65776 Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 50 

[NRC–2009–0359; NRC–2013–0133] 

RIN 3150–AI72 

Approval of American Society of 

Mechanical Engineers’ Code Cases 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Final rule. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is amending its 

regulations to incorporate by reference 

the latest revisions of three NRC 

Regulatory Guides (RGs) approving new 

and revised Code Cases published by 

the American Society of Mechanical 

Engineers. This action allows nuclear 

power plant licensees, and applicants 

for construction permits, operating 

licenses, combined licenses, standard 

design certifications, standard design 

approvals, and manufacturing licenses, 

to use the Code Cases listed in these 

RGs, as alternatives to engineering 

standards for the construction, inservice 

inspection, and inservice testing of 

nuclear power plant components. This 

final rule changes NRC’s regulations to 

address a petition for rulemaking (PRM), 

PRM–50–89, submitted by Mr. Raymond 

West. The final rule also restructures the 

NRC’s requirements governing Codes 

and standards to align with the Office of 

the Federal Register’s guidelines for 

incorporating documents by reference. 

This final rule announces the 

availability of the final versions of the 

three RGs that are being incorporated by 

reference, and a related RG, not 

incorporated by reference into the 

NRC’s regulations, that lists Code Cases 

that the NRC has not approved for use. 

For additional information on these 

RGs, see Section XVII, Availability of 

Regulatory Guides, of this document. 

DATES: This final rule is effective on 

December 5, 2014. The incorporation by 

reference of RG 1.84, ‘‘Design, 

Fabrication, and Materials Code Case 

Acceptability, ASME Section III,’’ 

Revision 36 (May 2014); RG 1.147, 

‘‘Inservice Inspection Code Case 

Acceptability, ASME Section XI, 

Division 1,’’ Revision 17 (May 2014); 

and RG 1.192, ‘‘Operation and 

Maintenance Code Case Acceptability, 

ASME OM Code,’’ Revision 1 (May 

2014) is approved by the Director of the 

Office of the Federal Register as of 

December 5, 2014. 

ADDRESSES: Please refer to Docket ID 

NRC–2009–0359 when contacting the 

NRC about the availability of 

information for this final rule and RGs 

1.84, 1.147 and 1.192. Please refer to 

Docket ID NRC–2013–0133 when 

contacting the NRC about the 

availability of information for RG 1.193. 

You may obtain publicly-available 

information related to this final rule by 

any of the following methods: 

• Federal Rulemaking Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2009–0359. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–287–3422; 

email: Carol.Gallagher@nrc.gov. For 

technical questions, contact the 

individuals listed in the FOR FURTHER

INFORMATION CONTACT section of this 

final rule. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

http://www.nrc.gov/reading-rm/

adams.html. To begin the search, select 

‘‘ADAMS Public Documents’’ and then 

select ‘‘Begin Web-Based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to pdr.resource@nrc.gov. For the 

convenience of the reader, the ADAMS 

accession numbers are provided in a 

table in the ‘‘Availability of Documents’’ 

section of this document. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

FOR FURTHER INFORMATION CONTACT: 

Jenny Tobin, Office of Nuclear Reactor 

Regulation; telephone: 301–415–2328, 

email: Jennifer.Tobin@nrc.gov; or 

Wallace Norris, Office of Nuclear 

Regulatory Research, telephone: 301– 

251–7650; email: Wallace.Norris@

nrc.gov; both are staff of the U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001. 

Executive Summary 

The U.S. Nuclear Regulatory 

Commission (NRC) is amending its 

regulations to incorporate by reference 

the latest revisions of three NRC 

Regulatory Guides (RGs) approving new 

and revised Code Cases published by 

the American Society of Mechanical 

Engineers (ASME). The three RGs 

incorporated by reference are RG 1.84, 

Revision 36; RG 1.147, Revision 17; and 

RG 1.192, Revision 1. This action allows 

nuclear power plant licensees, and 

applicants for construction permits, 

operating licenses, combined licenses, 

standard design certifications, standard 

design approvals, and manufacturing 

licenses, to use the Code Cases listed in 

these RGs as alternatives to engineering 

standards for the construction, inservice 

inspection, and inservice testing of 

nuclear power plant components. 

The NRC is announcing the 

availability of the final versions of the 

three RGs that are being incorporated by 

reference, and a final version of RG 

1.193, Revision 4, not incorporated by 

reference into the NRC’s regulations, 

that lists Code Cases that the NRC has 

not approved for generic use. 

This final rule also includes changes 

to the NRC’s regulations that address a 

petition for rulemaking (PRM), PRM– 

50–89, submitted by Mr. Raymond 

West. Mr. West requested that the NRC 

amend its regulations to allow 

consideration of alternatives to NRCapproved ASME Boiler and Pressure 

Vessel and Operation and Maintenance 

of Nuclear Power Plants Code Cases. 

This final rule resolves Mr. West’s 

petition and represents the NRC’s final 

action on PRM–50–89. 

Lastly, this final rule resequences the 

NRC’s requirements in § 50.55a of Title 

10 of the Code of Federal Regulations 

(10 CFR), governing Codes and 

standards to align with Office of the 

Federal Register’s guidelines for 

incorporating published standards by 

reference. 

SUPPLEMENTARY INFORMATION:

Table of Contents 

I. Background 

II. Opportunity for Public Participation 

A. Overview of Public Comments 

Table I—Comment Submissions Received 

on the Proposed Rule and Draft 

Regulatory Guides 

III. Public Comment Analysis 

A. NRC Reponses to Public Comments on 

Proposed Rule 

B. NRC Responses to Public Comments on 

Draft Regulatory Guides 

IV. NRC Approval of New and Amended 

ASME Code Cases 

A. ASME Code Cases Approved for 

Unconditional Use 

Table II—Unconditionally Approved Code 

Cases 

B. ASME Code Case Approved for Use 

With Conditions 

Table III—Conditionally Approved Code 

Cases 

C. ASME Code Cases Not Approved for Use 

V. Petition for Rulemaking (PRM–50–89) 

VI. Changes Addressing the Office of the 

Federal Register’s Guidelines on 

Incorporation by Reference 

VII. Addition of Headings to Paragraphs 

A. NRC’s Convention for Headings and 

Subheadings 

B. Readers Aids 

VIII. Paragraph-by-Paragraph Discussion 

IX. Regulatory Flexibility Certification 

X. Regulatory Analysis 

XI. Backfitting and Issue Finality 

XII. Plain Writing 

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Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 65777 

1ASME Code Cases can be categorized as one of 

two types: New or revised. A new Code Case 

provides for a new alternative to specific ASME 

Code provisions or addresses a new need. A revised 

Code Case is a revision (modification) to an existing 

Code Case to address, for example, technological 

advancements in examination techniques or to 

address NRC conditions imposed in one of the 

regulatory guides that have been incorporated by 

reference into 10 CFR 50.55a. 

XIII. Finding of No Significant 

Environmental Impact: Environmental 

Assessment 

XIV. Paperwork Reduction Act Statement 

XV. Congressional Review Act 

XVI. Voluntary Consensus Standards 

XVII. Availability of Regulatory Guides 

XVIII. Availability of Documents 

I. Background 

The American Society of Mechanical 

Engineers (ASME) develops and 

publishes the ASME Boiler and Pressure 

Vessel (BPV) Code, which contains 

requirements for the design, 

construction, and inservice inspection 

(ISI) and examination of nuclear power 

plant components, and the ASME Code 

for Operation and Maintenance of 

Nuclear Power Plants (OM) Code, which 

contains requirements for inservice 

testing (IST) of nuclear power plant 

components. In response to BPV and 

OM Code user requests, the ASME 

develops ASME Code Cases that provide 

alternatives to BPV and OM Code 

requirements under special 

circumstances. 

The NRC approves and/or mandates 

the use of the ASME BPV and OM 

Codes in § 50.55a of Title 10 of the Code 

of Federal Regulations (10 CFR) through 

the process of incorporation by 

reference (IBR). As such, each provision 

of the ASME Codes incorporated by 

reference into, and mandated by, 

§ 50.55a, ‘‘Codes and standards,’’ 

constitutes a legally-binding NRC 

requirement imposed by rule. As noted 

previously, ASME Code Cases, for the 

most part, represent alternative 

approaches for complying with 

provisions of the ASME BPV and OM 

Codes. Accordingly, the NRC 

periodically amends § 50.55a to 

incorporate by reference NRC 

Regulatory Guides (RGs) listing 

approved ASME Code Cases that may be 

used as alternatives to the BPV and OM 

Codes. See Federal Register notice 

(FRN), ‘‘Incorporation by Reference of 

ASME BPV and OM Code Cases’’ (68 FR 

40469; July 8, 2003). 

This rulemaking is the latest in a 

series of rulemakings that incorporate 

by reference new versions of several 

RGs identifying new and revised 1

unconditionally or conditionally 

acceptable ASME Code Cases that are 

approved for use. In developing these 

RGs, the NRC staff reviews ASME BPV 

and OM Code Cases, determines the 

acceptability of each Code Case, and 

publishes its findings in the RGs. The 

RGs are revised periodically as new 

Code Cases are published by the ASME. 

The NRC incorporates by reference the 

RGs listing acceptable and conditionally 

acceptable ASME Code Cases into 

§ 50.55a. Currently, NRC RG 1.84, 

Revision 35, ‘‘Design, Fabrication, and 

Materials Code Case Acceptability, 

ASME Section III’’; RG 1.147, Revision 

16, ‘‘Inservice Inspection Code Case 

Acceptability, ASME Section XI, 

Division 1’’; and RG 1.192, Revision 0, 

‘‘Operation and Maintenance Code Case 

Acceptability, ASME OM Code,’’ are 

incorporated into the NRC’s regulations 

at 10 CFR 50.55a, ‘‘Codes and 

standards.’’ 

This final rule adds provisions that 

allow the NRC to authorize alternatives 

to NRC-approved ASME BPV and OM 

Code Cases, as requested in a petition 

for rulemaking (PRM) that was 

submitted to the NRC on December 14, 

2007, and revised on December 19, 

2007, by Mr. Raymond West (ADAMS 

Accession No. ML073600974). A 

detailed discussion of the PRM is 

provided in Section V, ‘‘Petition for 

Rulemaking (PRM–50–89),’’ of this 

document. 

II. Opportunity for Public Participation 

On June 24, 2013 (78 FR 37886), the 

NRC published a proposed rule in the 

Federal Register that would incorporate 

by reference RG 1.84, Revision 36; RG 

1.147, Revision 17; and RG 1.192, 

Revision 1. On the same date, the NRC 

published a parallel FRN announcing 

the availability of the three draft RGs 

and opportunity for public comment (78 

FR 37721; June 24, 2013). The NRC 

provided a 75-day public comment 

period for both the proposed rule and 

the draft RGs, which ended on 

September 9, 2013. 

A. Overview of Public Comments 

The NRC received a total of 10 

comment submissions. The submissions 

were received from three private 

citizens, four utility organizations, and 

three industry groups that provide 

engineering and inspection services to 

the utilities. Table I lists the 

commenter’s name and affiliation, 

ADAMS accession number for the 

comment submission, and the Code 

Case or subject of each comment. 

TABLE I—COMMENT SUBMISSIONS RECEIVED ON THE PROPOSED RULE AND DRAFT REGULATORY GUIDES

Commenter name Affiliation 

Comment submission ADAMS 

Accession No. 

Affected code cases/subject 

William Culp ............................................ Private Citizen ........................................ ML13210A143 Proposed Rule. 

Saige Stephens ....................................... Private Citizen ........................................ ML13210A151 General. 

Richard Swayne ...................................... ASME ..................................................... ML13253A076 N–60–5. 

ML13252A286 ** N–416–4. 

N–561–2. 

N–562–2. 

N–597–2. 

N–606–1. 

N–619. 

N–648–1. 

N–661–2. 

N–702. 

N–739–1. 

N–798. 

N–800. 

N–659–2. 

Proposed Rule. 

Mark Richter ............................................ Nuclear Energy Institute ........................ ML13259A040 Proposed Rule. 

ML13254A080 ** 

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65778 Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 

TABLE I—COMMENT SUBMISSIONS RECEIVED ON THE PROPOSED RULE AND DRAFT REGULATORY GUIDES—Continued 

Commenter name Affiliation 

Comment submission ADAMS 

Accession No. 

Affected code cases/subject 

Edward Colie ........................................... South Carolina Electric and Gas ........... ML13254A082 Proposed Rule. 

Patricia Campbell .................................... GE Hitachi Nuclear Energy .................... ML13259A038 1332–6. 

Devin Kelley ............................................ AREVA ................................................... ML13259A039 N–71–18. 

David Helker ............................................ Exelon Generation Company, LLC ........ ML13269A371 N–60–5. 

N–798. 

N–800. 

N–702. 

Shawn Comstock .................................... Private Citizen ........................................ ML13182A081 OMN–1 (2006 Addenda). 

OMN–11 (2006 Addenda). 

OMN–12 (2004 Edition). 

Roy Hall ................................................... Inservice Inspection Program Owners 

Group.

ML13197A239 N–805. 

** There are two ADAMS accession numbers for the submissions from ASME and the Nuclear Energy Institute because each submission contained comments on the proposed rule and the drafts RGs. Both accession numbers are for the same incoming submission, but one accession 

number is identified in ADAMS as a response to the Federal Register notice soliciting comments on the proposed rule and the other is identified 

as a response for the draft RGs. 

III. Public Comment Analysis 

The NRC has reviewed every 

comment submission and has identified 

42 unique comments requiring NRC 

consideration and response. Comment 

summaries and the NRC responses are 

presented in this section. Comment 

responses have been organized in two 

categories: (A) NRC Responses to Public 

Comments on Proposed Rule and (B) 

NRC Responses to Public Comments on 

Draft RGs, further delineated by 

individual RG (i.e., RG 1.84, RG 1.147, 

and RG 1.192). 

A. NRC Reponses to Public Comments 

on Proposed Rule 

Proposed Rule 

Comment: The commenter developed 

a proposed one-page revision to the 

overall Codes and standards rule in 

§ 50.55a that reflects the commenter’s 

view of the current regulatory process 

and suggested parsing the details of 

§ 50.55a to the appropriate RGs. The 

commenter provided the background 

and bases for his proposed rule 

structure, and stated that the purpose of 

his proposal is to simplify the overall 

structure of § 50.55a. (Culp–3) 

NRC Response: The main purpose of 

this rulemaking is to amend § 50.55a to 

incorporate by reference the latest 

revisions of three RGs approving new 

and revised Code Cases published by 

ASME. This rulemaking also proposes 

to: (1) Resolve a petition for rulemaking 

(PRM–50–89) submitted by Mr. 

Raymond West, (2) resequence the 

NRC’s requirements governing Codes 

and standards in order to align with the 

latest guidelines of the OFR for IBR, and 

(3) add headings (explanatory titles) to 

paragraphs and lower-level 

subparagraphs of § 50.55a. 

The NRC is not proposing a major 

restructuring or simplification of the 

requirements in § 50.55a. As explained 

in the statement of considerations in the 

proposed rule, the proposed editorial, 

non-substantive changes were made to 

align with the IBR guidance for multiple 

standards that is included in Chapter 6 

of the OFR’s, ‘‘Federal Register 

Document Drafting Handbook,’’ January 

2011 Revision. These changes will 

structure NRC’s regulations consistent 

with other Federal regulations that 

incorporate by reference multiple 

standards. Although NRC welcomes 

public comments on the revised 

structure of § 50.55a, the NRC is limited 

in the types of changes it can make in 

response to public comments on the 

revised structure and must align with 

the OFR’s guidance. 

Adding headings at the paragraph and 

subparagraph levels of § 50.55a will 

enhance the reader’s ability to identify 

the subject matter of each paragraph and 

subparagraph. These headings are a first 

step toward addressing longstanding 

complaints about the readability and 

complex structure of § 50.55a. The NRC 

is not making significant structural 

changes to the rule at this time, but may, 

in the future, consider doing so in a 

separate rulemaking. The NRC would 

consider the commenter’s suggestions 

and proposed rule language if and when 

NRC conducts that rulemaking. At this 

time, however, the NRC considers the 

commenter’s suggestion to be outside 

the scope of this proposed rulemaking. 

No change was made to the final rule 

as a result of this comment. 

Comment: The purpose and scope of 

the rule has changed over time, and no 

longer reflects the actual regulatory 

process for review of consensus 

industry Codes and standards that have 

been found acceptable to the NRC staff 

on a generic basis or as part of a plantspecific review process that covers more 

than the Codes and standards 

mentioned. It does not seem appropriate 

for § 50.55a to reference Codes and 

standards that have been withdrawn 

(e.g., IEEE 279). The content of § 50.55a 

represents an archive of once-upon-atime requirements, not contemporary 

Codes and standards. It is not necessary 

to recapitulate what Codes and 

standards were approved on individual 

applications; applicants retain design 

and safety responsibility (including 

identification of unreviewed safety 

questions) that might arise from new 

regulatory guides, Codes and standards, 

and operating experience. The following 

Codes, standards, and Code Cases in the 

proposed regulation are not the latest 

and conditions are imposed on the use 

of superseded documents which would 

preferably not be used for new design or 

ISI activities (the conditions are most 

likely fully documented in the licenses, 

safety analyses, and ISI programs for 

individual nuclear power plants as 

approved by the NRC): (Culp–3.1, 3.3, 

3.9) 

a. ASME III and Code Case N–729–1 (N– 

729–4 Is Approved by ASME) 

b. ASME XI 

c. IEEE 279 

NRC Response: The NRC disagrees 

with the assertion that the proposed rule 

does not reflect the actual regulatory 

process for review of consensus 

industry Codes and standards that have 

been found acceptable to the NRC staff. 

Section II, ‘‘Discussion,’’ of the 

proposed rule described the three-step 

process that the NRC follows to 

determine the acceptability of new and 

revised Code Cases and the need for 

regulatory positions on the uses of these 

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Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 65779 

Code Cases. The fundamental process 

has not changed over time. Also, the 

Code of Record for design and 

construction does not change over time 

unless there is a voluntary update by the 

licensee. As such, these codes and 

standards must be referenced in § 50.55a 

as long as they are in use. 

Any Code or standard still in use 

must continue to be listed in the 

regulation, or licensees would have to 

discontinue their use when the rule 

becomes effective and immediately 

implement the latest version. These 

Codes and Code Cases are still in use 

and, therefore, may not be removed 

from § 50.55a without unacceptably 

changing their legal status from 

mandatory requirements or approved for 

use, to guidance. 

No change was made to the final rule 

as a result of this comment. 

Comment: The current language and 

structure of § 50.55a blurs the lines 

between the requirements for a quality 

program and for safety. (Culp–3.2) 

NRC Response: The NRC believes this 

is an out of scope comment because it 

addresses the clarity of the requirements 

in § 50.55a in this rulemaking. The 

scope of this rulemaking is to: (1) 

Incorporate by reference the three 

Regulatory Guides identifying NRCapproved ASME Code Cases; and (2) to 

reorganize the section to address Office 

of the Federal Register requirements for 

incorporation by reference. 

However, the NRC provides the 

following response to the out of scope 

comment. The NRC notes that the 

commenter did not provide any 

rationale why the rulemaking blurs the 

distinction between quality assurance 

and safety. In addition, the NRC notes 

that the reorganization of § 50.55a 

fundamentally addressed the paragraph 

identifying the ASME and IEEE codes 

that are incorporated by reference. The 

reorganization did not change any of the 

NRC requirements with respect to 

quality assurance or safety. 

No change was made to the final rule 

as a result of this comment. 

Comment: The proposed 

reorganization of § 50.55a uses the 

unconventional numbering hierarchy 

(a), (1), (i), (A). This is difficult to follow 

in the existing rule which is very long. 

It is even more difficult to follow in the 

proposed regulation with or without 

added introductory statements. (Culp– 

3.4) 

NRC Response: The NRC has added 

headings to the paragraph and 

subparagraph levels of § 50.55a to aid 

the reader of this regulation. The 

hierarchy used in § 50.55a is that which 

is used throughout the Code of Federal 

Regulations and is dictated by the OFR. 

The NRC is also considering developing 

additional user aides. 

No change was made to the final rule 

as a result of this comment. 

Comment: The proposed regulation 

states that the regulation is consistent 

with a policy to review and accept 

industry standards instead of writing 

regulations; this is not achieved in 

practice due to delays in endorsing new 

Code editions and addenda. In at least 

some cases, the unendorsed newer Code 

revisions have been specifically made to 

incorporate the conditions, exceptions, 

and limitations in § 50.55a. (Culp–3.5) 

NRC Response: The NRC appreciates 

the ASME’s efforts to consider the 

NRC’s concerns as addressed in 

conditions to § 50.55a. The NRC agrees 

that delays in approving new ASME 

Code editions and Code Cases can be 

counterproductive with respect to 

implementation of improvements in 

ASME Code requirements. The NRC 

continues to assess ways to improve the 

rulemaking process to find schedule 

efficiencies. 

No change was made to the final rule 

as a result of this comment. 

Comment: There is too much detail in 

the proposed regulation; NRC concerns 

should be more appropriately organized 

and put into consensus Code and Code 

Case work and topical regulatory guides. 

The proposed regulation is excessively 

detailed and covers an extraordinary 

range of subjects; the diverse NRC 

conditions ranging from grease caps to 

relief valve testing facility capabilities 

could be better organized and 

documented in regulatory guides on the 

specific topic (e.g., RG 1.90). (Culp–3.6) 

NRC Response: The NRC agrees that 

there are many conditions in § 50.55a. It 

should be noted, that certain conditions 

are necessary because applicants and 

licensees continue to use many different 

Code editions and addenda. 

