent fuel, in 
a safe condition; and
    (2) Conduct activities in accordance with all other restrictions 
applicable to the facility in accordance with the NRC's regulations and 
the provisions of the combined license for the facility.


Sec.  52.110  Termination of license.

    (a)(1) When a licensee has determined to permanently cease 
operations the licensee shall, within 30 days, submit a written 
certification to the NRC, consistent with the requirements of Sec.  
52.3(b)(8);
    (2) Once fuel has been permanently removed from the reactor vessel, 
the licensee shall submit a written certification to the NRC that meets 
the requirements of Sec.  52.3(b)(9); and
    (3) For licensees whose licenses have been permanently modified to 
allow possession but not operation of the facility, before September 
27, 2007, the certification required in paragraph (a)(1) of this 
section shall be deemed to have been submitted.
    (b) Upon docketing of the certifications for permanent cessation of 
operations and permanent removal of fuel from the reactor vessel, or 
when a final legally effective order to permanently cease operations 
has come into effect, the 10 CFR part 52 license no longer authorizes 
operation of the reactor or emplacement or retention of fuel into the 
reactor vessel.
    (c) Decommissioning will be completed within 60 years of permanent 
cessation of operations. Completion of decommissioning beyond 60 years 
will be approved by the Commission only when necessary to protect 
public health and safety. Factors that will be considered by the 
Commission in evaluating an alternative that provides for completion of 
decommissioning beyond 60 years of permanent cessation of operations 
include unavailability of waste disposal capacity and other site-
specific factors affecting the licensee's capability to carry out 
decommissioning, including presence of other nuclear facilities at the 
site.
    (d)(1) Before or within 2 years following permanent cessation of 
operations, the licensee shall submit a post-shutdown decommissioning 
activities report (PSDAR) to the NRC, and a copy to the affected 
State(s). The report must include a description of the planned 
decommissioning activities along with a schedule for their 
accomplishment, an estimate of expected costs, and a discussion that 
provides the reasons for concluding that the environmental impacts 
associated with site-specific decommissioning activities will be 
bounded by appropriate previously issued environmental impact 
statements.
    (2) The NRC shall notice receipt of the PSDAR and make the PSDAR 
available for public comment. The NRC shall also schedule a public 
meeting in the vicinity of the licensee's facility upon receipt of the 
PSDAR. The NRC shall publish a document in the Federal Register and in 
a forum, such as local newspapers, that is readily accessible to 
individuals in the vicinity of the site, announcing the date, time and 
location of the meeting, along with a brief description of the purpose 
of the meeting.
    (e) Licensees shall not perform any major decommissioning 
activities, as defined in Sec.  50.2 of this chapter, until 90 days 
after the NRC has received the licensee's PSDAR submittal and until 
certifications of permanent cessation of operations and permanent 
removal of fuel from the reactor vessel, as required under Sec.  
52.110(a)(1), have been submitted.
    (f) Licensees shall not perform any decommissioning activities, as 
defined in Sec.  52.1, that--
    (1) Foreclose release of the site for possible unrestricted use;
    (2) Result in significant environmental impacts not previously 
reviewed; or
    (3) Result in there no longer being reasonable assurance that 
adequate funds will be available for decommissioning.
    (g) In taking actions permitted under Sec.  50.59 of this chapter 
following submittal of the PSDAR, the licensee shall notify the NRC in 
writing and send a copy to the affected State(s), before performing any 
decommissioning activity inconsistent with, or making any significant 
schedule change from, those actions and schedules described in the 
PSDAR, including changes that significantly increase the 
decommissioning cost.
    (h)(1) Decommissioning trust funds may be used by licensees if--
    (i) The withdrawals are for expenses for legitimate decommissioning 
activities consistent with the definition of decommissioning in Sec.  
52.1;
    (ii) The expenditure would not reduce the value of the 
decommissioning trust below an amount necessary to place and maintain 
the reactor in a safe storage condition if unforeseen conditions or 
expenses arise and;
    (iii) The withdrawals would not inhibit the ability of the licensee 
to complete funding of any shortfalls in the decommissioning trust 
needed to ensure the availability of funds to ultimately release the 
site and terminate the license.
    (2) Initially, 3 percent of the generic amount specified in Sec.  
50.75 of this chapter may be used for decommissioning planning. For 
licensees that have submitted the certifications required under Sec.  
52.110(a) and commencing 90 days after the NRC has received the PSDAR, 
an additional

[[Page 49538]]

20 percent may be used. A site-specific decommissioning cost estimate 
must be submitted to the NRC before the licensee may use any funding in 
excess of these amounts.
    (3) Within 2 years following permanent cessation of operations, if 
not already submitted, the licensee shall submit a site-specific 
decommissioning cost estimate.
    (4) For decommissioning activities that delay completion of 
decommissioning by including a period of storage or surveillance, the 
licensee shall provide a means of adjusting cost estimates and 
associated funding levels over the storage or surveillance period.
    (i) All power reactor licensees must submit an application for 
termination of license. The application for termination of license must 
be accompanied or preceded by a license termination plan to be 
submitted for NRC approval.
    (1) The license termination plan must be a supplement to the FSAR 
or equivalent and must be submitted at least 2 years before termination 
of the license date.
    (2) The license termination plan must include--
    (i) A site characterization;
    (ii) Identification of remaining dismantlement activities;
    (iii) Plans for site remediation;
    (iv) Detailed plans for the final radiation survey;
    (v) A description of the end use of the site, if restricted;
    (vi) An updated site-specific estimate of remaining decommissioning 
costs;
    (vii) A supplement to the environmental report, under Sec.  51.53 
of this chapter, describing any new information or significant 
environmental change associated with the licensee's proposed 
termination activities; and
    (viii) Identification of parts, if any, of the facility or site 
that were released for use before approval of the license termination 
plan.
    (3) The NRC shall notice receipt of the license termination plan 
and make the license termination plan available for public comment. The 
NRC shall also schedule a public meeting in the vicinity of the 
licensee's facility upon receipt of the license termination plan. The 
NRC shall publish a document in the Federal Register and in a forum, 
such as local newspapers, which is readily accessible to individuals in 
the vicinity of the site, announcing the date, time and location of the 
meeting, along with a brief description of the purpose of the meeting.
    (j) If the license termination plan demonstrates that the remainder 
of decommissioning activities will be performed in accordance with the 
regulations in this chapter, will not be inimical to the common defense 
and security or to the health and safety of the public, and will not 
have a significant effect on the quality of the environment and after 
notice to interested persons, the Commission shall approve the plan, by 
license amendment, subject to terms and conditions as it deems 
appropriate and necessary and authorize implementation of the license 
termination plan.
    (k) The Commission shall terminate the license if it determines 
that--
    (1) The remaining dismantlement has been performed in accordance 
with the approved license termination plan; and
    (2) The final radiation survey and associated documentation, 
including an assessment of dose contributions associated with parts 
released for use before approval of the license termination plan, 
demonstrate that the facility and site have met the criteria for 
decommissioning in subpart E to 10 CFR part 20.
    (l) For a facility that has permanently ceased operation before the 
expiration of its license, the collection period for any shortfall of 
funds will be determined, upon application by the licensee, on a case-
by-case basis taking into account the specific financial situation of 
each licensee.

Subpart D--Reserved

Subpart E--Standard Design Approvals


Sec.  52.131  Scope of subpart.

    This subpart sets out procedures for the filing, NRC staff review, 
and referral to the Advisory Committee on Reactor Safeguards of 
standard designs for a nuclear power reactor of the type described in 
Sec.  50.22 of this chapter or major portions thereof.


Sec.  52.133  Relationship to other subparts.

    (a) This subpart applies to a person that requests a standard 
design approval from the NRC staff separately from an application for a 
construction permit filed under 10 CFR part 50 or a combined license 
filed under subpart C of this part. An applicant for a construction 
permit or combined license may reference a standard design approval.
    (b) Subpart B of this part governs the certification by rulemaking 
of the design of a nuclear power plant. Subpart B may be used 
independently of the provisions in this subpart.
    (c) Subpart F of this part governs the issuance of licenses to 
manufacture nuclear power reactors to be installed and operated at 
sites not identified in the manufacturing license application. Subpart 
F of this part may be used independently of the provisions in this 
subpart.


Sec.  52.135  Filing of applications.

    (a) Any person may submit a proposed standard design for a nuclear 
power reactor of the type described in 10 CFR 50.22 to the NRC staff 
for its review. The submittal may consist of either the final design 
for the entire facility or the final design of major portions thereof.
    (b) The submittal for review of the proposed standard design must 
be made in the same manner and in the same number of copies as provided 
in 10 CFR 50.30 and 52.3 for license applications.
    (c) The fees associated with the filing and review of the 
application are set forth in 10 CFR part 170.


Sec.  52.136  Contents of applications; general information.

    The application must contain all of the information required by 10 
CFR 50.33(a) through (d) and (j).


Sec.  52.137  Contents of applications; technical information.

    If the applicant seeks review of a major portion of a standard 
design, the application need only contain the information required by 
this section to the extent the requirements are applicable to the major 
portion of the standard design for which NRC staff approval is sought.
    (a) The application must contain a final safety analysis report 
that describes the facility, presents the design bases and the limits 
on its operation, and presents a safety analysis of the structures, 
systems, and components and of the facility, or major portion thereof, 
and must include the following information:
    (1) The site parameters postulated for the design, and an analysis 
and evaluation of the design in terms of those site parameters;
    (2) A description and analysis of the SSCs of the facility, with 
emphasis upon performance requirements, the bases, with technical 
justification, upon which the requirements have been established, and 
the evaluations required to show that safety functions will be 
accomplished. It is expected that the standard plant will reflect 
through its design, construction, and operation an extremely low 
probability for accidents that could result in the release of 
significant quantities of radioactive fission products. The description 
shall be sufficient to permit understanding of the system designs and 
their

[[Page 49539]]

relationship to the safety evaluations. Items such as the reactor core, 
reactor coolant system, instrumentation and control systems, electrical 
systems, containment system, other engineered safety features, 
auxiliary and emergency systems, power conversion systems, radioactive 
waste handling systems, and fuel handling systems shall be discussed 
insofar as they are pertinent. The following power reactor design 
characteristics will be taken into consideration by the Commission:
    (i) Intended use of the reactor including the proposed maximum 
power level and the nature and inventory of contained radioactive 
materials;
    (ii) The extent to which generally accepted engineering standards 
are applied to the design of the reactor;
    (iii) The extent to which the reactor incorporates unique, unusual 
or enhanced safety features having a significant bearing on the 
probability or consequences of accidental release of radioactive 
materials; and
    (iv) The safety features that are to be engineered into the 
facility and those barriers that must be breached as a result of an 
accident before a release of radioactive material to the environment 
can occur. Special attention must be directed to plant design features 
intended to mitigate the radiological consequences of accidents. In 
performing this assessment, an applicant shall assume a fission product 
release \9\ from the core into the containment assuming that the 
facility is operated at the ultimate power level contemplated. The 
applicant shall perform an evaluation and analysis of the postulated 
fission product release, using the expected demonstrable containment 
leak rate and any fission product cleanup systems intended to mitigate 
the consequences of the accidents, together with applicable postulated 
site parameters, including site meteorology, to evaluate the offsite 
radiological consequences. The evaluation must determine that:
---------------------------------------------------------------------------

    \9\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. These accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (A) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 25 rem \10\ total effective dose equivalent (TEDE); and
---------------------------------------------------------------------------

    \10\ A whole body dose of 25 rem has been stated to correspond 
numerically to the once in a lifetime accidental or emergency dose 
for radiation workers which, according to NCRP recommendations at 
the time could be disregarded in the determination of their 
radiation exposure status (see NBS Handbook 69 dated June 5, 1959). 
However, its use is not intended to imply that this number 
constitutes an acceptable limit for an emergency dose to the public 
under accident conditions. Rather, this dose value has been set 
forth in this section as a reference value, which can be used in the 
evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accident.
---------------------------------------------------------------------------

    (B) An individual located at any point on the outer boundary of the 
low population zone, who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
TEDE;
    (3) The design of the facility including:
    (i) The principal design criteria for the facility. Appendix A to 
10 CFR part 50, general design criteria (GDC), establishes minimum 
requirements for the principal design criteria for water-cooled nuclear 
power plants similar in design and location to plants for which 
construction permits have previously been issued by the Commission and 
provides guidance to applicants in establishing principal design 
criteria for other types of nuclear power units;
    (ii) The design bases and the relation of the design bases to the 
principal design criteria; and
    (iii) Information relative to materials of construction, general 
arrangement, and approximate dimensions, sufficient to provide 
reasonable assurance that the design will conform to the design bases 
with adequate margin for safety;
    (4) An analysis and evaluation of the design and performance of SSC 
with the objective of assessing the risk to public health and safety 
resulting from operation of the facility and including determination of 
the margins of safety during normal operations and transient conditions 
anticipated during the life of the facility, and the adequacy of SSCs 
provided for the prevention of accidents and the mitigation of the 
consequences of accidents. Analysis and evaluation of ECCS cooling 
performance and the need for high-point vents following postulated 
loss-of-coolant accidents shall be performed in accordance with the 
requirements of 10 CFR 50.46 and 50.46a;
    (5) The kinds and quantities of radioactive materials expected to 
be produced in the operation and the means for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in part 20 of this chapter;
    (6) The information required by Sec.  20.1406 of this chapter;
    (7) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (8) The information necessary to demonstrate compliance with any 
technically relevant portions of the Three Mile Island requirements set 
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), 
and (f)(3)(v) of 10 CFR 50.34(f);
    (9) For applications for light-water-cooled nuclear power plants, 
an evaluation of the standard plant design against the Standard Review 
Plan (SRP) revision in effect 6 months before the docket date of the 
application. The evaluation required by this section shall include an 
identification and description of all differences in design features, 
analytical techniques, and procedural measures proposed for the design 
and those corresponding features, techniques, and measures given in the 
SRP acceptance criteria. Where a difference exists, the evaluation 
shall discuss how the proposed alternative provides an acceptable 
method of complying with the Commission's regulations, or portions 
thereof, that underlie the corresponding SRP acceptance criteria. The 
SRP is not a substitute for the regulations, and compliance is not a 
requirement;
    (10) The information with respect to the design of equipment to 
maintain control over radioactive materials in gaseous and liquid 
effluents produced during normal reactor operations described in 10 CFR 
50.34a(e);
    (11) The information pertaining to design features that affect 
plans for coping with emergencies in the operation of the reactor 
facility or a major portion thereof;
    (12) An analysis and description of the equipment and systems for 
combustible gas control as required by Sec.  50.44 of this chapter;
    (13) The list of electric equipment important to safety that is 
required by 10 CFR 50.49(d);
    (14) A description of protection provided against pressurized 
thermal shock events, including projected values of the reference 
temperature for reactor vessel beltline materials as defined in 10 CFR 
50.60 and 50.61;
    (15) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram (ATWS) events in Sec.  50.62;

[[Page 49540]]

    (16) The coping analysis, and any design features necessary to 
address station blackout, as described in Sec.  50.63 of this chapter;
    (17) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68(b)(2)-(b)(4);
    (18) A description and analysis of the fire protection design 
features for the standard plant necessary to comply with part 50, 
appendix A, GDC 3, and Sec.  50.48 of this chapter;
    (19) A description of the quality assurance program applied to the 
design of the SSCs of the facility. Appendix B to 10 CFR part 50, 
``Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants,'' sets forth the requirements for quality 
assurance programs for nuclear power plants. The description of the 
quality assurance program for a nuclear power plant shall include a 
discussion of how the applicable requirements of appendix B to 10 CFR 
part 50 were satisfied;
    (20) The information necessary to demonstrate that the standard 
plant complies with the earthquake engineering criteria in 10 CFR part 
50, appendix S;
    (21) Proposed technical resolutions of those Unresolved Safety 
Issues and medium- and high-priority generic safety issues which are 
identified in the version of NUREG-0933 current on the date up to 6 
months before the docket date of the application and which are 
technically relevant to the design;
    (22) The information necessary to demonstrate how operating 
experience insights have been incorporated into the plant design;
    (23) For light-water reactor designs, a description and analysis of 
design features for the prevention and mitigation of severe accidents, 
e.g., challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen combustion, and containment bypass;
    (24) A description, analysis, and evaluation of the interfaces 
between the standard design and the balance of the nuclear power plant; 
and
    (25) A description of the design-specific probabilistic risk 
assessment and its results.
    (b) An application for approval of a standard design, which differs 
significantly from the light-water reactor designs of plants that have 
been licensed and in commercial operation before April 18, 1989, or 
uses simplified, inherent, passive, or other innovative means to 
accomplish its safety functions, must meet the requirements of 10 CFR 
50.43(e).


Sec.  52.139  Standards for review of applications.

    Applications filed under this subpart will be reviewed for 
compliance with the standards set out in 10 CFR parts 20, 50 and its 
appendices, and 10 CFR parts 73 and 100.


Sec.  52.141  Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).

    The Commission shall refer a copy of the application to the ACRS. 
The ACRS shall report on those portions of the application which 
concern safety.


Sec.  52.143  Staff approval of design.

    Upon completion of its review of a submittal under this subpart and 
receipt of a report by the Advisory Committee on Reactor Safeguards 
under Sec.  52.141 of this subpart, the NRC staff shall publish a 
determination in the Federal Register as to whether or not the design 
is acceptable, subject to appropriate terms and conditions, and make an 
analysis of the design in the form of a report available at the NRC Web 
site, http://www.nrc.gov.


Sec.  52.145  Finality of standard design approvals; information 
requests.

    (a) An approved design must be used by and relied upon by the NRC 
staff and the ACRS in their review of any individual facility license 
application that incorporates by reference a standard design approved 
in accordance with this paragraph unless there exists significant new 
information that substantially affects the earlier determination or 
other good cause.
    (b) The determination and report by the NRC staff do not constitute 
a commitment to issue a permit or license, or in any way affect the 
authority of the Commission, Atomic Safety and Licensing Board Panel, 
or presiding officers in any proceeding under part 2 of this chapter.
    (c) Except for information requests seeking to verify compliance 
with the current licensing basis of the standard design approval, 
information requests to the holder of a standard design approval must 
be evaluated before issuance to ensure that the burden to be imposed on 
respondents is justified in view of the potential safety significance 
of the issue to be addressed in the requested information. Each 
evaluation performed by the NRC staff must be in accordance with 10 CFR 
50.54(f) and must be approved by the Executive Director for Operations 
or his or her designee before issuance of the request.


