49352 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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. 

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 

nvg@nrc.gov. 

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 


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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 § 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, § 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, 

§ 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, § 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 

§ 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 

§ 50.4), employee protection (analogous 

to § 50.7), completeness and accuracy of 

information (analogous to § 50.9), 

exemptions (analogous to § 50.12), 

combining licenses (analogous to 


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49354 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

1The 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 § 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 

§ 52.0 through 52.11. 

§ 50.52), jurisdictional limits (analogous 

to § 50.53), and attacks and destructive 

acts (analogous to § 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, § 50.13 (attacks 

and destructive acts); § 50.32 

(elimination of repetition); § 50.52 

(combining licenses)), unless the 

substantive provisions of part 52, 

beginning with § 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. 

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., 

§ 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 crossreference 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 (§§ 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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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 49355 

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, §§ 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 § 50.47 and the 

requirements of appendix E of 10 CFR 

part 50. The requirements in § 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 § 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 § 50.47(b), and (2) describe the 

emergency planning zones as required in 

§§ 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 § 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 § 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 § 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 


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49356 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 

§ 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 § 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 § 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 

§ 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 nearterm 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 


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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 

§ 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, § 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, § 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 § 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 

§ 52.103(a) and NRC staff scheduling of 

inspection and audit activities to 

support NRC staff determinations of the 

successful completion of ITAAC under 

§ 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 

§ 52.103(e), and any Commission 

determinations on petitions as provided 

under § 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 § 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 § 52.103(a), 

completion of hearings within the time 

frame allotted under § 52.103(e), and 

completion of any Commission 

determinations on petitions filed under 

§ 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 

§ 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 


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2The 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 

§ 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. 

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 § 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 § 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 § 52.79. If additional required 

operational programs are identified, the 

Commission is considering adding them 

to § 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 

§ 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 § 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 

[§ 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 [§ 52.79(a)(36)(ii)]; (3) a 

description of the radiation protection 

program required by § 20.1101 

[§ 52.79(a)(39)]; (4) a description of the 

fire protection program required by 

§ 50.48 [§ 52.79(a)(40)]; and (5) a 

description of the fitness-for-duty 

program required by 10 CFR part 26 

[§ 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 

§ 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 § 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 

§ 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., § 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 § 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 § 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 reproposed 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 


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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. 

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 § 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 §§ 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 § 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 § 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 

§ 52.39 (and possible conforming 

changes in § 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 


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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 § 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 § 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, § 52.17 and/or § 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 

§ 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 ESPapproved 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 

§ 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 

§ 52.39(c)(2), which would be processed 

in accordance with § 2.206. This 

approach would be most consistent with 

the existing provisions in § 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 § 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 


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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 § 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 § 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 

§ 50.90 for review and approval, but 

rather should be considered to be 

information equivalent to that provided 

under § 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 §§ 50.90 and 50.92 to 

include ESPs within the scope of these 

requirements. The NRC is also adding a 

new provision to § 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-bycase basis. 

Question 10: The Commission is 

considering adopting in the final part 52 

rulemaking a new provision in § 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 § 50.59(c)(2) or, in the case of a 

license that references a certified design, 

in accordance with § 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 plantspecific 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 


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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 plantspecific 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 NEIrecommended 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 § 52.63(a)(1); 

in particular, the Commission believes 

that the proposed changes would satisfy 

the ‘‘reduces unnecessary regulatory 

burden’’ criterion in proposed 

§ 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 § 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 § 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 plantspecific 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 plantspecific 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 plantspecific 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 plantspecific 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 § 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 


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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 plantspecific 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 

§ 2.335) in order to require plantspecific 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 § 2.335, are 

present in order to require a plantspecific 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 

§ 52.99(e) and the Commission’s ITAAC 

finding under proposed § 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 


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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 § 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 § 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 

§§ 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 

§ 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 § 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 ‘‘§ 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 


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sufficient because of the manner in 

which these existing DCDs were 

submitted to the NRC. Therefore, 

Section IV.A.2 of the DCRs overrides 

§§ 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 § 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 plantspecific 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 § 50.109 in addition to the special 

circumstances in § 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 plantspecific departures should not have to 

meet the backfit requirements in 

§ 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 § 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 § 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 


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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 

§ 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 § 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 § 52.103(g) 

finding promptly following the 

completion of the last ITAAC. 

NRC Response: The NRC has decided 

to amend § 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 § 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 § 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 § 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 

§ 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 

§ 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 § 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 


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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 § 52.103(f), a contention which alleges 

the deficiency of the ITAAC is not admissible under 

§ 52.103(b). 

the public the notifications to be 

submitted under § 52.99(c)(1) and (c)(2), 

no later than the Federal Register notice 

of intended operation and opportunity 

for hearing on ITAAC under § 52.103(a). 

A conforming change is included in 

§ 2.105(b)(3) to require that the 

§ 52.103(a) notice reference the public 

availability of the § 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 § 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 § 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 § 52.103(b)(2).3

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 § 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 §§ 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 

§ 2.104(f), revises § 2.105 (governing the 

content of a Federal Register notice of 

proposed action where a mandatory 

hearing is not held under § 2.104) to 

add, as appropriate, references to 

issuance of manufacturing licenses, and 

revised § 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 


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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 

[§ 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 § 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 

§ 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 plantspecific 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 

§ 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 

§ 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., § 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 § 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 

§ 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 


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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 safetyrelated 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 § 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 § 206 of the ERA, and it was 

inconsistent with the Act to expand the 

scope of § 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 

§ 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 § 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 

§ 52.11), correct typographic errors, 

adopt consistent terminology (e.g., 

‘‘makes the finding under § 52.103(g)’’), 

correct grammar, and adopt plain 

English. These changes are not 

discussed further. 

B. Testing Requirements for Advanced 

Reactors 

This rule amends §§ 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 

§ 52.47(c)(2). The codification of testing 

requirements in the original § 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 lightwater 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 


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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-ofperformance 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-ofperformance 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 § 50.43(e)(2) for 

licensing a prototype plant, as defined 

in §§ 50.2 and 52.1, if the plant is used 

to meet the testing requirements in 

§ 50.43(e)(1). The new § 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 §§ 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 

§§ 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 § 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 

§ 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 § 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 § 52.1. The NRC is making 

similar changes to §§ 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 


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4As discussed in the section-by-section 

discussion for § 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 § 52.7 

even though the departure itself may not involve an 

exemption. 

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 (§§ 52.97 and 

52.167) 

Current § 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 § 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: § 50.50 could be revised to 

refer to combined licenses and 

manufacturing licenses, or provisions 

analogous to § 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 § 52.24. Accordingly, the 

NRC is adding the language in § 52.97(c) 

for combined licenses, and § 52.167(b) 

for manufacturing licenses, which are 

analogous to § 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 (§§ 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 

§ 2.102(a), which provides the NRC staff 

with overall authority to request 

information to support their review of 

an application. Accordingly, 

§§ 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 

§ 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, § 52.63(a) 

requires that the change (‘‘modification’’

under § 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: 

• Concerns certified design 

information or manufacturing license 

information. 

• Is requested by the applicant/ 

licensee referencing a design 

certification or the use of a 

manufactured reactor. 

• 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. 

• 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 § 52.7 

(which references the same criteria for 

the granting of exemptions that are set 

forth in § 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 plantspecific. 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. 


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5This 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. 

d. Exemption 

An exemption is a Commissiongranted 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 § 52.7. 

Exemption from an underlying technical 

requirement in part 50 would be 

provided under § 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 § 52.1, Scope, as § 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 § 52.1 as § 52.0. 

Paragraph (a) of § 52.0 is derived from 

the text of former § 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. 

Paragraph (b) of § 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 § 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 

§ 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 § 52.5 would provide greater 

clarity. On the contrary, some 

commenters indicated that § 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 § 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 § 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 

§ 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, § 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 § 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 § 52.103(g). That 

provision is added to part 52 to facilitate 

the licensing of nuclear plants, such as 

the Modular High Temperature GasCooled Reactor (MHTGR) and Power 

Reactor Innovative Small Module 


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(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 § 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 

§§ 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 § 50.47(b), or (2) describe 

the emergency planning zones as 

required in § 50.33(g). The goal of the 

‘‘major features’’ option in § 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 § 52.39(b). 

A definition of prototype plant is 

added to explain the type of nuclear 

power plant that the NRC is addressing 

in §§ 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-akind 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 

§ 52.4, Deliberate Misconduct 

The former section on interpretations 

in § 52.5 is retained and redesignated 

without change as § 52.2. The former 

section on deliberate misconduct in 

§ 52.9 is retained and redesignated 

without change as § 52.4. 

d. Section 52.3, Written 

Communications; § 52.5, Employee 

Protection; § 52.6, Completeness and 

Accuracy of Information; § 52.7, 

Specific Exemptions; § 52.8, Combining 

Licenses; § 52.9, Jurisdictional Limits; 

and § 52.10, Attacks and Destructive 

Acts 

Section 52.3, Written 

communications, which is essentially 

identical with the current § 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 § 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 § 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 § 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 § 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 (§ 50.12) for 

obtaining specific exemptions, § 50.12 

by its terms applies only to exemptions 

from part 50. Although it would be 

possible to revise § 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 § 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 § 52.8, which is 

analogous to § 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, § 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 

§ 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 § 50.31 only after the 

publication of the March 2006 proposed 

rule. 

Paragraph (b) of § 52.8, which is 

analogous to § 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, 


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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 

§ 50.32 only after the publication of the 

March 2006 proposed rule. 

Paragraph (c) of § 52.8, which is 

analogous to § 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 § 50.31, the 

language in § 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 § 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 

§ 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 § 50.13. Although it would 

be possible to revise § 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 § 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 § 52.10 to 

part 52, which is essentially identical 

with § 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 § 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 §§ 52.17(b), 

52.18, and 52.39 to address changes to 

emergency preparedness requirements 

for early site permit applicants. The 

NRC is amending § 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 

§ 52.18 to require that the Commission 

determine whether the information 

required of the applicant by 

§ 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 

§§ 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 §§ 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 

§ 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 § 52.17(b)(2). 

In the final rule, the NRC is making the 

additional conforming changes in 

§§ 52.17(b)(1) and (b)(4). 

The NRC is adding new § 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 


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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 

§ 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 § 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 

§ 52.17, Contents of Applications; 

Technical Information 

The NRC is adding § 52.16 to include 

the general content requirements from 

§ 52.17(a)(1). 

The title of § 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 § 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 § 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, 

§ 52.17(a)(1) was revised to address this 

situation. 

The NRC is amending § 52.17(a)(1) to 

eliminate all references to § 50.34. The 

references to § 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 § 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 § 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 § 50.34(a)(1) is 

being removed and the requirements are 

included in § 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 § 52.17(a)(1). A 

requirement is added to § 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 § 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 § 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 

§ 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 

§ 52.17(a)(1) that is taken from 

§ 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 


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applications and that use of this 

guidance will facilitate the NRC’s 

review. 

The NRC is making a change to 

§ 52.17(a)(1) based on several comments 

on the proposed rule. The NRC is 

deleting the requirement in proposed 

§ 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 § 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 

§ 52.79(a)(31). 

The NRC is moving the environmental 

provisions in former § 52.17(a)(2) to 

§ 51.50(b). Revised § 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 § 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 § 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 § 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 § 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 § 52.U0 

(c), those additional activities will have 

to be requested separately under 

§ 50.10(e)(1). Some minor changes were 

made to the rule language in § 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 § 50.10(e) 

address both safety and environmental 

issues. Additional changes were made 

to §§ 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 § 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 § 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 

§ 52.39(a)(2), and in particular 

§ 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 § 52.17(c) 

where the applicant must specify the 

activities that it is requesting 

authorization to perform at the site 

under § 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 § 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 § 52.27(b)(1) [final 

§ 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 


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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 

§ 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 § 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 

§ 52.27 based on public comments. The 

NRC is deleting proposed § 52.27(b)(2) 

because it was inconsistent with 

proposed § 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 § 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 § 52.39. After issuance of the 

construction permit or combined 

license, § 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 

§ 52.98 (e.g., in accordance with 

§§ 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 § 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 § 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 § 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 crossreference to 10 CFR part 21 and 

§ 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 § 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 § 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 


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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 

§ 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 § 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 § 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 § 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 

§ 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 § 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 § 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 § 52.39 and corresponding 

provisions in the revisions to § 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 § 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 § 2.206. 

In addition, under §§ 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 

§ 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 § 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 § 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 § 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 § 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 § 50.54(q) should be 

treated as an enforcement matter. Note 

that the NRC is making some 

adjustments to this position in the final 


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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 § 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 § 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 § 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 § 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 §§ 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 

§§ 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 

§ 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 §§ 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 

§ 52.39(c)(2) as challenges to the permit 

and processed in accordance with 

§ 2.206. The NRC is retaining the former 

provision in § 52.39(a)(2)(iii) requiring 

that the Commission consider a petition 

filed under § 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 § 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 § 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 § 52.39(b), is 

addressed in § 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 

§ 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 § 50.54(f). For early site 

permits, this provision is contained in 

§ 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 

§ 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 

§§ 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 

§§ 50.4 and 50.30 were replaced with 

references to subpart H of part 2. A new 


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§ 52.45(c) on design certification review 

fees, was moved from § 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 §§ 52.47(a) 

and 52.47(c), and the technical 

requirements for the remainder of the 

design certification application are in 

§ 52.47(b). The former § 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., 

§ 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 

§ 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 § 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 § 50.49(d) [now 

§ 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 § 50.49(d). Therefore, the 

NRC revised the wording in 

§ 52.47(a)(13) to be consistent with the 

wording for the same provision in 

§ 52.79(a), which requires that 

applicants provide the list of electrical 

equipment important to safety required 

by § 50.49(d). 

Some commenters recommended that 

the requirement in § 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 § 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 § 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 § 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 § 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 § 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 

§ 52.47(b)(1)] was conformed with the 

statutory language in the AEA. This 

clarification of the language in the 

former § 52.47(a)(1)(vi), which was a 

condensed version of the language in 

the former § 52.97(b)(1), was intended to 

avoid any misunderstandings regarding 

the statutory requirement. Some 

commenters recommended that the rule 

language in § 52.47(b)(1) be modified to 

maintain the language in the former 

§ 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 

§ 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 


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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 § 52.47(b) was reorganized by 

separating the requirements on scope of 

design and modular configuration [now 

located in § 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 § 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 §§ 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 § 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 § 52.47(a)(19), which 

is consistent with the requirement for 

licensees. 