Accordingly, it is necessary to continue 

to list conditions that may have been 

addressed by a later Code edition 

because the earlier Code edition is still 

in use. The NRC determined that other 

conditions, such as those addressing 

grease caps, are necessary to ensure that 

safety-related concerns are adequately 

addressed. 

With respect to the suggestion to use 

RGs, the NRC notes that RGs normally 

provide guidance and describe 

approaches that would be acceptable to 

the NRC for implementing a rule. Under 

the approach suggested in the comment, 

the RG would have to be incorporated 

by reference into § 50.55a in order for 

the provisions in the regulatory guides 

to continue to be legally-binding. In 

enclosure 5 to the comments submitted 

by the ASME, the ASME encouraged the 

NRC to consider alternative methods for 

endorsing ASME Codes and standards, 

such as moving many of the 

requirements currently specified in 

§ 50.55a into a suitable regulatory guide 

that can be referenced within the 

regulation. The NRC agrees that the 

format and organization of § 50.55a 

could be improved, and the NRC may, 

in the future, conduct a rulemaking to 

restructure and simplify § 50.55a. The 

public would be given opportunity to 

comment before implementation. 

No change was made to the final rule 

as a result of this comment. 

Comment: There are multiple reviews 

and opportunities for staff review and 

public comment without necessarily 

also requiring comment on the proposed 

regulations to ‘‘incorporate by 

reference’’ what started as a simple 

reference to ASME III. The process of a 

comment in Code committee, comment 

on proposed regulatory guides, and 

comment on Code Cases seems 

adequate. Yet, comments from NRC 

representatives in Code meetings do not, 

according to their own words, ‘‘carry the 

weight of the NRC staff endorsement,’’ 

and some conditions have arisen after 

Code committees have finished reviews 

and published revisions. (Culp–3.7) 

NRC Response: The NRC staff 

representatives on ASME Code 

committees have the opportunity to 

participate during the consideration of 

the Code cases during the ASME 

standards process. These individuals 

can provide input to the cases both 

before and after ASME endorsement. 

However, this participation is not a 

substitute for the technical, legal, and 

management reviews that must be 

conducted with respect to a complete 

rulemaking prior to issuance. 

The second issue in this comment 

concerns public involvement in the 

rulemaking process involved in 

incorporating by reference those Code 

cases that the NRC has reviewed and 

approved. In accordance with the 

Administrative Procedures Act, the 

public is afforded an opportunity for 

review and comment, unless there is 

reasonable likelihood that there will be 

no ‘‘significant adverse comment’’ on a 

proposed rule. Past NRC experience 

suggests that the NRC will receive at 

least one ‘‘significant adverse comment’’ 

on each § 50.55a proposed rule. 

No change was made to the final rule 

as a result of this comment. 

Comment: The proposed revision to 

§ 50.55a is very complicated and seems 

to be contrary to multiple claims in the 

discussion points in the proposed rule 

regarding: (Culp–3.8) 

a. Paperwork reduction 

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b. Regulatory flexibility 

c. Plain writing 

d. Backfitting and issue finality 

NRC Response: The NRC does not 

agree with the comment. The comment 

did not explain why the proposed 

Paperwork Reduction Act statement, 

Regulatory Flexibility Certification, 

Plain Writing discussion, or Backfitting 

and Issue Finality discussion is contrary 

to the proposed regulation. Complexity 

by itself does not mean that the NRC’s 

proposed discussions on the four areas 

are inadequate or in error. Furthermore, 

the bulk of the changes in this 

rulemaking involve the reorganization 

of the rule. Therefore, the comment 

incorrectly implies that this rulemaking 

is the reason for the ‘‘complexity’’ of 

§ 50.55a. 

No change was made to the final rule 

as a result of this comment. 

Comment: Should Mechanical 

Engineers become the new regulated 

embodiment of manufacturing arms? 

Change administration using 

international standards. (Stephens–4.1) 

NRC Response: The NRC is unable to 

respond to this comment because of its 

ambiguous nature. 

No change was made to the final rule 

as a result of this comment. 

Comment: The NRC should amend its 

regulations to allow consideration of 

alternatives to the ASME BPV and OM 

Code Cases, as requested in a petition 

for rulemaking submitted by Mr. 

Raymond West (PRM–50–89) (ADAMS 

Accession No. ML073600974). The 

possibility of implementing an 

alternative to a Code Case approved by 

the Director of the Office of Nuclear 

Reactor Regulation will reduce the 

administrative burden on licensees and 

significantly reduce the lengthy process 

of proposing and gaining acceptance for 

a change or modification to a Code Case. 

The ASME supports the proposed 

changes in § 50.55a(z) to address PRM– 

50–89. (NEI–6.2, ASME–5.5.1) 

NRC Response: The NRC agrees. 

Authorizing an alternative to an NRCapproved ASME Code Case reduces the 

administrative burden on the NRC and 

licensees. A complete discussion of the 

bases is set forth in Section V, ‘‘Petition 

for Rulemaking (PRM–50–89).’’ 

The final rule includes a provision in 

50.55a(z) allowing the NRC to authorize 

alternatives to NRC-approved ASME 

Code Cases. 

Comment: The ASME believes 

changes for Federal Register guidelines 

have been crafted to minimize 

administrative burden. (ASME–5.5.2) 

NRC Response: No response is 

necessary. 

Comment: Paragraph headings will 

improve readability. (ASME–5.5.3) 

NRC Response: No response is 

necessary. 

Comment: In general, the proposed 

RGs and related documents are written 

in a clear and effective manner, 

consistent with the Plain Writing Act 

and the Presidential Memorandum, 

‘‘Plain Language in Government 

Writing.’’ Well-written regulatory 

guidance documents support their 

correct interpretation and 

implementation (NEI–6.2). 

NRC Response: No response 

necessary. 

Comment: The proposed changes to 

10 CFR 50.55a would place a large 

burden on licensees. As discussed in 

Section VI, these changes would 

‘‘require substantial rewriting of these 

procedures and documents to correct 

the references to the old (superseded) 

sections, paragraphs and 

subparagraphs.’’ For licensees, these 

revisions would include licensing 

documentation. None of the proposed 

organizational changes to 10 CFR 50.55a 

pertain to any of the provisions of 10 

CFR 50.109(a)(4), since no information 

is changing and is merely reorganized. 

This means that in order to reorganize 

10 CFR 50.55a, backfit analysis would 

have to be performed in accordance 

with 10 CFR 50.109. There is no need 

to change the location of the content in 

10 CFR 50.55a (South Carolina Electric 

and Gas–7.1). 

NRC Response: As indicated in 

Section V, ‘‘Changes Addressing Office 

of the Federal Register’s Guidelines on 

Incorporation by Reference,’’ of the 

proposed rule, the reorganization of 

content was made in accordance with 

the revised guidance for incorporation 

by reference of multiple standards that 

is included in Chapter 6 of the OFR’s, 

‘‘Federal Register Document Drafting 

Handbook,’’ January 2011 Revision. All 

Federal agencies were directed to align 

with the guidelines. The OFR’s 

guidance provided several options for 

incorporating by reference multiple 

standards into regulations. The NRC 

found moving the incorporation by 

reference of multiple standards into the 

first paragraph of § 50.55a(a) to be the 

least disruptive option. These changes, 

which are required by the OFR, are not 

within the purview of the backfit rule, 

and no further consideration of 

backfitting is needed to address the 

OFR-mandated reorganization. 

No change was made to the final rule 

as a result of this comment. 

Comment: The NRC should consider 

adding hyperlinks and indentation to 

§ 50.55a because it would aid readers in 

navigating the rule. (South Carolina 

Electric and Gas–7.2) 

NRC Response: The NRC appreciates 

these practical suggestions and agrees 

that adding hyperlinks or indentation 

would aid the readers in navigating 

§ 50.55a. However, the NRC is unable to 

add hyperlinks or indentation to a rule 

published in the Code of Federal 

Regulations. Format requirements for 

the Code of Federal Regulations are 

established and enforced by the OFR, 

and do not permit inclusion of 

hyperlinks or a different indentation 

scheme. Please note that the NRC has 

prepared two documents to aid the 

reader in navigating § 50.55a: ‘‘Final 

Reorganization of Paragraphs and 

Subparagraphs in 10 CFR 50.55a, ‘Codes 

and standards’ ’’ (ADAMS Accession 

No. ML14015A191) and ‘‘CrossReference Tables’’ (ADAMS Accession 

No. ML14211A050—package with two 

tables). The NRC is currently 

considering developing several 

alternatives to improve the format and 

organization of § 50.55a in a potential 

future rulemaking. The NRC plans to 

seek public interaction as part of the 

rulemaking process. 

No change was made to the final rule 

as a result of this comment. 

B. NRC Responses to Public Comments 

on Draft Regulatory Guides 

Regulatory Guide 1.84, Revision 36 

(DG–1230) 

Code Case N–60–5 

Comment: Text in the proposed 

condition should be corrected to change 

‘‘stain-hardened’’ to ‘‘strain-hardened.’’ 

(ASME–5.1.1, Exelon–10.1) 

NRC Response: The NRC agrees with 

the comment. 

RG 1.84, Revision 36 has been 

corrected in accordance with the 

comment. 

Code Case 1332–6 

Comment: Appendix C of DG–1230 

states that Code Case 1332–6 is 

contained in Table 5. However, Code 

Case 1332–6 does not appear in Table 5. 

(GE Hitachi Nuclear Energy–8.1) 

NRC Response: The NRC agrees with 

this comment. Code Case 1332–6 has 

been added to Table 5 in RG 1.84, 

Revision 36, which lists those Section 

III Code Cases that have been 

superseded by revised Code Cases. 

Code Case N–71–18 

Comment: The American Welding 

Society (AWS) Code D1.1 was 

reformatted, and the provisions in 

paragraph 4.5.2.2 were relocated to 

paragraph 5.3.2.3 in the AWS Code. The 

paragraph references for AWS D1.1 in 

condition No. 3 to Code Case N–71–18 

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should be revised accordingly. 

(AREVA–9.1) 

NRC Response: The NRC agrees with 

this comment. The reference in 

condition 3 to Code Case N–71–18 has 

been corrected in RG 1.84, Revision 36 

by referring to paragraph ‘‘5.3.2.3.’’ 

Regulatory Guide 1.147, Revision 17 

(DG–1231) 

Code Case N–416–4 

Comment: The NRC condition on this 

Code Case requiring nondestructive 

examination of welded or brazed 

repairs, and fabricated and installed 

joints, in accordance with the 

construction code of record, imposes an 

unnecessary burden on licensees and is 

not necessary to ensure safe operation. 

The BPV Code has long relied on a 

specified relationship between NDE and 

allowable stresses, i.e., vintage codes, 

such as American National Standards 

Institute (ANSI) B31.1 or Section III, 

have lower allowable stresses, due to 

the fact that NDE is generally not 

required, whereas nuclear codes (ASME 

Section III and B31.7) have higher 

allowable stress intensities for Class 1 

components relative to Class 2 and 3 

components (due mostly to the 

additional examinations required for 

Class 1 components). 

The NRC stated that ‘‘A system 

pressure test or hydrostatic pressure test 

does not verify the structural integrity of 

the repaired piping components.’’ The 

ASME has never established any 

relationship between the test pressure to 

which a component is subjected and 

any other material or design 

characteristic. The primary technical 

consideration in development of the 

required test pressure is to ensure that 

it is low enough to prevent yielding of 

the material. Hydrostatic testing does 

not prove structural integrity; it proves 

only leak tightness. Similarly, NDE 

alone does not ensure structural 

integrity. The ASME Code ensures 

structural integrity through a 

combination of many factors, including 

material testing, design formulas, design 

factors, and qualification of personnel. 

Adding more NDE than required by the 

Construction Code (be it ASME Section 

III or B31.1) is not required to ensure 

structural integrity. (ASME–5.2.1) 

NRC Response: The NRC disagrees 

with the comment that the additional 

NDE requirements imposed when using 

Code Case N–416–4 are unnecessary 

and imply that existing components are 

unsuitable. The NRC does agree that 

hydrostatic pressure testing or NDE 

alone does not ensure structural 

integrity. The original Construction 

Codes ensured structural integrity 

through a combination of many factors 

including material testing, design 

formulas, design factors, qualification of 

procedures, qualification of personnel, 

NDE, and hydrostatic testing. Code Case 

N–416–4 would allow a system leakage 

test to be performed in lieu of (1) a 

hydrostatic pressure test prior to return 

to service of Class 1, 2, and 3 welded or 

brazed repairs; (2) fabrication welds or 

brazed joints for replacement parts and 

piping subassemblies; or (3) installation 

of replacement items by welding or 

brazing. 

The NRC believes that the rigorous 

NDE requirements of Section III should 

be performed when the hydrostatic 

pressure test is not performed. The 

reason for this condition is that some 

earlier Construction Codes have less 

stringent NDE requirements than 

Section III; however, they require a 

greater pressure for the Code Case N– 

416–4 required hydrostatic test. Section 

III NDE requirements for Class 1, 2, and 

3 components generally require either 

surface or volumetric examinations or 

possibly both. The NRC believes that 

these NDE requirements along with a 

system leakage test provide the same 

level of quality and safety as the higher 

pressure hydrostatic test and reduced 

NDE requirements of earlier 

Construction Codes. 

No changes were made to RG 1.147, 

Revision 17, as a result of this comment. 

Code Case N–561–2 

Comment: Proposed Conditions (1) 

and (3) should be eliminated. Proposed 

Conditions (1) and (3) limit the life of 

the repair ‘‘until the next refueling 

outage’’ for repairs performed on a wet 

surface or if the cause of the degradation 

has not been determined. The Code Case 

already limits the life of the repair to 

‘‘one fuel cycle’’ for these same 

situations. The ASME Code committee 

considered both phrases when revising 

this Code Case to add these restrictions, 

and intentionally chose ‘‘one fuel cycle’’ 

instead of ‘‘next refueling outage’’ so as 

not to imply that such weld overlays 

could not be performed while a plant is 

shut down for a refueling outage. In 

such a case, literal application of ‘‘next 

refueling outage’’ could mean the 

current refueling outage, which could be 

an extreme hardship, depending on the 

timing of the discovery of the need for 

a weld overlay. Use of the term ‘‘one 

fuel cycle’’ clearly requires that the 

overlay be removed during the 

subsequent fuel cycle no later than the 

same point in the cycle at which the 

overlay was applied. In the vast majority 

of cases, this will happen during the 

next refueling outage; otherwise, a 

special outage or a special limiting 

condition of operation would be 

required mid-cycle in order to effect its 

removal. (ASME–5.2.2.a) 

NRC Response: The NRC disagrees 

with the comment on the ‘‘next 

refueling outage.’’ The NRC finds that 

the suggested phrase, ‘‘next fuel cycle,’’ 

is not as conservative as ‘‘the next 

refueling outage’’ phrase because the 

‘‘next fuel cycle’’ condition would 

permit longer service time to the repair 

that is performed on a wet surface, or 

the cause of the degradation has not 

been determined. 

To clarify the difference between the 

‘‘next refueling outage’’ vs. ‘‘one fuel 

cycle,’’ the NRC staff uses the following 

example. Assume fuel cycle No. 1 is 

followed by refueling outage No. 1, fuel 

cycle No. 2, and refueling outage No. 2. 

Under the ‘‘next refueling outage’’ 

condition, if a repair is performed 

during fuel cycle No. 1, regardless 

whether on the first day or last day of 

fuel cycle No. 1, the ‘‘next refueling 

outage’’ would be refueling outage No. 

1 during which time the repair needs to 

be removed. If the repair is performed 

during refueling outage No. 1, the next 

refueling outage would be refueling 

outage No. 2 during which time the 

repair needs to be removed. Under the 

‘‘next fuel cycle’’ condition, if a repair 

is performed in the middle of fuel cycle 

No. 1, the next fuel cycle would mean 

fuel cycle No. 2 during which time the 

repair needs to be removed. However, 

this condition does not specify exactly 

when in the next fuel cycle (fuel cycle 

No. 2) the repair must be removed. A 

licensee could interpret the next fuel 

cycle as the entire fuel cycle No. 2 and 

remove the repair after fuel cycle No. 2 

is completed. This means that the 

licensee could remove the repair during 

refueling outage No. 2. Some licensees 

may choose to remove the overlay 

during refueling outage No. 1 as the 

comment stated, but based on the 

interpretation described earlier, the 

repair does not need to be removed 

during refueling outage No. 1. 

No changes were made to RG 1.147, 

Revision 17, as a result of this comment. 

Code Case N–561–2 

Comment: Proposed Condition (2) on 

Code Case N–561–2 should be 

eliminated. Proposed Condition (2) 

prohibits the use of the exemption listed 

in paragraph 6(c)(1) of this case. The 

provisions in paragraph 6(c)(1) are 

identical to existing, approved 

provisions of IWA 4520, Examination, 

in the 2001 Edition of ASME Section XI. 

Weld overlays are base metal repairs, 

and are therefore already exempt by 

Section XI, IWA–4520 (2001 and later 

editions and addenda). This exemption 

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was only included in revision 2 of Code 

Cases N–561 and N–562; and also in 

Revision 1 of Code Case N–661–2 which 

was approved by Regulatory Guide 

1.147, Rev. 16, without this condition, 

to enable plants not yet implementing 

the 2001 or later edition and addenda to 

apply the exemption which had been 

accepted by the NRC in § 50.55a. 

Paragraph 6(a) of the case requires a 

surface examination of the completed 

weld overlay to provide additional 

assurance of the quality of the repair 

weld. ASME believes that this 

requirement is sufficient for Class 3 

applications in locations where the 

Construction Code would not require 

volumetric examination of full 

penetration butt welds in that location. 

Further, with the added condition of 

ultrasonically examining the base metal 

to verify absence of cracking, the benefit 

of/need for volumetric examination is 

significantly reduced. (ASME–5.2.2.b) 

NRC Response: The NRC agrees that 

proposed condition (2) can be 

eliminated. Paragraph 6(c)(1) of the 

Code Case states that ‘‘Class 3 weld 

overlays are exempt from volumetric 

examination when the Construction 

Code does not require the full 

penetration butt welds in the same 

location be volumetrically examined.’’ 

Section XI, paragraph IWA–4520(a)(1), 

2001 Edition and later, states that ‘‘Base 

metal repairs on Class 3 items are not 

required to be volumetrically examined 

when the Construction Code does not 

require that full-penetration butt welds 

in the same location be volumetrically 

examined.’’ As indicated in the 

comment, the exemptions are identical. 

The NRC unconditionally approved 

paragraph IWA–4520(a)(1) in the 2001 

Edition through 2008 Addenda. 

Therefore, it would be inconsistent to 

retain the condition on the Code Case. 

The NRC has removed proposed 

Condition (2) on Code Case N–561–2 

from the final RG 1.147, Revision 17. 

Code Case N–561–2 and N–661.2 

Comment: Proposed Condition (5) on 

Code Case N–561–2 is unwarranted and 

should be removed or modified. 

The rationale for this condition is to 

reduce the chances of producing a 

suspect weld (i.e., one made on a wet 

surface). Additionally, proposed 

Conditions (1), (2), (3), and (5) are 

unwarranted for reasons listed in 

comments provided on Code Case N 

561–2. 

Footnote 6 in Code Cases N–561–2 

and N–661–2 (and footnote 5 in N–562– 

2) states: ‘‘Testing has shown that 

piping with areas of wall thickness less 

than the diameter of the electrode may 

burn-through during application of a 

water-backed weld overlay.’’ Testing 

performed by the Electric Power 

Research Institute (EPRI) and described 

in EPRI Report TR–108131, ‘‘Weld 

Repair of Class 2 and 3 Ferritic Piping,’’ 

demonstrated that this criteria applies to 

application of weld overlays under both 

pressurized (up to 500 psi during the 

testing) and non-pressurized conditions 

(during this testing, specimens that 

burned-through were successfully 

welded-up using the shielded metal arc 

welding process with water leaking 

from the pipe; and those specimens 

passed the subsequent burst testing at 

pressures beyond the minimum burst 

pressure of new pipe). The results were 

the same in both situations—if the 

electrode diameter exceeded the 

thickness being welded, burn-through 

was likely—irrespective of internal 

pressure. If the thickness of the base 

metal equaled the thickness of the 

electrode, burn through would not 

occur, regardless of internal pressure. 

To require depressurization in such 

cases—in order to reduce the chances of 

producing a suspect weld—would cause 

extreme hardships, with no technical 

justification. 

Code Cases N–561–1, N–562–1, and 

N–661–1 each contained the statement: 

‘‘4(b) Piping with wall thickness less 

than the diameter of the electrode shall 

be depressurized before welding.’’ This 

was changed to a footnote for editorial 

purposes in revision 2 of each Code 

Case. If the NRC believes that Condition 

(5) must be retained in Table 2 of RG 

1.147, the ASME recommends that this 

condition be revised to read ‘‘Piping 

with wall thickness less than the 

diameter of the electrode shall be 

depressurized before welding.’’ This 

wording is consistent with that 

specified in paragraph 4(b) of Code Case 

N–661–1, which is currently listed in 

Table 2 of RG 1.147. (ASME–5.2.2.c and 

ASME–5.2.7) 

NRC Response: The NRC agrees with 

the comment. 

The NRC staff has reviewed the EPRI 

report and finds that the ASME 

recommendation has merit because it is 

supported by experimental data. The 

results of the research shows that if the 

thickness of the base metal equals the 

thickness of the electrode then burn 

through will not occur regardless of 

internal pressure. There were five 

conditions in the draft regulatory guide 

issued for public comment. The NRC 

agreed in a response to a separate 

comment (follows below) to remove 

condition (2) regarding the exemption 

from volumetric examination of Class 3 

weld overlays. Condition (5) in the draft 

regulatory guide has therefore been 

renumbered as condition (4) in the final 

regulatory guide, and the NRC has 

revised it consistent with the ASME 

recommendation. 