Sec.  52.147  Duration of design approval.

    A standard design approval issued under this subpart is valid for 
15 years from the date of issuance and may not be renewed. A design 
approval continues to be valid beyond the date of expiration in any 
proceeding on an application for a construction permit or an operating 
license under part 50 or a combined license or manufacturing license 
under part 52 that references the final design approval and is docketed 
before the date of expiration of the design approval.

Subpart F--Manufacturing Licenses


Sec.  52.151  Scope of subpart.

    This subpart sets out the requirements and procedures applicable to 
Commission issuance of a license authorizing manufacture of nuclear 
power reactors to be installed at sites not identified in the 
manufacturing license application.


Sec.  52.153  Relationship to other subparts.

    (a) A nuclear power reactor manufactured under a manufacturing 
license issued under this subpart may only be transported to and 
installed at a site for which either a construction permit under part 
50 of this chapter or a combined license under subpart C of this part 
has been issued.
    (b) Subpart B of this part governs the certification by rulemaking 
of the design of standard nuclear power facilities. Subpart E of this 
part governs the NRC staff review and approval of standard designs for 
a nuclear power facility. A manufacturing license applicant may 
reference a standard design certification or a standard design approval 
in its application. These subparts may also be used independently of 
the provisions in this subpart.


Sec.  52.155  Filing of applications.

    (a) Any person, except one excluded by 10 CFR 50.38, may file an 
application for a manufacturing license under this subpart with the 
Director of New Reactors or the Director of Nuclear Reactor Regulation, 
as appropriate.
    (b) The application must comply with the applicable filing 
requirements of Sec. Sec.  52.3 and 50.30 of this chapter.
    (c) The fees associated with the filing and review of the 
application are set forth in 10 CFR part 170.


Sec.  52.156  Contents of applications; general information.

    The application must contain all of the information required by 10 
CFR 50.33(a) through (d), and (j).


Sec.  52.157  Contents of applications; technical information in final 
safety analysis report.

    The application must contain a final safety analysis report 
containing the information set forth below, with a level

[[Page 49541]]

of design information sufficient to enable the Commission to judge the 
applicant's proposed means of assuring that the manufacturing conforms 
to the design and to reach a final conclusion on all safety questions 
associated with the design, permit the preparation of construction and 
installation specifications by an applicant who seeks to use the 
manufactured reactor, and permit the preparation of acceptance and 
inspection requirements by the NRC:
    (a) The principal design criteria for the reactor to be 
manufactured. Appendix A of 10 CFR part 50, ``General Design Criteria 
for Nuclear Power Plants,'' establishes minimum requirements for the 
principal design criteria for water-cooled nuclear power plants similar 
in design and location to plants for which construction permits have 
previously been issued by the Commission and provides guidance to 
applicants in establishing principal design criteria for other types of 
nuclear power units;
    (b) The design bases and the relation of the design bases to the 
principal design criteria;
    (c) A description and analysis of the structures, systems, and 
components of the reactor to be manufactured, with emphasis upon the 
materials of manufacture, performance requirements, the bases, with 
technical justification therefor, upon which the performance 
requirements have been established, and the evaluations required to 
show that safety functions will be accomplished. The description shall 
be sufficient to permit understanding of the system designs and their 
relationship to safety evaluations. Items such as the reactor core, 
reactor coolant system, instrumentation and control systems, electrical 
systems, containment system, other engineered safety features, 
auxiliary and emergency systems, power conversion systems, radioactive 
waste handling systems, and fuel handling systems shall be discussed 
insofar as they are pertinent. The following power reactor design 
characteristics will be taken into consideration by the Commission:
    (1) Intended use of the manufactured reactor including the proposed 
maximum power level and the nature and inventory of contained 
radioactive materials;
    (2) The extent to which generally accepted engineering standards 
are applied to the design of the reactor; and
    (3) The extent to which the reactor incorporates unique, unusual or 
enhanced safety features having a significant bearing on the 
probability or consequences of accidental release of radioactive 
materials;
    (d) The safety features that are engineered into the reactor and 
those barriers that must be breached as a result of an accident before 
a release of radioactive material to the environment can occur. Special 
attention must be directed to reactor design features intended to 
mitigate the radiological consequences of accidents. In performing this 
assessment, an applicant shall assume a fission product release \11\ 
from the core into the containment assuming that the facility is 
operated at the ultimate power level contemplated. The applicant shall 
perform an evaluation and analysis of the postulated fission product 
release, using the expected demonstrable containment leak rate and any 
fission product cleanup systems intended to mitigate the consequences 
of the accidents, together with applicable postulated site parameters, 
including site meteorology, to evaluate the offsite radiological 
consequences. The evaluation must determine that:
---------------------------------------------------------------------------

    \11\ The fission product release assumed for this evaluation 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events. These accidents have generally been assumed to 
result in substantial meltdown of the core with subsequent release 
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------

    (1) An individual located at any point on the boundary of the 
exclusion area for any 2 hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 25 rem \12\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------

    \12\ A whole body dose of 25 rem has been stated to correspond 
numerically to the once in a lifetime accidental or emergency dose 
for radiation workers which, according to NCRP recommendations at 
the time could be disregarded in the determination of their 
radiation exposure status (see NBS Handbook 69 dated June 5, 1959). 
However, its use is not intended to imply that this number 
constitutes an acceptable limit for an emergency dose to the public 
under accident conditions. Rather, this dose value has been set 
forth in this section as a reference value, which can be used in the 
evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accident.
---------------------------------------------------------------------------

    (2) An individual located at any point on the outer boundary of the 
low population zone, who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
TEDE; and
    (e) The kinds and quantities of radioactive materials expected to 
be produced in the operation and the means for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in part 20 of this chapter.
    (f) Information necessary to establish that the design of the 
reactor to be manufactured complies with the technical requirements in 
10 CFR Chapter I, including:
    (1) An analysis and evaluation of the design and performance of 
structures, systems, and components with the objective of assessing the 
risk to public health and safety resulting from operation of the 
facility and including determination of the margins of safety during 
normal operations and transient conditions anticipated during the life 
of the facility, and the adequacy of structures, systems, and 
components provided for the prevention of accidents and the mitigation 
of the consequences of accidents. Analysis and evaluation of ECCS 
cooling performance and the need for high-point vents following 
postulated loss-of-coolant accidents shall be performed in accordance 
with the requirements of Sec. Sec.  50.46 and 50.46a of this chapter;
    (2) A description and analysis of the fire protection design 
features for the reactor necessary to comply with 10 CFR part 50, 
appendix A, GDC 3 and Sec.  50.48 of this chapter;
    (3) A description of protection provided against pressurized 
thermal shock events, including projected values of the reference 
temperature for reactor vessel beltline materials as defined in 
Sec. Sec.  50.60 and 50.61 of this chapter;
    (4) An analysis and description of the equipment and systems for 
combustible gas control as required by Sec.  50.44 of this chapter;
    (5) The coping analysis, and any design features necessary to 
address station blackout, as described in Sec.  50.63 of this chapter;
    (6) The list of electric equipment important to safety that is 
required by 10 CFR 50.49(d);
    (7) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram (ATWS) events in Sec.  50.62;
    (8) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68(b)(2)-(b)(4);
    (9) The information required by Sec.  20.1406 of this chapter;
    (10) [Reserved];
    (11) The information with respect to the design of equipment to 
maintain control over radioactive materials in gaseous and liquid 
effluents produced during normal reactor operations, as described in 
Sec.  50.34a(e) of this chapter;

[[Page 49542]]

    (12) The information necessary to demonstrate compliance with any 
technically relevant portions of the Three Mile Island requirements set 
forth in Sec.  50.34(f) of this chapter, except paragraphs (f)(1)(xii), 
(f)(2)(ix), and (f)(3)(v);
    (13) If the applicant seeks to use risk-informed treatment of SSCs 
in accordance with Sec.  50.69 of this chapter, the information 
required by Sec.  50.69(b)(2) of this chapter;
    (14) The information necessary to demonstrate that the manufactured 
reactor complies with the earthquake engineering criteria in appendix S 
to 10 CFR part 50;
    (15) Information sufficient to demonstrate compliance with the 
applicable requirements regarding testing, analysis, and prototypes as 
set forth in Sec.  50.43(e) of this chapter;
    (16) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (17) A description of the quality assurance program applied to the 
design, and to be applied to the manufacture of, the structures, 
systems, and components of the reactor. Appendix B to 10 CFR part 50, 
``Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants,'' sets forth the requirements for quality 
assurance programs for nuclear power plants. The description of the 
quality assurance program must include a discussion of how the 
applicable requirements of appendix B to 10 CFR part 50 have been and 
will be satisfied; and
    (18) Proposed technical specifications applicable to the reactor 
being manufactured, prepared in accordance with the requirements of 
Sec. Sec.  50.36 and 50.36a of this chapter;
    (19) The site parameters postulated for the design, and an analysis 
and evaluation of the reactor design in terms of those site parameters;
    (20) The interface requirements between the manufactured reactor 
and the remaining portions of the nuclear power plant. These 
requirements must be sufficiently detailed to allow for completion of 
the final safety analysis;
    (21) Justification that compliance with the interface requirements 
of paragraph (f)(20) of this section is verifiable through inspections, 
testing, or analysis. The method to be used for verification of 
interface requirements must be included as part of the proposed ITAAC 
required by Sec.  52.158(a);
    (22) A representative conceptual design for a nuclear power 
facility using the manufactured reactor, to aid the NRC in its review 
of the final safety analysis required by this section and to permit 
assessment of the adequacy of the interface requirements in paragraph 
(f)(20) of this section;
    (23) For light-water reactor designs, a description and analysis of 
design features for the prevention and mitigation of severe accidents, 
e.g., challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen combustion, and containment bypass;
    (24) [Reserved];
    (25) If the reactor is to be used in modular plant design, a 
description of the possible operating configurations of the reactor 
modules with common systems, interface requirements, and system 
interactions. The final safety analysis must also account for 
differences among the configurations, including any restrictions that 
will be necessary during the construction and startup of a given module 
to ensure the safe operation of any module already operating;
    (26) A description of the management plan for design and 
manufacturing activities, including:
    (i) The organizational and management structure singularly 
responsible for direction of design and manufacture of the reactor;
    (ii) Technical resources directed by the applicant, and the 
qualifications requirements;
    (iii) Details of the interaction of design and manufacture within 
the applicant's organization and the manner by which the applicant will 
ensure close integration of the architect engineer and the nuclear 
steam supply vendor, as applicable;
    (iv) Proposed procedures governing the preparation of the 
manufactured reactor for shipping to the site where it is to be 
operated, the conduct of shipping, and verifying the condition of the 
manufactured reactor upon receipt at the site; and
    (v) The degree of top level management oversight and technical 
control to be exercised by the applicant during design and manufacture, 
including the preparation and implementation of procedures necessary to 
guide the effort;
    (27) Necessary parameters to be used in developing plans for 
preoperational testing and initial operation;
    (28) Proposed technical resolutions of those Unresolved Safety 
Issues and medium- and high-priority generic safety issues which are 
identified in the version of NUREG-0933 current on the date up to 6 
months before the docket date of the application and which are 
technically relevant to the design;
    (29) The information necessary to demonstrate how operating 
experience insights have been incorporated into the manufactured 
reactor design;
    (30) For applications for light-water-cooled nuclear power plants, 
an evaluation of the design to be manufactured against the Standard 
Review Plan (SRP) revision in effect 6 months before the docket date of 
the application. The evaluation required by this section shall include 
an identification and description of all differences in design 
features, analytical techniques, and procedural measures proposed for 
the design and those corresponding features, techniques, and measures 
given in the SRP acceptance criteria. Where a difference exists, the 
evaluation shall discuss how the proposed alternative provides an 
acceptable method of complying with the Commission's regulations, or 
portions thereof, that underlie the corresponding SRP acceptance 
criteria. The SRP is not a substitute for the regulations, and 
compliance is not a requirement; and
    (31) A description of the design-specific probabilistic risk 
assessment and its results.


Sec.  52.158  Contents of application; additional technical 
information.

    The application must contain:
    (a)(1) Inspections, tests, analyses, and acceptance criteria 
(ITAAC). The proposed inspections, tests, and analyses that the 
licensee who will be operating the reactor shall perform, and the 
acceptance criteria that are necessary and sufficient to provide 
reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met:
    (i) The reactor has been manufactured in conformity with the 
manufacturing license; the provisions of the Act, and the Commission's 
rules and regulations; and
    (ii) The manufactured reactor will be operated in conformity with 
the approved design and any license authorizing operation of the 
manufactured reactor.
    (2) If the application references a standard design certification, 
the ITAAC contained in the certified design must apply to those 
portions of the facility design which are covered by the design 
certification.
    (3) If the application references a standard design certification, 
the application may include a notification that a required inspection, 
test, or analysis in the design certification ITAAC has been 
successfully completed and that the corresponding acceptance

[[Page 49543]]

criterion has been met. The Federal Register notification required by 
Sec.  52.163 must indicate that the application includes this 
notification.
    (b)(1) An environmental report as required by 10 CFR 51.54.
    (2) If the manufacturing license application references a standard 
design certification, the environmental report need not contain a 
discussion of severe accident mitigation design alternatives for the 
reactor.


Sec.  52.159  Standards for review of application.

    Applications filed under this subpart will be reviewed according to 
the applicable standards set out in 10 CFR parts 20, 50 and its 
appendices, 51, 73, and 100 and its appendices.


Sec.  52.161  Reserved.


Sec.  52.163  Administrative review of applications; hearings.

    A proceeding on a manufacturing license is subject to all 
applicable procedural requirements contained in 10 CFR part 2, 
including the requirements for docketing in Sec.  2.101(a)(1) through 
(4) of this chapter, and the requirements for issuance of a notice of 
proposed action in Sec.  2.105 of this chapter, provided, however, that 
the designated sections may not be construed to require that the 
environmental report or draft or final environmental impact statement 
include an assessment of the benefits of constructing and/or operating 
the manufactured reactor or an evaluation of alternative energy 
sources. All hearings on manufacturing licenses are governed by the 
hearing procedures contained in 10 CFR part 2, subparts C, G, L, and N.


Sec.  52.165  Referral to the Advisory Committee on Reactor Safeguards 
(ACRS).

    The Commission shall refer a copy of the application to the ACRS. 
The ACRS shall report on those portions of the application which 
concern safety.


Sec.  52.167  Issuance of manufacturing license.

    (a) After completing any hearing under Sec.  52.163, and receiving 
the report submitted by the ACRS, the Commission may issue a 
manufacturing license if the Commission finds that:
    (1) Applicable standards and requirements of the Act and the 
Commission's regulations have been met;
    (2) There is reasonable assurance that the reactor(s) will be 
manufactured, and can be transported, incorporated into a nuclear power 
plant, and operated in conformity with the manufacturing license, the 
provision of the Act, and the Commission's regulations;
    (3) The proposed reactor(s) can be incorporated into a nuclear 
power plant and operated at sites having characteristics that fall 
within the site parameters postulated for the design of the 
manufactured reactor(s) without undue risk to the health and safety of 
the public;
    (4) The applicant is technically qualified to design and 
manufacture the proposed nuclear power reactor(s);
    (5) The proposed inspections, tests, analyses and acceptance 
criteria are necessary and sufficient, within the scope of the 
manufacturing license, to provide reasonable assurance that the 
manufactured reactor has been manufactured and will be operated in 
conformity with the license, the provisions of the Act, and the 
Commission's regulations;
    (6) The issuance of a license to the applicant will not be inimical 
to the common defense and security or to the health and safety of the 
public; and
    (7) The findings required by subpart A of part 51 of this chapter 
have been made.
    (b) Each manufacturing license issued under this subpart shall 
specify:
    (1) Terms and conditions as the Commission deems necessary and 
appropriate;
    (2) Technical specifications for operation of the manufactured 
reactor, as the Commission deems necessary and appropriate;
    (3) Site parameters and design characteristics for the manufactured 
reactor; and
    (4) The interface requirements to be met by the site-specific 
elements of the facility, such as the service water intake structure 
and the ultimate heat sink, not within the scope of the manufactured 
reactor.
    (c)(1) A holder of a manufacturing license may not transport or 
allow to be removed from the place of manufacture the manufactured 
reactor except to the site of a licensee with either a construction 
permit under part 50 of this chapter or a combined license under 
subpart C of this part. The construction permit or combined license 
must authorize the construction of a nuclear power facility using the 
manufactured reactor(s).
    (2) A holder of a manufacturing license shall include, in any 
contract governing the transport of a manufactured reactor from the 
place of manufacture to any other location, a provision requiring that 
the person or entity transporting the manufactured reactor to comply 
with all NRC-approved shipping requirements in the manufacturing 
license.


Sec.  52.169  [Reserved].


Sec.  52.171  Finality of manufacturing licenses; information requests.

    (a)(1) Notwithstanding any provision in 10 CFR 50.109, during the 
term of a manufacturing license the Commission may not modify, rescind, 
or impose new requirements on the design of the nuclear power reactor 
being manufactured, or the requirements for the manufacture of the 
nuclear power reactor, unless the Commission determines that a 
modification is necessary to bring the design of the reactor or its 
manufacture into compliance with the Commission's requirements 
applicable and in effect at the time the manufacturing license was 
issued, or to provide reasonable assurance of adequate protection to 
public health and safety or common defense and security.
    (2) Any modification to the design of a manufactured nuclear power 
reactor which is imposed by the Commission under paragraph (a)(1) of 
this section will be applied to all reactors manufactured under the 
license, including those that have already been transported and sited, 
except those reactors to which the modification has been rendered 
technically irrelevant by action taken under paragraph (b) of this 
section.
    (3) In making the findings required for issuance of a construction 
permit, operating license, combined license, in any hearing under Sec.  
52.103, or in any enforcement hearing other than one initiated by the 
Commission under paragraph (a)(1) of this section, for which a nuclear 
power reactor manufactured under this subpart is referenced or used, 
the Commission shall treat as resolved those matters resolved in the 
proceeding on the application for issuance or renewal of the 
manufacturing license, including the adequacy of design of the 
manufactured reactor, the costs and benefits of severe accident 
mitigation design alternatives, and the bases for not incorporating 
severe accident mitigation design alternatives into the design of the 
reactor to be manufactured.
    (b)(1) The holder of a manufacturing license may not make changes 
to the design of the nuclear power reactor authorized to be 
manufactured without prior Commission approval. The request for a 
change to the design must be in the form of an application for a 
license amendment, and must meet the requirements of 10 CFR 50.90 and 
50.92.