A new § 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 § 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 § 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 § 52.63(a) to provide 

processes for amending design 

certification information without 

meeting the special backfit requirement 

in § 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, 

§ 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 § 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 § 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 § 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 § 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 § 52.63(a)(1)(iv) 

that allows for generic resolutions of 

DAC without meeting the special backfit 

requirement. These amendments will 


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apply to all plants that have or will 

reference the DCR under § 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 

§ 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 

§ 52.63(a)(2), which sets forth 

procedures for rulemakings conducted 

under § 52.63(a)(1). Paragraph (a)(2)(i) 

requires that for rulemakings under 

§ 52.63(a)(1), except for rulemakings 

under § 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 

§ 52.63(a)(2) [now § 52.63(a)(3)] to delete 

the reference to the former § 52.63(a)(4) 

[now § 52.63(a)(5)]. The reference to the 

former § 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 § 52.63(a)(3) deals with 

modifications that the NRC may impose 

on a DCR under §§ 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 § 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 

§ 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 §§ 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 §§ 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 § 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 § 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 


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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 § 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 §§ 52.39 and 52.79 

should contain an updating requirement 

to be imposed upon the applicant 

referencing an early site permit. 

A new § 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 § 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 

§ 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 § 52.39(a)(1). New information that 

materially changes the NRC’s 

determination of the matters in 

§ 52.17(b), or results in modifications of 

existing terms and conditions under 

§ 52.39(a)(1) will be subject to litigation 

during the construction permit, 

operating license, or combined license 

proceedings in accordance with 

§ 52.39(c). 

Not all new information on 

emergency preparedness will be subject 

to challenge in a hearing under 

§ 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 

§ 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 

§ 52.39(a)(1), as well as the basis for a 

hearing contention under § 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 § 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 

§ 52.99(c)(1) in the final rule to more 

closely follow the language of Section 

185b. of the AEA (in response to a latefiled 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 


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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 § 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 § 52.99(c)(1) and (c)(2), 

no later than the Federal Register notice 

of intended operation and opportunity 

for hearing on ITAAC under § 52.103(a). 

A conforming change is included in 

§ 2.105(b)(3) to require that the 

§ 52.103(a) notice reference the public 

availability of the § 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 § 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 § 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 § 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 § 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 § 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 

§ 52.99(d) in the final rule as a result of 

comments made on the proposed rule. 

Proposed § 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 § 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., 

§ 52.39(d)). Therefore, the NRC has 

adopted the commenters’ suggestion to 

delete references to ESP ITAAC and ESP 

variances from § 52.99(d). 


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Paragraph (e)(1) requires the NRC to 

publish, at appropriate intervals until 

the last date for submission of requests 

for hearing under § 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 § 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, § 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 § 52.97 (proposed 

§ 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 § 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 § 52.99 in the final 

rule to delete the requirements in 

proposed § 52.99(a). Proposed § 52.99(a) 

required holders of COLs to comply 

with the provisions of §§ 50.70 and 

50.71. Because the language in proposed 

§§ 50.70 and 50.71 requires COL holders 

to comply with their provisions, and 

because of the applicability provisions 

in § 52.0(b), this duplicate requirement 

in § 52.99 is unnecessary. 

The NRC has added a new paragraph 

(a) in § 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 § 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 § 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 § 52.73(b) the 

requirement in the current § 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 

§ 52.63(c) be deleted and the remaining 

provision be added to the former § 52.79 

rather than the former § 52.73. The NRC 

agrees with NEI that 10 CFR part 52 

should be modified to clarify that the 

requirement in former § 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 

§ 52.73(b), and not to § 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 §§ 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 §§ 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 § 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 

§ 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 § 52.80 covers requirements 

for the remainder of the technical 

content of a combined license 

application. 

Former § 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 § 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., §§ 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, 


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the NRC has attempted to capture all 

relevant requirements regarding 

contents of the FSAR for a combined 

license application. 

In addition, § 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 § 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 § 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 

[§ 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 [§ 52.79(a)(36)(ii)]; (3) a 

description of the radiation protection 

program required by § 20.1101 

[§ 52.79(a)(39)]; (4) a description of the 

fire protection program required by 

§ 50.48 [§ 52.79(a)(40)]; and (5) a 

description of the fitness-for-duty 

program required by 10 CFR part 26 

[§ 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 § 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 § 52.79(a) in the final rule 

to address the application requirements 

in current § 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 § 52.79 

contains a complete list of the 

requirements for the contents of a COL 

application, the NRC added paragraph 

(a)(45) to § 52.79 to require COL 

applications to include the information 

required by § 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 § 50.34 

which requires that emergency plans be 

included in the FSAR for operating 

license applications. 

The NRC is adding a new provision in 

§ 52.79(a)(29)(ii) that the applicant 

submit plans for coping with 

emergencies, other than the plans 

required by § 52.79(a)(21). Paragraph 

52.79(a)(21) requires the applicant to 

submit emergency plans complying 

with the requirements of § 50.47 and 10 

CFR part 50, appendix E. This 

requirement was drawn from the 

existing requirement in § 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 § 50.34(b)(6)(v), namely, the 

requirement in proposed § 52.79(a)(21) 

to submit emergency plans. The NRC 

has corrected this omission in the final 

rule by including the new provision in 

§ 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 § 52.79(a)(23) 

that addressed a request to conduct 

activities under § 50.10(e) and added 

them in a new § 52.80(c). The NRC 

concluded that it is preferable to 

include both the list of proposed 

§ 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 § 50.10(e) address both 

safety and environmental issues. 

Additional changes were made to 

§§ 51.50 and 52.17 to implement this 

concept. 

Some commenters recommended that 

the requirement in § 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 


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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 plantspecific PRA that was in proposed 

§ 52.80(a) to a new § 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 plantspecific 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 § 52.79(a) for a 

combined license application that 

references an early site permit. Former 

§ 52.79(a) did not explicitly require the 

application to address whether the 

terms and conditions specified in the 

early site permit under § 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, 

§ 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 

§ 52.79(b)(3) in the final rule based on 

public comments to add an exclusion 

for terms and conditions imposed under 

§ 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 

§§ 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 § 52.79(b)(1) in 

the final rule. Proposed § 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 § 50.59 after 

issuance of the COL. For these reasons, 

the NRC is modifying the language in 

§ 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 §§ 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 § 52.79(c) are 

essentially the same as those for a 

combined license application that 

references a standard design 

certification in § 52.79(d). 

Section 52.79(d) describes the 

requirements for combined license 

applications that reference a standard 


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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 plantspecific 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 

§ 52.79(d)(1). Section 52.79(d) also 

requires that the FSAR demonstrate that 

the interface requirements established 

for the design under § 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, § 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 § 52.79(d). In 

addition, § 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 

§ 50.10(e) with the associated redress 

plan. This last item was moved to 

§ 52.80(c) in the final rule from its 

location in § 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 § 50.10(e) address both 

safety and environmental issues. 

Additional changes were made to 

§§ 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 § 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 § 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, § 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 § 52.99. The NRC 

believes that the former § 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 § 52.83 also provides 

provisions that address the duration of 

a combined license and these provisions 

would be moved to proposed § 52.104, 

Duration of combined license. 

The new § 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 §§ 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 § 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 § 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 § 51.95(a). This provision is 

removed because the requirements for 

compliance with NEPA are now 

captured in § 52.79(a) and in the 

revisions to part 51. 


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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 

§ 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 § 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 § 52.17(c). This 

amendment is consistent with 

§ 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 § 50.10(e)(3)(i) activities, even 

if the NRC’s regulations are met. 

k. Section 52.93, Exemptions and 

Variances 

Paragraph (a) of § 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 § 52.63 

if there are no applicable exemption 

provisions in the referenced design 

certification rule. This provision 

formerly referred to compliance with 

§ 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 

§§ 52.98 and 52.171. However, the 

proposed rule incorrectly used the term 

‘‘variance’’ to describe an applicationspecific 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 § 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 § 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 

§ 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 § 52.97 to be 

more consistent with the parallel 

provision in § 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 

§ 50.50, which states that a combined 

license shall contain conditions and 

limitations, including technical 

specifications, as the NRC deems 

necessary and appropriate. Former 

§ 52.97(b)(2) is moved to new § 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 


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license, except in accordance with the 

provisions of §§ 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 § 52.171. Note that the 

proposed rule incorrectly used the term 

‘‘variance’’ to describe an applicationspecific 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., 

§§ 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 

§ 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 § 50.54(f). For combined licenses, this 

proposed provision is in § 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, 

§ 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 § 52.103(g) to continue the 

historical practice. The NRC believes 

that this is the common interpretation of 

§ 52.103(g). 

o. Section 52.104, Duration of Combined 

License; § 52.105, Transfer of Combined 

License; § 52.107, Application for 

Renewal; § 52.109, Continuation of 

Combined License; and § 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 

§ 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 § 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 § 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 § 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, §§ 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: §§ 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 (§§ 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 


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discussion on 10 CFR 52.43 in this 

document). 

When the NRC adopted part 52 in 

1989, the Commission did not reexamine 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 § 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 mediumand 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 

(§ 52.137) were renumbered to be 

consistent with the numbering of 

requirements in § 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 § 52.47). Some 

commenters recommended that the 

requirement for coping with 

emergencies [§ 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 § 52.139, which specifies the 

standards that will be used to review 

applications for design approvals and 

new §§ 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 § 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 §§ 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 


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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 reexamination 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 

§§ 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, § 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 § 50.43. As 

discussed separately with respect to 

§ 50.43, these testing and prototype 

requirements incorporated into § 50.43 

were derived from the former 

requirements in § 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 

§ 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 § 52.158(a) to a new 

§ 52.157(f)(31) which requires that 

manufacturing license applicants 

submit a description of the designspecific 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 costeffective 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 §§ 52.158(c)(1), 51.54, and 

51.75(c)(3). However, in the final rule, 

the Commission has removed from 

proposed §§ 52.158(c)(1) and (2) (final 

§§ 52.158(b)(1) and (2)) the rule 

language addressing the content of the 

environmental report, and integrated 

that language into §§ 51.54 and 

51.75(c)(3). Proposed § 52.158(c)(2) 

(final § 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 

§§ 2.104 and 2.105, the NRC has 

decided not to require a ‘‘mandatory’’

hearing for initial issuance of a 

manufacturing license, and § 52.163 is 

revised in the final rule to refer to a 

publication of a notice of proposed 

action under § 2.105, rather than a 

notice of hearing under § 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 § 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 


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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 § 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 § 50.59, and 

§ 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 § 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 § 52.171 

has been corrected in the final rule by 

replacing the reference in paragraph 

(b)(1) to § 50.12 with a reference to 

§ 52.7, and replacing the term, 

‘‘exemption,’’ in paragraph (b)(2) with 

‘‘departure.’’

In proposed § 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 § 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 


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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, 

§ 52.301 and § 52.303, which are 

comparable to former § 52.111 and 

§ 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 

§ 50.59-like process to that used in the 

current § 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 § 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 § 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 § 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 exvessel 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 semiannually, 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 § 52.103(g) is to 

facilitate NRC’s monitoring of changes 

to the nuclear power plant, to achieve 

a common understanding of how the asbuilt 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 


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under § 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 § 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 


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6The 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. 

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

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, § 50.2, Definitions 

New definitions are added as 

conforming changes to § 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 § 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, 

§ 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 § 50.10(b)) 

are authorized under a combined 

license, § 50.10(b) is revised to refer to 

combined licenses. 

Formerly § 52.17(c) authorized an 

early site permit applicant to request 

authority to perform the activities 

allowed under § 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 

§ 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 

§ 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 § 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 § 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 § 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 § 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, 

§ 50.30(a) is revised to ensure that the 

submission requirements governing 

applications (and amendments to these 

applications) in § 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 


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authority to conduct the activities 

permitted under § 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 § 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 nonlicensed 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 § 52.17(b)(2)(ii) to 

seek NRC review and approval of 

complete emergency plans. In addition, 

§ 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 § 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 § 50.33(g). This is 

necessary for the NRC to be able to find 

that major features describing the EPZs 

are acceptable under § 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 

§§ 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 § 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 § 50.75. 

c. Section 50.34, Contents of 

Construction Permit and Operating 

License Applications; Technical 

Information 

The NRC is changing the heading of 

§ 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 § 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 § 50.34. For 

example, former § 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 sitespecific. This would encompass 

requirements in both §§ 50.34(a) and (b). 

Also, former § 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 § 50.34(b). In 

addition to the requirements for 

technical information in §§ 50.34(a) and 

(b), §§ 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 § 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 § 50.34 

to make it applicable to construction 

permit and operating license 

applications only and to provide 

separate sections for the technical 


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contents of applications for the other 

types of licenses or regulatory approvals 

in 10 CFR part 52 (early site permits in 

§ 52.17, design certifications in § 52.47, 

combined licenses in § 52.79, design 

approvals in § 52.137, and 

manufacturing licenses in § 52.157). In 

its revisions to 10 CFR part 52, the NRC 

has brought forward the requirements 

from § 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 § 50.34(f) 

related to compliance with TMI related 

requirements. Due to the length and 

complexity of the requirements in this 

paragraph, § 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 § 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 § 50.34(f) based on 

public comments. The commenters 

pointed out that proposed § 50.34(f) was 

inconsistent with proposed 

§§ 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 § 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 § 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, § 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 § 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 

§ 50.34a in their applications by virtue 

of the provision in former 

§ 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 § 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 § 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 

§ 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 § 50.36. The existing 

language in § 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 § 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 § 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 § 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 

§ 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 § 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 


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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 § 52.157(e)(18) did require 

manufacturing license applicants to 

include proposed technical 

specifications in accordance with 

§ 50.36a. Therefore, it was clearly the 

NRC’s intent that the provisions of 

§ 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 § 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 § 52.103(g) 

that allows fuel load and operation. This 

would apply the requirements in 

§ 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 

§§ 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 

§ 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 § 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 § 50.36b. The 

NRC provided proposed requirements 

for identifying environmental 

conditions on early site permits and 

combined licenses in the proposed rule 

in §§ 51.50(b) and (c). Requirements for 

identifying environmental conditions 

for construction permits were contained 

in former § 51.50 and proposed 

§ 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 nonaquatic environment, proposed for 

possible inclusion in the license as 

environmental conditions in accordance 

with § 50.36b of this chapter.’’ However, 

the NRC neglected to make the 

additional conforming changes to 

§ 50.36b in the proposed rule. To correct 

this oversight, the NRC has modified 

§ 50.36b in the final rule to make the 

requirements in this section consistent 

with the requirements in § 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 § 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 § 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 § 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 


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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 

§ 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 § 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 § 50.34 was moved to the 

contents of application section for each 

of the licensing processes in part 52. 