Comment: Proposed Conditions (1), 

(2), (3), and (5) are unwarranted for 

reasons listed in comments provided on 

Code Case N–561–2. However, if the 

NRC believes that Condition (5) must be 

retained in Table 2 of RG 1.147, this 

condition be revised to read ‘‘Piping 

with wall thickness less than the 

diameter of the electrode shall be 

depressurized before welding.’’ This 

wording is consistent with that 

specified in paragraph 4(b) of Code Case 

N–661–1, which is currently listed in 

Table 2 of RG 1.147. (ASME–5.2.3) 

NRC Response: Code Case N–562–2 is 

similar to Code Case N–561–2. 

Therefore, the NRC’s position on 

conditions in Code Case N–561–2 are 

also applicable to Code Case N–562–2. 

Therefore, the NRC has determined to 

retain Conditions (1) and (3) as 

proposed. Proposed Condition (2) has 

been removed; paragraph 6(c)(1) of the 

Code Case states that ‘‘Class 3 weld 

overlays are exempt from volumetric 

examination when the Construction 

Code does not require the full 

penetration butt welds in the same 

location be volumetrically examined.’’ 

Section XI, paragraph IWA–4520(a)(1), 

2001 Edition and later, states that ‘‘Base 

metal repairs on Class 3 items are not 

required to be volumetrically examined 

when the Construction Code does not 

require that full-penetration butt welds 

in the same location be volumetrically 

examined.’’ As indicated in the 

comment, the exemptions are identical. 

The NRC unconditionally approved 

paragraph IWA–4520(a)(1) in the 2001 

Edition through 2008 Addenda. 

Therefore, it would be inconsistent to 

retain the condition on the Code Case. 

Due to the removal of Condition (2), 

proposed Conditions (3), (4), and (5) 

have been renumbered as Conditions 

(2), (3), and (4). Proposed Condition (5) 

has been revised as recommended in the 

comment. 

Code Case N–597–2 

Comment: It is unclear whether 

proposed Condition (6) prohibits the use 

of the Code Case for moderate-energy 

Class 2 and 3 piping. If the intent of this 

condition is to allow the use of this case 

only until the next refueling outage for 

moderate-energy Class 2 and 3 piping, 

this condition should be clarified. In 

addition, the reference to Code Case N– 

513–2 should be removed from the 

proposed condition since Code Case N– 

513–3 is listed in Table 2 of RG 1.147. 

Because the condition imposed on the 

use of Code Case N–513–3 already 

restricts the use of N–513–3 until a 

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repair/replacement activity can be 

performed during the next refueling 

outage, the proposed condition is not 

needed for Code Case N–597–2. 

Proposed Condition (6) should, 

therefore, be removed or revised to 

clarify the intent. (ASME–5.2.4) 

NRC Response: The NRC disagrees 

with this comment. As discussed in the 

statement of considerations for the 

proposed rule (78 FR 37886; June 24, 

2013), the NRC had received a comment 

in a previous rulemaking (74 FR 26303; 

June 2, 2009), suggesting that the 

method described in Code Case N–513– 

2 for the temporary acceptance of flaws 

in moderate energy piping be added to 

Code Case N–597–2. The NRC agreed 

that it should be permissible under 

certain circumstances for licensees to 

evaluate local pipe wall thinning under 

Code Case N–597–2 without the NRC 

review and acceptance. The intent of 

Condition (6) was to reference the 

method in Code Case N–513–2 so that 

all of the provisions, formulas, graphs, 

and figures would not have to be 

duplicated in conditions to Code Case 

N–597–2. 

As also discussed in the statement of 

considerations for the proposed rule, the 

circumstances under which such an 

evaluation is conducted must be 

limited, because Code Case N–597–2 is 

applicable to all the ASME Code class 

piping (including high energy piping), 

whereas Code Case N–513–2 is limited 

to Class 2 and 3 moderate energy piping. 

The NRC has only approved temporary 

acceptance of flaws for moderate energy 

Class 2 or 3 piping (maximum operating 

temperature does not exceed 200 °F (93 

°C) and maximum operating pressure 

does not exceed 275 psig (1.9 MPa)). In 

addition, it is not appropriate to apply 

the method under Code Case N–597–2 

to evaluate through-wall leakage 

conditions. 

Condition (6) in the proposed rule 

stated, ‘‘For moderate-energy Class 2 

and 3 piping, wall thinning acceptance 

criteria may be determined on a 

temporary basis (until the next refueling 

outage) based on the provisions of Code 

Case N–513–2. Moderate-energy piping 

is defined as Class 2 and 3 piping whose 

maximum operating temperature does 

not exceed 200 °F (93 °C) and whose 

maximum operating pressure does not 

exceed 275 psig (1.9 MPa). Code Case 

N–597–2 shall not be used to evaluate 

through-wall leakage conditions.’’ 

This condition has been revised in RG 

1.147, Revision 17, to read as follows: 

‘‘The evaluation criteria in Code Case 

N–513–2 may be applied to Code Case 

N–597–2 for the temporary acceptance 

of wall thinning (until the next refueling 

outage) for moderate-energy Class 2 and 

3 piping. Moderate-energy piping is 

defined as Class 2 and 3 piping whose 

maximum operating temperature does 

not exceed 200 °F (93 °C) and whose 

maximum operating pressure does not 

exceed 275 psig (1.9 MPa). Code Case 

N–597–2 shall not be used to evaluate 

through-wall leakage conditions.’’ 

Code Case N–606–1 

Comment: The proposed condition to 

Code Case N–606–1 is already 

inherently required. 

The surface preparation and cleaning 

prior to welding are considered to be 

standard requirements by Welding 

Programs complying with § 50.55a 

specified Codes and 10 CFR part 50, 

appendix B Quality Assurance 

Programs. Furthermore, these 

requirements are already required/

implied by the reference to the ASME 

Section IX and paragraph 3(e) of the 

Case. Many other instances where 

welding is performed, even temper bead 

welding, can be found in Code Cases 

and in Code that do not explicitly 

specify this level of detail since such 

details are included in the Owner’s or 

the Owner’s Repair Organization’s 

Welding Procedure Specification/

Welding Program. Therefore, this 

condition should be removed from the 

regulatory guide. (ASME–5.2.5) 

NRC Response: The NRC agrees that, 

the second sentence of the proposed 

condition is redundant with 

requirements in Section III NB–4412. 

The NRC removed the second sentence 

of the condition. 

The NRC disagrees with the 

comment’s suggestion to remove the 

first and third sentences of the 

condition. The original version of Code 

Case N–606, and other temper bead 

Code Cases (such as N–638–5), require 

that prior to welding base metal, a 

surface examination shall be performed 

on the area to be welded, so there is 

precedent for this level of detail in 

temper bead Code Cases. This 

verification is not required by Section 

IX of the ASME Code. The NRC has 

determined that this verification is 

necessary to assure the necessary 

quality level for temper bead welding. 

Therefore, the condition is necessary. 

No change was made to the first and 

third sentences of the condition in 

response to this comment. 

Code Case N–619 and N–648–1 

Comment: The NRC should not 

include the condition to Code Case N– 

619 and N–648–1 which requires the 1- 

mil wire standard for qualification of 

visual examinations for components 

within the scope of these code cases. 

Research has shown that characters on 

a printed chart are a better resolution 

standard than the use of 1-mil wire. 

The use of printed characters for 

qualification will improve the 

resolution of visual examinations, thus 

improving the capability of the 

technique in detecting indications for 

which the examinations are performed. 

(ASME–5.2.6.a, ASME–5.2.6.b) 

NRC Response: Visual resolution 

sensitivity techniques are used to ensure 

the capabilities of the examiner, and 

that a camera, when used, is operating 

properly. The NRC conducted a 

preliminary assessment of remote visual 

testing at Pacific Northwest National 

Laboratory. The results were published 

in NUREG/CR–6860, ‘‘An Assessment of 

Visual Testing,’’ which is available on 

the NRC’s public Web site at http://

www.nrc.gov/reading-rm/doccollections/nuregs/contract/. The 1-mil 

wire standard had been implemented in 

response to the requirement in the 

condition for a resolution sensitivity of 

1-mil. The preliminary assessment 

identified issues with respect to the 

accuracy of using a wire as a 

performance demonstration standard. 

Other issues were also identified. This 

led to the development of a cooperative 

research program between the NRC and 

the EPRI. This is the research effort 

referenced in ASME’s comment. While 

issues had been identified with the use 

of a wire standard, the NRC decided to 

not consider changes in the condition to 

Code Case N–619 until the cooperative 

research had progressed, and it could be 

determined if there were other issues 

that should be considered regarding 

visual examination. 

The research has not identified any 

issues calling into question the use of 

characters as a resolution standard. In 

addition as described in NUREG/CR– 

6860, the research demonstrated that the 

character resolution standard was 

superior to the wire standard. The NRC 

finds the ASME’s suggestion to remove 

the requirement for a 1-mil wire for VT– 

1 procedure demonstration acceptable. 

The condition has been revised to 

remove the 1-mil wire standard and to 

allow the use of printed characters. 

Code Case N–702 

Comment: The proposed condition for 

Code Case N–702 should be modified to 

reference BWRVIP–241: BWR Vessel 

and Internals Project, ‘‘Probabilistic 

Fracture Mechanics Evaluation for the 

Boiling Water Reactor Nozzle-to-Vessel 

Shell Welds and Nozzle Blend Radii,’’ 

EPRI Technical Report 1021005, 

October 2010 (ADAMS Accession No. 

ML11119A041). The proposed 

condition should be revised to read as 

follows: (ASME–5.2.8) 

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The technical basis supporting the 

implementation of this Code Case is 

addressed by BWRVIP–108, and BWRVIP– 

241. The applicability of Code Case N–702 

must be shown by demonstrating that the 

criteria in Section 5.0 of NRC Safety 

Evaluation regarding BWRVIP–108 dated 

December 18, 2007 (ADAMS Accession No. 

ML073600374), or Section 5.0 of NRC Safety 

Evaluation regarding BWRVIP–241 dated 

April 19, 2013 (ADAMS Accession No. 

ML13071A240), are met. The evaluation 

demonstrating the applicability of the Code 

Case shall be reviewed and approved by the 

NRC prior to the application of the Code 

Case. 

NRC Response: The NRC agrees with 

the suggestion to reference BWRVIP– 

241 in the condition. By letter dated 

April 19, 2013 (ADAMS Accession No. 

ML13071A233), to the Chairman of the 

BWR Vessel and Internals Project, the 

NRC stated that BWRVIP–241 was 

acceptable for referencing subject to the 

limitations specified in the technical 

report and in the NRC Safety 

Evaluation. The BWRVIP–241 was not 

referenced in the proposed condition to 

ASME Code Case N–702 because the 

draft RG was already in the review 

process when the NRC Safety 

Evaluation for BWRVIP–241 was 

released. The basis for including 

BWRVIP–241 in the reference is as 

follows. 

The BWRVIP–108 provides the 

technical basis document for ASME 

Code Case N–702 regarding reduction of 

the inspection of reactor pressure vessel 

(RPV) nozzle-to-vessel shell welds and 

nozzle inner radius areas from 100 

percent to 25 percent for each nozzle 

type every 10 years. The BWRVIP–241 

provides additional probabilistic 

fracture mechanics (PFM) analyses to 

support its proposed changes to the 

NRC staff’s criteria specified in the 

Safety Evaluation on BWRVIP–108. 

Based on the additional PFM results 

supporting the revised criteria, along 

with BWR RPV inspection results which 

show no indications of inservice 

degradation, the NRC staff determined 

that the inspection of 25 percent of each 

RPV nozzle type each 10-year interval is 

justified. 

Licensees who plan to request relief 

from the ASME Code, Section XI 

requirements for RPV nozzle-to-vessel 

shell welds and nozzle inner radius 

sections may reference the BWRVIP–241 

report as the technical basis for the use 

of ASME Code Case N–702 as an 

alternative. However, licensees should 

demonstrate the plant-specific 

applicability of the BWRVIP–241 report 

to their units in the relief request by 

addressing the conditions and 

limitations specified in Section 5.0 of 

the NRC Safety Evaluation for BWRVIP– 

241. The suggested condition is 

identical to the proposed condition in 

the draft RG other than adding the 

reference to BWRVIP–241 in two places. 

Therefore, the NRC finds the comment’s 

proposal to be acceptable. 

The condition on ASME Code Case 

N–702 has been revised to reference 

BWRVIP–241. 

Code Case N–739–1 

Comment: The American Concrete 

Institute (ACI) report referenced in the 

condition to Code Case N–739–1 should 

be clarified to reference ACI 201.1R. 

Note that the ASME has taken action to 

issue an erratum to correct this error in 

the Code Case and Section XI. The 

reference to ACI 201.1 R is correctly 

shown in Table IWA–1600–1. (ASME– 

5.2.9) 

NRC Response: The NRC agrees with 

the comment. The letter ‘‘R’’ was 

missing in the reference in Code Case 

N–739–1. The ACI uses the letter ‘‘R’’ to 

distinguish reports from standards. With 

the ASME approval of an erratum to the 

Code Case restoring the letter ‘‘R,’’ the 

NRC can remove the condition in final 

RG 1.147, Revision 17. 

The NRC has unconditionally 

approved Code Case N–739–1 in RG 

1.147, Revision 17. 

Code Cases N–798 and N–800 

Comment: Although Code Cases N– 

798 and N–800 have not been included 

in DG–1231, the NRC should include 

both of these cases in the next draft 

revision to RG 1.147. Until such time 

that N–798 and N–800 are included in 

RG 1.147, owners will continue to seek 

relief pursuant to § 50.55a(a)(3) 

[§ 50.55a(z) in the draft rule] to use 

provisions of these cases or similar 

alternatives. (ASME–5.2.10) 

NRC Response: The NRC agrees with 

the comment and plans to address these 

code cases in Supplement 11 to the 

2007 Edition through Supplement 10 to 

the 2010 Edition in draft Revision 18 to 

RG 1.147. Code Cases N–798 and N–800 

were not included in the draft 

regulatory guide because they were 

issued in Supplement 4 to the 2010 

Edition, which was not considered for 

this regulatory guide. 

No change was made to this final rule 

as a result of this comment. 

Regulatory Guide 1.192, Revision 1 

(DG–1232) 

Code Case OMN–1 

Comment: DG–1232 incorrectly 

identifies ASME Code Case OMN–1 

(2006 Addenda) as ‘‘Revision 0.’’ The 

version of OMN–1 published with the 

2006 Addenda does not include the 

identifier, ‘‘Revision 0.’’ (Comstock-2.1) 

NRC Response: The NRC agrees with 

this comment. The ASME OMN–1 Code 

Case published with the 2006 Addenda 

did not include the identifier ‘‘Revision 

0.’’ Accordingly, RG 1.192, Revision 1, 

has been revised to remove the words 

‘‘Revision 0’’ from the first sentence of 

the first paragraph in Table 2, under 

OMN–1 conditions. 

Comment: The descriptions in the 

first and second sentence say OMN–1 

may be used in lieu of the provisions for 

stroke time testing. However, OMN–1 

says it may be used in place of all 

provisions with the exception of leak 

testing. The conditions placed on the 

use of OMN–1 restrict its use in place 

of existing other ISTC requirements, 

such as position indication verification 

and periodic (quarterly, cold shutdown, 

refueling outage) exercising. All 

provisions of ISTC are implemented in 

OMN–1 with the exception of leak 

testing. The leak testing requirement of 

ISTC is referenced as a necessary 

requirement by the Code Case. Strike 

out the words ‘‘stroke-time’’ in the first 

and second sentences of Table 2 in DG– 

1232 to resolve this problem. 

(Comstock-2.2) 

NRC Response: The NRC disagrees 

with this comment. The general 

discrepancy noted in the comment is 

that draft RG 1.192 (DG–1232) states 

OMN–1 ‘‘may be used in lieu of the 

provisions for stroke time testing’’ 

versus OMN–1, which states ‘‘it may be 

used in place of all provisions.’’ After 

evaluating the comment, the NRC 

believes both statements are correct and 

the same for the following reasons. 

The requirements of the ASME OM 

Code, Subsection ISTC, can be 

simplified as having three test 

requirements: 

1. ISTC–3500—‘‘Valve Testing 

Requirements’’ 

2. ISTC–3600—‘‘Leak Testing 

Requirements’’ 

3. ISTC–3700—‘‘Position Verification 

Testing’’ 

Section ISTC–3500 of the ASME OM 

Code describes valve test requirements, 

such as exercise test frequency and 

obturator movement verification. 

Specific instructions for the different 

valve types can be found in Section 

ISTC–5000, ‘‘Specific Testing 

Requirements,’’ of the ASME OM Code. 

The ASME OM Code section for specific 

test requirements for motor-operated 

valves (MOVs) is ISTC–5120. The first 

specific instruction for an MOV test is 

ISTC–5121(a), ‘‘Valve Stroke Testing,’’ 

which states, ‘‘Active valves shall have 

their stroke times measured when 

exercised in accordance with ISTC– 

3500.’’ The specific instruction for the 

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stroke-time test encompasses all the 

requirements of ISTC–3500. Leak testing 

requirement ISTC–3600 remains the 

same. The position verification test is 

not specifically spelled out in the ASME 

OM Code Case OMN–1, but credit is 

given on the basis that OMN–1 requires 

diagnostic testing of MOVs to verify that 

they are set up correctly and will meet 

their design basis function. 

The comment also stated that all 

provisions of ISTC are implemented in 

OMN–1. This statement is not fully 

accurate. After a recent industry valve 

failure, it has been noted by the ASME 

OM Code Subgroup committee on 

MOVs that the ASME OM Code Case 

OMN–1 does not directly address the 

issue of verifying obturator movement, 

which is required in Section ISTC–3530. 

The subgroup committees for ISTC and 

MOVs are currently working on 

addressing this issue. Also, a review of 

past NRC documents, regulatory guides, 

and safety evaluations were completed. 

The majority of the NRC 

correspondence refers to ASME OM 

Code requirements for MOVs as being 

‘‘stroke time testing.’’ 

No change has been made to RG 

1.192, Revision 1, as a result of this 

comment. 

Code Case OMN–11 

Comment: In DG–1232, delete the first 

sentence in Condition (2) on OMN–11 

(2006 Addenda). It exceeds the NRC’s 

authority. 

In DG–1232, the conditions on OMN– 

11 (2006 addenda) add an unnecessary 

administrative burden. 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), Condition (1) 

should be deleted. This defeats the 

purpose of alternate requirements. 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), Condition (2) 

should be deleted. The OMN–11 3(b) 

rule requires the same treatment to be 

applied as OMN–1 3.5(b) by requiring 

an evaluation of all test results for every 

MOV in the group. The OMN–11 3(d) 

rule requires all low safety significant 

components (LSSC) to be tested over a 

10-year period. This requires the same 

treatment to be applied as OMN–1 

3.5(d) over a 10-year period, which 

requires testing for all valves in the 

group. The OMN–1 3.5(e) simply says 

the test results for a representative MOV 

from the group shall be applied to all 

MOVs in the group when doing the 

section 6 analyses and evaluation. This 

is the same rule described within the 

OMN–11 3(b) requirement that requires 

test results from an individual valve 

within a group to be applied to all 

MOVs within the group. 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), Condition (3) 

should be deleted. It is already imposed 

for OMN–1 (required for OMN–11). 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), note 1 should 

be deleted because it is circular and 

provides no guidance or information. 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), note 2 directs 

the reader to the wrong edition (2004) 

for OMN–1. If it referenced 2006, it 

would not provide any new 

information. 

In DG–1232, in the discussion of 

OMN–11 (2006 addenda), note 3 should 

be incorporated into Table 2 OMN–1 

note 2 or deleted. (Comstock-2.3) 

NRC Response: The NRC agrees that 

the specification of conditions in Table 

2 of RG 1.192 on Code Case OMN–11 in 

the 2006 Addenda of the ASME OM 

Code is not necessary because OMN–1 

in the 2006 Addenda has incorporated 

the provisions from OMN–11. 

Therefore, OMN–11 has been deleted 

from Table 2 of RG 1.192. A new Note 

2 has been included for OMN–1 in 

Table 2 of RG 1.192 explaining the 

incorporation of OMN–11 into OMN–1 

such that the use of OMN–11 in the 

2006 Addenda is no longer appropriate. 

Table 3 of RG 1.192 continues to specify 

conditions for the use of OMN–11 in the 

2001 Edition, 2003 Addenda, and 2004 

Edition of the OM Code for those 

superseded versions of OMN–11. In 

particular, Condition (1) on OMN–11 

indicates that all provisions in OMN–1 

must be satisfied, except those allowed 

to be relaxed by the risk-informed 

provisions in OMN–11. Condition (2) on 

OMN–11 indicates that only specific 

provisions for grouping of MOVs in 

OMN–1 may be relaxed through the use 

of OMN–11. Condition (3) on OMN–11 

is repeated from a similar condition on 

OMN–1 because OMN–11 has a specific 

section on high risk MOVs. Note 1 on 

OMN–11 in Table 3 of RG 1.192 

indicates that the permission to use 

allowable risk ranking methodologies 

applies to both OMN–1 and OMN–11. 

There are no additional notes on OMN– 

11 in Table 3 of RG 1.192. 

Code Case OMN–12 

Comment: Code Case OMN–12 should 

be removed from DG–1232 since its 

application will always require NRC 

permission to implement due to the 

ASME OM Code for which it applies. 