[[Page 49544]]

    (2) An applicant or licensee who references or uses a nuclear power 
reactor manufactured under a manufacturing license under this subpart 
may request a departure from the design characteristics, site 
parameters, terms and conditions, or approved design of the 
manufactured reactor. The Commission may grant a request only if it 
determines that the departure will comply with the requirements of 10 
CFR 52.7, and that the special circumstances outweigh any decrease in 
safety that may result from the reduction in standardization caused by 
the departure. The granting of a departure on request of an applicant 
is subject to litigation in the same manner as other issues in the 
construction permit or combined license hearing.
    (c) Except for information requests seeking to verify compliance 
with the current licensing basis of either the manufacturing license or 
the manufactured reactor, information requests to the holder of a 
manufacturing license or an applicant or licensee using a manufactured 
reactor must be evaluated before issuance to ensure that the burden to 
be imposed on respondents is justified in view of the potential safety 
significance of the issue to be addressed in the requested information. 
Each evaluation performed by the NRC staff must be in accordance with 
10 CFR 50.54(f) and must be approved by the Executive Director for 
Operations or his or her designee before issuance of the request.


Sec.  52.173  Duration of manufacturing license.

    A manufacturing license issued under this subpart may be valid for 
not less than 5, nor more than 15 years from the date of issuance. A 
holder of a manufacturing license may not initiate the manufacture of a 
reactor less than 3 years before the expiration of the license even 
though a timely application for renewal has been docketed with the NRC. 
Upon expiration of the manufacturing license, the manufacture of any 
uncompleted reactors must cease unless a timely application for renewal 
has been docketed with the NRC.


Sec.  52.175  Transfer of manufacturing license.

    A manufacturing license may be transferred in accordance with Sec.  
50.80 of this chapter.


Sec.  52.177  Application for renewal.

    (a) Not less than 12 months, nor more than 5 years before the 
expiration of the manufacturing license, or any later renewal period, 
the holder of the manufacturing license may apply for a renewal of the 
license. An application for renewal must contain all information 
necessary to bring up to date the information and data contained in the 
previous application.
    (b) The filing of an application for a renewed license must be in 
accordance with subpart A of 10 CFR part 2 and 10 CFR 52.3 and 50.30.
    (c) A manufacturing license, either original or renewed, for which 
a timely application for renewal has been filed, remains in effect 
until the Commission has made a final determination on the renewal 
application, provided, however, that in accordance with Sec.  52.173, 
the holder of a manufacturing license may not begin manufacture of a 
reactor less than 3 years before the expiration of the license.
    (d) Any person whose interest may be affected by renewal of the 
permit may request a hearing on the application for renewal. The 
request for a hearing must comply with 10 CFR 2.309. If a hearing is 
granted, notice of the hearing will be published in accordance with 10 
CFR 2.104.
    (e) The Commission shall refer a copy of the application for 
renewal to the Advisory Committee on Reactor Safeguards (ACRS). The 
ACRS shall report on those portions of the application which concern 
safety and shall apply the criteria set forth in Sec.  52.159.


Sec.  52.179  Criteria for renewal.

    The Commission may grant the renewal if the Commission determines:
    (a) The manufacturing license complies with the Atomic Energy Act 
and the Commission's regulations and orders applicable and in effect at 
the time the manufacturing license was originally issued; and
    (b) Any new requirements the Commission may wish to impose are:
    (1) Necessary for adequate protection to public health and safety 
or common defense and security;
    (2) Necessary for compliance with the Commission's regulations and 
orders applicable and in effect at the time the manufacturing license 
was originally issued; or
    (3) A substantial increase in overall protection of the public 
health and safety or the common defense and security to be derived from 
the new requirements, and the direct and indirect costs of 
implementation of those requirements are justified in view of this 
increased protection.


Sec.  52.181  Duration of renewal.

    A renewed manufacturing license may be issued for a term of not 
less than 5, nor more than 15 years, plus any remaining years on the 
manufacturing license then in effect before renewal. The renewed 
license shall be subject to the requirements of Sec. Sec.  52.171 and 
52.175.

Subpart G--Reserved

Subpart H--Enforcement


Sec.  52.301  Violations.

    (a) The Commission may obtain an injunction or other court order to 
prevent a violation of the provisions of--
    (1) The Atomic Energy Act of 1954, as amended;
    (2) Title II of the Energy Reorganization Act of 1974, as amended; 
or
    (3) A regulation or order issued under those Acts.
    (b) The Commission may obtain a court order for the payment of a 
civil penalty imposed under Section 234 of the Atomic Energy Act:
    (1) For violations of--
    (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of 
the Atomic Energy Act of 1954, as amended;
    (ii) Section 206 of the Energy Reorganization Act;
    (iii) Any regulation, or order issued under the sections specified 
in paragraph (b)(1)(i) of this section;
    (iv) Any term, condition, or limitation of any license issued under 
the sections specified in paragraph (b)(1)(i) of this section.
    (2) For any violation for which a license may be revoked under 
Section 186 of the Atomic Energy Act of 1954, as amended.


Sec.  52.303  Criminal penalties.

    (a) Section 223 of the Atomic Energy Act of 1954, as amended, 
provides for criminal sanctions for willful violation of, attempted 
violation of, or conspiracy to violate, any regulation issued under 
Sections 161b, 161i, or 161o of the Act. For purposes of Section 223, 
all the regulations in part 52 are issued under one or more of Sections 
161b, 161i, or 160o, except for the sections listed in paragraph (b) of 
this section.
    (b) The regulations in part 52 that are not issued under Sections 
161b, 161i, or 161o for the purposes of Section 223 are as follows: 
Sec. Sec.  52.0, 52.1, 52.2, 52.3, 52.7, 52.8, 52.9, 52.10, 52.11, 
52.12, 52.13, 52.15, 52.16, 52.17, 52.18, 52.21, 52.23, 52.24, 52.27, 
52.28, 52.29, 52.31, 52.33, 52.39, 52.41, 52.43, 52.45, 52.46, 52.47, 
52.48, 52.51, 52.53, 52.54, 52.55, 52.57, 52.59, 52.61, 52.63, 52.71, 
52.73, 52.75, 52.77, 52.79, 52.80, 52.81, 52.83, 52.85, 52.87, 52.93, 
52.97, 52.98, 52.103, 52.104, 52.105, 52.107, 52.109, 52.131,

[[Page 49545]]

52.133, 52.135, 52.136, 52.137, 52.139, 52.141, 52.143, 52.145, 52.147, 
52.151, 52.153, 52.155, 52.156, 52.157, 52.158, 52.159, 52.161, 52.163, 
52.165, 52.167, 52.171, 52.173, 52.175, 52.177, 52.179, 52.181, 52.301, 
and 52.303.

Appendix A to Part 52--Design Certification Rule for the U.S. Advanced 
Boiling Water Reactor

I. Introduction

    Appendix A constitutes the standard design certification for the 
U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance 
with 10 CFR part 52, subpart B. The applicant for certification of 
the U.S. ABWR design was GE Nuclear Energy.

II. Definitions

    A. Generic design control document (generic DCD) means the 
document containing the Tier 1 and Tier 2 information and generic 
technical specifications that is incorporated by reference into this 
appendix.
    B. Generic technical specifications means the information, 
required by 10 CFR 50.36 and 50.36a, for the portion of the plant 
that is within the scope of this appendix.
    C. Plant-specific DCD means the document, maintained by an 
applicant or licensee who references this appendix, consisting of 
the information in the generic DCD, as modified and supplemented by 
the plant-specific departures and exemptions made under Section VIII 
of this appendix.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (hereinafter Tier 1 information). The design descriptions, 
interface requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;
    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAAC);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must meet the requirement in Section III.B of this appendix to 
reference Tier 2 when referencing Tier 1. Tier 2 information 
includes:
    1. Information required by Sec. Sec.  52.47(a) and 52.47(c), 
with the exception of generic technical specifications and 
conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAAC have been met; and
    3. Combined license (COL) action items (COL license 
information), which identify certain matters that must be addressed 
in the site-specific portion of the final safety analysis report 
(FSAR) by an applicant who references this appendix. These items 
constitute information requirements but are not the only acceptable 
set of information in the FSAR. An applicant may depart from or omit 
these items, provided that the departure or omission is identified 
and justified in the FSAR. After issuance of a construction permit 
or COL, these items are not requirements for the licensee unless 
such items are restated in the FSAR.
    F. Tier 2* means the portion of the Tier 2 information, 
designated as such in the generic DCD, which is subject to the 
change process in Section VIII.B.6 of this appendix. This 
designation expires for some Tier 2* information under Section 
VIII.B.6.
    G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety 
analyses means:
    (1) Changing any of the elements of the method described in the 
plant-specific DCD unless the results of the analysis are 
conservative or essentially the same; or
    (2) Changing from a method described in the plant-specific DCD 
to another method unless that method has been approved by NRC for 
the intended application.
    H. All other terms in this appendix have the meaning set out in 
10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, 
as amended, as applicable.

III. Scope and Contents

    A. Tier 1, Tier 2, and the generic technical specifications in 
the U.S. ABWR Design Control Document, GE Nuclear Energy, Revision 4 
dated March 1997, are approved for incorporation by reference by the 
Director of the Office of the Federal Register in accordance with 5 
U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be 
obtained from the National Technical Information Service, 5285 Port 
Royal Road, Springfield, Virginia 22161. A copy is available for 
examination and copying at the NRC Public Document Room located at 
One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852. Copies are also available for examination 
at the NRC Library located at Two White Flint North, 11545 Rockville 
Pike, Rockville, Maryland 20582 and the Office of the Federal 
Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.
    B. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix, 
including Tier 1, Tier 2, and the generic technical specifications 
except as otherwise provided in this appendix. Conceptual design 
information, as set forth in the generic DCD, and the ``Technical 
Support Document for the ABWR'' are not part of this appendix. Tier 
2 references to the probabilistic risk assessment (PRA) in the ABWR 
standard safety analysis report do not incorporate the PRA into Tier 
2.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for design certification of the U.S. ABWR design or 
NUREG-1503, ``Final Safety Evaluation Report related to the 
Certification of the Advanced Boiling Water Reactor Design'' (FSER), 
and Supplement No. 1, then the generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a combined license that wishes to reference 
this appendix shall, in addition to complying with the requirements 
of 10 CFR 52.77, 52.79, and 52.80, comply with the following 
requirements:
    1. Incorporate by reference, as part of its application, this 
appendix;
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and using the same organization and numbering as the generic DCD for 
the U.S. ABWR design, as modified and supplemented by the 
applicant's exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific technical specifications, consisting of the 
generic and site-specific technical specifications, that are 
required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating compliance with the site parameters 
and interface requirements;
    e. Information that addresses the COL action items; and
    f. Information required by 10 CFR 52.47 that is not within the 
scope of this appendix.
    3. Include, in the plant-specific DCD, the proprietary 
information and safeguards information referenced in the U.S. ABWR 
DCD.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to the U.S. ABWR design are in 10 CFR parts 
20, 50, 73, and 100, codified as of May 2, 1997, that are applicable 
and technically relevant, as described in the FSER (NUREG-1503) and 
Supplement No. 1.
    B. The U.S. ABWR design is exempt from portions of the following 
regulations:
    1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety 
Parameter Display Console;
    2. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident 
Sampling for Boron, Chloride, and Dissolved Gases; and
    3. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment 
Penetration.

[[Page 49546]]

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
components, and design features of the U.S. ABWR design comply with 
the provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for the 
U.S. ABWR design.
    B. The Commission considers the following matters resolved 
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings 
for issuance of a combined license, amendment of a combined license, 
or renewal of a combined license, proceedings held under 10 CFR 
52.103, and enforcement proceedings involving plants referencing 
this appendix:
    1. All nuclear safety issues, except for the generic technical 
specifications and other operational requirements, associated with 
the information in the FSER and Supplement No. 1, Tier 1, Tier 2 
(including referenced information which the context indicates is 
intended as requirements), and the rulemaking record for 
certification of the U.S. ABWR design;
    2. All nuclear safety and safeguards issues associated with the 
information in proprietary and safeguards documents, referenced and 
in context, are intended as requirements in the generic DCD for the 
U.S. ABWR design;
    3. All generic changes to the DCD under and in compliance with 
the change processes in Sections VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, 
but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.f of this appendix, 
all departures from Tier 2 pursuant to and in compliance with the 
change processes in paragraph VIII.B.5 of this appendix that do not 
require prior NRC approval, but only for that plant;
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's final environmental assessment for the U.S. ABWR design 
and Revision 1 of the technical support document for the U.S. ABWR, 
dated December 1994, for plants referencing this appendix whose site 
parameters are within those specified in the technical support 
document.
    C. The Commission does not consider operational requirements for 
an applicant or licensee who references this appendix to be matters 
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission 
reserves the right to require operational requirements for an 
applicant or licensee who references this appendix by rule, 
regulation, order, or license condition.
    D. Except in accordance with the change processes in Section 
VIII of this appendix, the Commission may not require an applicant 
or licensee who references this appendix to:
    1. Modify structures, systems, components, or design features as 
described in the generic DCD;
    2. Provide additional or alternative structures, systems, 
components, or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, components, or design features discussed in the generic 
DCD.
    E.1. Persons who wish to review proprietary and safeguards 
information or other secondary references in the DCD for the U.S. 
ABWR design, in order to request or participate in the hearing 
required by 10 CFR 52.85 or the hearing provided under 10 CFR 
52.103, or to request or participate in any other hearing relating 
to this appendix in which interested persons have adjudicatory 
hearing rights, shall first request access to such information from 
GE Nuclear Energy. The request must state with particularity:
    a. The nature of the proprietary or other information sought;
    b. The reason why the information currently available to the 
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC 
Public Document Room, is insufficient;
    c. The relevance of the requested information to the hearing 
issue(s) which the person proposes to raise; and
    d. A showing that the requesting person has the capability to 
understand and utilize the requested information.
    2. If a person claims that the information is necessary to 
prepare a request for hearing, the request must be filed no later 
than 15 days after publication in the Federal Register of the notice 
required either by 10 CFR 52.85 or 10 CFR 52.103. If GE Nuclear 
Energy declines to provide the information sought, GE Nuclear Energy 
shall send a written response within 10 days of receiving the 
request to the requesting person setting forth with particularity 
the reasons for its refusal. The person may then request the 
Commission (or presiding officer, if a proceeding has been 
established) to order disclosure. The person shall include copies of 
the original request (and any subsequent clarifying information 
provided by the requesting party to the applicant) and the 
applicant's response. The Commission and presiding officer shall 
base their decisions solely on the person's original request 
(including any clarifying information provided by the requesting 
person to GE Nuclear Energy), and GE Nuclear Energy's response. The 
Commission and presiding officer may order GE Nuclear Energy to 
provide access to some or all of the requested information, subject 
to an appropriate non-disclosure agreement.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
June 11, 1997, except as provided for in 10 CFR 52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

    A. Tier 1 information.
    1. Generic changes to Tier 1 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in 10 CFR 52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will 
deny a request for an exemption from Tier 1, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design.
    B. Tier 2 information.
    1. Generic changes to Tier 2 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order while this appendix is in effect 
under Sec. Sec.  52.55 or 52.61, unless:
    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to assure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 52.7 are present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 52.7. The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The grant of an exemption 
to an applicant must be subject to litigation in the same manner as 
other issues material to the license hearing. The grant of an 
exemption to a licensee must be subject to an opportunity for a 
hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the technical specifications, 
or requires a license amendment under paragraphs B.5.b or B.5.c

[[Page 49547]]

of this section. When evaluating the proposed departure, an 
applicant or licensee shall consider all matters described in the 
plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
(SSC) important to safety previously evaluated in the plant-specific 
DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of a SSC important to safety previously evaluated 
in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of an SSC important 
to safety with a different result than any evaluated previously in 
the plant-specific DCD;
    (7) Result in a design basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2 affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe 
accident previously reviewed and determined to be not credible could 
become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. If a departure requires a license amendment pursuant to 
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR 
50.90.
    e. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    f. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an applicant or licensee who 
references this appendix has not complied with paragraph VIII.B.5 of 
this appendix when departing from Tier 2 information, may petition 
the NRC to admit into the proceeding such a contention. In addition 
to compliance with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a 10 CFR 52.103 
preoperational hearing, or that the change bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such a 
contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.
    6.a. An applicant who references this appendix may not depart 
from Tier 2* information, which is designated with italicized text 
or brackets and an asterisk in the generic DCD, without NRC 
approval. The departure will not be considered a resolved issue, 
within the meaning of Section VI of this appendix and 10 CFR 
52.63(a)(5).
    b. A licensee who references this appendix may not depart from 
the following Tier 2* matters without prior NRC approval. A request 
for a departure will be treated as a request for a license amendment 
under 10 CFR 50.90.
    (1) Fuel burnup limit (4.2).
    (2) Fuel design evaluation (4.2.3).
    (3) Fuel licensing acceptance criteria (appendix 4B).
    c. A licensee who references this appendix may not, before the 
plant first achieves full power following the finding required by 10 
CFR 52.103(g), depart from the following Tier 2* matters except in 
accordance with paragraph B.6.b of this section. After the plant 
first achieves full power, the following Tier 2* matters revert to 
Tier 2 status and are thereafter subject to the departure provisions 
in paragraph B.5 of this section.
    (1) ASME Boiler & Pressure Vessel Code, Section III.
    (2) ACI 349 and ANSI/AISC-690.
    (3) Motor-operated valves.
    (4) Equipment seismic qualification methods.
    (5) Piping design acceptance criteria.
    (6) Fuel system and assembly design (4.2), except burnup limit.
    (7) Nuclear design (4.3).
    (8) Equilibrium cycle and control rod patterns (App. 4A).
    (9) Control rod licensing acceptance criteria (App. 4C).
    (10) Instrument setpoint methodology.
    (11) EMS performance specifications and architecture.
    (12) SSLC hardware and software qualification.
    (13) Self-test system design testing features and commitments.
    (14) Human factors engineering design and implementation 
process.
    d. Departures from Tier 2* information that are made under 
paragraph B.6 of this section do not require an exemption from this 
appendix.
    C. Operational requirements.
    1. Generic changes to generic technical specifications and other 
operational requirements that were completely reviewed and approved 
in the design certification rulemaking and do not require a change 
to a design feature in the generic DCD are governed by the 
requirements in 10 CFR 50.109. Generic changes that do require a 
change to a design feature in the generic DCD are governed by the 
requirements in paragraphs A or B of this section.
    2. Generic changes to generic TS and other operational 
requirements are applicable to all applicants who reference this 
appendix, except those for which the change has been rendered 
technically irrelevant by action taken under paragraphs C.3 or C.4 
of this section.
    3. The Commission may require plant-specific departures on 
generic technical specifications and other operational requirements 
that were completely reviewed and approved, provided a change to a 
design feature in the generic DCD is not required and special 
circumstances as defined in 10 CFR 2.335 are present. The Commission 
may modify or supplement generic technical specifications and other 
operational requirements that were not completely reviewed and 
approved or require additional technical specifications and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic technical specifications or other 
operational requirements. The Commission may grant such a request 
only if it determines that the exemption will comply with the 
requirements of 10 CFR 52.7. The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing.
    5. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an operational requirement 
approved in the DCD or a technical specification derived from the 
generic technical specifications must be changed may petition to 
admit into the proceeding such a contention. Such petition must 
comply with the general requirements of 10 CFR 2.309 and must 
demonstrate why special circumstances as defined in 10 CFR 2.335 are 
present, or for compliance with the Commission's regulations in 
effect at the time this appendix was approved, as set forth in 
Section V of this appendix. Any other party may file a response 
thereto. If, on the basis of the petition and any response, the 
presiding officer determines that a sufficient showing has been 
made, the presiding officer shall certify the matter directly to the 
Commission for determination of the admissibility of the contention. 
All other issues with respect to the plant-specific technical 
specifications or other operational requirements are subject to a 
hearing as part of the license proceeding.
    6. After issuance of a license, the generic technical 
specifications have no further effect on the plant-specific 
technical specifications and changes to the plant-specific technical 
specifications will be treated as license amendments under 10 CFR 
50.90.