Therefore, § 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 

§ 52.103(g) of this chapter. The 

standards for issuance of a combined 

license are set forth in § 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 § 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, 

§ 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 § 50.47 is necessary for 

issuance of a renewed nuclear power 

reactor operating license. The NRC is 

revising § 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 § 50.47(a)(1) in the final rule. 

Proposed § 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 § 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 § 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 

§ 50.47(b). The NRC is revising 

§ 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 lowpower testing and training may be 


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

§ 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 §§ 50.47 and 50.54. The 

introduction to appendix E to part 50 

states that each applicant for an 

operating license is required by 

§ 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 

§ 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 § 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 § 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 § 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 § 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 § 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 exerciserelated 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 


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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 § 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 

§ 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 § 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 § 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 

§ 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 

§ 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 § 50.55, which 

indicates that these conditions are 

applicable until the date that the 

Commission makes the finding under 

§ 52.103(g). Conditions which are 

applicable during construction and 

operation or only during operation are 

set forth in § 50.54. The NRC is revising 

the introductory paragraph of § 50.54 to 

refer to combined licenses, and to 

exclude manufacturing licenses from its 

provisions. The NRC is making 

revisions to § 50.54 in the final rule 

based on public comments. In the 

proposed rule, the NRC did not 

distinguish which provisions in § 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 § 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 

§ 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 § 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 § 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 § 50.54(i-1) to 

indicate its applicability to combined 

licenses. Specifically, § 50.54(i-1) 

requires that within 3 months after the 

date that the Commission makes the 

finding under § 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 § 55.59(c) of this 

chapter. 

The NRC has added changes to 

§ 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 

§ 50.47(b) and the requirements in 

appendix E of part 50 is only applicable 

after the Commission makes the finding 

under § 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 § 50.54 

discussed earlier in this section. 

The NRC is adding a new § 50.54(gg). 

These revisions are discussed with 

related requirements in Section IV.D.4.f 

of this document, ‘‘Section 50.47, 

Emergency plans, § 50.54(gg), and 

appendix E to part 50.’’

Although the NRC generally views 

§ 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 

§ 50.55 should apply equally to all of 

the part 52 processes. Accordingly, the 

introductory text to § 50.55 is revised to 

specify which paragraphs apply to a 

construction permit, early site permit, 

combined license, and manufacturing 

license. 


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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 § 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 § 50.55(c) 

is necessary because the introduction to 

§ 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 § 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 § 50.55(e) from the date 

of issuance of the combined license 

until the Commission makes the finding 

under § 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 § 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 § 50.55(e)(1) to provide that the 

reporting requirements in § 50.55(e) 

apply to a holder for a combined license 

(until the NRC makes the finding under 

§ 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 

§ 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 

§ 50.54(a); requirements governing the 

SAR’s description of the QA program 

are contained in § 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 § 50.55a to clarify 

how the regulations in § 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 § 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, § 50.55a 

formerly applied to design certifications 

by virtue of the provision in former 

§ 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 

§ 50.55a, the NRC is requiring design 

approval holders to comply with 

§ 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 

§ 50.55a to state that each combined 

license for a utilization facility is subject 

to the conditions in § 50.55a, but is only 

subject to the conditions in §§ 50.55a(f) 

and (g) after the NRC makes the finding 

under § 52.103. The modifications to 

§ 50.55a also state that each 

manufacturing license, design approval, 

and design certification application is 

subject to the conditions in §§ 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. 


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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 § 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 

§ 50.62. Similarly, the NRC is adding 

provisions to §§ 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, § 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 § 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 

§ 50.70(a) encompasses combined 

license holders and manufacturing 

license holders because they are 

licensees. In addition, the provision in 

former § 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 

§ 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 § 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 

§ 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 

§ 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, 


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

§ 50.71(e) to specify the FSAR updating 

requirements for combined license 

applicants and holders. In addition, 

current § 50.71(f) is redesignated as 

§ 50.71(g), and a new § 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 § 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 § 52.103(g), the frequency would 

be governed by § 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 § 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 

§ 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 § 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 

§ 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 § 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 

§ 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 NRCendorsed 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 § 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 NRCendorsed 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 


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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 

§ 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 § 50.72 

only apply to a combined license holder 

after the Commission makes the finding 

under § 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 § 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 § 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 § 50.75(b)(4) that 

the operating license applicant submit a 

copy of the financial instrument 

obtained to satisfy the requirements of 

§ 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 § 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 § 50.33(k), but it need not 

obtain a financial instrument to fund 

decommissioning or to submit a copy to 

the NRC. Instead, under § 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 § 52.103(a). See 

§ 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 § 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 

§ 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 § 50.75(e). 

Once the Commission has made the 

finding under § 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 


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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 § 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 § 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 

§ 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 

§ 52.103(a), which are applicable to the 

combined license holder after the 

Commission has made the finding under 

§ 52.103, are adopted in the final rule 

without change from the proposed rule. 

The § 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 § 50.75, to subject holders of 

manufacturing licenses to the 

requirements of that section. It is clear 

from the words of former § 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 NonProliferation 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, 

§ 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 § 50.78 should also apply 

to a holder of a combined license under 

part 52. Accordingly, § 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 § 75.6, and 

§§ 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 § 50.80(a) 

to both permits under parts 50 and 52 

and all licenses under part 52, § 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 § 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 §§ 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 § 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 § 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. 


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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 § 50.4, following the form prescribed 

for the original application. Although 

the term, license, as amended in § 50.2 

includes combined licenses, 

manufacturing licenses, and early site 

permits under part 52, § 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 § 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 § 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, § 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 

§ 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 

§ 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 § 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 § 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, § 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, § 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 § 50.109 to reflect the 

licensing and regulatory approval 

processes in part 52, the Commission 

should adopt a general backfitting 

provision, analogous to § 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 

§ 50.109 to include the conforming 

changes, rather than adopting a 

backfitting provision in part 52. 


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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 

§ 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 § 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 § 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 §§ 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 § 50.55(f), and QA for 

operation is governed by § 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 § 50.55(f) from the date 

of issuance of the combined license 

until the Commission makes the finding 

under § 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 § 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 

§ 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 § 50.55(f), and 

§ 50.55(f) is revised accordingly. 

Appendix B to part 50 was formerly 

applicable to combined licenses under 

the provisions of § 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 § 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 §§ 50.34(a) and 

(b), as appropriate, and stated that the 

information required by § 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 


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7Although early site permit applicants may seek 

the authority to conduct activities allowed under 10 

CFR 50.10(e)(1) (but not activities allowed under 

§ 50.10(e)(3), see § 52.17(c)), these activities are not 

considered ‘‘construction.’’

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 § 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. 

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 §§ 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 §§ 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 

§ 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 

§ 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 

§ 50.82(a)(1) or § 52.110 as guidance in 

developing technical specifications 

under § 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 lightwater-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-WaterCooled 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 watercooled 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 


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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 § 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, § 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 § 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, § 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 

§ 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 § 50.30(f) if the 

application is for a construction permit 

or § 52.80(b) if the application is for a 

combined license. The other part must 


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contain the information required by 

§§ 50.34(a) and 50.34a if the application 

is for a construction permit, or § 52.79 

and § 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 § 50.33, 

§ 50.34(a)(1) if the application is for a 

construction permit, § 52.79(a)(1) if the 

application is for a combined license, 

and § 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 § 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 § 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 § 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 § 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 § 2.104. 

Accordingly, the language of § 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 § 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, 

§ 2.105(a)(4) is revised to clarify that the 

procedures in § 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 § 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 § 2.105 by adding 

§ 2.105(a)(12) which addresses the 

information to be contained in the 

Federal Register notice required by 

§ 52.103(a). 

Because the Commission’s 

authorization for a combined license 

holder to operate under § 52.103 does 

not constitute ‘‘issuance’’ of a license or 

amendment under § 2.106, § 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 § 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, § 2.105(b)(3) of the 

final rule does not include a 

requirement for such a finding to be 


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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 § 52.103(g). Although 

the AEA does not require publication of 

a notice of the Commission finding 

under § 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, § 2.106(a) is revised to 

require Federal Register publication of 

the Commission finding under § 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 § 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 § 2.110 are revised to refer 

to subpart E of part 52 and appendix Q 

of part 50. Paragraph (c) is corrected by 

adding § 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 § 2.202(e) as § 2.202(e)(1), 

and adding §§ 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 

§ 52.103(c), to make clear that in a 

proceeding under § 52.103, the 

Commission itself will act as the 

presiding officer, will consider and act 

upon a request for a hearing under 

§ 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 

§ 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 

§ 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 § 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 § 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 § 52.99. 

Furthermore, if the requestor contends 

that the § 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 

§ 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 


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8The 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. 

showing. The NRC believes that these 

changes to § 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 § 52.103 

hearing. Under the revised paragraph 

(g), a request for hearing under § 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 § 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 § 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 

§ 52.103(c), paragraph (i) is revised to 

prohibit any ‘‘appeal’’ under § 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 § 2.345, 

and it would be duplicative to afford an 

adversely-affected party a § 2.311 

‘‘review’’ right in addition to the 

opportunity to seek reconsideration 

under § 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 § 2.310 now provides that the 

Commission will determine the hearing 

procedures to be utilized in its 

determination on a hearing request 

under § 52.103, as well as the hearing 

procedures to be utilized in resolving 

admitted contentions under § 52.103(c) 

and (g).8

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 § 52.103(g) 

with respect to whether the acceptance 

criteria in a combined license ITAAC 

have been, or will be met. 

The § 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 § 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 


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9 31 FR 12774 (September 30, 1966). 

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 § 2.340(a)/§ 2.760a limitation to 

construction permit (including early site 

permit) proceedings. Nor should the 

§ 2.340(a)/§ 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 § 52.103(g), and in 

a manufacturing license proceeding 

(including a renewal or amendment 

proceeding). 

As discussed previously, the former 

§ 2.340(a)/§ 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 policybased, or legal reason why the § 2.340(a) 

limitation should not be extended to 

non-production or utilization facility 

proceedings. Accordingly, the NRC is 

revising § 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 

§ 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 § 2.340 as an integrated whole. 

Immediate Effectiveness of Presiding 

Officer’s Initial Decision in Production 

and Utilization Facility Proceedings 

The remainder of former § 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 § 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 § 2.764, and 

providing for both Atomic Safety and 

Licensing Appeal Board review and 

Commission review of the presiding 

officer’s initial decision. 

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 § 2.764 were transferred 

without substantive change to a new 

§ 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 § 2.345; (2) 

a petition for Commission review under 

§ 2.341; and (3) a motion for a stay with 

the presiding officer or the Commission 

under § 2.342. 

By removing the ‘‘automatic stay’’

provisions in former § 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 


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49416 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 § 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, NonProduction 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 § 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 lowlevel 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 

§ 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 § 52.103(g) Findings 

Former paragraph (c) of § 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 

§ 52.103(g). To resolve these concerns, 

new paragraphs (i), (j), and (k) are added 

to § 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 

§ 52.103(c), paragraph (a)(1) of § 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 § 52.103(c), separate from and in 

addition to a request for reconsideration 

under § 2.345. Inasmuch as the 

Commission is acting as the presiding 

officer in the § 52.103(c) determination, 

there can be no further ‘‘appeal’’ to a 

higher agency decisionmaker. Moreover, 

it would be duplicative to afford a 

§ 2.341 ‘‘review’’ or ‘‘appeal’’ right in 

addition to the opportunity to seek 

reconsideration under § 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, § 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). 


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Accordingly, § 2.347 is revised to 

explicitly provide that ex parte 

restrictions do not apply to uncontested 

proceedings. 

Second, § 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, § 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 § 2.340(a), and the 

presiding officer’s examination of that 

matter following Commission approval 

under § 2.340(a) (referred to as ‘‘sua 

sponte’’ issues at 53 FR 10361; March 

31, 1988). The application of ex parte 

restrictions to § 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 § 2.340(a) ‘‘sua 

sponte’’ matters as a ‘‘disputed issue’’

for purposes of applying § 2.347. 

Accordingly, the NRC is revising § 2.347 

to explicitly state that consideration of 

§ 2.340(a) ‘‘sua sponte’’ matters are to be 

subject to ex parte restrictions. 

Inasmuch as these § 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 § 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 § 2.340(a), and the 

presiding officer’s examination of that 

matter following Commission approval 

under § 2.340(a). As with the change in 

§ 2.347 with respect to ex parte 

restrictions, this change in § 2.348 does 

not depart from the NRC’s current 

practice of applying separation of 

function restrictions to ‘‘sua sponte’’

matters under § 2.340(a). The NRC 

believes that it is more accurate to 

explicitly state that sua sponte matters 

under § 2.340(a) are subject to 

separation of functions restrictions, 

rather than characterizing such matters 

as ‘‘disputed issues.’’

Inasmuch as these § 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 


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(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, § 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 § 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 § 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 § 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 § 2.104(f). As 

previously discussed, the Commission 

is consolidating in § 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 

§ 2.101(a–1). As discussed earlier, the 

NRC has revised § 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 § 2.600, Scope of subpart). 

Requirements applicable to partial 

decisions in construction permit 

proceedings continue to be addressed in 

§§ 2.602 through 2.606; a new 

subheading is added before § 2.602 to 

reflect the subject matter of these 

sections. The new requirements 

applicable to partial decisions in 

combined license proceedings are in 

§§ 2.621 through 2.629; a new 

subheading is also added before § 2.621 

to reflect the subject matter covered by 

these sections. Section 2.629, which has 

no analogous provisions in §§ 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., §§ 52.39, 

52.63, 52.98. 


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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 preapplication 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 § 2.800 is revised 

to reflect the additional function of this 

section. New §§ 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 §§ 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 

§§ 2.801 through 2.810. 

34. Section 2.801, Initiation of 

Rulemaking 

In a conforming change, § 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 § 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 § 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 

§ 50.30(a)(1) and (3), provides that a 

standard design certification application 

must meet the written communications 

requirements in § 2.813. Section 

2.811(c), which is analogous to 

§ 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 § 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 preapplication 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 § 2.802(a)(1)(i) through 

(iii), including the procedure and 

process for filing and processing an 

application for a design certification. 

However, § 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 

§ 2.802(a)(2) do not apply to these 

consultations. 

36. Section 2.813, Written 

Communications 

New § 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 § 50.4(a). Section 

2.813(b) is analogous to § 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 § 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 § 2.815 is analogous to 

§ 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 § 2.817 is analogous to § 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 § 2.819 is analogous to § 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 


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49420 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 § 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 § 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 nonoperational activities. Finally, part 19 

applied only to a ‘‘licensee’’ and 

activities authorized by a ‘‘license’’ (see, 

e.g., §§ 19.1, 19.2, 19.11, 19.20, 19.32), 

and did not extend to part 52’s nonlicensing 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 


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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 nonlicensing regulatory approvals in part 

52. 