The conditions described for the use of 

ASME Code Case OMN–12 do not allow 

it to be applied to any other ASME OM 

Code for which it was written (ASME 

OM Code 1998). In light of the current 

10 CFR 50.55a regulations, this renders 

the Code Case unusable for anyone in 

the USA through the application of RG 

1.192. The extra conditions also make 

the application of OMN–12 so 

burdensome, that no one would be 

willing to incur the extra expense and 

administrative burden associated with 

implementing this process under the 

Inservice Testing Program. (Comstock2.4) 

NRC Response: The NRC disagrees 

with this comment. The comment seems 

to be interpreting that the NRC is 

endorsing the use of OMN–12 only if 

the licensee’s IST Program is based on 

the 1998 Code. That is not the case. The 

NRC accepts with conditions the use of 

OMN–12 with any Code from 1998 up 

to and including the 2006 Addenda. 

No change has been made to the final 

rule as a result of this comment. 

Table 3—Code Cases That Have Been 

Superseded by Revised Code Cases 

Comment: Table 3 of DG–1232 should 

be deleted. It serves no useful purpose. 

The information is available via other 

sources. It delays the rule. (Comstock– 

2.5) 

NRC Response: The NRC disagrees 

with this comment. Table 3 in RG 1.192 

lists those OM Code Cases that have 

been superseded by revised Code Cases. 

Similar tables exist in RGs 1.84 and 

1.147 addressing Section III and Section 

XI Code Cases respectively. Section 

50.55a allows applicants and licensees 

to continue to apply superseded Code 

Cases for the remainder of an inservice 

inspection or testing interval. The 

ASME procedures require that the latest 

version of a Code Case be implemented. 

If not for the provision in the regulation, 

licensees would be required to update 

their inservice inspection and testing 

programs for every Code Case that is 

revised (i.e., that the licensee or 

applicant had previously implemented). 

Accordingly, any Code and standard 

that has been incorporated by reference 

into § 50.55a and is still in use must 

continue to be listed in the regulation. 

No change has been made to RG 

1.192, Revision 1, as a result of this 

comment. 

Regulatory Guide 1.193, Revision 4 

(DG–1233) 

Code Case N–659–2 

Comment: In DG–1233, in the 

discussion of N–659–2, there is a 

typographical error on page 7. It should 

say ‘‘radiography,’’ not ‘‘radiology.’’ 

(ASME–5.4.1) 

NRC Response: The NRC agrees with 

this comment. 

The NRC corrected the title of Code 

Case N–659–2 in RG 1.193, Revision 4. 

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65786 Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 

N–805 

Comment: The U.S. Nuclear 

Regulatory Commission (NRC) should 

consider including in this rulemaking 

Code Case N–805, ‘‘Alternative to Class 

1 Extended Boundary End of lnterval or 

Class 2 System Leakage Testing of the 

Reactor Vessel Head Flange O-Ring 

Leak-Detection System Section XI, 

Division 1.’’ (Inservice Inspection 

Program Owners Group–1.1) 

NRC Response: The NRC declines to 

adopt the suggestion to adopt Code Case 

N–805 in the final rulemaking and final 

regulatory guide. Code Case N–805 was 

published by the ASME in Supplement 

6 to the 2010 Edition which was not 

considered for inclusion in this 

rulemaking and draft regulatory guide. 

The NRC plans to include Code Case N– 

805 in draft Revision 18 to RG 1.147 

which is scheduled for public comment 

in spring 2015. 

No change was made to the final rule 

as a result of this comment. 

IV. NRC Approval of New and 

Amended ASME Code Cases 

This final rule incorporates by 

reference the latest revisions of the 

NRC’s RGs that list ASME BPV and OM 

Code Cases the NRC finds to be 

acceptable or ‘‘conditionally 

acceptable’’ (i.e., NRC-specified 

conditions). Regulatory Guide 1.84, 

Revision 36 (ADAMS Accession No. 

ML13339A515), supersedes the 

incorporation by reference of Revision 

35; RG 1.147, Revision 17 (ADAMS 

Accession No. ML13339A689), 

supersedes the incorporation by 

reference of Revision 16; and RG 1.192, 

Revision 1 (ADAMS Accession No. 

ML13340A034), supersedes the 

incorporation by reference of Revision 

0. 

This final rule addresses two 

categories of ASME Code Cases. The 

first category of Code Cases are the new 

and revised Section III and Section XI 

Code Cases listed in Supplements 1 

through 10 to the 2007 Edition of the 

BPV Code, and the OM Code Cases 

published with the 2002 Addenda 

through the 2006 Addenda. The second 

category is the Code Cases that were not 

addressed in the final rule published in 

the Federal Register on October 5, 2010 

(75 FR 61321). The 2010 final rule 

addressed the new and revised Section 

III and Section XI Code Cases listed in 

Supplements 2 through 11 to the 2004 

Edition and Supplement 0 to the 2007 

Edition of BPV Code. Public comments 

were received during the proposed rule 

stage (June 2, 2009; 74 FR 26303) on 

(Code Cases N–508–4, N–597–2, N–619, 

N–648, N–702, and N–748) requesting 

that the NRC include certain revised 

Code Cases in the final guides that were 

not listed in the draft guides. The NRC 

determined that the revised Code Cases 

represented changes significant enough 

to warrant broader public participation 

prior to the NRC making a final 

determination of them. Accordingly, the 

NRC requested comment on these Code 

Cases in the proposed rule (June 24, 

2013; 78 FR 37886). The comment 

responses shown earlier include 

responses to those Code Cases. 

The latest editions and addenda of the 

ASME BPV and OM Codes that the NRC 

has approved for use are referenced in 

§ 50.55a. The ASME also publishes 

Code Cases that provide alternatives to 

existing Code requirements developed 

and approved by ASME. The final rule 

incorporated by reference RGs 1.84, 

1.147, and 1.192. The NRC, by 

incorporating by reference these three 

RGs, allows nuclear power plant 

licensees and applicants for standard 

design certifications, standard design 

approvals, manufacturing licenses, 

applicants for OLs, CPs, and COLs 

under the regulations that govern 

license certifications, to use the Code 

Cases listed in these RGs as suitable 

alternatives to the ASME BPV and OM 

Codes for the construction, ISI, and IST 

of nuclear power plant components. 

This action is consistent with the 

provisions of the National Technology 

Transfer and Advancement Act of 1995, 

Public Law 104–113, which encourages 

Federal regulatory agencies to consider 

adopting industry consensus standards 

as an alternative to de novo agency 

development of standards affecting an 

industry. This action is also consistent 

with the NRC’s policy of evaluating the 

latest versions of consensus standards in 

terms of their suitability for 

endorsement by regulations or 

regulatory guides. 

The NRC follows a three-step process 

to determine the acceptability of new 

and revised Code Cases and the need for 

regulatory positions on the uses of these 

Code Cases. This process was employed 

in the review of the Code Cases in 

Supplements 1 through 10 to the 2007 

Edition of the BPV Code and the 2002 

Addenda through the 2006 Addenda of 

the OM Code. The Code Cases in these 

supplements are the subject of this final 

rule. First, the ASME develops Code 

Cases through a consensus development 

process, as administered by ANSI, 

which ensures that the various technical 

interests (e.g., utility, manufacturing, 

insurance, regulatory) are represented 

on standards development committees 

and that their viewpoints are addressed 

fairly. This process includes 

development of a technical justification 

in support of each new or revised Code 

Case. The ASME committee meetings 

are open to the public, and attendees are 

encouraged to participate. Task groups, 

working groups, and subgroups report to 

a standards committee. The standards 

committee is the decisive consensus 

committee and ensures that the 

development process fully complies 

with the ANSI consensus process. The 

NRC actively participates through full 

involvement in discussions and 

technical debates of the task groups, 

working groups, subgroups, and 

standards committee regarding the 

development of new and revised 

standards. 

Second, the standards committee 

transmits to its members a first 

consideration letter ballot requesting 

comment or approval of new and 

revised Code Cases. To be approved, 

Code Cases from the first consideration 

letter ballot must receive the following: 

(1) Approval votes from at least two 

thirds of the eligible consensus 

committee membership, (2) no 

disapprovals from the standards 

committee, and (3) no substantive 

comments from ASME oversight 

committees such as the Technical 

Oversight Management Committee 

(TOMC). The TOMC’s duties, in part, 

are to oversee various standards 

committees to ensure technical 

adequacy and provide recommendations 

in the development of Codes and 

standards, as required. The Code Cases 

that are disapproved or receive 

substantive comments from the first 

consideration ballot are reviewed by the 

working level group(s) responsible for 

their development to consider the 

comments received. These Code Cases 

may be approved by the standards 

committee on second consideration 

with an approval vote by at least two 

thirds of the eligible consensus 

committee membership, with no more 

than three disapprovals from the 

consensus committee. 

Third, the NRC reviews new and 

revised Code Cases to determine their 

acceptability for incorporation by 

reference in § 50.55a through the subject 

RGs. This rulemaking process, when 

considered together with the ANSI 

process for developing and approving 

ASME codes and standards and ASME 

Code Cases, constitutes the NRC’s basis 

that the Code Cases (with conditions as 

necessary) provide reasonable assurance 

of adequate protection to public health 

and safety. 

The NRC reviewed the new and 

revised Code Cases identified in this 

final rule and concluded, in accordance 

with the process previously described, 

that the Code Cases are technically 

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adequate (with conditions as necessary) 

and consistent with current NRC 

regulations. Therefore, the new and 

revised Code Cases listed in the subject 

RGs are approved for use subject to any 

specified conditions. 

A. ASME Code Cases Approved for 

Unconditional Use 

The NRC determined, in accordance 

with the process previously described 

for review of ASME Code Cases, that 

each ASME Code Case listed in Table II 

is appropriate for incorporation by 

reference and has been newly added to 

the RGs 

TABLE II—UNCONDITIONALLY APPROVED CODE CASES

Code case No. Code supplement Code case title 

ASME BPV Code Case, Section III 

N–4–13 ............................................ 5 ..................................................... Special Type 403 Modified Forgings or Bars, Section III, Division 1, 

Class 1 and CS. 

N–570–2 .......................................... 7 ..................................................... Alternative Rules for Linear Piping and Linear Standard Supports for 

Classes 1, 2, 3, and MC, Section III, Division 1. 

N–580–2 .......................................... 4 ..................................................... Use of Alloy 600 With Columbium Added, Section III, Division 1. 

N–655–1 .......................................... 2 ..................................................... Use of SA–738, Grade B, for Metal Containment Vessels, Class MC, 

Section III, Division 1. 

N–708 .............................................. 2 ..................................................... Use of JIS G–4303, Grades SUS304, SUS304L, SUS316, and 

SUS316L, Section III, Division 1. 

N–759–2 .......................................... 4 ..................................................... Alternative Rules for Determining Allowable External Pressure and 

Comprehensive Stress for Cylinders, Cones, Spheres, and Formed 

Heads, Section III, Division 1. 

N–760–2 .......................................... 7 ..................................................... Welding of Valve Plugs to Valve Stem Retainers, Classes 1, 2, and 

3, Section III, Division 1. 

N–767 .............................................. 4 ..................................................... Use of 21 Cr-6Ni-9Mn (Alloy UNS S21904) Grade GXM–11 (Conforming to SA 182/SA–182M and SA–336/SA–336M), Grade 

TPXM–11 (Conforming to SA 312/SA–312M) and Type XM–11 

(Conforming to SA–666) Material, for Class 1 Construction, Section 

III, Division 1. 

N–774 .............................................. 7 ..................................................... Use of 13Cr-4Ni (Alloy UNS S41500) Grade F6NM Forgings Weighing in Excess of 10,000 lb (4,540 kg) and Otherwise conforming to 

the Requirements of SA–336/SA–336M for Class 1, 2, and 3 Construction, Section III, Division 1. 

N–782 .............................................. 9 ..................................................... Use of Editions, Addenda, and Cases, Section III, Division 1. 

N–801 .............................................. 4 (2010 Edition) ............................. Rules for Repair of N-Stamped Class 1, 2, and 3 Components by Organization Other Than the N Certificate Holder That Originally 

Stamped the Component Being Repaired, Section III, Division 1. 

N–802 .............................................. 4 (2010 Edition) ............................. Rules for Repair of Stamped Components by the N Certificate Holder 

That Originally Stamped the Component, Section III, Division 1. 

ASME BPV Code Case, Section XI 

N–532–5 .......................................... 5 ..................................................... Alternative Requirements to Repair and Replacement Documentation 

Requirements and Inservice Summary Report Preparation and 

Submission as Required by IWA–4000 and IWA–6000, Section XI, 

Division 1. 

N–716–1 .......................................... 1 (2013 Edition) ............................. Alternative Piping Classification and Examination Requirements, Section XI, Division 1. 

N–739–1 .......................................... 1 ..................................................... Alternative Qualification Requirements for Personnel Performing 

Class CC Concrete and Post-Tensioning System Visual Examinations, Section XI, Division 1. 

N–747 .............................................. 9 ..................................................... Reactor Vessel Head-to-Flange Weld Examinations, Section XI, Division 1. 

N–762 .............................................. 1 ..................................................... Temper Bead Procedure Qualification Requirements for Repair/Replacement Activities Without Post Weld Heat Treatment, Section 

XI, Division 1. 

N–765 .............................................. 8 ..................................................... Alternative to Inspection Interval Scheduling Requirements of IWA– 

2430, Section XI, Division 1. 

N–769 .............................................. 8 ..................................................... Roll Expansion of Class 1 In-Core Housing Bottom Head Penetrations in BWRs, Section XI, Division 1. 

N–773 .............................................. 8 ..................................................... Alternative Qualification Criteria for Eddy Current Examinations of 

Piping Inside Surfaces, Section XI, Division 1. 

ASME OM Code Case 

OMN–6 ............................................ 2006 Addenda ............................... Alternate Rules for Digital Instruments. 

OMN–8 ............................................ 2006 Addenda ............................... Alternative Rules for Preservice and Inservice Testing of Power-Operated Valves That Are Used for System Control and Have a Safety Function per OM–10, ISTC–1.1, or ISTA–1100. 

OMN–14 .......................................... 2004 Addenda ............................... Alternative Rules for Valve Testing Operations and Maintenance, Appendix I: BWR CRD Rupture Disk Exclusion. 

OMN–16 .......................................... 2006 Addenda ............................... Use of a Pump Curve for Testing. 

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B. ASME Code Cases Approved for Use 

With Conditions 

The NRC has determined that certain 

Code Cases, as issued by ASME, are 

generally acceptable for use, but that the 

alternative requirements specified in 

those Code Cases must be supplemented 

to provide an acceptable level of quality 

and safety. Accordingly, the NRC 

proposes to impose conditions on the 

use of these Code Cases to modify, limit 

or clarify their requirements. For each 

applicable Code Case, the conditions 

would specify the additional activities 

that must be performed, the limits on 

the activities specified in the Code Case, 

and/or the supplemental information 

needed to provide clarity. These ASME 

Code Cases are included in Table III of 

the following: RG 1.84 (DG–1230), RG 

1.147 (DG–1231), and RG 1.192 (DG– 

1232). The NRC’s evaluation of the Code 

Cases and the reasons for the NRC’s 

conditions are discussed in the 

following paragraphs. 

TABLE III—CONDITIONALLY APPROVED CODE CASES

Code case No. Code supplement Code case title Conditions 

ASME BPV Code Case, Section III 

N–60–5 .......................... Reinstating condition ....................................... Material for Core Support Structures, Section 

III, Division I, Class 1.

The maximum yield strength of strain-hardened austenitic stainless 

steel shall not exceed 90,000 psi in view of the susceptibility of 

this material to environmental cracking. 

N–208–2 ........................ 4 ....................................................................... Fatigue Analysis for Precipitation Hardening 

Nickel Alloy Bolting Material to Specification 

SB–637 N07718 for Class 1 Construction, 

Section III, Division 1.

(1) In Figure A, the words ‘‘No mean stress’’ shall be implemented 

with the understanding that it denotes ‘‘Maximum mean stress.’’ 

(2) In Figure A, sy shall be implemented with the understanding that 

it denotes smax. 

N–520–2 ........................ 4 ....................................................................... Alternative Rules for Renewal of Active or Expired N-type Certificates for Plants Not in 

Active Construction, Section III, Division 1.

The Code Case is considered acceptable with one clarification: an 

AIA is an Authorized Inspection Agency and the AIA employs the 

Authorized Nuclear Inspector (ANI). 

N–757–1 ........................ 2 ....................................................................... Alternative Rules for Acceptability for Class 2 

and 3 Valves (DN 25) and Smaller with 

Welded and Nonwelded End Connections 

Other than Flanges, Section III, Division 1.

The design provisions of ASME Section III, Division 1, Appendix XIII, 

shall not be used for Class 3 valves. 

ASME BPV Code Case, Section XI 

N–508–4 ........................ 8 ....................................................................... Rotation of Serviced Snubbers and Pressure 

Retaining Items for the Purpose of Testing, 

Section XI, Division 1.

When Section XI requirements are used to govern the examination 

and testing of snubbers and the ISI Code of Record is earlier than 

Section XI, 2006 Addenda, Footnote 1 shall not be applied. 

N–561–2 ........................ 1 ....................................................................... Alternative Requirements for Wall Thickness 

Restoration of Class 2 and High Energy 

Class 3 Carbon Steel Piping, Section XI, 

Division 1.

(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage. 

(2) Paragraph 7(c): if the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage. 

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like 

defects exist. 

(4) Piping with wall thickness less than the diameter of the electrode 

shall be depressurized before welding. 

N–562–2 ........................ 1 ....................................................................... Alternative Requirements for Wall Thickness 

Restoration of Class 3 Moderate Energy 

Carbon Steel Piping, Section XI, Division 1.

(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage. 

(2) Paragraph 7(c): if the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage. 

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like 

defects exist. 

(4) Piping with wall thickness less than the diameter of the electrode 

shall be depressurized before welding. 

N–597–2 ........................ Previously approved Code Case. NRC had 

proposed one new condition in response to 

public comment on last rulemaking.

Requirements for Analytical Evaluation of 

Pipe Wall Thinning, Section XI, Division 1.

New condition (6): The evaluation criteria in Code Case N–513–2 

may be applied to Code Case N–597–2 for temporary acceptance 

of wall thinning (until the next refueling outage) for moderate-energy Class 2 and 3 piping. Moderate-energy piping is defined as 

Class 2 and 3 piping whose maximum operating temperature does 

not exceed 200 °F (93 °C) and whose maximum operating pressure does not exceed 275 psig (1.9MPa). Code Case N-597–2 

shall not be used to evaluate through-wall leakage conditions. 

N–606–1 ........................ Public comment received on previously approved rule requesting revision to condition. 

Condition was revised.

Similar and Dissimilar Metal Welding Using 

Ambient Temperature Machine GTAW 

Temper Bead Technique for BWR CRD 

Housing/Stub Tube Repairs, Section XI, Division 1.

Prior to welding, an examination or verification must be performed to 

ensure proper preparation of the base metal, and that the surface 

is properly contoured so that an acceptable weld can be produced. 

This verification is to be required in the welding procedures. 

N–619 ............................ Responding to comment on previously approved Code Case.Alternative Requirements for Nozzle Inner Radius Inspections for Class 1 Pressurizer 

and Steam Generator Nozzles, Section XI, 

Division 1.

In lieu of a UT examination, licensees may perform a VT–1 examination in accordance with the code of record for the Inservice Inspection Program utilizing the allowable flaw length criteria of Table 

IWB–3512–1 with limiting assumptions on the flaw aspect ratio. 

N–648–1 ........................ Responding to comment on previously approved Code Case.Alternative Requirements for Inner Radius Inspections for Class 1 Reactor Vessel Nozzles, Section XI, Division 1.

In lieu of a UT examination, licensees may perform a VT–1 examination in accordance with the code of record for the Inservice Inspection Program utilizing the allowable flaw length criteria of Table 

IWB–3512–1 with limiting assumptions on the flaw aspect ratio. 

N–661–2 ........................ 1 ....................................................................... Alternative Requirements for Wall Thickness 

Restoration of Classes 2 and 3 Carbon 

Steel Piping for Raw Water Service, Section XI, Division 1.

(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage. 

(2) Paragraph 7(c): if the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage. 

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like 

defects exist. 

(4) Piping with wall thickness less than the diameter of the electrode 

shall be depressurized before welding. 

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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued 

Code case No. Code supplement Code case title Conditions 

N–702 ............................ Responding to comment on previously approved Code Case.Alternative Requirements for Boiling Water 

Reactor (BWR) Nozzle Inner Radius and 

Nozzle-to-Shell Welds, Section XI, Division 

1.

The technical basis supporting the implementation of this Code Case 

is addressed by BWRVIP–108: BWR Vessel and Internals Project, 

‘‘Technical Basis for the Reduction of Inspection Requirements for 

the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,’’ EPRI Technical Report 1003557, October 2002 

(ADAMS Accession No. ML023330203); and BWRVIP–241: BWR 

Vessels and Internals Project, ‘‘Probabilistic Fracture Mechanics 

Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell 

Welds and Nozzle Blend Radii,’’ EPRI Technical Report 1021005, 

October 2010 (ADAMS Accession No. ML11119A041). The applicability of Code Case N–702 must be shown by demonstrating 

that the criteria in Section 5.0 of NRC Safety Evaluation regarding 

BWRVIP–108 dated December 18, 2007 (ADAMS Accession No. 

ML073600374), or Section 5.0 of NRC Safety Evaluation regarding 

BWRVIP–241 dated April 19, 2013 (ADAMS Accession No. 

ML13071A240), are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the 

NRC prior to the application of the Code Case. 

ASME OM Code Cases 

OMN–1 .......................... 2006 Addenda ................................................. Alternative Rules for Preservice and Inservice 

Testing of Active Electric Motor-Operated 

Valve Assemblies in Light-Water Reactor 

Power Plants.