[[Page 49548]]

IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)

    A.1. An applicant or licensee who references this appendix shall 
perform and demonstrate conformance with the ITAAC before fuel load. 
With respect to activities subject to an ITAAC, an applicant for a 
license may proceed at its own risk with design and procurement 
activities, and a licensee may proceed at its own risk with design, 
procurement, construction, and preoperational activities, even 
though the NRC may not have found that any particular ITAAC has been 
met.
    2. The licensee who references this appendix shall notify the 
NRC that the required inspections, tests, and analyses in the ITAAC 
have been successfully completed and that the corresponding 
acceptance criteria have been met.
    3. In the event that an activity is subject to an ITAAC, and the 
applicant or licensee who references this appendix has not 
demonstrated that the ITAAC has been met, the applicant or licensee 
may either take corrective actions to successfully complete that 
ITAAC, request an exemption from the ITAAC in accordance with 
Section VIII of this appendix and 10 CFR 52.97(b), or petition for 
rulemaking to amend this appendix by changing the requirements of 
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes 
to the ITAAC must meet the requirements of paragraph VIII.A.1 of 
this appendix.
    B.1. The NRC shall ensure that the required inspections, tests, 
and analyses in the ITAAC are performed. The NRC shall verify that 
the inspections, tests, and analyses referenced by the licensee have 
been successfully completed and, based solely thereon, find the 
prescribed acceptance criteria have been met. At appropriate 
intervals during construction, the NRC shall publish notices of the 
successful completion of ITAAC in the Federal Register.
    2. In accordance with 10 CFR 52.103(g), the Commission shall 
find that the acceptance criteria in the ITAAC for the license are 
met before fuel load.
    3. After the Commission has made the finding required by 10 CFR 
52.103(g), the ITAAC do not, by virtue of their inclusion within the 
DCD, constitute regulatory requirements either for licensees or for 
renewal of the license; except for specific ITAAC, which are the 
subject of a Sec.  52.103(a) hearing, their expiration will occur 
upon final Commission action in such proceeding. However, subsequent 
modifications must comply with the Tier 1 and Tier 2 design 
descriptions in the plant-specific DCD unless the licensee has 
complied with the applicable requirements of 10 CFR 52.98 and 
Section VIII of this appendix.

X. Records and Reporting

    A. Records.
    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes to Tier 1, Tier 2, and 
the generic TS and other operational requirements. The applicant 
shall maintain the proprietary and safeguards information referenced 
in the generic DCD for the period that this appendix may be 
referenced, as specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section VIII of this appendix throughout the period of application 
and for the term of the license (including any period of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).
    B. Reporting.
    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each. This report must be filed in accordance with the 
filing requirements applicable to reports in 10 CFR 52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its DCD, which reflect the generic changes and the 
plant-specific departures from the generic DCD made under Section 
VIII of this appendix. These updates must be filed under the filing 
requirements applicable to final safety analysis report updates in 
10 CFR 52.3 and 50.71(e).
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes the finding required by 10 
CFR 52.103(g), the report must be submitted semiannually. Updates to 
the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by 10 CFR 
52.103(g), reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 10 CFR 50.71(e)(4), respectively, 
or at shorter intervals as specified in the license.

Appendix B to Part 52--Design Certification Rule for the System 80+ 
Design

I. Introduction

    Appendix B constitutes design certification for the System 80+ 
\1\ standard plant design, in accordance with 10 CFR part 52, 
subpart B. The applicant for certification of the System 80+ design 
was Combustion Engineering, Inc. (ABB-CE), which is now Westinghouse 
Electric Company LLC.
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    \1\ ``System 80+'' is a trademark of Westinghouse Electric 
Company LLC.
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II. Definitions

    A. Generic design control document (generic DCD) means the 
document containing the Tier 1 and Tier 2 information and generic 
technical specifications that is incorporated by reference into this 
appendix.
    B. Generic technical specifications means the information, 
required by 10 CFR 50.36 and 50.36a, for the portion of the plant 
that is within the scope of this appendix.
    C. Plant-specific DCD means the document, maintained by an 
applicant or licensee who references this appendix, consisting of 
the information in the generic DCD, as modified and supplemented by 
the plant-specific departures and exemptions made under Section VIII 
of this appendix.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (hereinafter Tier 1 information). The design descriptions, 
interface requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;
    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAAC);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must meet the requirement in Section III.B of this appendix to 
reference Tier 2 when referencing Tier 1. Tier 2 information 
includes:
    1. Information required by Sec. Sec.  52.47(a) and 52.47(c), 
with the exception of generic technical specifications and 
conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAAC have been met; and
    3. Combined license (COL) action items (COL license 
information), which identify certain matters that must be addressed 
in the site-specific portion of the final safety analysis report 
(FSAR) by an applicant who references this appendix. These items 
constitute information requirements but are not the only acceptable 
set of information in the FSAR. An applicant may depart from or omit 
these items, provided that the departure or omission is identified 
and justified in the FSAR. After issuance of a construction permit 
or COL, these items are not requirements for the licensee unless 
such items are restated in the FSAR.
    F. Tier 2* means the portion of the Tier 2 information, 
designated as such in the generic DCD, which is subject to the 
change process in Section VIII.B.6 of this appendix.

[[Page 49549]]

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 or 52.1, or Section 11 of the Atomic Energy Act of 1954, 
as amended, as applicable.

III. Scope and Contents

    A. Tier 1, Tier 2, and the generic technical specifications in 
the System 80+ Design Control Document, ABB-CE, with revisions dated 
January 1997, are approved for incorporation by reference by the 
Director of the Office of the Federal Register in accordance with 5 
U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be 
obtained from the National Technical Information Service, 5285 Port 
Royal Road, Springfield, Virginia 22161. A copy is available for 
examination and copying at the NRC Public Document Room located at 
One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852. Copies are also available for examination 
at the NRC Library located at Two White Flint North, 11545 Rockville 
Pike, Rockville, Maryland 20582 and the Office of the Federal 
Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.
    B. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix, 
including Tier 1, Tier 2, and the generic technical specifications 
except as otherwise provided in this appendix. Conceptual design 
information, as set forth in the generic DCD, and the Technical 
Support Document for the System 80+ design are not part of this 
appendix.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for design certification of the System 80+ design or 
NUREG-1462, ``Final Safety Evaluation Report Related to the 
Certification of the System 80+ Design,'' (FSER) and Supplement No. 
1, then the generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a combined license that wishes to reference 
this appendix shall, in addition to complying with the requirements 
of 10 CFR 52.77, 52.79, and 52.80, comply with the following 
requirements:
    1. Incorporate by reference, as part of its application, this 
appendix;
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and using the same organization and numbering as the generic DCD for 
the System 80+ design, as modified and supplemented by the 
applicant's exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific technical specifications, consisting of the 
generic and site-specific technical specifications, that are 
required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating compliance with the site parameters 
and interface requirements;
    e. Information that addresses the COL action items; and
    f. Information required by 10 CFR 52.47 that is not within the 
scope of this appendix.
    3. Include, in the plant-specific DCD, the proprietary 
information referenced in the System 80+ DCD.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to the System 80+ design are in 10 CFR parts 
20, 50, 73, and 100, codified as of May 9, 1997, that are applicable 
and technically relevant, as described in the FSER (NUREG-1462) and 
Supplement No. 1.
    B. The System 80+ design is exempt from portions of the 
following regulations:
    1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety 
Parameter Display Console;
    2. Paragraphs (f)(2) (vii), (viii), (xxvi), and (xxviii) of 10 
CFR 50.34--Accident Source Terms;
    3. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident 
Sampling for Hydrogen, Boron, Chloride, and Dissolved Gases;
    4. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment 
Penetration; and
    5. Paragraphs III.A.1(a) and III.C.3(b) of Appendix J to 10 CFR 
50--Containment Leakage Testing.

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
components, and design features of the System 80+ design comply with 
the provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for the 
System 80+ design.
    B. The Commission considers the following matters resolved 
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings 
for issuance of a combined license, amendment of a combined license, 
or renewal of a combined license, proceedings held under 10 CFR 
52.103, and enforcement proceedings involving plants referencing 
this appendix:
    1. All nuclear safety issues, except for the generic technical 
specifications and other operational requirements, associated with 
the information in the FSER and Supplement No. 1, Tier 1, Tier 2 
(including referenced information which the context indicates is 
intended as requirements), and the rulemaking record for 
certification of the System 80+ design;
    2. All nuclear safety and safeguards issues associated with the 
information in proprietary and safeguards documents, referenced and 
in context, are intended as requirements in the generic DCD for the 
System 80+ design;
    3. All generic changes to the DCD under and in compliance with 
the change processes in Sections VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, 
but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.f of this appendix, 
all departures from Tier 2 under and in compliance with the change 
processes in paragraph VIII.B.5 of this appendix that do not require 
prior NRC approval, but only for that plant;
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's final environmental assessment for the System 80+ design 
and the technical support document for the System 80+ design, dated 
January 1995, for plants referencing this appendix whose site 
parameters are within those specified in the technical support 
document.
    C. The Commission does not consider operational requirements for 
an applicant or licensee who references this appendix to be matters 
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission 
reserves the right to require operational requirements for an 
applicant or licensee who references this appendix by rule, 
regulation, order, or license condition.
    D. Except in accordance with the change processes in Section 
VIII of this appendix, the Commission may not require an applicant 
or licensee who references this appendix to:
    1. Modify structures, systems, components, or design features as 
described in the generic DCD;
    2. Provide additional or alternative structures, systems, 
components, or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, components, or design features discussed in the generic 
DCD.
    E.1. Persons who wish to review proprietary information or other 
secondary references in the DCD for the System 80+ design, in order 
to request or participate in

[[Page 49550]]

the hearing required by 10 CFR 52.85 or the hearing provided under 
10 CFR 52.103, or to request or participate in any other hearing 
relating to this appendix in which interested persons have 
adjudicatory hearing rights, shall first request access to such 
information from Westinghouse. The request must state with 
particularity:
    a. The nature of the proprietary or other information sought;
    b. The reason why the information currently available to the 
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC 
Public Document Room, is insufficient;
    c. The relevance of the requested information to the hearing 
issue(s) which the person proposes to raise; and
    d. A showing that the requesting person has the capability to 
understand and utilize the requested information.
    2. If a person claims that the information is necessary to 
prepare a request for hearing, the request must be filed no later 
than 15 days after publication in the Federal Register of the notice 
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse 
declines to provide the information sought, Westinghouse shall send 
a written response within ten (10) days of receiving the request to 
the requesting person setting forth with particularity the reasons 
for its refusal. The person may then request the Commission (or 
presiding officer, if a proceeding has been established) to order 
disclosure. The person shall include copies of the original request 
(and any subsequent clarifying information provided by the 
requesting party to the applicant) and the applicant's response. The 
Commission and presiding officer shall base their decisions solely 
on the person's original request (including any clarifying 
information provided by the requesting person to Westinghouse), and 
Westinghouse's response. The Commission and presiding officer may 
order Westinghouse to provide access to some or all of the requested 
information, subject to an appropriate non-disclosure agreement.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
June 20, 1997, except as provided for in 10 CFR 52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

    A. Tier 1 information.
    1. Generic changes to Tier 1 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in 10 CFR 52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will 
deny a request for an exemption from Tier 1, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design.

B. Tier 2 Information

    1. Generic changes to Tier 2 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order while this appendix is in effect 
under Sec. Sec.  52.55 or 52.61, unless:
    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to assure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 52.7 are present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 52.7. The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The grant of an exemption 
to an applicant must be subject to litigation in the same manner as 
other issues material to the license hearing. The grant of an 
exemption to a licensee must be subject to an opportunity for a 
hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the technical specifications, 
or requires a license amendment under paragraphs B.5.b or B.5.c of 
this section. When evaluating the proposed departure, an applicant 
or licensee shall consider all matters described in the plant-
specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
(SSC) important to safety previously evaluated in the plant-specific 
DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of an SSC important to safety previously evaluated 
in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of an SSC important 
to safety with a different result than any evaluated previously in 
the plant-specific DCD;
    (7) Result in a design basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2 affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe 
accident previously reviewed and determined to be not credible could 
become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. If a departure requires a license amendment under paragraph 
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
    e. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    f. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an applicant or licensee who 
references this appendix has not complied with paragraph VIII.B.5 of 
this appendix when departing from Tier 2 information, may petition 
the NRC to admit into the proceeding such a contention. In addition 
to compliance with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a 10 CFR 52.103 
preoperational hearing, or that the change bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such a 
contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.
    6.a. An applicant who references this appendix may not depart 
from Tier 2*

[[Page 49551]]

information, which is designated with italicized text or brackets 
and an asterisk in the generic DCD, without NRC approval. The 
departure will not be considered a resolved issue, within the 
meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
    b. A licensee who references this appendix may not depart from 
the following Tier 2* matters without prior NRC approval. A request 
for a departure will be treated as a request for a license amendment 
under 10 CFR 50.90.
    (1) Maximum fuel rod average burnup.
    (2) Control room human factors engineering.
    c. A licensee who references this appendix may not, before the 
plant first achieves full power following the finding required by 10 
CFR 52.103(g), depart from the following Tier 2* matters except in 
accordance with paragraph B.6.b of this section. After the plant 
first achieves full power, the following Tier 2* matters revert to 
Tier 2 status and are thereafter subject to the departure provisions 
in paragraph B.5 of this section.
    (1) ASME Boiler & Pressure Vessel Code, Section III.
    (2) ACI 349 and ANSI/AISC-690.
    (3) Motor-operated valves.
    (4) Equipment seismic qualification methods.
    (5) Piping design acceptance criteria.
    (6) Fuel and control rod design, except burnup limit.
    (7) Instrumentation and controls setpoint methodology.
    (8) Instrumentation and controls hardware and software changes.
    (9) Instrumentation and controls environmental qualification.
    (10) Seismic design criteria for non-seismic Category I 
structures.
    d. Departures from Tier 2* information that are made under 
paragraph B.6 of this section do not require an exemption from this 
appendix.
    C. Operational requirements.
    1. Generic changes to generic technical specifications and other 
operational requirements that were completely reviewed and approved 
in the design certification rulemaking and do not require a change 
to a design feature in the generic DCD are governed by the 
requirements in 10 CFR 50.109. Generic changes that do require a 
change to a design feature in the generic DCD are governed by the 
requirements in paragraphs A or B of this section.
    2. Generic changes to generic TS and other operational 
requirements are applicable to all applicants who reference this 
appendix, except those for which the change has been rendered 
technically irrelevant by action taken under paragraphs C.3 or C.4 
of this section.
    3. The Commission may require plant-specific departures on 
generic technical specifications and other operational requirements 
that were completely reviewed and approved, provided a change to a 
design feature in the generic DCD is not required and special 
circumstances as defined in 10 CFR 2.335 are present. The Commission 
may modify or supplement generic technical specifications and other 
operational requirements that were not completely reviewed and 
approved or require additional technical specifications and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic technical specifications or other 
operational requirements. The Commission may grant such a request 
only if it determines that the exemption will comply with the 
requirements of 10 CFR 52.7. The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing.
    5. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an operational requirement 
approved in the DCD or a technical specification derived from the 
generic technical specifications must be changed may petition to 
admit into the proceeding such a contention. Such a petition must 
comply with the general requirements of 10 CFR 2.309 and must 
demonstrate why special circumstances as defined in 10 CFR 2.335 are 
present, or for compliance with the Commission's regulations in 
effect at the time this appendix was approved, as set forth in 
Section V of this appendix. Any other party may file a response 
thereto. If, on the basis of the petition and any response, the 
presiding officer determines that a sufficient showing has been 
made, the presiding officer shall certify the matter directly to the 
Commission for determination of the admissibility of the contention. 
All other issues with respect to the plant-specific technical 
specifications or other operational requirements are subject to a 
hearing as part of the license proceeding.
    6. After issuance of a license, the generic technical 
specifications have no further effect on the plant-specific 
technical specifications and changes to the plant-specific technical 
specifications will be treated as license amendments under 10 CFR 
50.90.

IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)

    A.1 An applicant or licensee who references this appendix shall 
perform and demonstrate conformance with the ITAAC before fuel load. 
With respect to activities subject to an ITAAC, an applicant for a 
license may proceed at its own risk with design and procurement 
activities, and a licensee may proceed at its own risk with design, 
procurement, construction, and preoperational activities, even 
though the NRC may not have found that any particular ITAAC has been 
met.
    2. The licensee who references this appendix shall notify the 
NRC that the required inspections, tests, and analyses in the ITAAC 
have been successfully completed and that the corresponding 
acceptance criteria have been met.
    3. In the event that an activity is subject to an ITAAC, and the 
applicant or licensee who references this appendix has not 
demonstrated that the ITAAC has been met, the applicant or licensee 
may either take corrective actions to successfully complete that 
ITAAC, request an exemption from the ITAAC in accordance with 
Section VIII of this appendix and 10 CFR 52.97(b), or petition for 
rulemaking to amend this appendix by changing the requirements of 
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes 
to the ITAAC must meet the requirements of Section VIII.A.1 of this 
appendix.
    B.1 The NRC shall ensure that the required inspections, tests, 
and analyses in the ITAAC are performed. The NRC shall verify that 
the inspections, tests, and analyses referenced by the licensee have 
been successfully completed and, based solely thereon, find the 
prescribed acceptance criteria have been met. At appropriate 
intervals during construction, the NRC shall publish notices of the 
successful completion of ITAAC in the Federal Register.
    2. In accordance with 10 CFR 52.103(g), the Commission shall 
find that the acceptance criteria in the ITAAC for the license are 
met before fuel load.
    3. After the Commission has made the finding required by 10 CFR 
52.103(g), the ITAAC do not, by virtue of their inclusion within the 
DCD, constitute regulatory requirements either for licensees or for 
renewal of the license; except for specific ITAAC, which are the 
subject of a Sec.  52.103(a) hearing, their expiration will occur 
upon final Commission action in such proceeding. However, subsequent 
modifications must comply with the Tier 1 and Tier 2 design 
descriptions in the plant-specific DCD unless the licensee has 
complied with the applicable requirements of 10 CFR 52.98 and 
Section VIII of this appendix.

X. Records and Reporting

    A. Records.
    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes to Tier 1, Tier 2, and 
the generic TS and other operational requirements. The applicant 
shall maintain the proprietary and safeguards information referenced 
in the generic DCD for the period that this appendix may be 
referenced, as specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section VIII of this appendix throughout the period of application 
and for the term of the license (including any period of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).
    B. Reporting.
    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each. This report must be filed in accordance with the 
filing requirements applicable to reports in 10 CFR 52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its

[[Page 49552]]

DCD, which reflect the generic changes to and plant-specific 
departures from the generic DCD made under Section VIII of this 
appendix. These updates must be filed under the filing requirements 
applicable to final safety analysis report updates in 10 CFR 52.3 
and 50.71(e).
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes the finding required by 10 
CFR 52.103(g), the report must be submitted semi-annually. Updates 
to the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by 10 CFR 
52.103(g), the reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at 
shorter intervals as specified in the license.

Appendix C to Part 52--Design Certification Rule for the AP600 Design

I. Introduction

    Appendix C constitutes the standard design certification for the 
AP600 \1\ design, in accordance with 10 CFR part 52, subpart B. The 
applicant for certification of the AP600 design is Westinghouse 
Electric Company LLC.
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    \1\ AP600 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions

    A. Generic design control document (generic DCD) means the 
document containing the Tier 1 and Tier 2 information and generic 
technical specifications that is incorporated by reference into this 
appendix.
    B. Generic technical specifications means the information, 
required by 10 CFR 50.36 and 50.36a, for the portion of the plant 
that is within the scope of this appendix.
    C. Plant-specific DCD means the document, maintained by an 
applicant or licensee who references this appendix, consisting of 
the information in the generic DCD, as modified and supplemented by 
the plant-specific departures and exemptions made under Section VIII 
of this appendix.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (hereinafter Tier 1 information). The design descriptions, 
interface requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;
    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAAC);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must meet the requirement in Section III.B of this appendix to 
reference Tier 2 when referencing Tier 1. Tier 2 information 
includes:
    1. Information required by Sec. Sec.  52.47(a) and 52.47(c), 
with the exception of generic technical specifications and 
conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAAC have been met; and
    3. Combined license (COL) action items (COL license 
information), which identify certain matters that must be addressed 
in the site-specific portion of the final safety analysis report 
(FSAR) by an applicant who references this appendix. These items 
constitute information requirements but are not the only acceptable 
set of information in the FSAR. An applicant may depart from or omit 
these items, provided that the departure or omission is identified 
and justified in the FSAR. After issuance of a construction permit 
or COL, these items are not requirements for the licensee unless 
such items are restated in the FSAR.
    4. The investment protection short-term availability controls in 
Section 16.3 of the DCD.
    F. Tier 2* means the portion of the Tier 2 information, 
designated as such in the generic DCD, which is subject to the 
change process in Section VIII.B.6 of this appendix. This 
designation expires for some Tier 2* information under Section 
VIII.B.6.
    G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety 
analyses means:
    (1) Changing any of the elements of the method described in the 
plant-specific DCD unless the results of the analysis are 
conservative or essentially the same; or
    (2) Changing from a method described in the plant-specific DCD 
to another method unless that method has been approved by NRC for 
the intended application.
    H. All other terms in this appendix have the meaning set out in 
10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954, 
as amended, as applicable.

III. Scope and Contents

    A. Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic 
technical specifications in the AP600 DCD (12/99 revision) are 
approved for incorporation by reference by the Director of the 
Office of the Federal Register on January 24, 2000, in accordance 
with 5 U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD 
may be obtained from Ronald P. Vijuk, Manager, Passive Plant 
Engineering, Westinghouse Electric Company, P.O. Box 355, 
Pittsburgh, Pennsylvania 15230-0355. A copy of the generic DCD is 
available for examination and copying at the NRC Public Document 
Room located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. Copies are also available for 
examination at the NRC Library located at Two White Flint North, 
11545 Rockville Pike, Rockville, Maryland 20582; and the Office of 
the Federal Register, 800 North Capitol Street, NW., Suite 700, 
Washington, DC.
    B. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix, 
including Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic 
technical specifications except as otherwise provided in this 
appendix. Conceptual design information in the generic DCD and the 
evaluation of severe accident mitigation design alternatives in 
Appendix 1B of the generic DCD are not part of this appendix.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for design certification of the AP600 design or NUREG-
1512, ``Final Safety Evaluation Report Related to Certification of 
the AP600 Standard Design,'' (FSER), then the generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a combined license that wishes to reference 
this appendix shall, in addition to complying with the requirements 
of 10 CFR 52.77, 52.79, and 52.80, comply with the following 
requirements:
    1. Incorporate by reference, as part of its application, this 
appendix;
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and utilizing the same organization and numbering as the generic DCD 
for the AP600 design, as modified and supplemented by the 
applicant's exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific technical specifications, consisting of the 
generic and site-specific technical specifications, that are 
required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating compliance with the site parameters 
and interface requirements;
    e. Information that addresses the COL action items; and

[[Page 49553]]

    f. Information required by 10 CFR 52.47 that is not within the 
scope of this appendix.
    3. Include, in the plant-specific DCD, the proprietary 
information and safeguards information referenced in the AP600 DCD.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to the AP600 design are in 10 CFR parts 20, 
50, 73, and 100, codified as of December 16, 1999, that are 
applicable and technically relevant, as described in the FSER 
(NUREG-1512) and the supplementary information for this section.
    B. The AP600 design is exempt from portions of the following 
regulations:
    1. Paragraph (a)(1) of 10 CFR 50.34--whole body dose criterion;
    2. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter 
Display Console;
    3. Paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10 
CFR 50.34--Accident Source Term in TID 14844;
    4. Paragraph (a)(2) of 10 CFR 50.55a--ASME Boiler and Pressure 
Vessel Code;
    5. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency) 
feedwater system;
    6. Appendix A to 10 CFR part 50, GDC 17--Offsite Power Sources; 
and
    7. Appendix A to 10 CFR part 50, GDC 19--whole body dose 
criterion.

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
components, and design features of the AP600 design comply with the 
provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for the 
AP600 design.
    B. The Commission considers the following matters resolved 
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings 
for issuance of a combined license, amendment of a combined license, 
or renewal of a combined license, proceedings held under 10 CFR 
52.103, and enforcement proceedings involving plants referencing 
this appendix:
    1. All nuclear safety issues, except for the generic technical 
specifications and other operational requirements, associated with 
the information in the FSER and Supplement No. 1, Tier 1, Tier 2 
(including referenced information which the context indicates is 
intended as requirements and the investment protection short-term 
availability controls in Section 16.3), and the rulemaking record 
for certification of the AP600 design;
    2. All nuclear safety and safeguards issues associated with the 
information in proprietary and safeguards documents, referenced and 
in context, are intended as requirements in the generic DCD for the 
AP600 design;
    3. All generic changes to the DCD under and in compliance with 
the change processes in Sections VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, 
but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.f of this appendix, 
all departures from Tier 2 under and in compliance with the change 
processes in paragraph VIII.B.5 of this appendix that do not require 
prior NRC approval, but only for that plant;
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's environmental assessment for the AP600 design and appendix 
1B of the generic DCD, for plants referencing this appendix whose 
site parameters are within those specified in the severe accident 
mitigation design alternatives evaluation.
    C. The Commission does not consider operational requirements for 
an applicant or licensee who references this appendix to be matters 
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission 
reserves the right to require operational requirements for an 
applicant or licensee who references this appendix by rule, 
regulation, order, or license condition.
    D. Except in accordance with the change processes in Section 
VIII of this appendix, the Commission may not require an applicant 
or licensee who references this appendix to:
    1. Modify structures, systems, components, or design features as 
described in the generic DCD;
    2. Provide additional or alternative structures, systems, 
components, or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, components, or design features discussed in the generic 
DCD.
    E.1. Persons who wish to review proprietary and safeguards 
information or other secondary references in the AP600 DCD, in order 
to request or participate in the hearing required by 10 CFR 52.85 or 
the hearing provided under 10 CFR 52.103, or to request or 
participate in any other hearing relating to this appendix in which 
interested persons have adjudicatory hearing rights, shall first 
request access to such information from Westinghouse. The request 
must state with particularity:
    a. The nature of the proprietary or other information sought;
    b. The reason why the information currently available to the 
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC 
Public Document Room, is insufficient;
    c. The relevance of the requested information to the hearing 
issue(s) which the person proposes to raise; and
    d. A showing that the requesting person has the capability to 
understand and utilize the requested information.
    2. If a person claims that the information is necessary to 
prepare a request for hearing, the request must be filed no later 
than 15 days after publication in the Federal Register of the notice 
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse 
declines to provide the information sought, Westinghouse shall send 
a written response within 10 days of receiving the request to the 
requesting person setting forth with particularity the reasons for 
its refusal. The person may then request the Commission (or 
presiding officer, if a proceeding has been established) to order 
disclosure. The person shall include copies of the original request 
(and any subsequent clarifying information provided by the 
requesting party to the applicant) and the applicant's response. The 
Commission and presiding officer shall base their decisions solely 
on the person's original request (including any clarifying 
information provided by the requesting person to Westinghouse), and 
Westinghouse's response. The Commission and presiding officer may 
order Westinghouse to provide access to some or all of the requested 
information, subject to an appropriate non-disclosure agreement.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
January 24, 2000, except as provided for in 10 CFR 52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

    A. Tier 1 information.
    1. Generic changes to Tier 1 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in 10 CFR 52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will 
deny a request for an exemption from Tier 1, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design.
    B. Tier 2 information.
    1. Generic changes to Tier 2 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order while this appendix is in effect 
under Sec. Sec.  52.55 or 52.61, unless:

[[Page 49554]]

    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to assure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 52.7 are present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 52.7. The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The grant of an exemption 
to an applicant must be subject to litigation in the same manner as 
other issues material to the license hearing. The grant of an 
exemption to a licensee must be subject to an opportunity for a 
hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the technical specifications, 
or requires a license amendment under paragraphs B.5.b or B.5.c of 
this section. When evaluating the proposed departure, an applicant 
or licensee shall consider all matters described in the plant-
specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
(SSC) important to safety previously evaluated in the plant-specific 
DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of an SSC important to safety previously evaluated 
in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of an SSC important 
to safety with a different result than any evaluated previously in 
the plant-specific DCD;
    (7) Result in a design basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2 affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe 
accident previously reviewed and determined to be not credible could 
become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. If a departure requires a license amendment under paragraphs 
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
    e. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    f. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an applicant or licensee who 
references this appendix has not complied with paragraph VIII.B.5 of 
this appendix when departing from Tier 2 information, may petition 
the NRC to admit into the proceeding such a contention. In addition 
to compliance with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a 10 CFR 52.103 
preoperational hearing, or that the change bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such a 
contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.
    6a. An applicant who references this appendix may not depart 
from Tier 2* information, which is designated with italicized text 
or brackets and an asterisk in the generic DCD, without NRC 
approval. The departure will not be considered a resolved issue, 
within the meaning of Section VI of this appendix and 10 CFR 
52.63(a)(5).
    b. A licensee who references this appendix may not depart from 
the following Tier 2* matters without prior NRC approval. A request 
for a departure will be treated as a request for a license amendment 
under 10 CFR 50.90.
    (1) Maximum fuel rod average burn-up.
    (2) Fuel principal design requirements.
    (3) Fuel criteria evaluation process.
    (4) Fire areas.
    (5) Human factors engineering.
    c. A licensee who references this appendix may not, before the 
plant first achieves full power following the finding required by 10 
CFR 52.103(g), depart from the following Tier 2* matters except in 
accordance with paragraph B.6.b of this section. After the plant 
first achieves full power, the following Tier 2* matters revert to 
Tier 2 status and are thereafter subject to the departure provisions 
in paragraph B.5 of this section.
    (1) Nuclear Island structural dimensions.
    (2) ASME Boiler and Pressure Vessel Code, Section III, and Code 
Case--284.
    (3) Design Summary of Critical Sections.
    (4) ACI 318, ACI 349, and ANSI/AISC--690.
    (5) Definition of critical locations and thicknesses.
    (6) Seismic qualification methods and standards.
    (7) Nuclear design of fuel and reactivity control system, except 
burn-up limit.
    (8) Motor-operated and power-operated valves.
    (9) Instrumentation and control system design processes, 
methods, and standards.
    (10) PRHR natural circulation test (first plant only).
    (11) ADS and CMT verification tests (first three plants only).
    d. Departures from Tier 2* information that are made under 
paragraph B.6 of this section do not require an exemption from this 
appendix.
    C. Operational requirements.
    1. Generic changes to generic technical specifications and other 
operational requirements that were completely reviewed and approved 
in the design certification rulemaking and do not require a change 
to a design feature in the generic DCD are governed by the 
requirements in 10 CFR 50.109. Generic changes that do require a 
change to a design feature in the generic DCD are governed by the 
requirements in paragraphs A or B of this section.
    2. Generic changes to generic TS and other operational 
requirements are applicable to all applicants who reference this 
appendix, except those for which the change has been rendered 
technically irrelevant by action taken under paragraphs C.3 or C.4 
of this section.
    3. The Commission may require plant-specific departures on 
generic technical specifications and other operational requirements 
that were completely reviewed and approved, provided a change to a 
design feature in the generic DCD is not required and special 
circumstances as defined in 10 CFR 2.335 are present. The Commission 
may modify or supplement generic technical specifications and other 
operational requirements that were not completely reviewed and 
approved or require additional technical specifications and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic technical specifications or other 
operational requirements. The Commission may grant such a request 
only if it determines that the exemption will comply with the 
requirements of 10 CFR 52.7. The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing.

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    5. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR 52.103(a), who believes that an operational requirement 
approved in the DCD or a technical specification derived from the 
generic technical specifications must be changed may petition to 
admit into the proceeding such a contention. Such petition must 
comply with the general requirements of 10 CFR 2.309 and must 
demonstrate why special circumstances as defined in 10 CFR 2.335 are 
present, or for compliance with the Commission's regulations in 
effect at the time this appendix was approved, as set forth in 
Section V of this appendix. Any other party may file a response 
thereto. If, on the basis of the petition and any response, the 
presiding officer determines that a sufficient showing has been 
made, the presiding officer shall certify the matter directly to the 
Commission for determination of the admissibility of the contention. 
All other issues with respect to the plant-specific technical 
specifications or other operational requirements are subject to a 
hearing as part of the license proceeding.
    6. After issuance of a license, the generic technical 
specifications have no further effect on the plant-specific 
technical specifications and changes to the plant-specific technical 
specifications will be treated as license amendments under 10 CFR 
50.90.

IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)

    A.1 An applicant or licensee who references this appendix shall 
perform and demonstrate conformance with the ITAAC before fuel load. 
With respect to activities subject to an ITAAC, an applicant for a 
license may proceed at its own risk with design and procurement 
activities, and a licensee may proceed at its own risk with design, 
procurement, construction, and preoperational activities, even 
though the NRC may not have found that any particular ITAAC has been 
met.
    2. The licensee who references this appendix shall notify the 
NRC that the required inspections, tests, and analyses in the ITAAC 
have been successfully completed and that the corresponding 
acceptance criteria have been met.
    3. In the event that an activity is subject to an ITAAC, and the 
applicant or licensee who references this appendix has not 
demonstrated that the ITAAC has been met, the applicant or licensee 
may either take corrective actions to successfully complete that 
ITAAC, request an exemption from the ITAAC in accordance with 
Section VIII of this appendix and 10 CFR 52.97(b), or petition for 
rulemaking to amend this appendix by changing the requirements of 
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes 
to the ITAAC must meet the requirements of paragraph VIII.A.1 of 
this appendix.
    B.1. The NRC shall ensure that the required inspections, tests, 
and analyses in the ITAAC are performed. The NRC shall verify that 
the inspections, tests, and analyses referenced by the licensee have 
been successfully completed and, based solely thereon, find the 
prescribed acceptance criteria have been met. At appropriate 
intervals during construction, the NRC shall publish notices of the 
successful completion of ITAAC in the Federal Register.
    2. In accordance with 10 CFR 52.103(g), the Commission shall 
find that the acceptance criteria in the ITAAC for the license are 
met before fuel load.
    3. After the Commission has made the finding required by 10 CFR 
52.103(g), the ITAAC do not, by virtue of their inclusion within the 
DCD, constitute regulatory requirements either for licensees or for 
renewal of the license; except for specific ITAAC, which are the 
subject of a Sec.  52.103(a) hearing, their expiration will occur 
upon final Commission action in such proceeding. However, subsequent 
modifications must comply with the Tier 1 and Tier 2 design 
descriptions in the plant-specific DCD unless the licensee has 
complied with the applicable requirements of 10 CFR 52.98 and 
Section VIII of this appendix.