To implement the NRC’s broadening 

of the scope of part 19, §§ 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 

§ 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 § 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 §§ 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 § 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 § 50.12 and § 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., § 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, § 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 

§ 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 

§ 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 § 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 

§ 50.55(e). As discussed previously, 

§ 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 


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10Throughout this discussion, reference to 

entities, licensees and/or applicants includes the 

contractors and subcontractors of those entitles, 

licensees and/or applicants. 

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. 

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 decisionmaking 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 

§ 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 


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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 § 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 

§ 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 

§ 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 §§ 21.31 and 

50.55(e)(7). 


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Division of Implementing Requirements 

Between Part 21 and § 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 § 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 § 50.55(e). The revised part 21 

and § 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 § 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 § 52.103(g) are 

contained in § 50.55(e). The 

requirements in part 21 apply after the 

Commission makes the finding under 

§ 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 § 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

Part 52 licensing or approval processes Applicable NRC requirement implementing section 206 of the ERA 

Sanctions 

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 § 52.103 Authorization ....................................................... 50.55(e) ............................................ 50 .110 50 .111 

COL after § 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 


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11These 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 nonstandardized 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. 

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 § 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 §§ 21.2, 21.3, 21.21, 21.51, and 

21.61 to explicitly refer to an applicant 

for a standard design certification, 

rather than § 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 § 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, valueadded 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 § 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 

§ 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 § 50.55(e) and part 21 with 

respect to NRC requirements 

implementing Section 206 of the ERA, 

the NRC revised § 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 § 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 § 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 § 50.55(e). 


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49426 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

Accordingly, the NRC revised § 50.55(e) 

to specifically apply its provisions to 

holders of manufacturing licenses. 

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, § 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, § 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-forduty 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 §§ 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 § 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 § 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 § 50.10(e)(3) are also 

treated the same as a construction 

permit holder. The final rule 

accomplishes this by: (1) Revising 

§ 26.2(a) to refer to combined license 

holders after the date that the NRC 

makes the finding under § 52.103(g); (2) 

revising § 26.2(c) to refer to a holder of 

a combined license before the date that 

the NRC makes the finding under 

§ 52.103(g), a holder of a manufacturing 

license under subpart F of part 52, and 

a person authorized to conduct the 

activities under § 50.10(e)(3); (3) 

revising § 26.10(a) to refer to the 

personnel of a holder of a 

manufacturing license and those 

authorized to conduct the activities 

under § 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 

§ 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 fitnessfor-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 preapproval 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 costjustified 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 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 49427 

12The design certification applicant may have 

chosen to specify site parameters for the design 

certification safety review under § 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 § 52.63 to apply in the 

combined license proceeding. 

13A 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 § 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. 

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 § 51.55 as § 51.58, and is adding 

new § 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 costbeneficial, 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 § 51.30 to provide 

for a new § 51.30(d) establishing the 

scope of an environmental assessment 

for a design certification. The NRC is 

adding §§ 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 

§ 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 § 51.31 

by adding § 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 § 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

Finally, the NRC is adding 

§ 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. 

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 § 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 


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49428 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

14Analogous 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 § 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 § 52.171 to 

apply in the combined license proceeding. 

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 § 51.20(b)(6), 

which formerly required preparation of 

an EIS for issuance of a manufacturing 

license, and § 51.76, which formerly 

addressed the subject matter of an EIS 

for a manufacturing license, from part 

51. 

The NRC is revising § 51.30(e) to 

establish the scope of an environmental 

assessment prepared for a 

manufacturing license. The NRC is 

adding §§ 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 § 50.32(b)(1) and 

(2), the NRC is including an explicit 

statement of no significant 

environmental impact for 

manufacturing licenses in § 51.32(b)(3) 

and (4) to facilitate external 

stakeholders’ understanding of the 

nature of the Commission’s action. The 

NRC is adding § 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 § 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 

§ 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, § 51.75(c)(3) 

requires the combined license EIS to 

summarize the findings and conclusions 

of the environmental assessment with 

respect to SAMDAs. Finally, 

§ 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. 

NEPA Obligations Associated With 

§ 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 § 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 

§ 52.103(b). The NRC never intended to 

make an environmental finding in 

connection with the § 52.103(g) finding 

on ITAAC, and the NRC does not 

believe that NEPA requires such a 

finding. The § 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 § 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 

§ 52.103(g) findings). The safety 

consequences of operation are not 

considered when making the § 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 § 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 § 52.103(b). 

Accordingly, the NRC is adding 

§ 51.108 to clarify that: (1) The 

Commission will not make any 

environmental findings in connection 

with the finding under § 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 

§ 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 § 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 


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

§ 51.20(b)(1) to indicate that issuance of 

an early site permit requires an EIS. The 

NRC is revising § 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 § 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 §§ 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 § 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, ScopingEnvironmental 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 § 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 § 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 § 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 § 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 

§ 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 


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49430 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 § 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 § 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 § 51.50(b). 

The NRC is further modifying 

§ 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, 

§ 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 § 51.75(b) 

[proposed § 51.71(d)]. 

The NRC is making additional 

changes to § 51.50(b) to further clarify 

the scope of the environmental review 

at the early site permit stage. Final 

§ 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 

§§ 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 

§ 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 

§ 50.10(e) address both safety and 

environmental issues. Additional 

changes were made to §§ 52.17(c), 

52.79(a), and 52.80 to implement this 

concept. 

The NRC is also revising § 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 § 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 

§ 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 


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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 § 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 

§ 51.50(b), but the associated provision 

to ensure any conditions included in the 

permit would be met was inadvertently 

left out of § 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 


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49432 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 § 51.51 to require 

that every environmental report 

prepared for the early site permit stage 

or combined license stage of a lightwater-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 § 51.52 to require 

that every environmental report 

prepared for the early site permit stage 

or combined license stage of a lightwater-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 § 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 § 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 

§ 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 § 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 § 51.55 to a new § 51.58 

(discussed in § 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 § 51.55 is 

similar to that of § 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 

§ 51.55 are addressed in a new § 51.58. 

The NRC is adding conforming 

references to § 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 § 51.71(d) to 

include a reference to § 51.75 in the first 

sentence because § 51.75 also includes 

exceptions to the provisions in 

§ 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 

§ 51.71(d) to their associated paragraphs 

in § 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 §§ 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 

§ 51.71(d) to § 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 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 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 § 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, § 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, § 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 § 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 § 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 § 51.75(c)(1). Paragraph (b) 

of § 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 § 51.92(e). In 

the final rule, the NRC is moving 

information related to combined 

licenses that was contained in proposed 

§ 51.71(d) to § 51.92(e) and is revising 

the wording of this provision. In the 

proposed rule, § 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 § 51.92(e). The revised 

language in paragraph (e) provides a 

clearer link to the finality provisions in 

§ 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 § 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 

§ 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 § 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 § 51.95(d) to indicate that, 

unless otherwise required by the 

Commission, in accordance with the 

provisions of § 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. 


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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 § 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 § 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 

§ 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 

§ 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 

§ 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 § 51.105a to 

provide requirements for public 

hearings in proceedings for issuance of 

manufacturing licenses. Specifically, 

§ 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 § 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 §§ 52.39 and 52.103. The 

NRC is making changes to the language 

in § 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 § 51.75(c)(1) and § 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 

§ 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. 


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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, §§ 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 

§ 73.1, Purpose and scope, § 73.2, 

Definitions, § 73.50, Requirements for 

physical protection of licensed 

activities, § 73.56, Personnel access 

authorization requirements for nuclear 

power plants, and § 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 § 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, § 73.1 

is revised to clarify that the regulations 

in part 73 apply to persons who receive 

combined licenses under part 52, and 

§ 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 

§ 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 

§ 52.103 for a combined license. 

Additional conforming changes to 

include part 52 licenses are made for 

§§ 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, § 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, 

§ 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 §§ 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 § 95.39 to use the phrase, ‘‘NRC 


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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 § 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 

§ 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 § 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 

§ 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 § 52.4. The Commission 

notes, as discussed below in the sectionby-section analysis of § 52.4, that 

deliberate misconduct under § 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 § 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 

§ 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 § 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 § 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 § 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 


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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, 

§ 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, § 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 

§§ 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 § 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 § 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 § 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 § 52.47(b) (new 

§ 50.43), and § 52.157(e)). A prototype 

plant is a licensed nuclear reactor test 

facility that is similar to and 

representative of either the first-of-akind 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 

§ 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 § 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 

§ 50.4; they are reproduced in § 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 

§ 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 § 52.4 differs 

from former § 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 § 52.4. The 

relevant law and policy considerations 

for former § 52.9 are merely clarified 

and extended in § 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 § 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 § 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 § 52.5; the 

relevant law and policy considerations 

for former § 50.7 are merely clarified 

and extended in § 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 § 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 


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underlying policy considerations is 

being made by the Commission 

adoption of § 52.6; the relevant law and 

policy considerations underlying § 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, § 50.9 does not impose a 

positive obligation on licensees to seek 

out new information meeting the 

reporting thresholds in the rule. In 

applying § 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 § 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 § 52.7; the 

relevant law and policy considerations 

underlying § 50.12 are merely extended 

to part 52. 

The NRC notes that the exemption 

provisions in § 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 § 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 § 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 § 52.7. 

However, the exemption request must 

then demonstrate compliance with the 

additional criteria in § 52.7. 

The Commission also notes that § 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 § 50.55a(a)(3). However, 

in the absence of a more specific 

dispensation provision, the Commission 

intends to utilize § 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 § 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 § 52.7 with respect to the 

reasons for dispensation from 

compliance with § 50.48. The holder 

need not address dispensation from 

compliance with § 52.0, which 

otherwise makes applicable the 

provisions of § 50.48 on the licensee. 

Any exemption granted by the 

Commission would address the reasons 

for dispensation with the underlying 

requirement—in this case, § 50.48, and 

would also provide dispensation from 

compliance with § 52.0. 

Section 52.8 Combining Licenses; 

Elimination of Repetition 

This new section includes provisions 

analogous to §§ 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 § 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, § 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 

§ 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 

§ 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 § 50.31, the 

language in § 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 

§§ 50.31, 50.32, and 50.52 is being made 

by the Commission adoption of § 52.8; 

the relevant law and policy 

considerations underlying §§ 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 § 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 

§ 50.53 is being made by the 

Commission adoption of § 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 § 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 § 52.10; the 

relevant law and policy considerations 

for the § 50.13 exclusion are merely 

extended to cover applicants for and 

holders of permits, licenses, and 

regulatory processes that are contained 

in part 52. 


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Section 52.11 Information Collection 

Requirements: OMB Approval 

This section, formerly designated as 

§ 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 preapproval 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 (§ 52.17(a)(1)(v)), the 

applicant should describe the foot print 

for all structures and external safetyrelated 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 § 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, 


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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 § 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 § 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 § 52.39. After issuance of the 

construction permit or combined 

license, § 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 

§ 52.98 (e.g., in accordance with 

§§ 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 


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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 § 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 § 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 § 52.39(b); or issues a 

variance requested under § 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 

§§ 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 § 52.39(a)(1). New 

information which materially changes 

the Commission’s determination of the 

matters in § 52.17(b), or results in 

modifications of existing terms and 

conditions by the NRC under 

§ 52.39(a)(1) would be subject to 

litigation during the licensing 

proceedings in accordance with 

§ 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 § 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 


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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 

§ 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 § 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 § 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 § 50.54(f). 

For early site permits, this provision is 

contained in § 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 § 52.45 to 

this section. The scope of the standard 

plant design must be essentially 

complete as described in § 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 § 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 § 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), 


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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 § 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 § 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 


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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 

§ 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 § 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 § 52.63(c) or § 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 § 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 

§ 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 plantspecific 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 § 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 


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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 § 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 § 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 

§ 52.63(a)(1), except for § 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 plantspecific 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 

§ 52.7. In addition, the Commission 

shall consider whether the special 

circumstances that justify the plantspecific 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 


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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 siterelated 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 § 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 § 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 § 50.36b are to be addressed in 

the environmental report and not in the 

FSAR. 


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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 

§ 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 

§ 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 sitespecific 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, postcombined 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 


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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 § 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 § 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 


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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 sectionby-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 § 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 § 52.17 regarding 

early site permits, § 52.47 regarding 

design certifications and § 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 § 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 §§ 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 

§ 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 

§§ 52.39(f), 52.145(c), and 52.171(c), 


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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 § 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 § 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 § 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 § 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 


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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 

§ 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 § 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 § 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 § 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-bysection 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 § 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 postconstruction 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 § 52.80 

before issuance of the combined license. 

In such a case, if the Commission makes 

a finding in accordance with § 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 § 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 § 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 § 52.103(b)(2)(i) 

(§ 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. 


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49452 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 

§ 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 

§ 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 § 52.103(g). 

Where the Commission has allowed 

operation during an interim period 

under § 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 

§ 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 

§ 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. 


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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 § 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. 


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49454 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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., §§ 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., 

§§ 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., 

§§ 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 §§ 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., §§ 52.16, 

52.46, 52.77). Section 52.156 requires 

that the applicant include the 

information set forth in § 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 § 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 § 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., §§ 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 technologyneutral design criteria for non-light 

water cooled reactor designs in the 

future. This requirement was drawn 

from § 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 

§ 50.34(a)(3)(ii). 


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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 

§ 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 

§ 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 

§ 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 § 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 safeshutdown 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 § 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 § 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 


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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 § 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 sitespecific 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 § 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 

§ 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 § 52.163 and § 51.54. 

Under § 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, § 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 

§ 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., §§ 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., §§ 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 § 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 § 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 § 2.310 (and reiterated in 

§ 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., §§ 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., §§ 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 


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15The finality provision in § 52.83 performs a 

different function than the finality sections cited 

above, in that it points back to, and thereby reemphasizes, the primary finality provisions for each 

license or regulatory approval mechanism in part 

52, e.g., the finality provision in § 52.39 for early 

site permits. 

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 § 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., §§ 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. 

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 § 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 

§ 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 

§ 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 §§ 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, 


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16A 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. 

17Subpart C does not contain a ‘‘criteria for 

renewal’’ provision, inasmuch as the renewal 

would be governed by 10 CFR part 54, see § 52.107. 

Part 54 contains a provision, § 54.29, setting forth 

the standards for issuance of renewed licenses. 

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 § 52.7; however, the 

departure itself is not considered an 

exemption (unless, of course, the 

departure also involves a noncompliance 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 § 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 § 50.59. 