Licensees may use Code Case OMN–1, ‘‘Alternative Rules for 

Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants,’’ in 

lieu of the provisions for stroke-time testing in Subsection ISTC of 

the 1995 Edition up to and including the 2006 Addenda of the 

ASME OM Code when applied in conjunction with the provisions 

for leakage rate testing in, as applicable, ISTC 4.3 (1995 Edition 

with the 1996 and 1997 Addenda) and ISTC–3600 (1998 Edition 

through the 2006 Addenda). In addition, licensees who continue to 

implement Section XI of the ASME BPV Code as their Code of 

Record may use OMN–1 in lieu of the provisions for stroke-time 

testing specified in Paragraph 4.2.1 of ASME/ANSI OM Part 10 as 

required by 10 CFR 50.55a(b)(2)(vii) subject to the conditions in 

this regulatory guide. Licensees who choose to apply OMN–1 

must apply all its provisions. 

(1) The adequacy of the diagnostic test interval for each motor-operated valve (MOV) must be evaluated and adjusted as necessary, 

but not later than 5 years or three refueling outages (whichever is 

longer) from initial implementation of OMN–1. 

(2) When extending exercise test intervals for high risk MOVs beyond a quarterly frequency, licensees must ensure that the potential increase in Core Damage Frequency (CDF) and risk associated with the extension is small and consistent with the intent of 

the Commission’s Safety Goal Policy Statement. 

(3) When applying risk insights as part of the implementation of 

OMN–1, licensees must categorize MOVs according to their safety 

significance using the methodology described in Code Case 

OMN–3, ‘‘Requirements for Safety Significance Categorization of 

Components Using Risk Insights for Inservice Testing of LWR 

Power Plants,’’ with the conditions discussed in this regulatory 

guide or use other MOV risk ranking methodologies accepted by 

the NRC on a plant specific or industry-wide basis with the conditions in the applicable safety evaluations. 

Note 1: As indicated at 64 FR 51370–51386, licensees are cautioned 

that, when implementing OMN 1, the benefits of performing a particular test should be balanced against the potential adverse effects placed on the valves or systems caused by this testing. 

Note 2: RG 1.192, Rev. 0, conditionally accepted Code Case OMN– 

11 for use in conjunction with Code Case OMN–1. The provisions 

of Code Case OMN–11 were acceptably incorporated into Code 

Case OMN–1, 2006 Addenda, including the conditions in the RG 

on the use of Code Case OMN–11. Code Case OMN–11, 2006 

Addenda, is therefore no longer appropriate for use. Accordingly, 

applicants and licensees choosing to perform risk-informed testing 

of motor-operated valves (MOVs) as allowed by RG 1.192 must do 

so in accordance with the applicable provisions of Code Case 

OMN–1 together with the conditions specified for its use in Table 2 

of this regulatory guide. In accordance with 10 CFR 

50.55a(b)(6)(ii), applicants and licensees that have implemented 

versions of Code Cases OMN–1 and OMN–11 earlier than the 

2006 Addenda (i.e., with the conditions as specified in Table 3 of 

this RG) may continue to use those versions through the end of 

the current IST interval. If that applicant or licensee plans to continue to implement a risk-informed IST program for its MOVs in the 

subsequent IST interval, then OMN–1, 2006 Addenda, with the 

conditions specified in Table 2 of this RG will need to be implemented. 

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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued 

Code case No. Code supplement Code case title Conditions 

OMN–3 .......................... 2004 Edition ..................................................... Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power 

Plants.

In addition to those components identified in ASME IST Program 

Plan, implementation of Section 1, ‘‘Applicability,’’ of the Code 

Case must include within the scope of a licensee’s risk-informed 

IST Program non-ASME Code Components categorized as high 

safety significant components (HSSCs) that might not currently be 

included in the IST Program Plan. 

(2) The decision criteria discussed in Section 4.4.1, ‘‘Decision Criteria,’’ of the Code Case for evaluating the acceptability of aggregate risk effects (i.e., for Core Damage Frequency [CDF] and 

Large Early Release Frequency [LERF]) must be consistent with 

the guidance provided in Regulatory Guide 1.174, ‘‘An Approach 

for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.’’ 

(3) Section 4.4.4, ‘‘Defense in Depth,’’ of the Code Case must be 

consistent with the guidance contained in Sections 2.2.1, ‘‘Defense-in-Depth Evaluation’’; and 2.2.2, ‘‘Safety Margin Evaluation,’’ 

of Regulatory Guide 1.175, ‘‘An Approach for Plant-Specific, RiskInformed Decisionmaking: Inservice Testing.’’ 

(4) Implementation of Sections 4.5, ‘‘Inservice Testing Program’’; and 

4.6, ‘‘Performance Monitoring,’’ of the Code Case must be consistent with the guidance pertaining to inservice testing of pumps 

and valves provided in Section 3.2, ‘‘Program Implementation’’; 

and Section 3.3, ‘‘Performance Monitoring,’’ of Regulatory Guide 

1.175. Testing and performance monitoring of individual components must be performed as specified in the risk-informed components Code Cases (e.g., OMN–1, OMN–4, OMN–7, and OMN–12, 

as modified by the conditions discussed in this regulatory guide). 

(5) Implementation of Section 3.2, ‘‘Plant Specific PRA,’’ of the Code 

Case must be consistent with the guidance that the Owner is responsible for demonstrating and justifying the technical adequacy 

of the probabilistic risk assessment (PRA) analyses used as the 

basis to perform component risk ranking and for estimating the aggregate risk impact. Regulatory Guide 1.200, ‘‘An Approach for 

Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,’’ provides guidance for 

determining the technical adequacy of the PRA used in a risk-informed regulatory activity. Regulatory Guide 1.201, ‘‘Guidelines for 

Categorizing Structures, Systems, and Components in Nuclear 

Power Plants According to their Safety Significance,’’ describes 

one acceptable method to categorize the safety significance of an 

active component, including methods to use when a plant-specific 

PRA that meets the appropriate Regulatory Guide 1.200 capability 

for specific hazard group(s) (e.g., seismic and fire) is not available. 

(6) Section 4.2.4, ‘‘Reconciliation,’’ paragraph (b), is not endorsed. 

The expert panel may not classify components that are ranked 

HSSC by the results of a qualitative or quantitative PRA evaluation 

(excluding the sensitivity studies) or the defense-in-depth assessment to low safety significant component (LSSC). 

(7) Implementation of Section 3.3, ‘‘Living PRA,’’ must be consistent 

with the following: (1) To account for potential changes in failure 

rates and other changes that could affect the PRA, changes to the 

plant must be reviewed, and, as appropriate, the PRA updated; (2) 

When the PRA is updated, the categorization of structures, systems, and components must be reviewed and changed if necessary to remain consistent with the categorization process; and 

(3) The review of plant changes must be performed in a timely 

manner and must be performed once every two refueling outages 

or as required by 10 CFR 50.71(h)(2) for combined license holders. 

Note 1: The Code Case methodology for risk ranking uses two categories of safety significance. The NRC staff has determined that 

this is acceptable for ranking all component types. However, the 

NRC staff has accepted other methodologies for risk ranking 

MOVs, with certain conditions that use three categories of safety 

significance. 

OMN–4 .......................... 2004 Edition ..................................................... Requirements for Risk Insights for Inservice 

Testing of Check Valves at LWR Power 

Plants.

(1) Valve opening and closing functions must be demonstrated when 

flow testing or examination methods (nonintrusive, or disassembly 

and inspection) are used. 

(2) The initial interval for tests and associated examinations may not 

exceed two fuel cycles or 3 years, whichever is longer; any extension of this interval may not exceed one fuel cycle per extension 

with the maximum interval not to exceed 10 years. Trending and 

evaluation of existing data must be used to reduce or extend the 

time interval between tests. 

(3) If the Appendix II condition monitoring program is discontinued, 

the requirements of ISTC 4.5.1, ‘‘Exercising Test Frequency,’’ 

through ISTC 4.5.4, ‘‘Valve Obturator Movement,’’ (1996 and 1997 

Addenda) or ISTC 3510, 3520, 3540, and 5221 (1998 Edition with 

the 1999 and 2000 Addenda), as applicable, must be implemented. 

Note 1: The conditions with respect to allowable methodologies for 

OMN–3 risk ranking specified for the use of OMN–1 also apply to 

OMN–4. 

OMN–9 .......................... 2004 Edition ..................................................... Use of a Pump Curve for Testing ................... (1) When a reference curve may have been affected by repair, replacement, or routine servicing of a pump, a new reference curve 

must be determined, or an existing reference curve must be reconfirmed, in accordance with Section 3 of this Code Case. 

(2) If it is necessary or desirable, for some reason other than that 

stated in Section 4 of this Code Case, to establish an additional 

reference curve or set of curves, these new curves must be determined in accordance with Section 3. 

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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued 

Code case No. Code supplement Code case title Conditions 

OMN–12 ........................ 2004 Edition ..................................................... Alternative Requirements for Inservice Testing 

Using Risk Insights for Pneumatically and 

Hydraulically Operated Valve Assemblies in 

Light-Water Reactor Power Plants (OMCode 1998, Subsection ISTC).

(1) Paragraph 4.2, ‘‘Inservice Test Requirements,’’ of OMN–12 specifies inservice test requirements for pneumatically and hydraulically 

operated valve assemblies categorized as high safety significant 

within the scope of the Code Case. The inservice testing program 

must include a mix of static and dynamic valve assembly performance testing. The mix of valve assembly performance testing may 

be altered when justified by an engineering evaluation of test data. 

(2) Paragraph 4.2.2.3 of OMN 12 specifies the periodic test requirements for pneumatically and hydraulically operated valve assemblies categorized as high safety significant within the scope of the 

code case. The adequacy of the diagnostic test interval for each 

high safety significant valve assembly must be evaluated and adjusted as necessary, but not later than 5 years or three refueling 

outages (whichever is longer) from initial implementation of OMN– 

12. 

(3) Paragraph 4.2.3, ‘‘Periodic Valve Assembly Exercising,’’ of OMN 

12 specifies periodic exercising for pneumatically and hydraulically 

operated valve assemblies categorized as high safety significant 

within the scope of the code case. Consistent with the requirement 

in OMN 3 to evaluate the aggregate change in risk associated with 

changes in test strategies, when extending exercise test intervals 

for high safety significant valve assemblies beyond a quarterly frequency, the potential increase in Core Damage Frequency (CDF) 

and risk associated with the extension must be evaluated and determined to be small and consistent with the intent of the Commission’s Safety Goal Policy Statement. 

(4) Paragraph 4.4.1, ‘‘Acceptance Criteria,’’ of OMN 12 specifies that 

acceptance criteria must be established for the analysis of test 

data for pneumatically and hydraulically operated valve assemblies 

categorized as high safety significant within the scope of the code 

case. When establishing these acceptance criteria, the potential 

degradation rate and available capability margin for each valve assembly must be evaluated and determined to provide assurance 

that the valve assemblies are capable of performing their design 

basis functions until the next scheduled test. 

(5) Paragraph 5, ‘‘Low Safety Significant Valve Assemblies,’’ of OMN 

12 specifies that the purpose of its provisions is to provide a high 

degree of confidence that pneumatically and hydraulically operated 

valve assemblies categorized as low safety significant within the 

scope of the code case will perform their intended safety function 

if called upon. The licensee must have reasonable confidence that 

low safety significant valve assemblies remain capable of performing their intended design-basis safety functions until the next 

scheduled test. The test and evaluation methods may be less rigorous than those applied to high safety significant valve assemblies. 

(6) Paragraph 5.1, ‘‘Set Points and/or Critical Parameters,’’ of OMN 

12 specifies requirements and guidance for establishing set points 

and critical parameters of pneumatically and hydraulically operated 

valve assemblies categorized as low safety significant within the 

scope of the code case. Setpoints for these valve assemblies must 

be based on direct dynamic test information, a test based methodology, or grouping with dynamically tested valves, and documented according to Paragraph 5.1.4. The setpoint justification 

methods may be less rigorous than provided for high risk significant valve assemblies. 

(7) Paragraph 5.4, ‘‘Evaluations,’’ of OMN–12, specifies evaluations 

to be performed of pneumatically and hydraulically operated valve 

assemblies categorized as low safety significant within the scope 

of the Code Case. Initial and periodic diagnostic testing must be 

performed to establish and verify the setpoints of these valve assemblies to ensure that they are capable of performing their design-basis safety functions. Methods for testing and establishing 

test frequencies may be less rigorous than applied to high risk significant valve assemblies. 

(8) Paragraph 5.6, ‘‘Corrective Action,’’ of OMN–12 specifies that 

corrective action must be initiated if the parameters monitored and 

evaluated for pneumatically and hydraulically operated valve assemblies categorized as low safety significant within the scope of 

the code case do not meet the established criteria. Further, if the 

valve assembly does not satisfy its acceptance criteria, the operability of the valve assembly must be evaluated. 

Note 1: Licensees are cautioned that, when implementing OMN–12, 

the benefits of performing a particular test should be balanced 

against the potential adverse effects placed on the valves or systems caused by this testing. 

Note 2: Paragraph 3.1 of OMN–12 states that ‘‘Valve assemblies 

shall be classified as either high safety significant or low safety 

significant in accordance with Code Case OMN–3.’’ This note as 

well as Note 2 to OMN–4 have been added to ensure the consistent consideration of risk insights. 

C. ASME Code Cases Not Approved for 

Use 

The ASME Code Cases which are 

currently issued by ASME but not 

approved for generic use by the NRC are 

listed in RG 1.193, ‘‘ASME Code Cases 

Not Approved for Use.’’ The Code Cases 

which are not approved for use include 

Code Cases on high-temperature gas 

cooled reactors; certain requirements in 

Section III, Division 2, not endorsed by 

the NRC, liquid metal; and submerged 

spent fuel waste casks. Regulatory 

Guide 1.193 is not incorporated by 

reference into § 50.55a. Regulatory 

Guide 1.193 is prepared by the NRC as 

a resource for stakeholders, allowing 

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65792 Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 

them to easily identify Code Cases 

which the NRC has not approved for use 

as a generic matter. Listing of a Code 

Case in RG 1.193 does not preclude an 

application or licensee for seeking 

individual, case-by-case NRC approval 

to use a listed Code Case. 

V. Petition for Rulemaking (PRM–50– 

89) 

On December 14, 2007, Mr. Raymond 

West (the petitioner) submitted a PRM 

requesting the NRC to amend § 50.55a to 

allow consideration of alternatives to 

the NRC-approved ASME BPV and OM 

Code Cases. The petitioner submitted an 

amended petition on December 19, 2007 

(ADAMS Accession No. ML073600974). 

The petition was docketed by the NRC 

as PRM–50–89. The petitioner requested 

that the regulations be amended to 

provide applicants and licensees a 

process for requesting NRC approval of 

changes or modifications to ASME Code 

Cases that are listed in the relevant 

NRC-approved RGs cited in the current 

regulations. The petitioner stated that 

the current requirements do not allow 

changes or modifications to be proposed 

as alternatives to NRC-approved ASME 

Code Cases, and asserted that such 

changes or modifications should be 

allowed as alternatives to NRC Code 

Cases. Overall, the petitioner requested 

that the regulations be amended to 

allow applicants and licensees to 

request authorization of NRC-approved 

Code Cases with proposed 

modifications directly through 

§ 50.55a(a)(3). 

The NRC determined that the issues 

raised in this PRM should be considered 

in the NRC’s rulemaking process, and 

the NRC published a FRN with this 

determination on April 22, 2009 (74 FR 

18303). 

The NRC believes that Code Cases 

often provide alternatives that have 

technical merit and, in many instances, 

are incorporated into future ASME Code 

editions. The ASME Code Case process 

itself constitutes a method of how an 

applicant or licensee can seek to obtain 

ASME approval for a variation of a 

previously-approved Code provision. 

Section 50.55a(a)(3) currently provides 

specific approaches for obtaining NRC 

authorization of alternatives to ASME 

Code provisions. Inasmuch as ASME 

Code Cases are analogous to ASME 

Code provisions, it is not unreasonable 

to provide an analogous regulatory 

approach for obtaining NRC 

authorization of alternatives to ASME 

Code Cases. Therefore, the NRC has 

included language in § 50.55a(z) 

(previously § 50.55a(a)(3)) that would 

allow applicants and licensees to 

request authorization of alternatives for 

changes to conditions on NRC-approved 

ASME Code Cases in current paragraphs 

(b)(4), (b)(5), and (b)(6) of § 50.55a. In 

addition, the NRC is extending the 

scope of the petitioner’s request for 

allowing alternatives to NRC-approved 

Code Case conditions to allow 

applicants and licensees to request 

authorization of alternatives for changes 

to conditions on Section III and XI of 

the ASME BPV Code and OM Code in 

current paragraphs (b)(1), (b)(2), and 

(b)(3). 

In the final rule, the requirements in 

former paragraph (a)(3) have been 

moved to newly created paragraph (z), 

making room in this section for the 

listing of all standards to be 

incorporated by reference in paragraph 

(a). The reasons for this change is 

discussed in the SUPPLEMENTARY

INFORMATION in Section VI. Changes 

addressing the Office of the Federal 

Register’s Guidelines on Incorporation 

by Reference. 

This final rule resolves and represents 

the NRC’s final action on PRM–50–89. 

VI. Changes Addressing the Office of 

the Federal Register’s Guidelines on 

Incorporation by Reference 

This final rule includes changes to 

§§ 50.54, 50.55, and 50.55a. These 

changes were made in accordance with 

the guidance for incorporation by 

reference of multiple standards that are 

included in Chapter 6 of the OFR’s 

‘‘Federal Register Document Drafting 

Handbook,’’ January 2011 Revision. 

This latest revision of the OFR’s 

guidance provides several options for 

incorporating by reference multiple 

standards into regulations. 

The NRC has incorporated by 

reference, in a single paragraph, the 

multiple standards mentioned in 

§ 50.55a. For the least disruption to the 

existing structure of the section, the 

NRC incorporated by reference the 

multiple standards into § 50.55a(a), the 

first paragraph of the section. Each 

national consensus standard that is 

being incorporated by reference in 

§ 50.55a has been listed separately. 

Accordingly, the regulatory language of 

§§ 50.54, 50.55, and 50.55a has been 

reorganized by moving existing 

paragraphs, creating new paragraphs, 

and revising introductory and regulatory 

texts. 

The NRC has made conforming 

changes to references throughout 

§ 50.55a to reflect this reorganization. A 

detailed discussion of the affected 

paragraphs, other than the 

aforementioned reference changes, is 

provided in Section VIII, ‘‘Paragraph-byParagraph Discussion,’’ of this 

document. The regulatory text of 

§ 50.55a has been set out in its entirety 

for the convenience of the reader. The 

NRC staff has also developed reader aids 

to help users understand these changes 

(see Section VII of this document). 

VII. Addition of Headings to 

Paragraphs 

The NRC has added headings 

(explanatory titles) to paragraphs and all 

lower-level subparagraphs of § 50.55a. 

These headings are intended to enhance 

the readers’ ability to identify the 

paragraphs (e.g., paragraphs (a), (b), (c)) 

and subparagraphs with the same 

subject matter. The NRC evaluated a 

range of solutions, including the 

creation of new regulations and 

relocation of existing requirements from 

§ 50.55a to the new regulations. 

Some alternatives the NRC considered 

were a new regulation adjacent to 

§ 50.55a (e.g., §§ 50.55b, 50.55c, 50.55d), 

a new subpart containing a new series 

of regulations at the end of 10 CFR part 

50 (e.g., subpart B beginning at § 50.200, 

and continuing with §§ 50.201, 50.202, 

50.203), or a new part (designated for 

Codes and standards) containing a new 

series of regulations addressing Codes 

and standards approved for 

incorporation by reference by the OFR. 

The relocation of each existing 

requirement to a new regulation (or set 

of regulations) would follow a set of 

organizing principles established by the 

NRC after consideration of public views. 

Upon consideration of these 

alternatives, the NRC decided that these 

alternatives should not be adopted—at 

least not at this time without further 

public input—and instead that the NRC 

should develop and adopt headings for 

paragraphs and subparagraphs. The 

primary reason for the NRC’s decision is 

external stakeholders’ objections to a 

previous attempt by the NRC to redesignate paragraphs in § 50.55a (75 FR 

24324; May 4, 2010). As the NRC 

understands it, many nuclear power 

plant licensees’ procedures reference 

specific paragraphs and subparagraphs 

of § 50.55a. It would require substantial 

rewriting of these procedures and 

documents to correct the references to 

the old (superseded) section, paragraphs 

and subparagraphs. In addition, 

currently-approved design certification 

rules may require conforming 

amendments to be made to correct 

references to ASME Code provisions on 

design (and possibly ISI and IST). As 

mentioned earlier in the response to 

Comment No. 1, the NRC received 

several public comments but deferred 

their consideration to a potential future 

rulemaking effort for reorganizing the 

entire § 50.55a with public input. The 

current reorganization of this 

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rulemaking is based upon two major 

issues- consideration of the OFR’s 

revised guidelines for incorporating by 

reference consensus standards in 

regulations and addition of headings 

(explanatory titles) to paragraphs and 

lower-level subparagraphs of § 50.55a as 

reader aids. 

A. NRC’s Convention for Headings and 

Subheadings 

The NRC has added headings to all 

first, second, third, fourth, and some 

fifth-level paragraphs for certain 

sections of § 50.55a to add clarity and a 

user-friendly method for following 

sublevel contents within a regulation. 

The heading for a fourth-level follows 

the same convention, but may designate 

the provision number only. Fifth-level 

paragraphs are only for newly 

incorporated Code Cases. Each firstlevel paragraph (designated using letters 

[e.g., (a), (b), (c)]) have a heading that 

concisely describes the general subject 

matter addressed in that paragraph. 