X. Records and Reporting

    A. Records.
    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes to Tier 1, Tier 2, and 
the generic TS and other operational requirements. The applicant 
shall maintain the proprietary and safeguards information referenced 
in the generic DCD for the period that this appendix may be 
referenced, as specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section VIII of this appendix throughout the period of application 
and for the term of the license (including any period of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).
    B. Reporting.
    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each. This report must be filed in accordance with the 
filing requirements applicable to reports in 10 CFR 52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its DCD, which reflect the generic changes to and 
plant-specific departures from the generic DCD made under Section 
VIII of this appendix. These updates must be filed under the filing 
requirements applicable to final safety analysis report updates in 
10 CFR 52.3 and 50.71(e).
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes the finding required by 10 
CFR 52.103(g), the report must be submitted semi-annually. Updates 
to the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by 10 CFR 
52.103(g), the reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 50.71(e), respectively, or at 
shorter intervals as specified in the license.

Appendix D to Part 52--Design Certification Rule for the AP1000 Design

I. Introduction

    Appendix D constitutes the standard design certification for the 
AP1000 \1\ design, in accordance with 10 CFR part 52, subpart B. The 
applicant for certification of the AP1000 design is Westinghouse 
Electric Company LLC.
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    \1\ AP1000 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions

    A. Generic design control document (generic DCD) means the 
document containing the Tier 1 and Tier 2 information and generic 
technical specifications that is incorporated by reference into this 
appendix.
    B. Generic technical specifications means the information 
required by 10 CFR 50.36 and 50.36a for the portion of the plant 
that is within the scope of this appendix.
    C. Plant-specific DCD means the document maintained by an 
applicant or licensee who references this appendix consisting of the 
information in the generic DCD as modified and supplemented by the 
plant-specific departures and exemptions made under Section VIII of 
this appendix.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (Tier 1 information). The design descriptions, interface 
requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;
    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAAC);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must

[[Page 49556]]

meet the requirement in Section III.B of this appendix to reference 
Tier 2 when referencing Tier 1. Tier 2 information includes:
    1. Information required by Sec. Sec.  52.47(a) and 52.47(c), 
with the exception of generic technical specifications and 
conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAAC have been met; and
    3. Combined license (COL) action items (COL license 
information), which identify certain matters that must be addressed 
in the site-specific portion of the final safety analysis report 
(FSAR) by an applicant who references this appendix. These items 
constitute information requirements but are not the only acceptable 
set of information in the FSAR. An applicant may depart from or omit 
these items, provided that the departure or omission is identified 
and justified in the FSAR. After issuance of a construction permit 
or COL, these items are not requirements for the licensee unless 
such items are restated in the FSAR.
    4. The investment protection short-term availability controls in 
Section 16.3 of the DCD.
    F. Tier 2* means the portion of the Tier 2 information, 
designated as such in the generic DCD, which is subject to the 
change process in Section VIII.B.6 of this appendix. This 
designation expires for some Tier 2* information under paragraph 
VIII.B.6.
    G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety 
analyses means:
    1. Changing any of the elements of the method described in the 
plant-specific DCD unless the results of the analysis are 
conservative or essentially the same; or
    2. Changing from a method described in the plant-specific DCD to 
another method unless that method has been approved by the NRC for 
the intended application.
    H. All other terms in this appendix have the meaning set out in 
10 CFR 50.2, or 52.1, or Section 11 of the Atomic Energy Act of 
1954, as amended, as applicable.

III. Scope and Contents

    A. Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic TS in 
the AP1000 DCD (Revision 15, dated December 8, 2005) are approved 
for incorporation by reference by the Director of the Office of the 
Federal Register on February 27, 2006, under 5 U.S.C. 552(a) and 1 
CFR part 51. Copies of the generic DCD may be obtained from Ronald 
P. Vijuk, Manager, Passive Plant Engineering, Westinghouse Electric 
Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355. A copy 
of the generic DCD is also available for examination and copying at 
the NRC Public Document Room, One White Flint North, 11555 Rockville 
Pike, Rockville, Maryland 20852. Copies are available for 
examination at the NRC Library, Two White Flint North, 11545 
Rockville Pike, Rockville, Maryland, telephone (301) 415-5610, e-
mail <a href="/cdn-cgi/l/email-protection#25696c677764777c656b77660b626a73"><span class="__cf_email__" data-cfemail="efa3a6adbdaebdb6afa1bdacc1a8a0b9">[email protected]</span></a> or at the National Archives and Records 
Administration (NARA). For information on the availability of this 
material at NARA, call (202) 741-6030 or go to http://www.archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html.
    B. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix, 
including Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3 of the DCD), and the 
generic TS except as otherwise provided in this appendix. Conceptual 
design information in the generic DCD and the evaluation of severe 
accident mitigation design alternatives in appendix 1B of the 
generic DCD are not part of this appendix.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for design certification of the AP1000 design or NUREG-
1793, ``Final Safety Evaluation Report Related to Certification of 
the AP1000 Standard Design,'' (FSER) and Supplement No. 1, then the 
generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a combined license that wishes to reference 
this appendix shall, in addition to complying with the requirements 
of 10 CFR 52.77, 52.79, and 52.80, comply with the following 
requirements:
    1. Incorporate by reference, as part of its application, this 
appendix.
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and using the same organization and numbering as the generic DCD for 
the AP1000 design, as modified and supplemented by the applicant's 
exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific TS, consisting of the generic and site-
specific TS that are required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating compliance with the site parameters 
and interface requirements;
    e. Information that addresses the COL action items; and
    f. Information required by 10 CFR 52.47(a) that is not within 
the scope of this appendix.
    3. Include, in the plant-specific DCD, the proprietary 
information and safeguards information referenced in the AP1000 DCD.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to the AP1000 design are in 10 CFR parts 20, 
50, 73, and 100, codified as of January 23, 2006, that are 
applicable and technically relevant, as described in the FSER 
(NUREG-1793) and Supplement No. 1.
    B. The AP1000 design is exempt from portions of the following 
regulations:
    1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter 
Display Console;
    2. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency) 
feedwater system; and
    3. Appendix A to 10 CFR part 50, GDC 17--Second offsite power 
supply circuit.

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
components, and design features of the AP1000 design comply with the 
provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for the 
AP1000 design.
    B. The Commission considers the following matters resolved 
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings 
for issuance of a COL, amendment of a COL, or renewal of a COL, 
proceedings held under 10 CFR 52.103, and enforcement proceedings 
involving plants referencing this appendix:
    1. All nuclear safety issues, except for the generic TS and 
other operational requirements, associated with the information in 
the FSER and Supplement No. 1, Tier 1, Tier 2 (including referenced 
information, which the context indicates is intended as 
requirements, and the investment protection short-term availability 
controls in Section 16.3 of the DCD), and the rulemaking record for 
certification of the AP1000 design;
    2. All nuclear safety and safeguards issues associated with the 
information in proprietary and safeguards documents, referenced and 
in context, are intended as requirements in the generic DCD for the 
AP1000 design;
    3. All generic changes to the DCD under and in compliance with 
the change processes in Sections VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix, 
but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.f of this appendix, 
all departures from Tier 2 under and in compliance with the change 
processes in paragraph VIII.B.5 of this appendix that do not require 
prior NRC approval, but only for that plant;
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's EA for the AP1000 design and Appendix 1B

[[Page 49557]]

of the generic DCD, for plants referencing this appendix whose site 
parameters are within those specified in the severe accident 
mitigation design alternatives evaluation.
    C. The Commission does not consider operational requirements for 
an applicant or licensee who references this appendix to be matters 
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission 
reserves the right to require operational requirements for an 
applicant or licensee who references this appendix by rule, 
regulation, order, or license condition.
    D. Except under the change processes in Section VIII of this 
appendix, the Commission may not require an applicant or licensee 
who references this appendix to:
    1. Modify structures, systems, components, or design features as 
described in the generic DCD;
    2. Provide additional or alternative structures, systems, 
components, or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, components, or design features discussed in the generic 
DCD.
    E.1. Persons who wish to review proprietary and safeguards 
information or other secondary references in the AP1000 DCD, in 
order to request or participate in the hearing required by 10 CFR 
52.85 or the hearing provided under 10 CFR 52.103, or to request or 
participate in any other hearing relating to this appendix in which 
interested persons have adjudicatory hearing rights, shall first 
request access to such information from Westinghouse. The request 
must state with particularity:
    a. The nature of the proprietary or other information sought;
    b. The reason why the information currently available to the 
public in the NRC's public document room is insufficient;
    c. The relevance of the requested information to the hearing 
issue(s) which the person proposes to raise; and
    d. A showing that the requesting person has the capability to 
understand and utilize the requested information.
    2. If a person claims that the information is necessary to 
prepare a request for hearing, the request must be filed no later 
than 15 days after publication in the Federal Register of the notice 
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse 
declines to provide the information sought, Westinghouse shall send 
a written response within 10 days of receiving the request to the 
requesting person setting forth with particularity the reasons for 
its refusal. The person may then request the Commission (or 
presiding officer, if a proceeding has been established) to order 
disclosure. The person shall include copies of the original request 
(and any subsequent clarifying information provided by the 
requesting party to the applicant) and the applicant's response. The 
Commission and presiding officer shall base their decisions solely 
on the person's original request (including any clarifying 
information provided by the requesting person to Westinghouse), and 
Westinghouse's response. The Commission and presiding officer may 
order Westinghouse to provide access to some or all of the requested 
information, subject to an appropriate non-disclosure agreement.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
February 27, 2006, except as provided for in 10 CFR 52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

    A. Tier 1 information.
    1. Generic changes to Tier 1 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in 10 CFR 52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will 
deny a request for an exemption from Tier 1, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design.
    B. Tier 2 information.
    1. Generic changes to Tier 2 information are governed by the 
requirements in 10 CFR 52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order while this appendix is in effect 
under 10 CFR 52.55 or 52.61, unless:
    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to ensure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 50.12(a) are 
present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 50.12(a). The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The grant of an exemption 
to an applicant must be subject to litigation in the same manner as 
other issues material to the license hearing. The grant of an 
exemption to a licensee must be subject to an opportunity for a 
hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, Tier 2* information, or the TS, or requires a license 
amendment under paragraphs B.5.b or B.5.c of this section. When 
evaluating the proposed departure, an applicant or licensee shall 
consider all matters described in the plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of an SSC important to safety previously evaluated 
in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of an SSC important 
to safety with a different result than any evaluated previously in 
the plant-specific DCD;
    (7) Result in a design basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2 affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe 
accident previously reviewed and determined to be not credible could 
become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. If a departure requires a license amendment under paragraph 
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
    e. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    f. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
10 CFR

[[Page 49558]]

52.103(a), who believes that an applicant or licensee who references 
this appendix has not complied with paragraph VIII.B.5 of this 
appendix when departing from Tier 2 information, may petition to 
admit into the proceeding such a contention. In addition to 
compliance with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a 10 CFR 52.103 
preoperational hearing, or that the change bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such a 
contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.
    6.a. An applicant who references this appendix may not depart 
from Tier 2* information, which is designated with italicized text 
or brackets and an asterisk in the generic DCD, without NRC 
approval. The departure will not be considered a resolved issue, 
within the meaning of Section VI of this appendix and 10 CFR 
52.63(a)(5).
    b. A licensee who references this appendix may not depart from 
the following Tier 2* matters without prior NRC approval. A request 
for a departure will be treated as a request for a license amendment 
under 10 CFR 50.90.
    (1) Maximum fuel rod average burn-up.
    (2) Fuel principal design requirements.
    (3) Fuel criteria evaluation process.
    (4) Fire areas.
    (5) Human factors engineering.
    (6) Small-break loss-of-coolant accident (LOCA) analysis 
methodology.
    c. A licensee who references this appendix may not, before the 
plant first achieves full power following the finding required by 10 
CFR 52.103(g), depart from the following Tier 2* matters except 
under paragraph B.6.b of this section. After the plant first 
achieves full power, the following Tier 2* matters revert to Tier 2 
status and are subject to the departure provisions in paragraph B.5 
of this section.
    (1) Nuclear Island structural dimensions.
    (2) American Society of Mechanical Engineers Boiler & Pressure 
Vessel Code (ASME Code), Section III, and Code Case-284.
    (3) Design Summary of Critical Sections.
    (4) American Concrete Institute (ACI) 318, ACI 349, American 
National Standards Institute/American Institute of Steel 
Construction (ANSI/AISC)-690, and American Iron and Steel Institute 
(AISI), ``Specification for the Design of Cold Formed Steel 
Structural Members, Part 1 and 2,'' 1996 Edition and 2000 
Supplement.
    (5) Definition of critical locations and thicknesses.
    (6) Seismic qualification methods and standards.
    (7) Nuclear design of fuel and reactivity control system, except 
burn-up limit.
    (8) Motor-operated and power-operated valves.
    (9) Instrumentation and control system design processes, 
methods, and standards.
    (10) Passive residual heat removal (PRHR) natural circulation 
test (first plant only).
    (11) Automatic depressurization system (ADS) and core make-up 
tank (CMT) verification tests (first three plants only).
    (12) Polar crane parked orientation.
    (13) Piping design acceptance criteria.
    (14) Containment vessel design parameters.
    d. Departures from Tier 2* information that are made under 
paragraph B.6 of this section do not require an exemption from this 
appendix.
    C. Operational requirements.
    1. Generic changes to generic TS and other operational 
requirements that were completely reviewed and approved in the 
design certification rulemaking and do not require a change to a 
design feature in the generic DCD are governed by the requirements 
in 10 CFR 50.109. Generic changes that require a change to a design 
feature in the generic DCD are governed by the requirements in 
paragraphs A or B of this section.
    2. Generic changes to generic TS and other operational 
requirements are applicable to all applicants who reference this 
appendix, except those for which the change has been rendered 
technically irrelevant by action taken under paragraphs C.3 or C.4 
of this section.
    3. The Commission may require plant-specific departures on 
generic TS and other operational requirements that were completely 
reviewed and approved, provided a change to a design feature in the 
generic DCD is not required and special circumstances as defined in 
10 CFR 2.335 are present. The Commission may modify or supplement 
generic TS and other operational requirements that were not 
completely reviewed and approved or require additional TS and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic technical specifications or other 
operational requirements. The Commission may grant such a request 
only if it determines that the exemption will comply with the 
requirements of 10 CFR 52.7. The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing.
    5. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license, or for operation under 
10 CFR 52.103(a), who believes that an operational requirement 
approved in the DCD or a TS derived from the generic TS must be 
changed may petition to admit such a contention into the proceeding. 
The petition must comply with the general requirements of 10 CFR 
2.309 and must demonstrate why special circumstances as defined in 
10 CFR 2.335 are present, or demonstrate compliance with the 
Commission's regulations in effect at the time this appendix was 
approved, as set forth in Section V of this appendix. Any other 
party may file a response to the petition. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. All other issues with respect 
to the plant-specific TS or other operational requirements are 
subject to a hearing as part of the license proceeding.
    6. After issuance of a license, the generic TS have no further 
effect on the plant-specific TS. Changes to the plant-specific TS 
will be treated as license amendments under 10 CFR 50.90.

IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)

    A.1. An applicant or licensee who references this appendix shall 
perform and demonstrate conformance with the ITAAC before fuel load. 
With respect to activities subject to an ITAAC, an applicant for a 
license may proceed at its own risk with design and procurement 
activities. A licensee may also proceed at its own risk with design, 
procurement, construction, and preoperational activities, even 
though the NRC may not have found that any particular ITAAC has been 
met.
    2. The licensee who references this appendix shall notify the 
NRC that the required inspections, tests, and analyses in the ITAAC 
have been successfully completed and that the corresponding 
acceptance criteria have been met.
    3. If an activity is subject to an ITAAC and the applicant or 
licensee who references this appendix has not demonstrated that the 
ITAAC has been met, the applicant or licensee may either take 
corrective actions to successfully complete that ITAAC, request an 
exemption from the ITAAC under Section VIII of this appendix and 10 
CFR 52.97(b), or petition for rulemaking to amend this appendix by 
changing the requirements of the ITAAC, under 10 CFR 2.802 and 
52.97(b). Such rulemaking changes to the ITAAC must meet the 
requirements of paragraph VIII.A.1 of this appendix.
    B.1. The NRC shall ensure that the required inspections, tests, 
and analyses in the ITAAC are performed. The NRC shall verify that 
the inspections, tests, and analyses referenced by the licensee have 
been successfully completed and, based solely thereon, find that the 
prescribed acceptance criteria have been met. At appropriate 
intervals during construction, the NRC shall publish notices of the 
successful completion of ITAAC in the Federal Register.
    2. In accordance with 10 CFR 52.103(g), the Commission shall 
find that the acceptance criteria in the ITAAC for the license are 
met before fuel load.
    3. After the Commission has made the finding required by 10 CFR 
52.103(g), the ITAAC do not, by virtue of their inclusion within the 
DCD, constitute regulatory requirements either for licensees or for 
renewal of the license; except for specific ITAAC, which are the 
subject of a Sec.  52.103(a) hearing, their expiration will occur 
upon final Commission action in such a

[[Page 49559]]

proceeding. However, subsequent modifications must comply with the 
Tier 1 and Tier 2 design descriptions in the plant-specific DCD 
unless the licensee has complied with the applicable requirements of 
10 CFR 52.98 and Section VIII of this appendix.

X. Records and Reporting

    A. Records
    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes to Tier 1, Tier 2, and 
the generic TS and other operational requirements. The applicant 
shall maintain the proprietary and safeguards information referenced 
in the generic DCD for the period that this appendix may be 
referenced, as specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section VIII of this appendix throughout the period of application 
and for the term of the license (including any period of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations which provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any period of renewal).
    B. Reporting
    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each. This report must be filed in accordance with the 
filing requirements applicable to reports in 10 CFR 52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its DCD, which reflect the generic changes to and 
plant-specific departures from the generic DCD made under Section 
VIII of this appendix. These updates must be filed under the filing 
requirements applicable to final safety analysis report updates in 
10 CFR 52.3 and 50.71(e).
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes its findings required by 10 
CFR 52.103(g), the report must be submitted semi-annually. Updates 
to the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by 10 CFR 
52.103(g), the reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at 
shorter intervals as specified in the license.