Paragraph (c), which is analogous to 

§§ 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., §§ 52.33, 52.61, 52.104). Under 

§ 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 § 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 

§ 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., 

§§ 52.28, 52.105).16 Section 52.175 

provides that a manufacturing license 

may be transferred in accordance with 

§ 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 § 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 § 50.80(b). 

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., 

§§ 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., 

§§ 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. 

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. 


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18Subpart 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 § 52.107. Part 54 contains a provision, 

§ 54.31, governing the duration of renewed licenses. 

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., 

§§ 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. 

Subpart G—Reserved 

This subpart is reserved for future use 

by the Commission. 

Subpart H—Enforcement 

This subpart contains two provisions, 

§ 52.301 and § 52.303, which are 

comparable to former § 52.111 and 

§ 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 § 52.113 as not being 

subject to criminal sanctions. The 

regulations in part 52 which are subject 

to criminal sanctions are §§ 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 plantspecific 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 § 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 § 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 

§§ 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. 


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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, § 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 §§ 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 § 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 2003 rule .............................. ML032810026 

Regulatory History Index for the March 13, 2006, proposed rule ................... ML062080575 

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 


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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

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

10 CFR Part 1 ................. Statement of Organization and General Information.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 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 

NRC.

These provisions are designated Compatibility 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 the 

definition are NRC.

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. 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. 


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LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

2.100 thru 2.390 ....... All of the sections covered by Subparts A, 

B, and C.

D, except portions of 

these provisions are 

NRC.

These provisions are designated Compatibility 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 covered by Subparts D, 

E, and F.

NRC, for all of the sections.These provisions are designated 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. 

2.800 ......................... Scope and applicability D, except portions of 

these provisions are 

NRC.

These provisions are designated Compatibility 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 rulemaking .. D, except portions of 

these provisions are 

NRC.

These provisions are designated Compatibility 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 design 

certification application, required copies.

NRC ............................... This provision is designated 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. 

2.813 ......................... Written communications NRC ............................... This provision is designated 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 acceptance review.NRC ............................... This provision is designated 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. 


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LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

2.817 ......................... Withdrawal of application.NRC ............................... This provision is designated a 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 for 

failure to supply information.

NRC ............................... This provision is designated 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. 

2.1202 ....................... Authority and role of 

NRC staff.

NRC ............................... This provision is designated 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 procedures 

for determining eligibility for access to restricted data or national security information or an employment 

clearance.

NRC for all sections ...... These provisions are designated 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 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 the 

provisions in (a)(1), 

(a)(2), (a)(3), and 

(a)(4) are designated 

as NRC.

This provision is designated Compatibility Category D because it does 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, 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 the 

definition are NRC.

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. 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. 


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LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

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 the 

definition are NRC.

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. 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 

workers.

C, except portions of 

paragraph (a), and all 

of paragraphs (b) and 

(e) are designated as 

NRC.

This provision is designated Compatibility Category C because it is needed to provide a minimum level of 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 representatives of licensees and 

workers during inspections.

C, except paragraph (a) 

is designated as NRC.

This provision is designated Compatibility Category C because it is needed to provide a minimum level of 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 the 

provision are NRC.

This provision is designated Compatibility 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. 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. 


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LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

19.31 ......................... Application for exemptions.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 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 prohibited 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 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 the 

provision are designated as NRC.

This provision is designated Compatibility 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. 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 and 

scope.

C, except portions of the 

provision are designated as NRC.

This provision is designated Compatibility Category C 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 contamination.C, except portions of 

paragraph (a) and all 

of paragraph (b) are 

designated as NRC.

This provision is designated Compatibility Category C because it is needed to provide a minimum level of 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. 


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LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

20.2203 ..................... Reports of exposures, 

etc., exceeding the 

limits.

C paragraphs (a) and 

(b).

NRC paragraphs (c) and 

(d).

Paragraphs (a) and (b) are designated Compatibility Category 

C, because they are 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 

and Noncompliance.

Not applicable for all 

sections.

The provisions in Part 21 are 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 Programs.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 50 ............... Domestic Licensing of 

Production and Utilization Facilities.

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 51 ............... Environmental Protection Regulation for Domestic Licensing and 

Related Regulatory 

Functions.

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 52 ............... Licenses, Certifications, 

and Approvals For 

Nuclear Power Plants.

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 54 ............... Requirements for Renewal of Operating License for Nuclear 

Power Plants.

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. 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 49467 

LIST OF CHANGES 10 CFR PART 52 FINAL RULEMAKING—Continued 

Sections Description new, 

changes Compatibility designation Comments regarding compatibility designation 

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 Requirements 

for Independent Storage of Spent Nuclear 

Fuel and High-level 

Radioactive Waste 

and Greater than 

Class C.

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 73 ............... Physical Protection of 

Plants and Materials.

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 75 ............... Safeguards on Nuclear 

Material—Implementation of US/IAEA 

Agreement.

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 95 ............... Facility Security Clearance and Safeguarding of National 

Security Information 

and Restricted Data.

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 140 ............. Financial Protection Requirements and Indemnity Agreements.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 170 ............. Fees for Facilities, Materials, Import and Export Licenses, and 

Other Regulatory 

Services under the 

Atomic Energy Act of 

1954, as Amended.

D .................................... These provisions are designated a 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. 

10 CFR Part 171 ............. Annual Fees: For Reactor Licenses and Fuel 

Cycle Licenses and 

Material Licenses, Including Holders of 

Certificates of Compliance, Registrations, 

and Quality Assurance 

Program Approvals 

and Government 

Agencies Licensed by 

NRC.

D .................................... These provisions are designated a 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. 

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 


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49468 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

19When 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 § 51.22(c)(3). 

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

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 

INFOCOLLECTS@NRC.GOV; 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 §§ 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. 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 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 § 50.104 or § 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 


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49470 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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. 

■ 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 

■ 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. 

■ 2. In § 1.43, paragraph (a)(2) is revised 

to read as follows: 

§ 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 

■ 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). 

■ 4. In § 2.1, paragraphs (c) and (d) are 

revised and a new paragraph (e) is 

added to read as follows: 

§ 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. 

■ 5. In § 2.4, the definitions of contested 

proceeding, license and licensee are 

revised to read as follows: 

§ 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. 

* * * * * 

■ 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 

■ 7. Section 2.100 is revised to read as 

follows: 

§ 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. 

■ 8. In § 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: 

§ 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 


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

§ 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 § 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 § 51.20(b) of 

this chapter, and is of the type specified 

in § 50.21(b)(2) or (3) or § 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 § 50.30(f) of this chapter, or 

§ 52.80(b) of this chapter, as applicable. 

The other part shall include any 

information required by § 50.34(a) and, 

if applicable, § 50.34a of this chapter, or 

§§ 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 

§§ 50.30(e) and 170.21 of this chapter 

and the information required by 

§§ 50.33, 50.34(a)(1) or 52.79(a)(1), as 

applicable, and § 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 § 51.20(b) of this chapter, 

and is of the type specified in 

§ 50.21(b)(2) or (3) or § 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 


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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 § 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. 

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 

§ 51.20(b) of this chapter and is of the 

type specified in § 50.21(b)(2) or (3) or 

§ 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 §§ 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 §§ 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 

§ 2.603(b)(1). 

(ii) Part two shall include or be 

accompanied by the remaining 

information required by §§ 50.30(f), 

50.33, and 50.34(a)(1) of this chapter. 

(iii) Part three shall include the 

remaining information required by 

§§ 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 §§ 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 §§ 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 

§ 2.621(b)(1). 

(ii) Part two shall include or be 

accompanied by the remaining 

information required by §§ 50.30(f), 

50.33, and 52.79(a)(1) of this chapter. 

(iii) Part three shall include the 

remaining information required by 

§§ 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. 

* * * * * 

■ 9. In § 2.102, paragraph (a) is revised 

to read as follows: 

§ 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. 

* * * * * 

■ 10. Section 2.104 is revised to read as 

follows: 

§ 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 

§ 50.21(b) or § 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 § 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 § 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. 

(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 


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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 § 72.16(e) of 

this chapter. 

■ 11. In § 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: 

§ 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 

§ 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 §§ 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 § 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 § 50.10(e)(1) and the 

Commission determines under §§ 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 § 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 § 52.103(a) of this chapter, the 

following information: 

(i) The identification of the NRC 

action as making the finding required 

under § 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 

§ 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. 

* * * * * 

■ 12. In § 2.106, paragraphs (a) and (b) 

are revised to read as follows: 

§ 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 § 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 

§ 50.21(b) or § 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 § 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 

§ 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. 

* * * * * 

■ 13. Section 2.109 is revised to read as 

follows: 

§ 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 § 52.29 of this chapter 

at least 12 months before the expiration 

of the existing early site permit, the 


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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 § 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. 

■ 14. Section 2.110 is revised to read as 

follows: 

§ 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 

§§ 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 §§ 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. 

■ 15. Section 2.111 is revised to read as 

follows: 

§ 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. 

■ 16. In § 2.202, paragraph (e) is revised 

to read as follows: 

§ 2.202 Orders. 

* * * * * 

(e)(1) If the order involves the 

modification of a part 50 license and is 

a backfit, the requirements of § 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 § 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 § 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 § 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 § 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 § 52.171 of this chapter 

must be followed, unless the applicant 

or licensee has consented to the action 

required. 

■ 17. In § 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: 

§ 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) * * * 


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(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 § 52.99(c)). If the requestor 

identifies a specific portion of the 

§ 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 § 2.310. If a request/ 

petition relies upon § 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. 

■ 18. In § 2.310, paragraph (j) is 

redesignated as paragraph (k), and a 

new paragraph (j) is added to read as 

follows: 

§ 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. 

* * * * * 

■ 19. In § 2.339, paragraph (d) is revised 

to read as follows: 

§ 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. 

■ 20. Section 2.340 is revised to read as 

follows: 

§ 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 § 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 


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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 § 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 

§ 2.345, a petition for review under 

§ 2.341, or a motion for stay under 

§ 2.342, or the filing of a petition under 

§ 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 § 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 

§ 2.345, a petition for review under 

§ 2.341, or a motion for stay under 

§ 2.342, or the filing of a petition under 

§ 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 

§ 2.345, a petition for review under 

§ 2.341, or a motion for stay under 

§ 2.342, or the filing of a petition under 

§ 2.206. 

■ 21. In § 2.341, paragraph (a)(1) is 

revised to read as follows: 

§ 2.341 Review of decisions and actions of 

a presiding officer. 

(a)(1) Except for requests for review or 

appeals under § 2.311 or in a proceeding 

on the high-level radioactive waste 

repository (which are governed by 

§ 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). 

* * * * * 

■ 22. In § 2.347, paragraph (a) is revised, 

and new paragraph (f)(5) is added to 

read as follows: 

§ 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; 


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(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 § 2.340(a). 

* * * * * 

(f) * * * 

(5) Communications, in contested 

proceedings and uncontested mandatory 

proceeding, regarding an undisputed 

issue. 

■ 23. In § 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: 

§ 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 § 2.340(a). 

* * * * * 

(3) Separation of functions does not 

apply to uncontested proceedings, or to 

an undisputed issue in contested initial 

licensing proceedings. 

* * * * * 

■ 24. In § 2.390, the introductory text of 

paragraph (a) is revised to read as 

follows: 

§ 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: 

* * * * * 

■ 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. 

§ 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. 

§ 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 § 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 § 2.104. 

(b) The notice of hearing will also 

state the time and place of the hearings 

on any separate phase of the proceeding. 

§ 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. 

§ 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 § 2.105, a 

notice of proposed action with respect 

to each application as soon as 

practicable after the applications have 

been docketed. 

§ 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 

§ 2.402. 


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1The 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. 

§ 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 § 2.341. If the 

proceedings have also been 

consolidated with respect to matters 

other than the common design under 

§ 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. 

§ 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. 

§ 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. 

■ 26. Section 2.500 is revised to read as 

follows: 

§ 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. 

■ 27. In § 2.501, the section heading, the 

introductory text of paragraph (a) and 

paragraph (b) are revised to read as 

follows: 

§ 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 § 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: 

* * * * * 

(b) The notice of hearing shall comply 

with the requirements of § 2.104(f) of 

this chapter. 

* * * * * 

§ 2.502 [Removed] 

■ 28. Remove and reserve § 2.502. 

§ 2.503 [Removed] 

■ 29. Remove and reserve § 2.503. 

§ 2.504 [Removed] 

■ 30. Remove and reserve § 2.504. 

■ 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. 

§ 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 § 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 § 51.20(b) of this chapter and 

is of the type specified in § 50.21(b)(2) 

or (3) or § 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 §§ 2.601 

through 2.609 apply to all applications 

under this subpart. 

(b) The procedures in §§ 2.611 

through 2.619 apply to applications for 

a permit to construct a utilization 

facility which is subject to § 51.20(b) of 

this chapter and is of the type specified 

in § 50.21(b)(2) or (3) or § 50.22 of this 

chapter or is a testing facility. 

(c) The procedures in §§ 2.621 

through 2.629 apply to applications for 

combined license under part 52 of this 

chapter for a nuclear power facility. 

§ 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 

§ 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 § 50.30(e) and part 

170 of this chapter. 

§ 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 § 2.101(a–1) of this part 

will be initially treated as a tendered 


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application. If it is determined that any 

one of the parts as described in 

§ 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 § 51.20(b) of 

this chapter and is of the type specified 

in § 50.21(b)(2) or (3) or § 50.22 of this 

chapter, or is a testing facility where 

part one of the application as described 

in § 2.101(a–1) is complete. Part one of 

any application will not be considered 

complete unless it contains proposed 

findings as required by § 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 longrange plans for ultimate development of 

the site. Upon assignment of a docket 

number, the procedures in § 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. 

§ 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 § 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 § 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 §§ 2.101(a–

1) and 2.603, a supplementary notice of 

hearing will be published under § 2.104 

with respect to the remaining 

unresolved issues in the proceeding 

within the scope of § 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 § 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 § 2.315(c) and for 

limited appearances under § 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 nontimely 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 §§ 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. 

§ 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 § 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 § 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. 

§ 2.606 Partial decision on site suitability 

issues in construction permit proceeding. 

(a) The provisions of §§ 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 


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3The 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. 

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 § 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. 

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. 

(a) Each part of an application 

submitted in accordance with § 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 § 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 

§ 2.101(a–1) is complete. Part one of any 

application will not be considered 

complete unless it contains proposed 

findings as required by § 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 longrange plans for ultimate development of 

the site. Upon assignment of a docket 

number, the procedures in § 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. 