Each second-level paragraph 

(designated using numbers [e.g., (1), (2), 

(3)] have a heading comprised of a 

summary of the first-level paragraph’s 

heading and a semicolon (‘‘;’’), followed 

by a concise description of the subject 

matter addressed in the second 

paragraph. The heading for a third-level 

paragraph follows the same convention 

(i.e., a heading comprised of a summary 

level of the higher-level paragraph’s title 

and a semicolon, followed by a concise 

description of the subject matter 

addressed in that subparagraph). The 

heading for a fourth-level paragraph 

follows the same convention, but 

designate the provision number only. 

The fifth-level paragraph is applied to 

only paragraph (a) for incorporation by 

reference of approved editions and 

addenda to the ASME BPV and OM 

Codes. 

B. Reader Aids 

The NRC staff has developed a table 

showing the structure of § 50.55a. This 

table, ‘‘Final Reorganization of 

Paragraphs and Subparagraphs in 10 

CFR 50.55a, ‘Codes and standards’’’ 

(ADAMS Accession No. ML14015A191), 

is available in a separate document and 

outlines the section showing all 

paragraph designations, including the 

new paragraph headings. The NRC staff 

has also developed cross-reference 

tables showing the current designations 

for §§ 50.54, 50.55, and 50.55a 

regulations and the new designations for 

these sections. These tables contain the 

new headings and a description of each 

change and are available in separate 

documents (ADAMS Accession No. 

ML14211A050- package contains two 

tables). 

VIII. Paragraph-by-Paragraph 

Discussion 

Overall Considerations on the Use of 

ASME Code Cases 

This rulemaking has amended 

§ 50.55a to incorporate by reference RG 

1.84, Revision 36, which supersedes 

Revision 35; RG 1.147, Revision 17, 

which supersedes Revision 16; and RG 

1.192, Revision 1, which supersedes 

Revision 0. The following general 

guidance applies to the use of the ASME 

Code Cases approved in the latest 

versions of the RGs that are 

incorporated by reference into § 50.55a 

as part of this rulemaking. 

The approval of a Code Case in the 

NRC RGs constitutes acceptance of its 

technical position for applications that 

are not precluded by regulatory or other 

requirements or by the 

recommendations in these or other RGs. 

The applicant and/or licensee are 

responsible for ensuring that use of the 

Code Case does not conflict with 

regulatory requirements or licensee 

commitments. The Code Cases listed in 

the RGs are acceptable for use within 

the limits specified in the Code Cases. 

If the RG states an NRC condition on the 

use of a Code Case, then the NRC 

condition supplements and does not 

supersede any condition(s) specified in 

the Code Case, unless otherwise stated 

in the NRC condition. 

The ASME Code Cases may be revised 

for many reasons (e.g., to incorporate 

operational examination and testing 

experience and to update material 

requirements based on research results). 

On occasion, an inaccuracy in an 

equation is discovered or an 

examination, as practiced, is found not 

to be adequate to detect a newly 

discovered degradation mechanism. 

Hence, when an applicant or a licensee 

initially implements a Code Case, 

§ 50.55a requires that the applicant or 

the licensee implement the most recent 

version of that Code Case as listed in the 

RGs incorporated by reference. Code 

Cases superseded by revision are no 

longer acceptable for new applications 

unless otherwise indicated. 

Section III of the ASME BPV Code 

applies only to new construction (i.e., 

the edition and addenda to be used in 

the construction of a plant are selected 

based on the date of the construction 

permit and are not changed thereafter, 

except voluntarily by the applicant or 

the licensee). Hence, if a Section III 

Code Case is implemented by an 

applicant or a licensee and a later 

version of the Code Case is incorporated 

by reference into § 50.55a and listed in 

the RGs, the applicant or the licensee 

may use either version of the Code Case 

(subject, however, to whatever change 

requirements apply to its licensing basis 

(e.g., § 50.59)). 

A licensee’s ISI and IST programs 

must be updated every 10 years to the 

latest edition and addenda of Section XI 

and the OM Code, respectively, that 

were incorporated by reference into 

§ 50.55a and in effect 12 months prior 

to the start of the next inspection and 

testing interval. Licensees who were 

using a Code Case prior to the effective 

date of its revision may continue to use 

the previous version for the remainder 

of the 120-month ISI or IST interval. 

This relieves licensees of the burden of 

having to update their ISI or IST 

program each time a Code Case is 

revised by the ASME and approved for 

use by the NRC. Code Cases apply to 

specific editions and addenda, and Code 

Cases may be revised if they are no 

longer accurate or adequate, so licensees 

choosing to continue using a Code Case 

during the subsequent ISI or IST 

interval must implement the latest 

version incorporated by reference into 

§ 50.55a and listed in the RGs. 

The ASME may annul Code Cases that 

are no longer required, are determined 

to be inaccurate or inadequate, or have 

been incorporated into the ASME BPV 

or OM Codes. If an applicant or a 

licensee applied a Code Case before it 

was listed as annulled, the applicant or 

the licensee may continue to use the 

Code Case until the applicant or the 

licensee updates its Construction Code 

of Record (in the case of an applicant, 

updates its application) or until the 

licensee’s 120 month ISI or IST update 

interval expires, after which the 

continued use of the Code Case is 

prohibited unless NRC authorization is 

given under the current § 50.55a(a)(3). If 

a Code Case is incorporated by reference 

into § 50.55a and later annulled by the 

ASME because experience has shown 

that the design analysis, construction 

method, examination method, or testing 

method is inadequate; the NRC will 

amend § 50.55a and the relevant RG to 

remove the approval of the annulled 

Code Case. Applicants and licensees 

should not begin to implement such 

annulled Code Cases in advance of the 

rulemaking. 

A Code Case may be revised, for 

example, to incorporate user experience. 

The older or superseded version of the 

Code Case cannot be applied by the 

licensee or applicant for the first time. 

If an applicant or a licensee applied 

a Code Case before it was listed as 

superseded, the applicant or the 

licensee may continue to use the Code 

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Case until the applicant or the licensee 

updates its Construction Code of Record 

(in the case of an applicant, updates its 

application) or until the licensee’s 120- 

month ISI or IST update interval 

expires, after which the continued use 

of the Code Case is prohibited unless 

NRC authorization is given under new 

§ 50.55a(z). If a Code Case is 

incorporated by reference into § 50.55a 

and later a revised version is issued by 

the ASME because experience has 

shown that the design analysis, 

construction method, examination 

method, or testing method is 

inadequate; the NRC will amend 

§ 50.55a and the relevant RG to remove 

the approval of the superseded Code 

Case. Applicants and licensees should 

not begin to implement such superseded 

Code Cases in advance of the 

rulemaking. 

Incorporation by Reference 

The final rule includes changes to 

§§ 50.54, 50.55, and 50.55a. This change 

brings the NRC’s requirements into 

compliance with the OFR’s revised 

guidelines for incorporating by 

reference consensus standards in 

regulations. 

Section 50.54 

In § 50.54, the introductory statement 

has been revised to include a reference 

to § 50.55a. This revision clarifies that 

nuclear power plant licensees, as 

described in the introductory paragraph 

of § 50.54, also are subject to the 

applicable requirements delineated in 

§ 50.55a. In addition, the NRC revised 

the introductory text of this section and 

added and reserved paragraph (ii), and 

added paragraph (jj) to include a 

condition of every license. This 

requirement is currently contained in 

§ 50.55a(a)(1), and no change to the 

requirement is intended by the transfer 

of this requirement from § 50.55a(a)(1) 

to § 50.54(jj), except for clarification of 

its applicability. 

Section 50.55 

In § 50.55, the introductory text has 

been revised to include references to 

existing § 50.55a, and paragraphs (g) and 

(h) have been added and reserved for 

future use. Further, existing 

§ 50.55a(a)(1) has been moved to a 

newly created § 50.55(i) enabling the 

removal of the current regulation from 

the current 50.55a(a)(1). No change to 

the requirement is intended by this 

transfer, except for clarification of its 

applicability. The introductory text of 

§ 50.55 has been revised to maintain the 

existing applicability of the requirement 

in the newly created § 50.55(i) to 

construction permits for utilization 

facilities. 

Section 50.55a 

The introductory text to § 50.55a was 

relocated to several other locations. 

There is no introductory text to § 50.55a 

in the new rule. The first sentence in the 

previous introductory text was relocated 

to the first sentence in § 50.55. The 

remaining sentences were relocated to 

§ 50.55a(b) (second sentence), 

§ 50.55a(b)(1) (first sentence), 

§ 50.55a(b)(4) (first sentence), § 50.55a(c) 

(second sentence), § 50.55a(d) (second 

sentence), § 50.55a(e) (second sentence), 

§ 50.55a(f) (second and third sentences), 

§ 50.55a(g) (second and third sentences), 

and § 50.55a(h) (second sentence). 

In addition to moving existing 

paragraphs, creating new paragraphs, 

and revising introductory and regulatory 

texts, the footnotes in § 50.55a have 

been reorganized to appear in sequential 

order. The NRC also has reserved 

footnote numbers so that the NRC may 

add a footnote in a future rulemaking 

without having to renumber the existing 

footnotes. 

Paragraph (a): A new paragraph (a) 

has been created in § 50.55a to 

incorporate by reference the multiple 

standards currently identified in 

existing § 50.55a. The heading has been 

revised to read ‘‘Documents approved 

for incorporation by reference.’’ 

Paragraph (a)(1): This paragraph, 

‘‘American Society of Mechanical 

Engineers (ASME),’’ has been added to 

group all ASME sections. 

Paragraph (a)(1)(i): This paragraph, 

‘‘ASME Boiler and Pressure Vessel 

Code, Section III,’’ has been added to 

discuss the availability of standards 

referenced in current paragraph (b)(1). 

Paragraph (a)(1)(i)(A): This paragraph, 

‘‘Rules for Construction of Nuclear 

Vessels,’’ has been added to group all 

the individual standards referenced 

regarding the subject matter included in 

current paragraph (b)(1). 

Paragraph (a)(1)(i)(B): This paragraph, 

‘‘Rules for Construction of Nuclear 

Power Plant Components,’’ has been 

added to group all the individual 

standards referenced regarding the 

subject matter included in current 

paragraph (b)(1). 

Paragraph (a)(1)(i)(C): This paragraph, 

‘‘Division 1 Rules for Construction of 

Nuclear Power Plant Components,’’ has 

been added to group all the individual 

standards referenced regarding the 

subject matter included in current 

paragraph (b)(1). 

Paragraph (a)(1)(i)(D): This paragraph, 

‘‘Rules for Construction of Nuclear 

Power Plant Components—Division 1,’’ 

has been added to group all the 

individual standards referenced 

regarding the subject matter included in 

current paragraph (b)(1). 

Paragraph (a)(1)(i)(E): This paragraph, 

‘‘Rules for Construction of Nuclear 

Facility Components—Division 1,’’ has 

been added to group all the individual 

standards referenced regarding the 

subject matter included in current 

paragraph (b)(1). 

Paragraph (a)(1)(ii): This paragraph, 

‘‘ASME Boiler and Pressure Vessel 

Code, Section XI,’’ has been added to 

discuss the availability of standards 

referenced in current paragraph (b)(2). 

Paragraph (a)(1)(ii)(A): This 

paragraph, ‘‘Rules for Inservice 

Inspection of Nuclear Reactor Coolant 

Systems,’’ has been added to discuss the 

availability of individual standards 

referenced regarding the subject matter 

included in current paragraph (b)(2). 

Paragraph (a)(1)(ii)(B): This 

paragraph, ‘‘Rules for Inservice 

Inspection of Nuclear Power Plant 

Components,’’ has been added to 

discuss the availability of individual 

standards referenced regarding the 

subject matter included in current 

paragraph (b)(2). 

Paragraph (a)(1)(ii)(C): This 

paragraph, ‘‘Rules for Inservice 

Inspection of Nuclear Power Plant 

Components—Division 1,’’ has been 

added to discuss the availability of 

individual standards referenced 

regarding the subject matter included in 

current paragraph (b)(2). 

Paragraph (a)(1)(iii): This paragraph, 

‘‘ASME Code Cases: Nuclear 

Components,’’ has been added to 

discuss the newly approved Code Cases 

referenced regarding the subject matter 

in current paragraph (b). 

Paragraph (a)(1)(iii)(A): This 

paragraph, ‘‘ASME Code Case N–722– 

1,’’ has been added to discuss the newly 

approved Code Case referenced 

regarding the subject matter in current 

paragraph (b). 

Paragraph (a)(1)(iii)(B): This 

paragraph, ‘‘ASME Code Case N–729– 

1,’’ has been added to discuss the newly 

approved Code Case referenced 

regarding the subject matter in current 

paragraph (b). 

Paragraph (a)(1)(iii)(C): This 

paragraph, ‘‘ASME Code Case N–770– 

1,’’ has been added to discuss the newly 

approved Code Case referenced 

regarding the subject matter in current 

paragraph (b). 

Paragraph (a)(1)(iv): This paragraph, 

‘‘ASME Operation and Maintenance 

Code,’’ has been added to group all the 

individual standards referenced in 

current paragraph (b). 

Paragraph (a)(1)(iv)(A): This 

paragraph, ‘‘Code for Operation and 

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Maintenance of Nuclear Power Plants,’’ 

has been added to group all the 

individual standards referenced in 

current paragraph (b). 

Paragraph (a)(1)(iv)(B): This paragraph 

has been added and reserved for future 

use. 

Paragraph (a)(2): This paragraph, 

‘‘Institute of Electrical and Electronics 

Engineers (IEEE) Service Center,’’ has 

been added to list all IEEE sections. 

Paragraph (a)(2)(i): This paragraph, 

‘‘IEEE Standard 279—1971,’’ has been 

added to discuss the availability of 

standards referenced in current 

paragraph (h)(2). 

Paragraph (a)(2)(ii): This paragraph, 

‘‘IEEE Standard 603—1991,’’ has been 

added to discuss the availability of the 

standard referenced in current 

paragraphs (h)(2) and (h)(3). 

Paragraph (a)(2)(iii): This paragraph, 

‘‘IEEE Standard 603—1991 correction 

sheet,’’ has been added to discuss the 

availability of the standard referenced in 

current paragraphs (h)(2) and (h)(3). 

Paragraph (a)(3): This paragraph, 

‘‘U.S. Nuclear Regulatory Commission 

(NRC) Reproduction and Distribution 

Services Section,’’ lists all RGs being 

incorporated by reference. 

Paragraph (a)(3)(i): This paragraph, 

‘‘NRC Regulatory Guide 1.84, Revision 

36,’’ has been added to discuss the 

availability of the standard. 

Paragraph (a)(3)(ii): This paragraph, 

‘‘NRC Regulatory Guide 1.147, Revision 

17,’’ has been added to discuss the 

availability of the standard. 

Paragraph (a)(3)(iii): This paragraph, 

‘‘NRC Regulatory Guide 1.192, Revision 

1,’’ has been added to discuss the 

availability of the standard. 

Paragraph (b): The paragraph heading 

has been revised to ‘‘Use and conditions 

on the use of standards.’’ The contents 

have been moved, in part, to § 50.55a(a) 

for compliance with the OFR’s revised 

guidelines for incorporating by 

reference consensus standards in 

regulations. 

Paragraphs (b)(4): Reference to the 

revision number for RG 1.84 has been 

changed from ‘‘Revision 35’’ to 

‘‘Revision 36.’’ 

Paragraphs (b)(5): Reference to the 

revision number for RG 1.147 has been 

changed from ‘‘Revision 16’’ to 

‘‘Revision 17.’’ 

Paragraphs (b)(6): Reference to the 

revision number for RG 1.192 has been 

changed from ‘‘Revision 0’’ to ‘‘Revision 

1.’’ 

Paragraph (c): Introductory text has 

been added to the existing paragraph (c). 

Explanatory headings have been added 

for subparagraphs. 

Paragraph (d): The new paragraph 

adds introductory text to ‘‘Quality 

Group B components,’’ as part of the 

NRC initiative of adding headings and 

providing clarity. Explanatory headings 

have been added for subparagraphs. 

Paragraph (e): The new paragraph 

adds introductory text to ‘‘Quality 

Group C components,’’ as part of the 

NRC initiative of adding headings and 

providing clarity. Explanatory headings 

have been added for subparagraphs. 

Paragraph (f): Introductory text has 

been revised and expanded in 

‘‘Inservice testing requirements,’’ as part 

of the NRC initiative of adding headings 

and providing clarity. Explanatory 

headings have been added for 

subparagraphs. 

Paragraph (g): Introductory text has 

been revised and expanded in 

‘‘Inservice inspection requirements,’’ as 

part of the NRC initiative of adding 

headings and providing clarity. 

Explanatory headings have been added 

for subparagraphs. 

Paragraphs (b)(5), (f)(2), (f)(3)(iii)(A), 

(f)(3)(iv)(A), (f)(4)(ii), (g)(2), (g)(3)(i), 

(g)(3)(ii), (g)(4)(i), and (g)(4)(ii): 

Reference to the revision number for RG 

1.147 has been changed from ‘‘Revision 

16’’ to ‘‘Revision 17.’’ 

Paragraph (h)(1): This paragraph has 

been designated as reserved because the 

informational content from current 

(h)(1) has been moved to paragraph 

(a)(2). 

Paragraphs (i)–(y): These paragraphs 

have been added and reserved for future 

use. 

Paragraph (z): This paragraph has 

been added to contain information that 

has been relocated from the 

introductory text of current paragraph 

(a)(3) and current subparagraphs 

(a)(3)(i)–(ii) as a result of the NRC’s 

compliance with the OFR’s revised 

guidelines for incorporating by 

reference consensus standards in 

regulations. Paragraph (z) has also been 

revised to allow applicants and 

licensees to request alternatives to the 

requirements in paragraph (b) of this 

section. 

IX. Regulatory Flexibility Certification 

Under the Regulatory Flexibility Act 

of 1980 (5 U.S.C. 605(b)), the 

Commission certifies that this final rule 

would not impose a significant 

economic impact on a substantial 

number of small entities. This final rule 

would affect only the licensing and 

operation of nuclear power plants. The 

companies that own these plants are not 

‘‘small entities’’ as defined in the 

Regulatory Flexibility Act or the size 

standards established by the NRC (10 

CFR 2.810). 

X. Regulatory Analysis 

The ASME Code Cases listed in the 

RGs to be incorporated by reference 

provide voluntary alternatives to the 

provisions in the ASME BPV and OM 

Codes for design, construction, ISI, and 

IST of specific structures, systems, and 

components used in nuclear power 

plants. Implementation of these Code 

Cases is not required. Licensees and 

applicants use NRC-approved ASME 

Code Cases to reduce unnecessary 

regulatory burden or gain additional 

operational flexibility. It would be 

difficult for the NRC to provide these 

advantages independently of the ASME 

Code Case publication process without 

expending considerable additional 

resources. The NRC has prepared a 

regulatory analysis addressing the 

qualitative benefits of the alternatives 

considered in this rulemaking and 

comparing the costs associated with 

each alternative (ADAMS Accession No. 

ML14010A426). Copies of the regulatory 

analysis are available to the public as 

indicated in Section XVIII, ‘‘Availability 

of Documents,’’ of this document. 

XI. Backfitting and Issue Finality 

The provisions in this final rule 

would allow licensees and applicants to 

voluntarily apply NRC-approved Code 

Cases, sometimes with NRC-specified 

conditions. The approved Code Cases 

are listed in three RGs that are 

incorporated by references into § 50.55a. 

An applicant’s and/or a licensee’s 

voluntary application of an approved 

Code Case does not constitute 

backfitting, inasmuch as there is no 

imposition of a new requirement or new 

position. Similarly, voluntary 

application of an approved Code Case 

by a 10 CFR part 52 applicant or 

licensee does not represent NRC 

imposition of a requirement or action, 

which is inconsistent with any issue 

finality provision in 10 CFR part 52. For 

these reasons, the NRC finds that this 

final rule does not involve any 

provisions requiring the preparation of 

a backfit analysis or documentation 

demonstrating that one or more of the 

issue finality criteria in 10 CFR part 52 

are met. 

XII. Plain Writing 

The Plain Writing Act of 2010 

(Pub. L. 111–274) requires Federal 

agencies to write documents in a clear, 

concise, and well-organized manner. 

The NRC has written this document to 

be consistent with the Plain Writing Act 

as well as the Presidential 

Memorandum, ‘‘Plain Language in 

Government Writing,’’ published June 

10, 1998 (63 FR 31883). 

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XIII. Finding of No Significant 

Environmental Impact: Environmental 

Assessment 

This action stems from the 

Commission’s practice of incorporating 

by reference the RGs listing the most 

recent set of NRC-approved ASME Code 

Cases. The purpose of this action is to 

allow licensees to use the Code Cases 

listed in the RGs as alternatives to 

requirements in the ASME BPV and OM 

Codes for the construction, ISI, and IST 

of nuclear power plant components. 

This action is intended to advance the 

NRC’s strategic goal of ensuring 

adequate protection of public health and 

safety and the environment. It also 

demonstrates the agency’s commitment 

to participate in the national consensus 

standards process under the National 

Technology Transfer and Advancement 

Act of 1995 (NTTAA), Public Law 104– 

113. 

The National Environmental Policy 

Act of 1969, as amended (NEPA), 

requires Federal government agencies to 

study the impacts of their ‘‘major 

Federal actions significantly affecting 

the quality of the human environment’’ 

and prepare detailed statements on the 

environmental impacts of the action and 

alternatives to the action (42 U.S.C. 

4332(C); Sec. 102(C) of NEPA). 