Appendices E Through M to Part 52 [Reserved]

Appendix N to Part 52--Standardization of Nuclear Power Plant Designs: 
Combined Licenses To Construct and Operate Nuclear Power Reactors of 
Identical Design at Multiple Sites

    The Commission's regulations in part 2 of this chapter 
specifically provide for the holding of hearings on particular 
issues separately from other issues involved in hearings in 
licensing proceedings, and for the consolidation of adjudicatory 
proceedings and of the presentations of parties in adjudicatory 
proceedings such as licensing proceedings (Sec. Sec.  2.316 and 
2.317 of this chapter).
    This appendix sets out the particular requirements and 
provisions applicable to situations in which applications for 
combined licenses under subpart C of this part are filed by one or 
more applicants for licenses to construct and operate nuclear power 
reactors of identical design (``common design'') to be located at 
multiple sites.\1\
---------------------------------------------------------------------------

    \1\ If the design for the power reactor(s) proposed in a 
particular application is not identical to the others, that 
application may not be processed under this appendix and subpart D 
of part 2 of this chapter.
---------------------------------------------------------------------------

    1. Except as otherwise specified in this appendix or as the 
context otherwise indicates, the provisions of subpart C of this 
part and subpart D of part 2 of this chapter apply to combined 
license applications subject to this appendix.
    2. Each combined license application submitted pursuant to this 
appendix must be submitted as specified in Sec.  52.75 and 10 CFR 
2.101. Each application must state that the applicant wishes to have 
the application considered under 10 CFR part 52, appendix N, and 
must list each of the applications to be treated together under this 
appendix.
    3. Each application must include the information required by 
Sec. Sec.  52.77, 52.79, and 52.80(a), provided however, that the 
application must identify the common design, and, if applicable, 
reference a standard design certification under subpart B of this 
part, or the use of a reactor manufactured under subpart F of this 
part. The final safety analysis report for each application must 
either incorporate by reference or include the final safety analysis 
of the common design, including, if applicable, the final safety 
analysis report for the referenced design certification or the 
manufactured reactor.\2\
---------------------------------------------------------------------------

    \2\ As used in this appendix, the design of a nuclear power 
reactor included in a single referenced safety analysis report means 
the design of those structures, systems, and components important to 
radiological health and safety and the common defense and security.
---------------------------------------------------------------------------

    4. Each combined license application submitted pursuant to this 
appendix must contain an environmental report as required by Sec.  
52.80(b), and which complies with the applicable provisions of 10 
CFR part 51, provided, however, that the application may incorporate 
by reference a single environmental report on the environmental 
impacts of the common design.
    5. Upon a determination that each application is acceptable for 
docketing under 10 CFR 2.101, each application will be docketed and 
a notice of docketing for each application will be published in the 
Federal Register, in accordance with 10 CFR 2.104, provided, 
however, that the notice must state that the application will be 
processed under the provisions of 10 CFR part 52, appendix N, and 
subpart D of part 2 of this chapter. As the discretion of the 
Commission, a single notice of docketing for multiple applications 
may be published in the Federal Register.
    6. The NRC staff shall prepare draft and final environmental 
impact statements for each of the applications under part 51 of this 
chapter. Scoping under 10 CFR 51.28 and 51.29 for each of the 
combined license applications may be conducted simultaneously and 
joint scoping may be conducted with respect to the environmental 
issues relevant to the common design.
    If the applications reference a standard design certification, 
then the environmental impact statement for each of the applications 
must incorporate by reference the design certification environmental 
assessment. If the applications do not reference a standard design 
certification, then the NRC staff shall prepare draft and final 
supplemental environmental impact statements which address severe 
accident mitigation design alternatives for the common design, which 
must be incorporated by reference into the environmental impact 
statement prepared for each application. Scoping under 10 CFR 51.28 
and 51.29 for the supplemental environmental impact statement may be 
conducted simultaneously, and may be part of the scoping for each of 
the combined license applications.
    7. The ACRS shall report on each of the applications as required 
by Sec.  52.87. Each report must be limited to those safety matters 
for each application which are not relevant to the common design. In 
addition, the ACRS shall separately report on the safety of the 
common design, provided, however, that the report need not address 
the safety of a referenced standard design certification or reactor 
manufactured under subpart F of this part.
    8. The Commission shall designate a presiding officer to conduct 
the proceeding with respect to the health and safety, common defense 
and security, and environmental matters relating to the common 
design. The hearing will be governed by the applicable provisions of 
subparts A, C, G, L, N, and O of part 2 of this chapter relating to 
applications for combined licenses. The presiding officer shall 
issue a partial initial decision on the common design.

PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR 
POWER PLANTS

0
151. The authority citation for part 54 continues to read as follows:

    Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83 
Stat. 1244, as amended (42 U.S.C. 2132, 2133,

[[Page 49560]]

2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs 201, 202, 206, 
88 Stat. 1242, 1244 as amended (42 U.S.C. 5841, 5842).
    Section 54.17 also issued under E.O. 12829, 3 CFR, 1993 Comp., 
p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333; E.O. 
12968, 3 CFR, 1995 Comp., p. 391.


0
152. Section 54.1 is revised to read as follows:


Sec.  54.1  Purpose.

    This part governs the issuance of renewed operating licenses and 
renewed combined licenses for nuclear power plants licensed pursuant to 
Sections 103 or 104b of the Atomic Energy Act of 1954, as amended, and 
Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242)

0
153. In Sec.  54.3, paragraph (a), the definition for Current licensing 
basis is revised, and the definition for Renewed combined license is 
added to read as follows:


Sec.  54.3  Definitions.

    (a) * * *
    Current licensing basis (CLB) is the set of NRC requirements 
applicable to a specific plant and a licensee's written commitments for 
ensuring compliance with and operation within applicable NRC 
requirements and the plant-specific design basis (including all 
modifications and additions to such commitments over the life of the 
license) that are docketed and in effect. The CLB includes the NRC 
regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 
51, 52, 54, 55, 70, 72, 73, 100 and appendices thereto; orders; license 
conditions; exemptions; and technical specifications. It also includes 
the plant-specific design-basis information defined in 10 CFR 50.2 as 
documented in the most recent final safety analysis report (FSAR) as 
required by 10 CFR 50.71 and the licensee's commitments remaining in 
effect that were made in docketed licensing correspondence such as 
licensee responses to NRC bulletins, generic letters, and enforcement 
actions, as well as licensee commitments documented in NRC safety 
evaluations or licensee event reports.
* * * * *
    Renewed combined license means a combined license originally issued 
under part 52 of this chapter for which an application for renewal is 
filed in accordance with 10 CFR 52.107 and issued under this part.
* * * * *

0
154. In Sec.  54.17, paragraph (c) is revised to read as follows:


Sec.  54.17  Filing of application.

* * * * *
    (c) An application for a renewed license may not be submitted to 
the Commission earlier than 20 years before the expiration of the 
operating license or combined license currently in effect.
* * * * *

0
155. Section 54.27 is revised to read as follows:


Sec.  54.27  Hearings.

    A notice of an opportunity for a hearing will be published in the 
Federal Register in accordance with 10 CFR 2.105. In the absence of a 
request for a hearing filed within 30 days by a person whose interest 
may be affected, the Commission may issue a renewed operating license 
or renewed combined license without a hearing upon 30-day notice and 
publication in the Federal Register of its intent to do so.

0
156. In Section 54.31, paragraphs (a), (b), and (c) are revised to read 
as follows:


Sec.  54.31  Issuance of a renewed license.

    (a) A renewed license will be of the class for which the operating 
license or combined license currently in effect was issued.
    (b) A renewed license will be issued for a fixed period of time, 
which is the sum of the additional amount of time beyond the expiration 
of the operating license or combined license (not to exceed 20 years) 
that is requested in a renewal application plus the remaining number of 
years on the operating license or combined license currently in effect. 
The term of any renewed license may not exceed 40 years.
    (c) A renewed license will become effective immediately upon its 
issuance, thereby superseding the operating license or combined license 
previously in effect. If a renewed license is subsequently set aside 
upon further administrative or judicial appeal, the operating license 
or combined license previously in effect will be reinstated unless its 
term has expired and the renewal application was not filed in a timely 
manner.
* * * * *

0
157. Section 54.35 is revised to read as follows:


Sec.  54.35  Requirements during term of renewed license.

    During the term of a renewed license, licensees shall be subject to 
and shall continue to comply with all Commission regulations contained 
in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72, 
73, and 100, and the appendices to these parts that are applicable to 
holders of operating licenses or combined licenses, respectively.

0
158. In Sec.  54.37, paragraph (a) is revised to read as follows:


Sec.  54.37  Additional records and recordkeeping requirements.

    (a) The licensee shall retain in an auditable and retrievable form 
for the term of the renewed operating license or renewed combined 
license all information and documentation required by, or otherwise 
necessary to document compliance with, the provisions of this part.
* * * * *

PART 55--OPERATORS' LICENSES

0
159. The authority citation for part 55 continues to read as follows:

    Authority: Secs. 107, 161, 182, 68 Stat. 939, 948, 953, as 
amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2137, 2201, 
2232, 2282); secs. 201, as amended, 202, 88 Stat. 1242, as amended, 
1244 (42 U.S.C. 5841, 5842); sec. 1704, 112 Stat. 2750 (44 U.S.C. 
3504 note). Sections 55.41, 55.43, 55.45, and 55.59 also issued 
under sec. 306, Pub. L. 97-425, 96 Stat. 2262 (42 U.S.C. 10226). 
Section 55.61 also issued under secs. 186, 187, 68 Stat. 955 (42 
U.S.C. 2236, 2237).


0
160. In Sec.  55.1, paragraph (a) is revised to read as follows:


Sec.  55.1  Purpose.

* * * * *
    (a) Establish procedures and criteria for the issuance of licenses 
to operators and senior operators of utilization facilities licensed 
under the Atomic Energy Act of 1954, as amended, or Section 202 of the 
Energy Reorganization Act of 1974, as amended, and part 50, part 52, or 
part 54 of this chapter,
* * * * *

0
161. In Sec.  55.2, paragraph (a) is revised to read as follows:


Sec.  55.2  Scope.

* * * * *
    (a) Any individual who manipulates the controls of any utilization 
facility licensed under parts 50, 52, or 54 of this chapter,
* * * * *

0
162. In Sec.  55.5, paragraph (b)(1) and the introductory text of 
paragraph (b)(2) are revised to read as follows:


Sec.  55.5  Communications.

* * * * *
    (b)(1) Except for test and research reactor facilities, the 
Director of New Reactors or the Director of Nuclear Reactor Regulation, 
as appropriate, has delegated to the Regional Administrators of Regions 
I, II, III, and IV authority and responsibility under the regulations 
in this part for the

[[Page 49561]]

issuance and renewal of licenses for operators and senior operators of 
nuclear power reactors licensed under 10 CFR part 50 or part 52 and 
located in these regions.
    (2) Any application for a license or license renewal filed under 
the regulations in this part involving a nuclear power reactor licensed 
under 10 CFR part 50 or part 52 and any related inquiry, communication, 
information, or report must be submitted to the Regional Administrator 
by an appropriate method listed in paragraph (a) of this section. The 
Regional Administrator or the Administrator's designee will transmit to 
the Director of New Reactors or the Director of Nuclear Reactor 
Regulation, as appropriate, any matter that is not within the scope of 
the Regional Administrator's delegated authority.
* * * * *

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR 
RELATED GREATER THAN CLASS C WASTE

0
163. 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); sec. 1704, 112 Stat. 2750 (44 U.S.C. 
3504 note).
    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).


0
164. 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 10 CFR part 52.

0
165. In Sec.  72.218, paragraph (b) is revised to read as follows:


Sec.  72.218  Termination of licenses.

* * * * *
    (b) An application for termination of a reactor operating license 
issued under 10 CFR part 50 and submitted under Sec.  50.82 of this 
chapter, or a combined license issued under 10 CFR part 52 and 
submitted under Sec.  52.110 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

0
166. 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 (42 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); sec. 1704, 112 Stat. 
2750 (44 U.S.C. 3504 note).
    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).


0
167. 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 under parts 50 or 52 of this chapter,
* * * * *

0
168. In Sec.  73.2, the introductory text of paragraph (a) is revised 
to read as follows:


Sec.  73.2  Definitions.

* * * * *
    (a) Terms defined in parts 50, 52, and 70 of this chapter have the 
same meaning when used in this part.
* * * * *

0
169. In Sec.  73.50, the introductory text is revised to read as 
follows:


Sec.  73.50  Requirements for physical protection of licensed 
activities.

    Each licensee who is not subject to Sec.  73.51, but who possesses, 
uses, or stores formula quantities of strategic special nuclear 
material that are not readily separable from other radioactive material 
and which have total external radiation dose rates in excess of 100 
rems per hour at a distance of 3 feet from any accessible surfaces 
without intervening shielding other than at nuclear reactor facility 
licensed under parts 50 or 52 of this chapter, shall comply with the 
following:
* * * * *

0
170. In Sec.  73.56, paragraph (a)(3) is revised to read as follows:


Sec.  73.56  Personnel access authorization requirements for nuclear 
power plants.

    (a) * * *
    (3) Each applicant for a license to operate a nuclear power reactor 
under Sec. Sec.  50.21(b) or 50.22 of this chapter, including an 
applicant for a combined license under part 52 of this chapter, whose 
application is submitted after April 25, 1991, shall include the 
required access authorization program as part of its Physical Security 
Plan. The applicant, upon receipt of an operating license or upon 
notice of the Commission's finding under Sec.  52.103(g) of this 
chapter, shall implement the required access authorization program as 
part of its site Physical Security Plan.
* * * * *

0
171. In Sec.  73.57, paragraphs (a)(1), (a)(2), and (a)(3) are revised 
to read as follows:


Sec.  73.57  Requirements for criminal history checks of individuals 
granted unescorted access to a nuclear power facility or access to 
Safeguards Information by power reactor licensees.

    (a) * * *
    (1) Each licensee who is authorized to operate a nuclear power 
reactor under part 50 of this chapter, or each holder of a combined 
license under part 52 of this chapter upon receipt of notice of the 
Commission's finding under Sec.  52.103(g), shall comply with the 
requirements of this section.
    (2) Each applicant for a license to operate a nuclear power reactor 
under part 50 of this chapter and each applicant for a combined license 
under part 52 of this chapter shall submit fingerprints for those 
individuals who have or will have access to Safeguards Information.
    (3) Before receiving its operating license under part 50 of this 
chapter or before the Commission makes its finding under Sec.  
52.103(g) of this chapter, each applicant for a license to

[[Page 49562]]

operate a nuclear power reactor (including an applicant for a combined 
license) may submit fingerprints for those individuals who will require 
unescorted access to the nuclear power facility.
* * * * *

0
172. In Appendix C to Part 73, the Introduction is revised to read as 
follows:

Appendix C to Part 73--Licensee Safeguards Contingency Plans

Introduction

    A licensee safeguards contingency plan is a documented plan to 
give guidance to licensee personnel in order to accomplish specific 
defined objectives in the event of threats, thefts, or radiological 
sabotage relating to special nuclear material or nuclear facilities 
licensed under the Atomic Energy Act of 1954, as amended. An 
acceptable safeguards contingency plan must contain:
    (1) A predetermined set of decisions and actions to satisfy 
stated objectives;
    (2) An identification of the data, criteria, procedures, and 
mechanisms necessary to efficiently implement the decisions; and
    (3) A stipulation of the individual, group, or organizational 
entity responsible for each decision and action.
    The goals of licensee safeguards contingency plans for 
responding to threats, thefts, and radiological sabotage are:
    (1) To organize the response effort at the licensee level;
    (2) To provide predetermined, structured responses by licensees 
to safeguards contingencies;
    (3) To ensure the integration of the licensee response with the 
responses by other entities; and
    (4) To achieve a measurable performance in response capability.
    Licensee safeguards contingency planning should result in 
organizing the licensee's resources in such a way that the 
participants will be identified, their several responsibilities 
specified, and the responses coordinated. The responses should be 
timely.
    It is important to note that a licensee's safeguards contingency 
plan is intended to be complementary to any emergency plans 
developed under appendix E to part 50 of this chapter, Sec.  52.17 
or Sec.  52.79, or to Sec.  70.22(i) of this chapter.
* * * * *

PART 75--SAFEGUARDS ON NUCLEAR MATERIAL--IMPLEMENTATION OF US/IAEA 
AGREEMENT

0
173. The authority citation for part 75 continues to read as follows

    Authority: Secs. 53, 63, 103, 104, 122, 161, 68 Stat. 930, 932, 
936, 937, 939, 948, as amended (42 U.S.C. 2073, 2093, 2133, 2134, 
2152, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C. 5841); 
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
    Section 75.4 also issued under secs. 135, 141, Pub. L. 97-425, 
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).


0
174. In Sec.  75.6, paragraph (b) is revised to read as follows:


Sec.  75.6  Maintenance of records and delivery of information, 
reports, and other communications.

* * * * *
    (b) If an installation is a nuclear power plant or a non-power 
reactor for which a construction permit, operating license or a 
combined license has been issued, whether or not a license to receive 
and possess nuclear material at the installation has been issued, the 
cognizant Director is either the Director, Office of New Reactors, or 
the Director, Office of Nuclear Reactor Regulation. For all other 
installations, the cognizant Director is the Director, Office of 
Nuclear Material Safety and Safeguards.
* * * * *

PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL 
SECURITY INFORMATION AND RESTRICTED DATA

0
175. The authority citation for Part 95 continues to read as follows:

    Authority: Secs. 145, 161, 193, 68 Stat. 942, 948, as amended 
(42 U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42 
U.S.C. 5841); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); E.O. 
10865, as amended, 3 CFR 1959-1963 COMP., p. 398 (50 U.S.C. 401, 
note); E.O. 12829, 3 CFR, 1993 Comp., p. 570; E.O. 12958, as 
amended, 3 CFR, 1995 Comp., p. 333, as amended by E.O. 13292, 3 CFR, 
2004 Comp., p. 196; E.O. 12968, 3 CFR, 1995 Comp., p. 391.


0
176. In Sec.  95.5, the definition of license is revised to read as 
follows:


Sec.  95.5  Definitions.

* * * * *
    License means a license issued under 10 CFR parts 50, 52, 54, 60, 
63, 70, or 72.
* * * * *

0
177. In Sec.  95.13, paragraph (b) is revised to read as follows:


Sec.  95.13  Maintenance of records.

* * * * *
    (b) Each record required by this part must be legible throughout 
the retention period specified by each Commission regulation. The 
record may be the original or a reproduced copy or a microform provided 
that the copy or microform is authenticated by authorized personnel and 
that the microform is capable of producing a clear copy throughout the 
required retention period. The record may also be stored in electronic 
media with the capability for producing legible, accurate, and complete 
records during the required retention period. Records such as letters, 
drawings, or specifications must include all pertinent information such 
as stamps, initials, and signatures. The licensee, certificate holder, 
or other person shall maintain adequate safeguards against tampering 
with and loss of records.


0
178. In Sec.  95.19, the introductory text of paragraph (b) is revised 
to read as follows:


Sec.  95.19  Changes to security practices and procedures.