§ 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 § 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 

§ 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 § 2.315(c) and for 

limited appearances under § 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 §§ 2.101(a–

1) and 2.603, a supplementary notice of 

hearing will be published under § 2.104 

with respect to the remaining 

unresolved issues in the proceeding 

within the scope of § 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 § 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 § 2.315(c) and for 

limited appearances under § 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 § 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 § 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 §§ 2.309(c)(1)(i) 


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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. 

§ 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 § 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 § 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. 

§ 2.627 Partial decision on site suitability 

issues in combined license proceeding. 

(a) The provisions of §§ 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 § 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. 

§ 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 § 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 

§ 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. 

■ 32. Section 2.800 is revised to read as 

follows: 

§ 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 §§ 2.804 

through 2.810 apply to all rulemakings. 

(c) The procedures in §§ 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 §§ 2.811 

through 2.819, as supplemented by the 


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provisions of subpart B of part 52, apply 

to standard design certification 

rulemaking. 

■ 33. Section 2.801 is revised to read as 

follows: 

§ 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. 

■ 34. In subpart H, §§ 2.811, 2.813, 

2.815, 2.817 and 2.819 are added to read 

as follows: 

§ 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, email, 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/esubmittals.html, by calling (301) 415–

0439, by e-mail at EIE@nrc.gov, 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 § 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 § 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 NRCREP@nrc.gov 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 § 2.802(a)(2) do not apply. 

§ 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/esubmittals.html, by calling (301) 415–

0439, by e-mail at EIE@nrc.gov, 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. 

§ 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. 

§ 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 


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

§ 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. 

■ 35. In § 2.1202, paragraph (a) is 

revised to read as follows: 

§ 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. 

* * * * * 

§ 2.1211 [Removed] 

■ 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 

■ 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. 

■ 38. In § 10.1, paragraphs (a)(1) and 

(a)(2) are revised and paragraph (a)(3) is 

added to read as follows: 

§ 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. 

* * * * * 

■ 39. In § 10.2, paragraph (b) is revised 

to read as follows: 

§ 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 

■ 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). 

■ 41. Section 19.1 is revised to read as 

follows: 

§ 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 


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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. 

■ 42. Section 19.2 is revised to read as 

follows: 

§ 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. 

■ 43. In § 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: 

§ 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. 

■ 44. In § 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: 

§ 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 

forms@nrc.gov, 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 NRClicensed 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. 

■ 45. Section 19.14 is revised to read as 

follows: 

§ 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 


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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 § 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 § 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. 

■ 46. Section 19.20 is revised to read as 

follows: 

§ 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 § 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 § 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. 

■ 47. Section 19.31 is revised to read as 

follows: 

§ 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. 

■ 48. Section 19.32 is revised to read as 

follows: 

§ 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 

■ 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). 

■ 50. Section 20.1002 is revised to read 

as follows: 

§ 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 § 35.75, or 

to exposure from voluntary 

participation in medical research 

programs. 

■ 51. In § 20.1401 paragraph (a) is 

revised to read as follows: 

§ 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 

§ 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. 

* * * * * 

■ 52. Section 20.1406 is revised to read 

as follows: 

§ 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 


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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. 

■ 53. In § 20.2203, paragraphs (c) and 

(d) are revised to read as follows: 

§ 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 §§ 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 § 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/esubmittals.html, by calling (301) 415–

0439, by e-mail to EIE@nrc.gov, 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 

■ 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). 

■ 55. In § 21.2, paragraphs (a), (b), and 

(c) are revised to read as follows: 

§ 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 § 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 

§ 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 § 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 §§ 50.72, 

50.73, or § 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. 

* * * * * 

■ 56. In § 21.3 the definitions of basic 

component, defect, deviation, and 

substantial safety hazard are revised to 

read as follows: 

§ 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 safeshutdown condition; or 


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(C) The capability to prevent or 

mitigate the consequences of accidents 

which could result in potential offsite 

exposures comparable to those referred 

to in §§ 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 safeshutdown 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 §§ 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. 

* * * * * 

■ 57. Section 21.5 is revised to read as 

follows: 

§ 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 

EIE@nrc.gov, 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. 

■ 58. In § 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: 

§ 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 


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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. 

* * * * * 

■ 59. In § 21.51 paragraphs (a)(4) and 

(a)(5) are added and paragraph (b) is 

revised to read as follows: 

§ 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. 

■ 60. In § 21.61, paragraph (b) is revised 

to read as follows: 

§ 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 § 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 

■ 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). 

■ 62. The heading of part 25 is revised 

to read as set forth above. 

■ 63. In § 25.35, paragraph (a) is revised 

to read as follows: 

§ 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 

■ 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). 

■ 65. In § 26.2, the introductory text of 

paragraph (a), and paragraph (c) are 

revised to read as follows: 

§ 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 

§ 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 fitnessfor-duty program which complies with 

this part. The provisions of the fitnessfor-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 

§ 52.103(g) of this chapter, holders of 

manufacturing licenses under part 52, 

and persons authorized to conduct the 

activities under § 50.10(e)(3) of this 

chapter. Each licensee with a 

construction permit, a combined license 

before the Commission makes the 

finding under § 52.103(g) of this 

chapter, a manufacturing license, or 

person authorized to conduct the 

activities under § 50.10(e)(3) of this 

chapter, with a plant or reactor under 

active construction or manufacture, 

shall—

(1) Comply with §§ 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. 

* * * * * 

■ 66. In § 26.10, paragraph (a) is revised 

to read as follows: 


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§ 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 § 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; 

* * * * * 

■ 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 § 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 

■ 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). 

■ 69. In Section 50.2, definitions of 

applicant, license, licensee, and 

prototype plant, are added to read as 

follows: 

§ 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 § 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. 

* * * * * 

■ 70. In § 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: 

§ 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 § 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 

§ 51.20(b) of this chapter, and is of the 

type specified in §§ 50.21(b)(2) or (3), or 

§ 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 §§ 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 


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Regulation, as appropriate, may 

authorize an applicant for a 

construction permit for a utilization 

facility which is subject to § 51.20(b) of 

this chapter, and is of the type specified 

in §§ 50.21(b)(2) or (3), or § 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. 

* * * * * 

■ 71. Section 50.23 is revised to read as 

follows: 

§ 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 § 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 

§ 50.91. 

■ 72. The undesignated center heading 

before § 50.30 is revised to read as 

follows: 

Applications for Licenses, 

Certifications, and Regulatory 

Approvals; Form; Contents; Ineligibility 

of Certain Applicants 

■ 73. In § 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: 

§ 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 § 50.4 or § 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 

§ 50.4 or § 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 

§ 2.101(a) of this chapter. Copies must 

be submitted to the Commission, as 

specified in § 50.4 or § 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 § 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. 

■ 74. In § 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: 

§ 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 


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4Emergency 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. 

(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 § 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. 

(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 § 50.75, indicating how 

reasonable assurance will be provided 

that funds will be available to 

decommission the facility. 

* * * * * 

■ 75. In § 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: 

§ 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 § 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 § 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 §§ 73.50, 73.55, or 

§ 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 


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9A physical security plan that contains all the 

information required in both § 73.55 and appendix 

C to part 73 of this chapter satisfies the requirement 

for a contingency plan. 

10Alphanumeric 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. 

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

(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 § 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-waterreactor 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. 

* * * * * 

(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 § 50.44 as a part of 

their application. 

(h) * * * 

(1) * * * 

(ii) Applications for light-watercooled 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. 

* * * * * 

■ 76. Section 50.34a is revised to read 

as follows: 

§ 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: 


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(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. 

■ 77. In § 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: 

§ 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. 

* * * * * 

■ 78. In § 50.36a, paragraph (a) is 

revised to read as follows: 

§ 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 § 20.1301 of this chapter, 

require that: 

(1) Operating procedures developed 

pursuant to § 50.34a(c) for the control of 

effluents be established and followed 

and that the radioactive waste system, 

pursuant to § 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 § 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 § 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. 

* * * * * 

■ 79. Section 50.36b is revised to read 

as follows: 

§ 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 § 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 

§ 50.82(a)(1) or § 52.110(a) of this 

chapter has been submitted, which is of 

a type described in § 50.21(b)(2) or (3) 

or § 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 §§ 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. 

■ 80. Section 50.37 is revised to read as 

follows: 

§ 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. 

■ 81. The undesignated center heading 

before § 50.40 is revised to read as 

follows: 

Standards for Licenses, Certifications, 

and Regulatory Approvals 

■ 82. Section 50.40 is revised to read as 

follows: 

§ 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 


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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 

§ 50.21(b) or § 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. 

■ 83. In § 50.43, the section heading, the 

introductory paragraph, and paragraph 

(d) are revised, and paragraph (e) is 

added to read as follows: 

§ 50.43 Additional standards and 

provisions affecting class 103 licenses and 

certifications for commercial power. 

In addition to applying the standards 

set forth in §§ 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. 

■ 84. Section 50.45 is revised to read as 

follows: 

§ 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 §§ 50.31 

through 50.38, and the standards of 

§§ 50.40 through 50.43, as applicable. 

(b) A holder of a combined license 

who proposes, after the Commission 

makes the finding under § 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 §§ 50.30 through 50.33, 

§ 50.34(f), §§ 50.34a through 50.38, the 

standards of §§ 50.40 through 50.43, as 

applicable, and §§ 52.79 and 52.80 of 

this chapter. 

■ 85. In § 50.46, paragraph (a)(3) is 

revised to read as follows: 

§ 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 °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 °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 § 50.4 or § 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 

§ 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 

§§ 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 § 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 


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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 § 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 § 50.46 requirements. 

The affected applicant or holder shall 

propose immediate steps to demonstrate 

compliance or bring plant design into 

compliance with § 50.46 requirements. 

* * * * * 

■ 86. In § 50.47, paragraph (a)(1) is 

revised and paragraph (e) is added to 

read as follows: 

§ 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 § 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 § 50.54(gg). 

■ 87. In § 50.48, the introductory text of 

paragraph (a)(1) is revised and 

paragraph (a)(4) is added to read as 

follows: 

§ 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. 

* * * * * 

■ 88. In § 50.49, paragraph (a) is revised 

to read as follows: 

§ 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 § 50.82(a)(1) or 

§ 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. 

* * * * * 

■ 89. In § 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: 

§ 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 § 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 § 50.34(b)(6)(ii) or § 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 § 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 

§ 55.59(c) of this chapter. 

Notwithstanding the provisions of 

§ 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 §§ 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 


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49496 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

qualification plan, prepared pursuant to 

§ 50.34(c) or § 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 § 50.34(d) or 

§ 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 § 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 § 50.4 or § 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 § 52.103(g) of this 

chapter, shall follow and maintain in 

effect emergency plans which meet the 

standards in § 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 § 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 § 50.4, a report of each 

proposed change for approval. If a 

change is made without approval, the 

licensee shall submit, as specified in 

§ 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 § 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. 


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■ 90. In § 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: 

§ 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 § 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 § 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 


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49498 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

obligations under part 21 of this 

chapter, and satisfies the responsibility 

of individual directors or responsible 

officers of holders of construction 

permits issued under § 50.23, holders of 

combined licenses (until the 

Commission makes the finding under 

§ 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 § 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 

§ 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 § 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 § 73.71 of this chapter, or under 

§§ 50.55(e) or 50.73. For holders of 

construction permits issued before 

October 29, 1991, evaluation, reporting 

and recordkeeping requirements of 

§ 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 § 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 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 49499 

9See footnotes at end of section. 

approval before implementation, as 

follows: 

(A) Changes to the safety analysis 

report must be submitted for review as 

specified in § 50.4. Changes made to 

NRC-accepted quality assurance topical 

report descriptions must be submitted 

as specified in § 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] 

■ 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: 

§ 50.55a Codes and standards. 

Each construction permit for a 

utilization facility is subject to the 

following conditions in addition to 

those specified in § 50.55. Each 

combined license for a utilization 

facility is subject to the following 

conditions in addition to those specified 

in § 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 § 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 § 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 B9 must meet the requirements 

for Class 2 Components in Section III of 

the ASME Boiler and Pressure Vessel 

Code. 

* * * * * 

(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 C9 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 watercooled 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) 


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2Changes to RTPTS 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. 

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 watercooled 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 watercooled 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. 

■ 92. In § 50.59, paragraphs (b), (d)(2), 

and (d)(3) are revised to read as follows: 

§ 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 

§ 50.82(a)(1) or § 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 § 50.4 or § 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. 

■ 93. In § 50.61, paragraph (b)(1) is 

revised to read as follows: 

§ 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 

§ 50.82(a)(1) have been submitted, the 

licensee shall have projected values of 

RTPTS, accepted by the NRC, for each 

reactor vessel beltline material for the 

EOL fluence of the material. The 

assessment of RTPTS 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 RTPTS 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 RTPTS, or upon request for a change 

in the expiration date for operation of 

the facility. 

* * * * * 

■ 94. In § 50.62, paragraph (d) is revised 

to read as follows: 

§ 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 lightwater-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 § 50.4, a 

proposed schedule for meeting the 


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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. 

■ 95. In § 50.63, the introductory text of 

paragraphs (a)(1) and (c)(1) are revised 

to read as follows: 

§ 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 

§ 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 § 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-watercooled 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. 

* * * * * 

■ 96. In § 50.65, paragraph (a)(1) is 

revised to read as follows: 

§ 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 § 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 industrywide 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 

§ 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. 

* * * * * 

■ 97. In § 50.70 paragraphs (a) and (b)(2) 

are revised to read as follows: 

§ 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. 

* * * * * 

■ 98. In § 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: 

§ 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 § 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 


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1Effects of changes includes appropriate 

revisions of descriptions in the FSAR such that the 

FSAR (as updated) is complete and accurate. 

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 § 50.21 or § 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 § 50.59(c)(2) or, in the 

case of a license that references a 

certified design, in accordance with 

§ 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. 

* * * * * 

(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 

§ 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 

§ 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 

§ 52.171 of this chapter or the analyses 

requested by the Commission under 

§ 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 §§ 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 § 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. 

■ 99. In § 50.72, the introductory text of 

paragraph (a)(1) is revised to read as 

follows: 

§ 50.72 Immediate notification 

requirements for operating nuclear power 

reactors. 

(a) * * * 

(1) Each nuclear power reactor 

licensee licensed under §§ 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 

§ 52.103(g), shall notify the NRC 

Operations Center via the Emergency 

Notification System of: 

* * * * * 

■ 100. In § 50.73, paragraph (a)(1) is 

revised to read as follows: 

§ 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 

§ 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 

§ 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. 

* * * * * 

■ 101. In § 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: 

§ 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. 