The Commission has determined 

under NEPA, as amended, and the 

Commission’s regulations in subpart A 

of 10 CFR part 51, that this rule would 

not be a major Federal action 

significantly affecting the quality of the 

human environment. Therefore, an 

environmental impact statement is not 

required. 

As alternatives to the ASME Code, 

NRC-approved Code Cases provide an 

equivalent level of safety. Therefore, the 

probability or consequences of accidents 

is not changed. There are also no 

significant, non-radiological impacts 

associated with this action because no 

changes would be made affecting nonradiological plant effluents and because 

no changes would be made in activities 

that would adversely affect the 

environment. The determination of this 

environmental assessment is that there 

will be no significant offsite impact to 

the public from this action. 

XIV. Paperwork Reduction Act 

Statement 

This final rule contains new or 

amended information collection 

requirements 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 (OMB), 

approval number 3150–0011. 

The burden to the public for these 

information collections is estimated to 

average a reduction of 80 hours per 

response, including 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 

further reducing the burden, to the 

FOIA, Privacy, and Information 

Collections Branch (T–5 F52), U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, or by 

email to INFOCOLLECTS.RESOURCE@

NRC.GOV; and to the Desk Officer, 

Office of Information and Regulatory 

Affairs, NEOB–10202 (3150–0011), 

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. 

XV. Congressional Review Act 

In accordance with the Congressional 

Review Act of 1996 (5 U.S.C. 801–808), 

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. 

XVI. Voluntary Consensus Standards 

Section 12(d)(3) of the NTTAA, Public 

Law 104–113, and implementing 

guidance in OMB Circular A–119 

(February 10, 1998), require each 

Federal government agency (should it 

decide that regulation is necessary) to 

use a voluntary consensus standard 

instead of developing a governmentunique standard. An exception to using 

a voluntary consensus standard is 

allowed where the use of such a 

standard is inconsistent with applicable 

law or is otherwise impractical. The 

NTTAA requires Federal agencies to use 

industry consensus standards to the 

extent practical; it does not require 

Federal agencies to endorse a standard 

in its entirety. Neither the NTTAA nor 

OMB Circular A–119 prohibit an agency 

from adopting a voluntary consensus 

standard while taking exception to 

specific portions of the standard, if 

those provisions are deemed to be 

‘‘inconsistent with applicable law or 

otherwise impractical.’’ Furthermore, 

taking specific exceptions furthers the 

Congressional intent of Federal reliance 

on voluntary consensus standards 

because it allows the adoption of 

substantial portions of consensus 

standards without the need to reject the 

standards in their entirety because of 

limited provisions that are not 

acceptable to the agency. 

In this rulemaking, the NRC is 

continuing its existing practice of 

approving the use of ASME BPV and 

OM Code Cases, which are ASMEapproved alternatives to compliance 

with various provisions of the ASME 

BPV and OM Codes. The NRC’s 

approval of the ASME Code Cases is 

accomplished by amending the NRC’s 

regulations to incorporate by reference 

the latest revisions of the following, 

which are the subject of this 

rulemaking, into § 50.55a: RG 1.84, 

‘‘Design, Fabrication, and Materials 

Code Case Acceptability, ASME Section 

III,’’ Revision 36; RG 1.147, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME Section XI, Division 1,’’ Revision 

17; and RG 1.192, ‘‘Operation and 

Maintenance Code Case Acceptability, 

ASME Code,’’ Revision 1. These RGs list 

the ASME Code Cases that the NRC has 

approved for use. The ASME Code 

Cases are national consensus standards 

as defined in the NTTAA and OMB 

Circular A–119. The ASME Code Cases 

constitute voluntary consensus 

standards, in which all interested 

parties (including the NRC and 

licensees of nuclear power plants) 

participate. Therefore, the NRC’s 

approval of the use of the ASME Code 

Cases identified in RGs 1.84, Revision 

36; RG 1.147, Revision 17; and RG 

1.192, Revision 1, which are the subject 

of this rulemaking, is consistent with 

the overall objectives of the NTTAA and 

OMB Circular A–119. 

The NRC reviews each Section III, 

Section XI, and OM Code Case 

published by the ASME to ascertain 

whether it is consistent with the safe 

operation of nuclear power plants. The 

Code Cases found to be generically 

acceptable are listed in the RGs that are 

incorporated by reference in § 50.55a. 

The Code Cases found to be 

unacceptable are listed in RG 1.193, but 

licensees may still seek the NRC’s 

approval to apply these Code Cases 

through the processes in § 50.55a for 

requesting the approval of alternatives 

or for relief. Code Cases that the NRC 

finds to be conditionally acceptable are 

also listed in RGs 1.84, 1.147, and 1.192, 

which are the subject of this 

rulemaking, together with the 

conditions that must be used if the Code 

Case is applied. The NRC believes that 

this rule complies with the NTTAA and 

OMB Circular A–119 despite these 

conditions. If the NRC did not 

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conditionally accept ASME Code Cases, 

it would disapprove these Code Cases 

entirely. The effect would be that 

licensees and applicants would submit 

a larger number of requests for use of 

alternatives under the current 

§ 50.55a(a)(3), requests for relief under 

§ 50.55a(f) and (g), or requests for 

exemptions under §§ 50.12 and/or 52.7. 

For these reasons, the final rule does not 

conflict with any policy on agency use 

of consensus standards specified in 

OMB Circular A–119. 

The NRC did not identify any other 

voluntary consensus standards 

developed by the United States 

voluntary consensus standards bodies 

for use within the United States that the 

NRC could approve instead of the 

ASME Code Cases. 

The NRC also did not identify any 

voluntary consensus standards 

developed by multinational voluntary 

consensus standards bodies for use on a 

multinational basis that the NRC could 

incorporate by reference instead of the 

ASME Code Cases. This is because no 

other multinational voluntary consensus 

body would develop alternatives to a 

voluntary consensus standard (i.e., 

either the ASME BPV Code or the ASME 

OM Code) for which they did not 

develop and do not maintain. 

In summary, this final rule satisfies 

the requirements of Section 12(d)(3) of 

the NTTAA and OMB Circular A–119. 

XVII. Availability of Regulatory Guides 

Regulatory Guides Being Incorporated 

by Reference 

The NRC is issuing three revisions to 

existing guides in the agency’s 

‘‘Regulatory Guide’’ series. This final 

rule is incorporating by reference these 

three RGs into 10 CFR 50.55a. 

Revision 36 of RG 1.84, ‘‘Design, 

Fabrication, and Materials Code Case 

Acceptability, ASME Section III,’’ is 

available electronically under ADAMS 

Accession No. ML13339A515. 

Revision 17 of RG 1.147, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME Section XI, Division 1,’’ is 

available electronically under ADAMS 

Accession No. ML13339A689. 

Revision 1 of RG 1.192, ‘‘Operation 

and Maintenance [OM] Code Case 

Acceptability, ASME OM Code,’’ is 

available electronically under ADAMS 

Accession No. ML13340A034. 

As discussed in Section II of this 

document, ‘‘Opportunities for Public 

Participation,’’ these three RGs were 

issued in draft form for public comment 

in June 2013. The NRC staff’s responses 

to the public comments received are 

located in Section III of this document, 

‘‘Public Comment Analysis.’’ 

Issuance of Regulatory Guide 1.193 

The NRC is issuing a revision to an 

existing guide in the NRC’s ‘‘Regulatory 

Guide’’ series. This RG is not being 

incorporated by reference in this final 

rule. 

Revision 4 of RG 1.193, ‘‘ASME Code 

Cases Not Approved for Use,’’ was 

issued with a temporary identification 

of Draft Regulatory Guide, DG–1233. 

This revision of RG 1.193 includes new 

information reviewed by the NRC in 

ASME BPV Code Section III and Section 

XI Code Cases listed in Supplements 1– 

10 to the 2007 Edition, and the OM 

Code Cases listed in the 2002 Addenda 

through the 2006 Addenda. This is an 

update to RG 1.193, Revision 3, which 

included information from Supplements 

2–11 to the 2004 Edition, and 

Supplement 0 to the 2007 Edition of the 

BPV Code. 

This RG does not approve the use of 

the Code Cases listed herein. Licensees 

may submit a plant-specific request to 

implement one or more of the Code 

Cases listed in this RG. The request 

must address the NRC’s concerns about 

the Code Case at issue. 

The NRC published DG–1233 in the 

Federal Register on June 24, 2013 (78 

FR 37848), for a 75-day public comment 

period. The public comment period 

closed on September 9, 2013. Public 

comments on DG–1233 and the NRC 

staff responses to the public comments 

are available in ADAMS under 

Accession No. ML14106A577. 

XVIII. Availability of Documents 

The NRC is making the documents 

identified in Table IV available to 

interested persons through one or more 

of the following methods, as indicated. 

To access documents related to this 

action, see the ADDRESSES section of this 

document. 

TABLE IV—AVAILABILITY OF DOCUMENTS

Proposed rule documents ADAMS 

Accession No. 

Proposed Rule–Regulatory Analysis ................................................................................................................................................ ML103060189 

Proposed Rule–Federal Register Notice ........................................................................................................................................ ML103060003 

Proposed Reorganization of Paragraphs and Subparagraphs ........................................................................................................ ML12289A121 

Draft RG 1.84, Revision 36 (DG–1230) ........................................................................................................................................... ML102590003 

Draft RG 1.147, Revision 17 (DG–1231) ......................................................................................................................................... ML102590004 

Draft RG 1.192, Revision 1 (DG–1232) ........................................................................................................................................... ML102600001 

Final rule documents ADAMS 

Accession No. 

Final Rule–Regulatory Analysis ........................................................................................................................................................ ML14010A426 

Final Rule–Federal Register Notice ................................................................................................................................................ ML14008A332 

Final Reorganization of Paragraphs and Subparagraphs ................................................................................................................ ML14015A191 

Cross-Reference Tables (package) .................................................................................................................................................. ML14211A050 

RG 1.84, ‘‘Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,’’ Revision 36 .................................... ML13339A515 

RG 1.147, ‘‘Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,’’ Revision 17 ....................................... ML13339A689 

RG 1.192, ‘‘Operation and Maintenance Code Case Acceptability, ASME OM Code,’’ Revision 1 ............................................... ML13340A034 

RG 1.193, ‘‘ASME Code Cases Not Approved for Use,’’ Revision 4 .............................................................................................. ML13350A001 

RG 1.200, ‘‘An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed 

Activities,’’ Revision 2.

ML090410014 

RG 1.201, ‘‘Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their 

Safety Significance,’’ Revision 1.

ML061090627 

2007/12/19—‘‘SECY—Petition for Rulemaking to amend 10 CFR 50.55a—Rev.1’’ submitted by Ray West ................................ ML073600974 

Hatch Plant Report—‘‘Hatch, Units 1 & 2, Farley, Units 1 & 2, Vogtle, Units 1 & 2, Safety Evaluation Re. Request to Use 

ASME Code Case N–661’’.

ML033280037 

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Final rule documents ADAMS 

Accession No. 

EPRI Technical Report—Project No. 704—BWRVIP–108: BWR Vessel & Internals Project, Technical Basis for Reduction of 

Inspection Requirements for Boiling Water Reactor Nozzle-to-Vessel Shell Welds & Nozzle Blend Radii.

ML023330203 

Safety Evaluation of Proprietary EPRI Report—BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP– 

108).

ML073600374 

Comment Letter—Comment (4) of Bryan A. Erler on Behalf of ASME Supporting Draft Regulatory Guides DG–1191, DG– 

1192, DG–1193, and the Proposed Rule Incorporating the Final Revisions of these Regulatory Guides into 10 CFR 50.55a.

ML092190138 

SRM–COMNJD–03–0002—Stabilizing the PRA Quality Expectations and Requirements ............................................................. ML033520457 

SECY–04–0118—Plan for the Implementation of the Commission’s Phased Approach to Probabilistic Risk Assessment Quality.ML041470505 

SRM–SECY–04–0118—Plan for the Implementation of the Commission’s Phased Approach to Probabilistic Risk Assessment 

Quality.

ML042800369 

NUREG–0800—Chapter 4, Section 4.5.1, Revision 3, Control Rod Drive Structural Materials, dated March 2007 ...................... ML070230007 

NUREG–0800—Chapter 5, Section 5.2.3, Revision 3, Reactor Coolant Pressure Boundary Materials, dated March 2007 ......... ML063190006 

NUREG/CR–6943—A Study of Remote Visual Methods to Detect Cracking in Reactor Components .......................................... ML073110060 

List of Subjects in 10 CFR Part 50 

Antitrust, Classified information, 

Criminal penalties, Fire protection, 

Incorporation by reference, 

Intergovernmental relations, Nuclear 

power plants and reactors, Radiation 

protection, Reactor siting criteria, 

Reporting and recordkeeping 

requirements. 

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 part 50. 

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES 

■ 1. The authority citation for part 50 is 

revised to read as follows: 

Authority: Atomic Energy Act secs. 102, 

103, 104, 105, 147, 149, 161, 181, 182, 183, 

186, 189, 223, 234 (42 U.S.C. 2132, 2133, 

2134, 2135, 2167, 2169, 2201, 2231, 2232, 

2233, 2236, 2239, 2273, 2282); Energy 

Reorganization Act secs. 201, 202, 206 (42 

U.S.C. 5841, 5842, 5846); Nuclear Waste 

Policy Act sec. 306 (42 U.S.C. 10226); 

Government Paperwork Elimination Act sec. 

1704 (44 U.S.C. 3504 note); Energy Policy Act 

of 2005, Pub. L. No. 109–58, 119 Stat. 194 

(2005). Section 50.7 also issued under Pub. 

L. 95–601, sec. 10, as amended by Pub. L. 

102–486, sec. 2902 (42 U.S.C. 5851). Section 

50.10 also issued under Atomic Energy Act 

secs. 101, 185 (42 U.S.C. 2131, 2235); 

National Environmental Protection Act sec. 

102 (42 U.S.C. 4332). Sections 50.13, 

50.54(d), and 50.103 also issued under 

Atomic Energy Act sec. 108 (42 U.S.C. 2138). 

Sections 50.23, 50.35, 50.55, and 50.56 also 

issued under Atomic Energy Act sec. 185 (42 

U.S.C. 2235). Appendix Q also issued under 

National Environmental Protection Act sec. 

102 (42 U.S.C. 4332). Sections 50.34 and 

50.54 also issued under sec. 204 (42 U.S.C. 

5844). Sections 50.58, 50.91, and 50.92 also 

issued under Pub. L. 97–415 (42 U.S.C. 

2239). Section 50.78 also issued under 

Atomic Energy Act sec. 122 (42 U.S.C. 2152). 

Sections 50.80–50.81 also issued under 

Atomic Energy Act sec. 184 (42 U.S.C. 2234). 

■ 2. In § 50.54, revise the introductory 

text, add reserved paragraph (ii), and 

add paragraph (jj) to read as follows: 

§ 50.54 Conditions of licenses. 

The following paragraphs of this 

section, with the exception of 

paragraphs (r) and (gg), and the 

applicable requirements of 10 CFR 

50.55a, 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) introductory text, (i)(1), 

(j), (k), (l), (m), (n), (q), (w), (x), (y), (z), 

and (hh) of this section are only 

applicable after the Commission makes 

the finding under § 52.103(g) of this 

chapter. 

* * * * * 

(ii) [Reserved] 

(jj) Structures, systems, and 

components subject to the codes and 

standards in 10 CFR 50.55a must be 

designed, fabricated, erected, 

constructed, tested, and inspected to 

quality standards commensurate with 

the importance of the safety function to 

be performed. 

■ 3. In § 50.55, revise the introductory 

text, add reserved paragraphs (g) and 

(h), and add paragraph (i) to read as 

follows: 

§ 50.55 Conditions of construction 

permits, early site permits, combined 

licenses, and manufacturing licenses. 

Each construction permit for a 

utilization facility is subject to the 

following terms and conditions and the 

applicable requirements of § 50.55a; 

each construction permit for a 

production facility is subject to the 

following terms and conditions with the 

exception of paragraph (i); 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), (f), and (i) of this section 

and the applicable requirements of 

§ 50.55a; and each combined license is 

subject to the terms and conditions in 

paragraphs (e), (f), and (i) of this section 

and the applicable requirements of 

§ 50.55a until the date that the 

Commission makes the finding under 

§ 52.103(g) of this chapter: 

* * * * * 

(g) [Reserved] 

(h) [Reserved] 

(i) Structures, systems, and 

components subject to the codes and 

standards in 10 CFR 50.55a must be 

designed, fabricated, erected, 

constructed, tested, and inspected to 

quality standards commensurate with 

the importance of the safety function to 

be performed. 

■ 4. Revise § 50.55a to read as follows: 

§ 50.55a Codes and standards. 

(a) Documents approved for 

incorporation by reference. The 

standards listed in this paragraph have 

been approved for incorporation by 

reference by the Director of the Federal 

Register pursuant to 5 U.S.C. 552(a) and 

1 CFR part 51. The standards are 

available for inspection at the NRC 

Technical Library, 11545 Rockville Pike, 

Rockville, Maryland 20852; telephone: 

301–415–6239; 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/cfr/

ibr-locations.html. 

(1) American Society of Mechanical 

Engineers (ASME), Three Park Avenue, 

New York, NY 10016; telephone: 

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1–800–843–2763; http://www.asme.org/

Codes/. 

(i) ASME Boiler and Pressure Vessel 

Code, Section III. The editions and 

addenda for Section III of the ASME 

Boiler and Pressure Vessel Code are 

listed below, but limited to those 

provisions identified in paragraph (b)(1) 

of this section. 

(A) ‘‘Rules for Construction of Nuclear 

Vessels:’’ 

(1) 1963 Edition, 

(2) Summer 1964 Addenda, 

(3) Winter 1964 Addenda, 

(4) 1965 Edition, 

(5) 1965 Summer Addenda, 

(6) 1965 Winter Addenda, 

(7) 1966 Summer Addenda, 

(8) 1966 Winter Addenda, 

(9) 1967 Summer Addenda, 

(10) 1967 Winter Addenda, 

(11) 1968 Edition, 

(12) 1968 Summer Addenda, 

(13)1968 Winter Addenda, 

(14) 1969 Summer Addenda, 

(15) 1969 Winter Addenda, 

(16) 1970 Summer Addenda, and 

(17) 1970 Winter Addenda. 

(B) ‘‘Rules for Construction of Nuclear 

Power Plant Components:’’ 

(1) 1971 Edition, 

(2) 1971 Summer Addenda, 

(3) 1971 Winter Addenda, 

(4) 1972 Summer Addenda, 

(5) 1972 Winter Addenda, 

(6) 1973 Summer Addenda, and 

(7) 1973 Winter Addenda. 

(C) ‘‘Division 1 Rules for Construction 

of Nuclear Power Plant Components:’’ 

(1) 1974 Edition, 

(2) 1974 Summer Addenda, 

(3) 1974 Winter Addenda, 

(4) 1975 Summer Addenda, 

(5) 1975 Winter Addenda, 

(6) 1976 Summer Addenda, and 

(7) 1976 Winter Addenda; 

(D) ‘‘Rules for Construction of Nuclear 

Power Plant Components—Division 1’’; 

(1) 1977 Edition, 

(2) 1977 Summer Addenda, 

(3) 1977 Winter Addenda, 

(4) 1978 Summer Addenda, 

(5) 1978 Winter Addenda, 

(6) 1979 Summer Addenda, 

(7) 1979 Winter Addenda, 

(8) 1980 Edition, 

(9) 1980 Summer Addenda, 

(10) 1980 Winter Addenda, 

(11) 1981 Summer Addenda, 

(12) 1981 Winter Addenda, 

(13) 1982 Summer Addenda, 

(14) 1982 Winter Addenda, 

(15) 1983 Edition, 

(16) 1983 Summer Addenda, 

(17) 1983 Winter Addenda, 

(18) 1984 Summer Addenda, 

(19) 1984 Winter Addenda, 

(20) 1985 Summer Addenda, 

(21) 1985 Winter Addenda, 

(22) 1986 Edition, 

(23) 1986 Addenda, 

(24) 1987 Addenda, 

(25) 1988 Addenda, 

(26) 1989 Edition, 

(27) 1989 Addenda, 

(28) 1990 Addenda, 

(29) 1991 Addenda, 

(30) 1992 Edition, 

(31) 1992 Addenda, 

(32) 1993 Addenda, 

(33) 1994 Addenda, 

(34) 1995 Edition, 

(35) 1995 Addenda, 

(36) 1996 Addenda, and 

(37) 1997 Addenda. 

(E) ‘‘Rules for Construction of Nuclear 

Facility Components—Division 1:’’ 

(1) 1998 Edition, 

(2) 1998 Addenda, 

(3) 1999 Addenda, 

(4) 2000 Addenda, 

(5) 2001 Edition, 

(6) 2001 Addenda, 

(7) 2002 Addenda, 

(8) 2003 Addenda, 

(9) 2004 Edition, 

(10) 2005 Addenda, 

(11) 2006 Addenda, 

(12) 2007 Edition, and 

(13) 2008 Addenda. 

(ii) ASME Boiler and Pressure Vessel 

Code, Section XI. The editions and 

addenda for Section XI of the ASME 

Boiler and Pressure Vessel Code are 

listed below, but limited to those 

provisions identified in paragraph (b)(2) 

of this section. 

(A) ‘‘Rules for Inservice Inspection of 

Nuclear Reactor Coolant Systems:’’ 

(1) 1970 Edition, 

(2) 1971 Edition, 

(3) 1971 Summer Addenda, 

(4) 1971 Winter Addenda, 

(5) 1972 Summer Addenda, 

(6) 1972 Winter Addenda, 

(7) 1973 Summer Addenda, and 

(8) 1973 Winter Addenda. 