* * * * *
    (b) A licensee, certificate holder, or other person may effect a 
minor, non-substantive change to an approved Standard Practice 
Procedures Plan for the safeguarding of classified information without 
receiving prior CSA approval. These minor changes that do not affect 
the security of the facility may be submitted to the addressees noted 
in paragraph (a) of this section within 30 days of the change. Page 
changes rather than a complete rewrite of the plan may be submitted. 
Some examples of minor, non-substantive changes to the Standard 
Practice Procedures Plan include--
* * * * *

0
179. Section 95.20 is revised to read as follows:


Sec.  95.20  Grant, denial or termination of facility clearance.

    The Division of Nuclear Security shall provide notification in 
writing (or orally with written confirmation) to the licensee, 
certificate holder, or other person of the Commission's grant, 
acceptance of another agency's facility clearance, denial, or 
termination of facility clearance. This information must also be 
furnished to representatives of the NRC, NRC contractors, licensees, 
certificate holders, or other person, or other Federal agencies having 
a need to transmit classified information to the licensees or other 
person.


0
180. In Sec.  95.23, paragraph (b) is revised to read as follows:


Sec.  95.23  Termination of facility clearance.

* * * * *
    (b) When facility clearance is terminated, the licensee, 
certificate holder, or other person will be notified in writing of the 
determination and the procedures outlined in Sec.  95.53 apply.


0
181. Section 95.31 is revised to read as follows:

[[Page 49563]]

Sec.  95.31  Protective personnel.

    Whenever protective personnel are used to protect classified 
information they shall:
    (a) Possess an ``L'' access authorization (or CSA equivalent) if 
the licensee, certificate holder, or other person possesses information 
classified Confidential National Security Information, Confidential 
Restricted Data or Secret National Security Information.
    (b) Possess a ``Q'' access authorization (or CSA equivalent) if the 
licensee, certificate holder, or other person possesses Secret 
Restricted Data related to nuclear weapons design, manufacturing and 
vulnerability information; and certain particularly sensitive Naval 
Nuclear Propulsion Program information (e.g., fuel manufacturing 
technology) and the protective personnel require access as part of 
their regular duties.


0
182. In Sec.  95.33, paragraph (c) is revised to read as follows:


Sec.  95.33  Security education.

* * * * *
    (c) Temporary Help Suppliers. A temporary help supplier, or other 
contractor who employs cleared individuals solely for dispatch 
elsewhere, is responsible for ensuring that required briefings are 
provided to their cleared personnel. The temporary help supplier or the 
using licensee's, certificate holder's, or other person's facility may 
conduct these briefings.
* * * * *

0
183. Section 95.34 is revised to read as follows:


Sec.  95.34  Control of visitors.

    (a) Uncleared visitors. Licensees, certificate holders, or other 
persons subject to this part shall take measures to preclude access to 
classified information by uncleared visitors.
    (b) Foreign visitors. Licensees, certificate holders, or other 
persons subject to this part shall take measures as may be necessary to 
preclude access to classified information by foreign visitors. The 
licensee, certificate holder, or other person shall retain records of 
visits for 5 years beyond the date of the visit.


0
184. In Sec.  95.35, the introductory text of paragraph (a), and 
paragraph (a)(3) are revised to read as follows:


Sec.  95.35  Access to matter classified as National Security 
Information and Restricted Data.

    (a) Except as the Commission may authorize, no licensee, 
certificate holder or other person subject to the regulations in this 
part may receive or may permit any other licensee, certificate holder, 
or other person to have access to matter revealing Secret or 
Confidential National Security Information or Restricted Data unless 
the individual has:
* * * * *
    (3) NRC-approved storage facilities if classified documents or 
material are to be transmitted to the licensee, certificate holder, or 
other person.
* * * * *

0
185. In Sec.  95.36, paragraphs (c), (d), and (e) are revised to read 
as follows:


Sec.  95.36  Access by representatives of the International Atomic 
Energy Agency or by participants in other international agreements.

* * * * *
    (c) In accordance with the specific disclosure authorization 
provided by the Division of Nuclear Security, licensees, certificate 
holders, or other persons subject to this part are authorized to 
release (i.e., transfer possession of) copies of documents that contain 
classified National Security Information directly to IAEA inspectors 
and other representatives officially designated to request and receive 
classified National Security Information documents. These documents 
must be marked specifically for release to IAEA or other international 
organizations in accordance with instructions contained in the NRC's 
disclosure authorization letter. Licensees, certificate holders, and 
other persons subject to this part may also forward these documents 
through the NRC to the international organization's headquarters in 
accordance with the NRC disclosure authorization. Licensees, 
certificate holders, and other persons may not reproduce documents 
containing classified National Security Information except as provided 
in Sec.  95.43.
    (d) Records regarding these visits and inspections must be 
maintained for 5 years beyond the date of the visit or inspection. 
These records must specifically identify each document released to an 
authorized representative and indicate the date of the release. These 
records must also identify (in such detail as the Division of Nuclear 
Security, by letter, may require) the categories of documents that the 
authorized representative has had access and the date of this access. A 
licensee, certificate holder, or other person subject to this part 
shall also retain Division of Nuclear Security disclosure 
authorizations for 5 years beyond the date of any visit or inspection 
when access to classified information was permitted.
    (e) Licensees, certificate holders, or other persons subject to 
this part shall take such measures as may be necessary to preclude 
access to classified matter by participants of other international 
agreements unless specifically provided for under the terms of a 
specific agreement.


0
186. In Sec.  95.37, paragraphs (a), (b), and (h) are revised to read 
as follows:


Sec.  95.37  Classification and preparation of documents.

    (a) Classification. Classified information generated or possessed 
by a licensee, certificate holder, or other person must be 
appropriately marked. Classified material which is not conducive to 
markings (e.g., equipment) may be exempt from this requirement. These 
exemptions are subject to the approval of the CSA on a case-by-case 
basis. If a person or facility generates or possesses information that 
is believed to be classified based on guidance provided by the NRC or 
by derivation from classified documents, but which no authorized 
classifier has determined to be classified, the information must be 
protected and marked with the appropriate classification markings 
pending review and signature of an NRC authorized classifier. This 
information shall be protected as classified information pending final 
determination.
    (b) Classification consistent with content. Each document 
containing classified information shall be classified Secret or 
Confidential according to its content. NRC licensees, certificate 
holders, or other persons subject to the requirements of 10 CFR part 95 
may not make original classification decisions.
* * * * *
    (h) Classification challenges. Licensees, certificate holders, or 
other persons in authorized possession of classified National Security 
Information who in good faith believe that the information's 
classification status (i.e., that the document), is classified at 
either too high a level for its content (overclassification) or too low 
for its content (underclassification) are expected to challenge its 
classification status. Licensees, certificate holders, or other persons 
who wish to challenge a classification status shall--
    (1) Refer the document or information to the originator or to an 
authorized NRC classifier for review. The authorized classifier shall 
review the document and render a written classification decision to the 
holder of the information.
    (2) In the event of a question regarding classification review, the

[[Page 49564]]

holder of the information or the authorized classifier shall consult 
the NRC Division of Facilities and Security, Information Security 
Branch, for assistance.
    (3) Licensees, certificate holders, or other persons who challenge 
classification decisions have the right to appeal the classification 
decision to the Interagency Security Classification Appeals Panel.
    (4) Licensees, certificate holders, or other persons seeking to 
challenge the classification of information will not be the subject of 
retribution.
* * * * *

0
187. In Sec.  95.39, paragraph (a) is revised to read as follows:


Sec.  95.39  External transmission of documents and material.

    (a) Restrictions. Documents and material containing classified 
information received or originated in connection with an NRC license, 
certificate, or standard design approval or standard design 
certification under part 52 of this chapter must be transmitted only to 
CSA approved security facilities.
* * * * *

0
188. In Sec.  95.43, paragraph (a) is revised to read as follows:


Sec.  95.43  Authority to reproduce.

    (a) Each licensee, certificate holder, or other person possessing 
classified information shall establish a reproduction control system to 
ensure that reproduction of classified material is held to the minimum 
consistent with operational requirements. Classified reproduction must 
be accomplished by authorized employees knowledgeable of the procedures 
for classified reproduction. The use of technology that prevents, 
discourages, or detects the unauthorized reproduction of classified 
documents is encouraged.
* * * * *

0
189. In Sec.  95.45, paragraph (d) is revised to read as follows:


Sec.  95.45  Changes in classification.

* * * * *
    (d) Any licensee, certificate holder, or other person making a 
change in classification or receiving notice of such a change shall 
forward notice of the change in classification to holders of all copies 
as shown on their records.

0
190. Section 95.49 is revised to read as follows:


Sec.  95.49  Security of automatic data processing (ADP) systems.

    Classified data or information may not be processed or produced on 
an ADP system unless the system and procedures to protect the 
classified data or information have been approved by the CSA. Approval 
of the ADP system and procedures is based on a satisfactory ADP 
security proposal submitted as part of the licensee's, certificate 
holder's, or other person's request for facility clearance outlined in 
Sec.  95.15 or submitted as an amendment to its existing Standard 
Practice Procedures Plan for the protection of classified information.

0
191. Section 95.51 is revised to read as follows:


Sec.  95.51  Retrieval of classified matter following suspension or 
revocation of access authorization.

    In any case where the access authorization of an individual is 
suspended or revoked in accordance with the procedures set forth in 
part 25 of this chapter, or other relevant CSA procedures, the 
licensee, certificate holder, or other person shall, upon due notice 
from the Commission of such suspension or revocation, retrieve all 
classified information possessed by the individual and take the action 
necessary to preclude that individual having further access to the 
information.

0
192. Section 95.53 is revised to read as follows:


Sec.  95.53  Termination of facility clearance.

    (a) If the need to use, process, store, reproduce, transmit, 
transport, or handle classified matter no longer exists, the facility 
clearance will be terminated. The licensee, certificate holder, or 
other person for the facility may deliver all documents and matter 
containing classified information to the Commission, or to a person 
authorized to receive them, or must destroy all classified documents 
and matter. In either case, the licensee, certificate holder, or other 
person for the facility shall submit a certification of nonpossession 
of classified information to the NRC Division of Nuclear Security 
within 30 days of the termination of the facility clearance.
    (b) In any instance where a facility clearance has been terminated 
based on a determination of the CSA that further possession of 
classified matter by the facility would not be in the interest of the 
national security, the licensee, certificate holder, or other person 
for the facility shall, upon notice from the CSA, dispose of classified 
documents in a manner specified by the CSA.

0
193. In Sec.  95.57, the introductory paragraph is revised to read as 
follows:


Sec.  95.57  Reports.

    Each licensee, certificate holder, or other person having a 
facility clearance shall report to the CSA and the Regional 
Administrator of the appropriate NRC Regional Office listed in 10 CFR 
part 73, appendix A:
* * * * *

0
194. Section 95.59 is revised to read as follows:


Sec.  95.59  Inspections.

    The Commission shall make inspections and reviews of the premises, 
activities, records and procedures of any licensee, certificate holder, 
or other person subject to the regulations in this part as the 
Commission and CSA deem necessary to effect the purposes of the Act, 
E.O. 12958 and/or NRC rules.

PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY 
AGREEMENTS

0
195. 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. 841, 5842); Sec. 1704, 112 Stat. 
2750 (44 U.S.C. 3504 note).


0
196. In Sec.  140.2, paragraphs (a)(1) and (a)(2) are 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 under 10 CFR parts 50, 52, or 54 to operate a nuclear reactor, 
and
    (2) With respect to an extraordinary nuclear occurrence, to each 
person who is an applicant for or holder of a license to operate a 
production facility or a utilization facility (including an operating 
license issued under part 50 of this chapter and a combined license 
under part 52 of this chapter), and to other persons indemnified with 
respect to the involved facilities.
* * * * *

0
197. Section 140.10 is revised to read as follows:


Sec.  140.10  Scope.

    This subpart applies to each person who is an applicant for or 
holder of a license issued under 10 CFR parts 50 or 54 to operate a 
nuclear reactor, or is the applicant for or holder of a combined 
license issued under parts 52 or 54 of this chapter, except licenses 
held by persons found by the Commission to be Federal agencies or 
nonprofit educational institutions licensed to conduct educational 
activities. This subpart also applies to persons licensed

[[Page 49565]]

to possess and use plutonium in a plutonium processing and fuel 
fabrication plant.

0
198. In Sec.  140.11, paragraph (b) is revised to read as follows:


Sec.  140.11  Amounts of financial protection for certain reactors.

* * * * *
    (b) In any case where a person is authorized under parts 50, 52, or 
54 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.

0
199. In Sec.  140.12, paragraph (c) is revised to read as follows:


Sec.  140.12  Amount of financial protection required for other 
reactors.

* * * * *
    (c) In any case where a person is authorized under parts 50, 52, or 
54 of this chapter to operate two or more nuclear reactors at the same 
location, the total 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 financial 
protection covers all reactors at the location.
* * * * *

0
200. 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 under 10 CFR part 
52.

    Each holder of a part 50 construction permit, or a holder of a 
combined license under part 52 of this chapter before the date that the 
Commission had made the finding under 10 CFR 52.103(g), who also holds 
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 either an operating license under 10 CFR part 50 or 
combined license under 10 CFR part 52, shall, during the period before 
issuance of a license authorizing operation under 10 CFR part 50, or 
the period before the Commission makes the finding under Sec.  
52.103(g) of this chapter, as applicable, 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 of this chapter before issuance of the license under part 70 of 
this chapter.

0
201. In Sec.  140.20, paragraph (a)(1)(ii) is revised, and paragraph 
(a)(1)(iii) is added to read as follows:


Sec.  140.20  Indemnity agreements and liens.

    (a) * * *
    (1) * * *
    (ii) The date that the Commission makes the finding under Sec.  
52.103(g) of this chapter; or
    (iii) The effective date of the license (issued under part 70 of 
this chapter) authorizing the licensee to possess and store 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 
for the reactor, whichever is earlier. No such agreement, however, 
shall be effective prior to September 26, 1957; or
* * * * *

0
202. In Sec.  140.81, paragraph (a) is revised to read as follows:


Sec.  140.81  Scope and purpose.

    (a) Scope. This subpart applies to applicants for and holders of 
licenses authorizing operation of production facilities and utilization 
facilities, including combined licenses under part 52 of this chapter, 
and to other persons indemnified with respect to such facilities.
* * * * *

0
203. In Sec.  140.93 Appendix C, Article VIII, paragraph 4 is revised 
to read as follows:


Sec.  140.93  Appendix C--Form of indemnity agreement with licensees 
furnishing proof of financial protection in the form of licensee's 
resources.

* * * * *

Article VIII

* * * * *
    4. If the Commission determines that the licensee is financially 
able to reimburse the Commission for a deferred premium payment made 
in its behalf, and the licensee, after notice of such determination 
by the Commission fails to make such reimbursement within 120 days, 
the Commission will take appropriate steps to suspend the license 
for 30 days. The Commission may take any further action as necessary 
if reimbursement is not made within the 30-day suspension period 
including, but not limited to, termination of the operating license 
or combined license.
* * * * *

0
204. Section 140.96 is revised to read as follows:


Sec.  140.96  Appendix F--Indemnity locations.

    (a) Geographical boundaries of indemnity locations.
    (1) In every indemnity agreement between the Commission and a 
licensee which affords indemnity protection for the preoperational 
storage of fuel at the site of a nuclear power reactor under 
construction, the geographical boundaries of the indemnity location 
will include the entire construction area of the nuclear power reactor, 
as determined by the Commission. Such area will not necessarily be 
coextensive with the indemnity location which will be established at 
the time an operating license or combined license under 10 CFR part 52 
is issued for such additional nuclear power reactors.
    (2) In every indemnity agreement between the Commission and a 
licensee which affords indemnity protection for an existing nuclear 
power reactor, the geographical boundaries of the indemnity location 
shall include the entire construction area of any additional nuclear 
power reactor as determined by the Commission, built as part of the 
same power station by the same licensee. Such area will not necessarily 
be coextensive with the indemnity location which will be established at 
the time an operating license or combined license is issued for such 
additional nuclear power reactors.
    (3) This section is effective May 1, 1973, as to construction 
permits issued before March 2, 1973, and, as to construction permits 
and combined licenses issued on or after March 2, 1973, the provisions 
of this section will apply no later than such time as a construction 
permit or combined license is issued authorizing construction of any 
additional nuclear power reactor.

PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT 
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT 
OF 1954, AS AMENDED

0
205. 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); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).


0
206. In Sec.  170.2, paragraph (j) is removed and reserved, and 
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 early site permit, standard design

[[Page 49566]]

certification, standard design approval, manufacturing license, or 
combined license issued under 10 CFR part 52;
* * * * *
    (j) [Reserved]
    (k) Applying for or already has applied for review, under appendix 
Q to 10 CFR part 50 of a facility site before the submission of an 
application for a construction permit;
* * * * *

PART 171--ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES 
AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF 
COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS 
AND GOVERNMENT AGENCIES LICENSED BY THE NRC

0
207. The authority citation for part 171 continues to read as follows:

    Authority: Sec. 7601, Pub. L. 99-272, 100 Stat. 146, as amended 
by sec. 5601, Pub. L. 100-203, 101 Stat. 1330 as amended by sec. 
3201, Pub. L. 101-239, 103 Stat. 2132, as amended by sec. 6101, Pub. 
L. 101-508, 104 Stat. 1388, as amended by sec. 2903a, Pub. L. 102-
486, 106 Stat. 3125 (42 U.S.C. 2213, 2214); 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. 1704, 112 Stat. 2750 
(44 U.S.C. 3504 note).

0
208. In Sec.  171.15, paragraph (a) is revised to read as follows:


Sec.  171.15  Annual Fees: Reactor licenses and independent spent fuel 
storage licenses.

    (a) Each person holding an operating license for a power, test, or 
research reactor; each person holding a combined license under part 52 
of this chapter after the Commission has made the finding under Sec.  
52.103(g); each person holding a part 50 or part 52 power reactor 
license that is in decommissioning or possession only status, except 
those that have no spent fuel onsite; and each person holding a part 72 
license who does not hold a part 50 or part 52 license shall pay the 
annual fee for each license held at any time during the Federal fiscal 
year in which the fee is due. This paragraph does not apply to test and 
research reactors exempted under Sec.  171.11(a).
* * * * *

    Dated at Rockville, Maryland, this 1st day of August 2007.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 07-3861 Filed 8-20-07; 8:45 am]
BILLING CODE 7590-01-P


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