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(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 § 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 § 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 § 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 § 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. 

* * * * * 

■ 102. Section 50.78 is revised to read 

as follows: 

§ 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 § 75.6 and §§ 75.11 through 

75.14 of this chapter. 

■ 103. In § 50.80, paragraphs (a) and (b) 

are revised to read as follows: 

§ 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 

§§ 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 §§ 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 §§ 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 §§ 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, 


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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 § 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. 

* * * * * 

■ 104. In § 50.81, paragraph (d)(1) is 

revised, and a new paragraph (d)(3) is 

added to read as follows: 

§ 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. 

■ 105. Section 50.90 is revised to read 

as follows: 

§ 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 §§ 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. 

■ 106. In § 50.91, the introductory text 

is revised to read as follows: 

§ 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 

§§ 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 § 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: 

* * * * * 

■ 107. Section 50.92 paragraph (a), and 

the introductory text of paragraph (c) are 

revised to read as follows: 

§ 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 § 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 § 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 

§ 50.91, that a proposed amendment to 

an operating license or a combined 

license for a facility or reactor licensed 

under §§ 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: 

* * * * * 

■ 108. Section 50.100 is revised to read 

as follows: 

§ 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 §§ 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 § 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. 

■ 109. In § 50.109, paragraph (a)(1) is 

revised to read as follows: 

§ 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 


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(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 § 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 § 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 § 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 

§ 52.171 of this chapter apply with 

respect to matters resolved in the 

manufacturing license proceeding. 

* * * * * 

■ 110. Section 50.120 is revised to read 

as follows: 

§ 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 § 50.21(b) 

or § 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. 

■ 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 § 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 § 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 § 50.2 for the duration of the 

accident. 

* * * * * 

■ 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 § 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 

§ 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 § 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 


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1While 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. 

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 safetyrelated 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 safetyrelated 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. 

* * * * * 

■ 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 

§ 50.21(b) or § 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 § 50.21(b) or § 50.22 or a testing 

facility, and each holder of a combined 

license issued under part 52 of this chapter, 

is required by § 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. 

* * * * * 

■ 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 § 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 § 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 § 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 § 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 § 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) 


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4Full 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. 

5Partial 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. 

6Co-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. 

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. 

(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 

§ 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 

§ 50.54(gg) apply. 

* * * * * 

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: 

(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 § 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 § 50.4, 

within 30 days of such changes. 

* * * * * 

■ 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-WaterCooled 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 

§ 50.34a(a), or by an applicant for a combined 

license under part 52 of this chapter as 

guidance in meeting the requirements of 

§ 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 § 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 lightwater-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 § 50.82(a)(1) or 

§ 52.110 of this chapter as guidance in 

developing technical specifications under 

§ 50.36a(a) to keep levels of radioactive 

materials in effluents to unrestricted areas as 

low as is reasonably achievable. 

* * * * * 


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3Specific guidance concerning a performancebased 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.’’

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-watercooled nuclear power reactor under part 52 

of this chapter. 

* * * * * 

■ 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 

§ 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 § 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

* * * * * 

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 performancebased 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] 

■ 117. Appendix M to Part 50 is 

removed and reserved. 

■ 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] 

■ 119. Appendix O to Part 50 is 

removed and reserved. 

■ 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 §§ 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 

§ 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. 


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Response spectrum is a plot of the 

maximum responses (acceleration, velocity, 

or displacement) of idealized single-degreeof-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 safeshutdown 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 § 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 

■ 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)). 

■ 122. In § 51.17, paragraph (b) is 

revised to read as follows: 

§ 51.17 Information collection 

requirements; OMB approval. 

* * * * * 

(b) The approved information 

collection requirements in this part 

appear in §§ 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. 

■ 123. In § 51.20, paragraphs (b)(1) and 

(b)(2) are revised, and paragraph (b)(6) 

is removed and reserved. 

The revisions read as follows: 

§ 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] 

* * * * * 

■ 124. In § 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: 

§ 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 § 52.103(g) of 

this chapter. 

* * * * * 

■ 125. In § 51.23 paragraphs (b) and (c) 

are revised to read as follows: 

§ 51.23 Temporary storage of spent fuel 

after cessation of reactor operation—

generic determination of no significant 

environmental impact. 

* * * * * 

(b) Accordingly, as provided in 

§§ 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 


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49510 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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. 

■ 126. In § 51.26, a new paragraph (d) is 

added to read as follows: 

§ 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 § 51.27, and will be 

published in the Federal Register as 

provided in § 51.116. The NRC staff 

need not conduct a scoping process (see 

§§ 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 

§ 51.92(e). 

■ 127. In § 51.27, the introductory text 

of paragraph (a) is revised, and a new 

paragraph (b) is added to read as 

follows: 

§ 51.27 Notice of intent. 

(a) The notice of intent required by 

§ 51.26(a) shall: 

* * * * * 

(b) The notice of intent required by 

§ 51.26(d) shall: 

(1) State that a supplement to a final 

environmental impact statement will be 

prepared in accordance with § 51.72 or 

§ 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 

§ 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. 

■ 128. In § 51.29, the section heading 

and paragraph (a)(1) are revised to read 

as follows: 

§ 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 § 51.92(e). 

* * * * * 

■ 129. In § 51.30, the introductory text 

of paragraph (a) is revised, and 

paragraphs (d) and (e) are added to read 

as follows: 

§ 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. 

■ 130. Section 51.31 is revised to read 

as follows: 

§ 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 


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Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 49511 

impact on the proposed action. As 

provided in § 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 

§ 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 § 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 

§ 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 § 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 

§ 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 § 2.315(c) of this chapter, 

may take a position and offer evidence 

on the matters within the scope of the 

environmental assessment. 

■ 131. In § 51.32, paragraph (b) is added 

to read as follows: 

§ 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. 

■ 132. In § 51.45, paragraphs (a) and (c) 

are revised to read as follows: 

§ 51.45 Environmental report. 

(a) General. As required by §§ 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 § 51.50(b), or environmental 

reports prepared at the license renewal 

stage under § 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 § 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 § 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. 

* * * * * 

■ 133. Section 51.50 is revised to read 

as follows: 

§ 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 § 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 §§ 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 


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environment, proposed for possible 

inclusion in the license as 

environmental conditions in accordance 

with § 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 §§ 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 § 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 §§ 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 § 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. 

■ 134. In § 51.51 paragraph (a) is revised 

to read as follows: 

§ 51.51 Uranium fuel cycle environmental 

data—Table S–3. 

(a) Under § 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. 

* * * * * 

■ 135. In § 51.52, the introductory 

paragraph is revised to read as follows: 

§ 51.52 Environmental effects of 

transportation of fuel and waste—Table 

S–4. 

Under § 51.50, every environmental 

report prepared for the construction 

permit stage or early site permit stage or 

combined license stage of a light-waterVerDate Aug<31>2005 17:54 Aug 27, 2007 Jkt 211001 PO 00000 Frm 00162 Fmt 4701 Sfmt 4700 E:\FR\FM\28AUR2.SGM 28AUR2

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

* * * * * 

■ 136. In § 51.53, paragraph (a) and the 

introductory text of paragraph (c)(3) are 

revised to read as follows: 

§ 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: 

* * * * * 

■ 137. Section 51.54 is revised to read 

as follows: 

§ 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. 

■ 138. Section 51.55 is redesignated as 

§ 51.58, and is revised to read as 

follows: 

§ 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 §§ 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 § 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/esubmittals.html, by calling (301) 415–

0439, by e-mail to EIE@nrc.gov, 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 § 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. 

■ 139. Section 51.55 is added to read as 

follows: 


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3Compliance 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. 

§ 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. 

■ 140. Section 51.66 is revised to read 

as follows: 

§ 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 

§§ 51.60(b)(1) through (6); or by §§ 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 § 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. 

■ 141. In § 51.71 paragraph (d) and 

Footnote 3 are revised to read as 

follows: 

§ 51.71 Draft environmental impact 

statement—contents. 

* * * * * 

(d) Analysis. Unless excepted in this 

paragraph or § 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 § 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 § 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. 

* * * * * 

■ 142. Section 51.75 is revised to read 

as follows: 

§ 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 

§§ 51.70, 51.71, 51.72, and 51.73. The 

contribution of the environmental 

effects of the uranium fuel cycle 

activities specified in § 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


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5 Values for releases of Rn-222 and Tc-99 are not 

given in the table. The amount and significance of 

Rn-222 releases from the fuel cycle and Tc-99 

releases from waste management or reprocessing 

activities shall be considered in the draft 

environmental impact statement and may be the 

subject of litigation in individual licensing 

proceedings. 

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 

§§ 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 § 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. 

(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 

§§ 51.70, 51.71, 51.72, and 51.73. The 

contribution of the environmental 

effects of the uranium fuel cycle 

activities specified in § 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 

§ 51.23(a) and in accordance with 

§ 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 § 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. 

§ 51.76 [Removed] 

■ 143. Section 51.76 is removed and 

reserved. 

■ 144. Section 51.92 is revised to read 

as follows: 

§ 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 

§ 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 § 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; 


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(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 

§ 51.73 and a notice of availability will 

be published in the Federal Register as 

provided in § 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 § 51.118. 

■ 145. In § 51.95, paragraph (a), the 

introductory text of paragraph (c), and 

paragraph (d) are revised to read as 

follows: 

§ 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 staffprepared 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 § 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 § 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 § 51.73. Unless otherwise 

required by the Commission in 

accordance with the generic 

determination in § 51.23(a) and the 

provisions of § 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. 

■ 146. Section 51.105 is revised to read 

as follows: 

§ 51.105 Public hearings in proceedings 

for issuance of construction permits or 

early site permits. 

(a) In addition to complying with 

applicable requirements of § 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. 

■ 147. Section 51.105a is added to read 

as follows: 

§ 51.105a Public hearings in proceedings 

for issuance of manufacturing licenses. 

In addition to complying with 

applicable requirements of § 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 


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should be issued as proposed by the 

NRC’s Director of New Reactors or 

Director of Nuclear Reactor Regulation. 

■ 148. Section 51.107 is added under 

the undesignated center heading 

‘‘Production and Utilization Facilities’’

to read as follows: 

§ 51.107 Public hearings in proceedings 

for issuance of combined licenses. 

(a) In addition to complying with the 

applicable requirements of § 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 § 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. 

■ 149. Section 51.108 is added under 

the undesignated center heading 

‘‘Production and Utilization Facilities,’’

to read as follows: 

§ 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). 

■ 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. 


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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 

§ 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. 

§ 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-akind 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. 


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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. 

§ 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. 

§ 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 

EIE@nrc.gov, 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 

§ 52.79 of this chapter; 

(ii) Safeguards contingency plan 

under § 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 § 50.54(p) of this chapter; 

(iv) Application for amendment of 

physical security plan, guard training 

and qualification plan, or safeguards 

contingency plan under § 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 § 52.17(b) or 

§ 52.79(a); 

(ii) Change to an emergency plan 

under § 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 § 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 


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Electronic Submissions to the 

Commission. 

(7) Quality assurance related 

submissions. (i) A change to the safety 

analysis report quality assurance 

program description under § 50.54(a)(3) 

or § 50.55(f)(4) of this chapter, or a 

change to a licensee’s NRC-accepted 

quality assurance topical report under 

§ 50.54(a)(3) or § 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 § 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 

§ 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. 

§ 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. 

§ 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 


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(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 forms@nrc.gov, 

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. 

§ 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. 

§ 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 § 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 


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1The 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. 

governed by § 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. 

§ 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. 

§ 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. 

§ 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. 

§ 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 §§ 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 

§ 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. 

§ 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. 

§ 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 

§§ 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. 

§ 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. 

§ 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 


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2A 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. 

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: 

(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). 

(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 §§ 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. 

§ 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 


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consultation with DHS, whether the 

information required of the applicant by 

§ 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 § 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 § 52.17(b)(2)(ii) provide 

reasonable assurance that adequate 

protective measures can and will be 

taken in the event of a radiological 

emergency. 

§ 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 § 2.101(a)(1) through (4) of 

this chapter, and the requirements for 

issuance of a notice of hearing in 

§§ 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. 

§ 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. 

§ 52.24 Issuance of early site permit. 

(a) After conducting a hearing under 

§ 52.21 and receiving the report to be 

submitted by the ACRS under § 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 § 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 § 52.17(c) that the 

permit holder is authorized to perform. 

§ 52.25 Extent of activities permitted. 

If the activities authorized by 

§ 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 § 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. 

§ 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. 

§ 52.28 Transfer of early site permit. 

An application to transfer an early site 

permit will be processed under 10 CFR 

50.80. 

§ 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 § 52.27(b). 

(d) The Commission shall refer a copy 

of the application for renewal to the 


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ACRS. The ACRS shall report on those 

portions of the application which 

concern safety and shall apply the 

criteria set forth in § 52.31. 

§ 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 § 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. 

§ 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. 

§ 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 

§ 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. 

§ 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 §§ 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 § 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 

§ 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 § 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 permitemergency 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 § 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 § 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 


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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 

§ 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. 

§ 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. 

§ 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 

§§ 52.3 and §§ 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. 

§ 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). 

§ 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 


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3The 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. 

4A 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. 

(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: 

(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); 

(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 watercooled 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 §§ 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 

§ 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-watercooled 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 §§ 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 § 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 

§ 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 

§ 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 


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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 designspecific 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. 

§ 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. 

§ 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 § 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. 

§ 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. 

§ 52.54 Issuance of standard design 

certification. 

(a) After conducting a rulemaking 

proceeding under § 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 

§ 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; 


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(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. 

§ 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 § 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. 

§ 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 § 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 

§ 52.59. 

§ 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. 

§ 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. 

§ 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 §§ 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 

§ 52.63(a)(1), except for § 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. 


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(ii) The rulemaking procedures for 

changes under § 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 § 52.55 or § 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 § 52.7, the 

Commission shall consider whether the 

special circumstances which § 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 § 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 § 52.7. In addition to 

the factors listed in § 52.7, the 

Commission shall consider whether the 

special circumstances that § 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 § 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 

§ 52.71 Scope of subpart. 

This subpart sets out the requirements 

and procedures applicable to 

Commission issuance of combined 

licenses for nuclear power facilities. 

§ 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. 

§ 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 

§§ 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. 

§ 52.77 Contents of applications; general 

information. 

The application must contain all of 

the information required by 10 CFR 

50.33. 

§ 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) 


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5The 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. 

6A 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. 

7The 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. 

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: 

(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). 

(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; 

(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 §§ 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 

§ 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 

§§ 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 § 50.44 of this 

chapter; 

(9) The coping analyses, and any 

design features necessary to address 

station blackout, as described in § 50.63 

of this chapter; 

(10) A description of the program, and 

its implementation, required by 

§ 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 


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8A physical security plan that contains all the 

information required in both § 73.55 of this chapter 

and appendix C to 10 CFR part 73 satisfies the 

requirement for a contingency plan. 

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 § 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 § 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 § 50.34(f) of this 

chapter, with the exception of 

§§ 50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v); 

(18) If the applicant seeks to use riskinformed treatment of SSCs in 

accordance with § 50.69 of this chapter, 

the information required by § 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 § 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 § 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 

§ 52.79(a)(21); 

(30) Proposed technical specifications 

prepared in accordance with the 

requirements of §§ 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 § 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 § 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 § 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.) 

(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 


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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 § 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 

§ 20.1101 of this chapter and its 

implementation. 

(40) A description of the fire 

protection program required by § 50.48 

of this chapter and its implementation. 

(41) For applications for light-watercooled 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 § 50.62 of this 

chapter; 

(43) Information demonstrating how 

the applicant will comply with 

requirements for criticality accidents in 

§ 50.68 of this chapter; 

(44) A description of the fitness-forduty program required by 10 CFR part 

26 and its implementation. 

(45) The information required by 

§ 20.1406 of this chapter. 

(46) A description of the plantspecific 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 §§ 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 § 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 § 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 § 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 sitespecific 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 plantspecific 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 § 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 

§ 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. 


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(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 § 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. 

§ 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 § 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. 

§ 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. 

§ 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 

§§ 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. 

§ 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 (§ 2.101 of this chapter) and 

issuance of a notice of hearing (§ 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. 

§ 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 

§ 52.81, in accordance with the finality 

provisions in § 52.83. 

§ 52.89 [Reserved]. 

§ 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. 

§ 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 


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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 § 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 § 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. 

§ 52.97 Issuance of combined licenses. 

(a)(1) After conducting a hearing in 

accordance with § 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 § 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. 

§ 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 § 52.103 or § 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., §§ 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 

§ 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 


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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. 

§ 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 

§ 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 

§ 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 § 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 

§ 52.103(a), make available all licensee 

notifications under paragraphs (c)(1) 

and (c)(2) of this section. 

§ 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 § 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. 


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(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 

§ 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 

§§ 52.98(e) or (f), as applicable. 

§ 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 § 52.103(g) or allowing 

operation during an interim period 

under the combined license under 

§ 52.103(c). 

§ 52.105 Transfer of combined license. 

A combined license may be 

transferred in accordance with § 50.80 

of this chapter. 

§ 52.107 Application for renewal. 

The filing of an application for a 

renewed license must be in accordance 

with 10 CFR part 54. 

§ 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. 

§ 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 

§ 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 § 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 sitespecific 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 § 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 § 52.110(a)(1), have been 

submitted. 

(f) Licensees shall not perform any 

decommissioning activities, as defined 

in § 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 

§ 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 § 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 § 50.75 of this 

chapter may be used for 

decommissioning planning. For 

licensees that have submitted the 

certifications required under § 52.110(a) 

and commencing 90 days after the NRC 

has received the PSDAR, an additional 


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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 § 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 

§ 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 § 50.22 of this chapter or 

major portions thereof. 

§ 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. 

§ 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. 

§ 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). 

§ 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 


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9The 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. 

10A 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. 

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: 

(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 

(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 watercooled 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 

§ 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-watercooled 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 

§ 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 § 50.62; 


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(16) The coping analysis, and any 

design features necessary to address 

station blackout, as described in § 50.63 

of this chapter; 

(17) Information demonstrating how 

the applicant will comply with 

requirements for criticality accidents in 

§ 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 § 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 designspecific 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). 

§ 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. 

§ 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. 

§ 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 § 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. 

§ 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. 

§ 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 

§ 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. 

§ 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. 

§ 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 

§§ 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. 

§ 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). 

§ 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 


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11The 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. 

12A 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. 

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: 

(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); 

(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 §§ 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 

§ 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 

§§ 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 § 50.44 of this 

chapter; 

(5) The coping analysis, and any 

design features necessary to address 

station blackout, as described in § 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 § 50.62; 

(8) Information demonstrating how 

the applicant will comply with 

requirements for criticality accidents in 

§ 50.68(b)(2)–(b)(4); 

(9) The information required by 

§ 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 § 50.34a(e) of this chapter; 


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(12) The information necessary to 

demonstrate compliance with any 

technically relevant portions of the 

Three Mile Island requirements set forth 

in § 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 riskinformed treatment of SSCs in 

accordance with § 50.69 of this chapter, 

the information required by § 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 § 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 §§ 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 § 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-watercooled 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 designspecific probabilistic risk assessment 

and its results. 

§ 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 


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criterion has been met. The Federal 

Register notification required by 

§ 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. 

§ 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. 

§ 52.161 Reserved. 

§ 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 

§ 2.101(a)(1) through (4) of this chapter, 

and the requirements for issuance of a 

notice of proposed action in § 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. 

§ 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. 

§ 52.167 Issuance of manufacturing 

license. 

(a) After completing any hearing 

under § 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 NRCapproved shipping requirements in the 

manufacturing license. 

§ 52.169 [Reserved]. 

§ 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 § 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. 


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(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. 

§ 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. 

§ 52.175 Transfer of manufacturing 

license. 

A manufacturing license may be 

transferred in accordance with § 50.80 

of this chapter. 

§ 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 § 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 

§ 52.159. 

§ 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. 

§ 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 §§ 52.171 and 

52.175. 

Subpart G—Reserved 

Subpart H—Enforcement 

§ 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. 

§ 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: §§ 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, 


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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 designrelated 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 designrelated 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 §§ 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. 


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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 nondisclosure 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 plantspecific order while this appendix is in effect 

under §§ 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 


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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 exvessel 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 plantspecific 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. 


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1 ‘‘System 80+’’ is a trademark of Westinghouse 

Electric Company LLC. 

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 § 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 plantspecific 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 sitespecific 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. 

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 designrelated 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 designrelated 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 §§ 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. 


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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 


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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 nondisclosure 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 plantspecific order while this appendix is in effect 

under §§ 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 exvessel 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* 


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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 plantspecific 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 § 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 plantspecific 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 


rwilkins on PROD1PC63 with RULES2
49552 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

1AP600 is a trademark of Westinghouse Electric 

Company LLC. 

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 sitespecific 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. 

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 designrelated 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 designrelated 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 §§ 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 


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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 nondisclosure 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 plantspecific order while this appendix is in effect 

under §§ 52.55 or 52.61, unless: 


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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 exvessel 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 plantspecific 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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1AP1000 is a trademark of Westinghouse Electric 

Company LLC. 

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 § 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 plantspecific 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 sitespecific 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. 

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 designrelated 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 designrelated 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 


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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 §§ 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 

LIBRARY@NRC.GOV 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 


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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 nondisclosure 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 plantspecific 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 exvessel 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 


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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 plantspecific 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 plantspecific 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 § 52.103(a) 

hearing, their expiration will occur upon 

final Commission action in such a 


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1 If the design for the power reactor(s) proposed 

in a particular application is not identical to the 

others, that application may not be processed under 

this appendix and subpart D of part 2 of this 

chapter. 

2As 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. 

proceeding. However, subsequent 

modifications must comply with the Tier 1 

and Tier 2 design descriptions in the plantspecific 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 plantspecific 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 sitespecific 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 (§§ 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. 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 § 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 §§ 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

4. Each combined license application 

submitted pursuant to this appendix must 

contain an environmental report as required 

by § 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 § 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 

■ 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, 


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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. 

■ 152. Section 54.1 is revised to read as 

follows: 

§ 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) 

■ 153. In § 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: 

§ 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 plantspecific 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. 

* * * * * 

■ 154. In § 54.17, paragraph (c) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 155. Section 54.27 is revised to read 

as follows: 

§ 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. 

■ 156. In Section 54.31, paragraphs (a), 

(b), and (c) are revised to read as 

follows: 

§ 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. 

* * * * * 

■ 157. Section 54.35 is revised to read 

as follows: 

§ 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. 

■ 158. In § 54.37, paragraph (a) is 

revised to read as follows: 

§ 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 

■ 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). 

■ 160. In § 55.1, paragraph (a) is revised 

to read as follows: 

§ 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, 

* * * * * 

■ 161. In § 55.2, paragraph (a) is revised 

to read as follows: 

§ 55.2 Scope. 

* * * * * 

(a) Any individual who manipulates 

the controls of any utilization facility 

licensed under parts 50, 52, or 54 of this 

chapter, 

* * * * * 

■ 162. In § 55.5, paragraph (b)(1) and the 

introductory text of paragraph (b)(2) are 

revised to read as follows: 

§ 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 


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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 

■ 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). 

■ 164. Section 72.210 is revised to read 

as follows: 

§ 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. 

■ 165. In § 72.218, paragraph (b) is 

revised to read as follows: 

§ 72.218 Termination of licenses. 

* * * * * 

(b) An application for termination of 

a reactor operating license issued under 

10 CFR part 50 and submitted under 

§ 50.82 of this chapter, or a combined 

license issued under 10 CFR part 52 and 

submitted under § 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 

■ 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). 

■ 167. In § 73.1, paragraph (b)(1)(i) is 

revised to read as follows: 

§ 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, 

* * * * * 

■ 168. In § 73.2, the introductory text of 

paragraph (a) is revised to read as 

follows: 

§ 73.2 Definitions. 

* * * * * 

(a) Terms defined in parts 50, 52, and 

70 of this chapter have the same 

meaning when used in this part. 

* * * * * 

■ 169. In § 73.50, the introductory text 

is revised to read as follows: 

§ 73.50 Requirements for physical 

protection of licensed activities. 

Each licensee who is not subject to 

§ 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: 

* * * * * 

■ 170. In § 73.56, paragraph (a)(3) is 

revised to read as follows: 

§ 73.56 Personnel access authorization 

requirements for nuclear power plants. 

(a) * * * 

(3) Each applicant for a license to 

operate a nuclear power reactor under 

§§ 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 § 52.103(g) 

of this chapter, shall implement the 

required access authorization program 

as part of its site Physical Security Plan. 

* * * * * 

■ 171. In § 73.57, paragraphs (a)(1), 

(a)(2), and (a)(3) are revised to read as 

follows: 

§ 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 

§ 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 § 52.103(g) of this 

chapter, each applicant for a license to 


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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. 

* * * * * 

■ 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, § 52.17 or § 52.79, or to 

§ 70.22(i) of this chapter. 

* * * * * 

PART 75—SAFEGUARDS ON 

NUCLEAR MATERIAL—

IMPLEMENTATION OF US/IAEA 

AGREEMENT 

■ 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). 

■ 174. In § 75.6, paragraph (b) is revised 

to read as follows: 

§ 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 

■ 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. 

■ 176. In § 95.5, the definition of license 

is revised to read as follows: 

§ 95.5 Definitions. 

* * * * * 

License means a license issued under 

10 CFR parts 50, 52, 54, 60, 63, 70, or 

72. 

* * * * * 

■ 177. In § 95.13, paragraph (b) is 

revised to read as follows: 

§ 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. 

■ 178. In § 95.19, the introductory text 

of paragraph (b) is revised to read as 

follows: 

§ 95.19 Changes to security practices and 

procedures. 

* * * * * 

(b) A licensee, certificate holder, or 

other person may effect a minor, nonsubstantive 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, nonsubstantive changes to the Standard 

Practice Procedures Plan include—

* * * * * 

■ 179. Section 95.20 is revised to read 

as follows: 

§ 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. 

■ 180. In § 95.23, paragraph (b) is 

revised to read as follows: 

§ 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 § 95.53 apply. 

■ 181. Section 95.31 is revised to read 

as follows: 


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§ 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. 

■ 182. In § 95.33, paragraph (c) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 183. Section 95.34 is revised to read 

as follows: 

§ 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. 

■ 184. In § 95.35, the introductory text 

of paragraph (a), and paragraph (a)(3) 

are revised to read as follows: 

§ 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. 

* * * * * 

■ 185. In § 95.36, paragraphs (c), (d), 

and (e) are revised to read as follows: 

§ 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 § 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. 

■ 186. In § 95.37, paragraphs (a), (b), 

and (h) are revised to read as follows: 

§ 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 


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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. 

* * * * * 

■ 187. In § 95.39, paragraph (a) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 188. In § 95.43, paragraph (a) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 189. In § 95.45, paragraph (d) is 

revised to read as follows: 

§ 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. 

■ 190. Section 95.49 is revised to read 

as follows: 

§ 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 

§ 95.15 or submitted as an amendment 

to its existing Standard Practice 

Procedures Plan for the protection of 

classified information. 

■ 191. Section 95.51 is revised to read 

as follows: 

§ 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. 

■ 192. Section 95.53 is revised to read 

as follows: 

§ 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. 

■ 193. In § 95.57, the introductory 

paragraph is revised to read as follows: 

§ 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: 

* * * * * 

■ 194. Section 95.59 is revised to read 

as follows: 

§ 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 

■ 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). 

■ 196. In § 140.2, paragraphs (a)(1) and 

(a)(2) are revised to read as follows: 

§ 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. 

* * * * * 

■ 197. Section 140.10 is revised to read 

as follows: 

§ 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 


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to possess and use plutonium in a 

plutonium processing and fuel 

fabrication plant. 

■ 198. In § 140.11, paragraph (b) is 

revised to read as follows: 

§ 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. 

■ 199. In § 140.12, paragraph (c) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 200. Section 140.13 is revised to read 

as follows: 

§ 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 

§ 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 § 140.15 of this chapter before 

issuance of the license under part 70 of 

this chapter. 

■ 201. In § 140.20, paragraph (a)(1)(ii) is 

revised, and paragraph (a)(1)(iii) is 

added to read as follows: 

§ 140.20 Indemnity agreements and liens. 

(a) * * * 

(1) * * * 

(ii) The date that the Commission 

makes the finding under § 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 

* * * * * 

■ 202. In § 140.81, paragraph (a) is 

revised to read as follows: 

§ 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. 

* * * * * 

■ 203. In § 140.93 Appendix C, Article 

VIII, paragraph 4 is revised to read as 

follows: 

§ 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. 

* * * * * 

■ 204. Section 140.96 is revised to read 

as follows: 

§ 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 

■ 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). 

■ 206. In § 170.2, paragraph (j) is 

removed and reserved, and paragraphs 

(g) and (k) are revised to read as follows: 

§ 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 


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49566 Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 / Rules and Regulations 

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 

■ 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). 

■ 208. In § 171.15, paragraph (a) is 

revised to read as follows: 

§ 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 § 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 § 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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