(B) ‘‘Rules for Inservice Inspection of 

Nuclear Power Plant Components:’’ 

(1) 1974 Edition, 

(2) 1974 Summer Addenda, 

(3) 1974 Winter Addenda, and 

(4) 1975 Summer Addenda. 

(C) ‘‘Rules for Inservice Inspection of 

Nuclear Power Plant Components— 

Division 1:’’ 

(1) 1977 Edition, 

(2) 1977 Summer Addenda, 

(3) 1977 Winter Addenda, 

(4) 1978 Summer Addenda, 

(5) 1978 Winter Addenda, 

(6) 1979 Summer Addenda, 

(7) 1979 Winter Addenda, 

(8) 1980 Edition, 

(9) 1980 Winter Addenda, 

(10) 1981 Summer Addenda, 

(11) 1981 Winter Addenda, 

(12) 1982 Summer Addenda, 

(13) 1982 Winter Addenda, 

(14) 1983 Edition, 

(15) 1983 Summer Addenda, 

(16) 1983 Winter Addenda, 

(17) 1984 Summer Addenda, 

(18) 1984 Winter Addenda, 

(19) 1985 Summer Addenda, 

(20) 1985 Winter Addenda, 

(21) 1986 Edition, 

(22) 1986 Addenda, 

(23) 1987 Addenda, 

(24) 1988 Addenda, 

(25) 1989 Edition, 

(26) 1989 Addenda, 

(27) 1990 Addenda, 

(28) 1991 Addenda, 

(29) 1992 Edition, 

(30) 1992 Addenda, 

(31) 1993 Addenda, 

(32) 1994 Addenda, 

(33) 1995 Edition, 

(34) 1995 Addenda, 

(35) 1996 Addenda, 

(36) 1997 Addenda, 

(37) 1998 Edition, 

(38) 1998 Addenda, 

(39) 1999 Addenda, 

(40) 2000 Addenda, 

(41) 2001 Edition, 

(42) 2001 Addenda, 

(43) 2002 Addenda, 

(44) 2003 Addenda, 

(45) 2004 Edition, 

(46) 2005 Addenda, 

(47) 2006 Addenda, 

(48) 2007 Edition, and 

(49) 2008 Addenda. 

(iii) ASME Code Cases: Nuclear 

Components—(A) ASME Code Case N– 

722–1. ASME Code Case N–722–1, 

‘‘Additional Examinations for PWR 

Pressure Retaining Welds in Class 1 

Components Fabricated with Alloy 600/ 

82/182 Materials, Section XI, Division 

1’’ (Approval Date: January 26, 2009), 

with the conditions in paragraph 

(g)(6)(ii)(E) of this section. 

(B) ASME Code Case N–729–1. ASME 

Code Case N–729–1, ‘‘Alternative 

Examination Requirements for PWR 

Reactor Vessel Upper Heads With 

Nozzles Having Pressure-Retaining 

Partial-Penetration Welds, Section XI, 

Division 1’’ (Approval Date: March 28, 

2006), with the conditions in paragraph 

(g)(6)(ii)(D) of this section. 

(C) ASME Code Case N–770–1. ASME 

Code Case N–770–1, ‘‘Additional 

Examinations for PWR Pressure 

Retaining Welds in Class 1 Components 

Fabricated with Alloy 600/82/182 

Materials, Section XI, Division 1’’ 

(Approval Date: December 25, 2009), 

with the conditions in paragraph 

(g)(6)(ii)(F) of this section. 

(iv) ASME Operation and 

Maintenance Code. The editions and 

addenda for the ASME Code for 

Operation and Maintenance of Nuclear 

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Power Plants are listed below, but 

limited to those provisions identified in 

paragraph (b)(3) of this section. 

(A) ‘‘Code for Operation and 

Maintenance of Nuclear Power Plants:’’ 

(1) 1995 Edition, 

(2) 1996 Addenda, 

(3) 1997 Addenda, 

(4) 1998 Edition, 

(5) 1999 Addenda, 

(6) 2000 Addenda, 

(7) 2001 Edition, 

(8) 2002 Addenda, 

(9) 2003 Addenda, 

(10) 2004 Edition, 

(11) 2005 Addenda, and 

(12) 2006 Addenda. 

(B) [Reserved] 

(2) Institute of Electrical and 

Electronics Engineers (IEEE) Service 

Center, 445 Hoes Lane, Piscataway, NJ 

08855; telephone: 1–800–678–4333; 

http://ieeexplore.ieee.org. 

(i) IEEE standard 279–1971. (IEEE Std 

279–1971), ‘‘Criteria for Protection 

Systems for Nuclear Power Generating 

Stations’’ (Approval Date: June 3, 1971), 

referenced in paragraph (h)(2) of this 

section. 

(ii) IEEE Standard 603–1991. (IEEE 

Std 603–1991), ‘‘Standard Criteria for 

Safety Systems for Nuclear Power 

Generating Stations’’ (Approval Date: 

June 27, 1991), referenced in paragraphs 

(h)(2) and (3) of this section. All other 

standards that are referenced in IEEE 

Std 603–1991 are not approved for 

incorporation by reference. 

(iii) IEEE standard 603–1991, 

correction sheet. (IEEE Std 603–1991 

correction sheet), ‘‘Standard Criteria for 

Safety Systems for Nuclear Power 

Generating Stations, Correction Sheet, 

Issued January 30, 1995, ’’ referenced in 

paragraphs (h)(2) and (3) of this section. 

(Copies of this correction sheet may be 

purchased from Thomson Reuters, 3916 

Ranchero Dr., Ann Arbor, MI 48108; 

http://www.techstreet.com.) 

(3) U.S. Nuclear Regulatory 

Commission (NRC) Public Document 

Room, 11555 Rockville Pike, Rockville, 

Maryland 20852; telephone: 1–800– 

397–4209; email: pdr.resource@nrc.gov; 

http://www.nrc.gov/reading-rm/doccollections/reg-guides/. 

(i) NRC Regulatory Guide 1.84, 

Revision 36. NRC Regulatory Guide 

1.84, Revision 36, ‘‘Design, Fabrication, 

and Materials Code Case Acceptability, 

ASME Section III,’’ dated August 2014, 

with the requirements in paragraph 

(b)(4) of this section. 

(ii) NRC Regulatory Guide 1.147, 

Revision 17. NRC Regulatory Guide 

1.147, Revision 17, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME Section XI, Division 1,’’ dated 

August 2014, which lists ASME Code 

Cases that the NRC has approved in 

accordance with the requirements in 

paragraph (b)(5) of this section. 

(iii) NRC Regulatory Guide 1.192, 

Revision 1. NRC Regulatory Guide 

1.192, Revision 1, ‘‘Operation and 

Maintenance Code Case Acceptability, 

ASME OM Code,’’ dated August 2014, 

which lists ASME Code Cases that the 

NRC has approved in accordance with 

the requirements in paragraph (b)(6) of 

this section. 

(b) Use and conditions on the use of 

standards. Systems and components of 

boiling and pressurized water-cooled 

nuclear power reactors must meet the 

requirements of the ASME Boiler and 

Pressure Vessel Code (BPV Code) and 

the ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(OM Code) as specified in this 

paragraph. Each combined license for a 

utilization facility is subject to the 

following conditions. 

(1) Conditions on ASME BPV Code 

Section III. Each manufacturing license, 

standard design approval, and design 

certification under part 52 of this 

chapter is subject to the following 

conditions. As used in this section, 

references to Section III refer to Section 

III of the ASME Boiler and Pressure 

Vessel Code and include the 1963 

Edition through 1973 Winter Addenda 

and the 1974 Edition (Division 1) 

through the 2008 Addenda (Division 1), 

subject to the following conditions: 

(i) Section III condition: 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) Section III condition: Weld leg 

dimensions. When applying the 1989 

Addenda through the latest edition and 

addenda, applicants or licensees may 

not apply subparagraphs NB– 

3683.4(c)(1) and NB–3683.4(c)(2) or 

Footnote 11 from the 1989 Addenda 

through the 2003 Addenda, or Footnote 

13 from the 2004 Edition through the 

2008 Addenda to Figures NC–3673.2(b)– 

1 and ND–3673.2(b)–1 for welds with 

leg size less than 1.09 tn.

(iii) Section III condition: Seismic 

design of piping. Applicants or licensees 

may use Subarticles NB–3200, NB– 

3600, NC–3600, and ND–3600 for 

seismic design of piping, up to and 

including the 1993 Addenda, subject to 

the condition specified in paragraph 

(b)(1)(ii) of this section. Applicants or 

licensees may not use these subarticles 

for seismic design of piping in the 1994 

Addenda through the 2005 Addenda 

incorporated by reference in paragraph 

(a)(1) of this section, except that 

Subarticle NB–3200 in the 2004 Edition 

through the 2008 Addenda may be used 

by applicants and licensees, subject to 

the condition in paragraph (b)(1)(iii)(A) 

of this section. Applicants or licensees 

may use Subarticles NB–3600, NC– 

3600, and ND–3600 for the seismic 

design of piping in the 2006 Addenda 

through the 2008 Addenda, subject to 

the conditions of this paragraph 

corresponding to those subarticles. 

(A) Seismic design of piping: First 

provision. When applying Note (1) of 

Figure NB–3222–1 for Level B service 

limits, the calculation of Pb stresses 

must include reversing dynamic loads 

(including inertia earthquake effects) if 

evaluation of these loads is required by 

NB–3223(b). 

(B) Seismic design of piping: Second 

provision. For Class 1 piping, the 

material and Do/t requirements of NB– 

3656(b) must be met for all Service 

Limits when the Service Limits include 

reversing dynamic loads, and the 

alternative rules for reversing dynamic 

loads are used. 

(iv) Section III condition: Quality 

assurance. When applying editions and 

addenda later than the 1989 Edition of 

Section III, the requirements of NQA–1, 

‘‘Quality Assurance Requirements for 

Nuclear Facilities,’’ 1986 Edition 

through the 1994 Edition, are acceptable 

for use, provided that the edition and 

addenda of NQA–1 specified in NCA– 

4000 is used in conjunction with the 

administrative, quality, and technical 

provisions contained in the edition and 

addenda of Section III being used. 

(v) Section III condition: 

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 

(a)(1) of this section. 

(vi) Section III condition: Subsection 

NH. The provisions in Subsection NH, 

‘‘Class 1 Components in Elevated 

Temperature Service,’’ 1995 Addenda 

through the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1) of this section, may only be used 

for the design and construction of Type 

316 stainless steel pressurizer heater 

sleeves where service conditions do not 

cause the components to reach 

temperatures exceeding 900 °F. 

(vii) Section III condition: Capacity 

certification and demonstration of 

function of incompressible-fluid 

pressure-relief valves. When applying 

the 2006 Addenda through the 2007 

Edition up to and including the 2008 

Addenda, applicants and licensees may 

use paragraph NB–7742, except that 

paragraph NB–7742(a)(2) may not be 

used. For a valve design of a single size 

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to be certified over a range of set 

pressures, the demonstration of function 

tests under paragraph NB–7742 must be 

conducted as prescribed in NB–7732.2 

on two valves covering the minimum set 

pressure for the design and the 

maximum set pressure that can be 

accommodated at the demonstration 

facility selected for the test. 

(2) Conditions on ASME BPV Code 

Section XI. As used in this section, 

references to Section XI refer to Section 

XI, Division 1, of the ASME Boiler and 

Pressure Vessel Code, and include the 

1970 Edition through the 1976 Winter 

Addenda and the 1977 Edition through 

the 2007 Edition with the 2008 

Addenda, subject to the following 

conditions: 

(i) [Reserved] 

(ii) Section XI condition: Pressureretaining welds in ASME Code Class 1 

piping (applies to Table IWB–2500 and 

IWB–2500–1 and Category B–J). If the 

facility’s application for a construction 

permit was docketed prior to July 1, 

1978, the extent of examination for Code 

Class 1 pipe welds may be determined 

by the requirements of Table IWB–2500 

and Table IWB–2600 Category B–J of 

Section XI of the ASME BPV Code in 

the 1974 Edition and Addenda through 

the Summer 1975 Addenda or other 

requirements the NRC may adopt. 

(iii) [Reserved] 

(iv) [Reserved] 

(v) [Reserved] 

(vi) Section XI condition: Effective 

edition and addenda of Subsection IWE 

and Subsection IWL. Applicants or 

licensees may use either the 1992 

Edition with the 1992 Addenda or the 

1995 Edition with the 1996 Addenda of 

Subsection IWE and Subsection IWL, as 

conditioned by the requirements in 

paragraphs (b)(2)(viii) and (ix) of this 

section, when implementing the initial 

120-month inspection interval for the 

containment inservice inspection 

requirements of this section. Successive 

120-month interval updates must be 

implemented in accordance with 

paragraph (g)(4)(ii) of this section. 

(vii) Section XI condition: Section XI 

references to OM Part 4, OM Part 6, and 

OM Part 10 (Table IWA–1600–1). When 

using Table IWA–1600–1, ‘‘Referenced 

Standards and Specifications,’’ in the 

Section XI, Division 1, 1987 Addenda, 

1988 Addenda, or 1989 Edition, the 

specified ‘‘Revision Date or Indicator’’ 

for ASME/ANSI OM part 4, ASME/

ANSI part 6, and ASME/ANSI part 10 

must be the OMa–1988 Addenda to the 

OM–1987 Edition. These requirements 

have been incorporated into the OM 

Code, which is incorporated by 

reference in paragraph (a)(1)(iv) of this 

section. 

(viii) Section XI condition: Concrete 

containment examinations. Applicants 

or licensees applying Subsection IWL, 

1992 Edition with the 1992 Addenda, 

must apply paragraphs (b)(2)(viii)(A) 

through (E) of this section. Applicants 

or licensees applying Subsection IWL, 

1995 Edition with the 1996 Addenda, 

must apply paragraphs (b)(2)(viii)(A), 

(b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of 

this section. Applicants or licensees 

applying Subsection IWL, 1998 Edition 

through the 2000 Addenda, must apply 

paragraphs (b)(2)(viii)(E) and (F) of this 

section. Applicants or licensees 

applying Subsection IWL, 2001 Edition 

through the 2004 Edition, up to and 

including the 2006 Addenda, must 

apply paragraphs (b)(2)(viii)(E) through 

(G) of this section. Applicants or 

licensees applying Subsection IWL, 

2007 Edition through the latest edition 

and addenda incorporated by reference 

in paragraph (a)(1)(ii) of this section, 

must apply paragraph (b)(2)(viii)(E) of 

this section. 

(A) Concrete containment 

examinations: First provision. Grease 

caps that are accessible must be visually 

examined to detect grease leakage or 

grease cap deformations. Grease caps 

must be removed for this examination 

when there is evidence of grease cap 

deformation that indicates deterioration 

of anchorage hardware. 

(B) Concrete containment 

examinations: Second provision. When 

evaluation of consecutive surveillances 

of prestressing forces for the same 

tendon or tendons in a group indicates 

a trend of prestress loss such that the 

tendon force(s) would be less than the 

minimum design prestress requirements 

before the next inspection interval, an 

evaluation must be performed and 

reported in the Engineering Evaluation 

Report as prescribed in IWL–3300. 

(C) Concrete containment 

examinations: Third provision. When 

the elongation corresponding to a 

specific load (adjusted for effective 

wires or strands) during retensioning of 

tendons differs by more than 10 percent 

from that recorded during the last 

measurement, an evaluation must be 

performed to determine whether the 

difference is related to wire failures or 

slip of wires in anchorage. A difference 

of more than 10 percent must be 

identified in the ISI Summary Report 

required by IWA–6000. 

(D) Concrete containment 

examinations: Fourth provision. The 

applicant or licensee must report the 

following conditions, if they occur, in 

the ISI Summary Report required by 

IWA–6000: 

(1) The sampled sheathing filler 

grease contains chemically combined 

water exceeding 10 percent by weight or 

the presence of free water; 

(2) The absolute difference between 

the amount removed and the amount 

replaced exceeds 10 percent of the 

tendon net duct volume; and 

(3) Grease leakage is detected during 

general visual examination of the 

containment surface. 

(E) Concrete containment 

examinations: Fifth provision. For Class 

CC applications, the applicant or 

licensee must evaluate the acceptability 

of inaccessible areas when conditions 

exist in accessible areas that could 

indicate the presence of or the result in 

degradation to such inaccessible areas. 

For each inaccessible area identified, 

the applicant or licensee must provide 

the following in the ISI Summary Report 

required by IWA–6000: 

(1) A description of the type and 

estimated extent of degradation, and the 

conditions that led to the degradation; 

(2) An evaluation of each area, and 

the result of the evaluation; and 

(3) A description of necessary 

corrective actions. 

(F) Concrete containment 

examinations: Sixth provision. 

Personnel that examine containment 

concrete surfaces and tendon hardware, 

wires, or strands must meet the 

qualification provisions in IWA–2300. 

The ‘‘owner-defined’’ personnel 

qualification provisions in IWL–2310(d) 

are not approved for use. 

(G) Concrete containment 

examinations: Seventh provision. 

Corrosion protection material must be 

restored following concrete containment 

post-tensioning system repair and 

replacement activities in accordance 

with the quality assurance program 

requirements specified in IWA–1400. 

(ix) Section XI condition: Metal 

containment examinations. Applicants 

or licensees applying Subsection IWE, 

1992 Edition with the 1992 Addenda, or 

the 1995 Edition with the 1996 

Addenda, must satisfy the requirements 

of paragraphs (b)(2)(ix)(A) through (E) of 

this section. Applicants or licensees 

applying Subsection IWE, 1998 Edition 

through the 2001 Edition with the 2003 

Addenda, must satisfy the requirements 

of paragraphs (b)(2)(ix)(A) and (B) and 

(b)(2)(ix)(F) through (I) of this section. 

Applicants or licensees applying 

Subsection IWE, 2004 Edition, up to and 

including the 2005 Addenda, must 

satisfy the requirements of paragraphs 

(b)(2)(ix)(A) and (B) and (b)(2)(ix)(F) 

through (H) of this section. Applicants 

or licensees applying Subsection IWE, 

2004 Edition with the 2006 Addenda, 

must satisfy the requirements of 

paragraphs (b)(2)(ix)(A)(2) and 

(b)(2)(ix)(B) of this section. Applicants 

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or licensees applying Subsection IWE, 

2007 Edition through the latest addenda 

incorporated by reference in paragraph 

(a)(1)(ii) of this section, must satisfy the 

requirements of paragraphs 

(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) 

of this section. 

(A) Metal containment examinations: 

First provision. For Class MC 

applications, the following apply to 

inaccessible areas. 

(1) The applicant or licensee must 

evaluate the acceptability of 

inaccessible areas when conditions exist 

in accessible areas that could indicate 

the presence of or could result in 

degradation to such inaccessible areas. 

(2) For each inaccessible area 

identified for evaluation, the applicant 

or licensee must provide the following 

in the ISI Summary Report as required 

by IWA–6000: 

(i) A description of the type and 

estimated extent of degradation, and the 

conditions that led to the degradation; 

(ii) An evaluation of each area, and 

the result of the evaluation; and 

(iii) A description of necessary 

corrective actions. 

(B) Metal containment examinations: 

Second provision. When performing 

remotely the visual examinations 

required by Subsection IWE, the 

maximum direct examination distance 

specified in Table IWA–2210–1 may be 

extended and the minimum 

illumination requirements specified in 

Table IWA–2210–1 may be decreased 

provided that the conditions or 

indications for which the visual 

examination is performed can be 

detected at the chosen distance and 

illumination. 

(C) Metal containment examinations: 

Third provision. The examinations 

specified in Examination Category E–B, 

Pressure Retaining Welds, and 

Examination Category E–F, Pressure 

Retaining Dissimilar Metal Welds, are 

optional. 

(D) Metal containment examinations: 

Fourth provision. This paragraph 

(b)(2)(ix)(D) may be used as an 

alternative to the requirements of IWE– 

2430. 

(1) If the examinations reveal flaws or 

areas of degradation exceeding the 

acceptance standards of Table IWE– 

3410–1, an evaluation must be 

performed to determine whether 

additional component examinations are 

required. For each flaw or area of 

degradation identified that exceeds 

acceptance standards, the applicant or 

licensee must provide the following in 

the ISI Summary Report required by 

IWA–6000: 

(i) A description of each flaw or area, 

including the extent of degradation, and 

the conditions that led to the 

degradation; 

(ii) The acceptability of each flaw or 

area and the need for additional 

examinations to verify that similar 

degradation does not exist in similar 

components; and 

(iii) A description of necessary 

corrective actions. 

(2) The number and type of additional 

examinations to ensure detection of 

similar degradation in similar 

components. 

(E) Metal containment examinations: 

Fifth provision. A general visual 

examination as required by Subsection 

IWE must be performed once each 

period. 

(F) Metal containment examinations: 

Sixth provision. VT–1 and VT–3 

examinations must be conducted in 

accordance with IWA–2200. Personnel 

conducting examinations in accordance 

with the VT–1 or VT–3 examination 

method must be qualified in accordance 

with IWA–2300. The ‘‘owner-defined’’ 

personnel qualification provisions in 

IWE–2330(a) for personnel that conduct 

VT–1 and VT–3 examinations are not 

approved for use. 

(G) Metal containment examinations: 

Seventh provision. The VT–3 

examination method must be used to 

conduct the examinations in Items 

E1.12 and E1.20 of Table IWE–2500–1, 

and the VT–1 examination method must 

be used to conduct the examination in 

Item E4.11 of Table IWE–2500–1. An 

examination of the pressure-retaining 

bolted connections in Item E1.11 of 

Table IWE–2500–1 using the VT–3 

examination method must be conducted 

once each interval. The ‘‘ownerdefined’’ visual examination provisions 

in IWE–2310(a) are not approved for use 

for VT–1 and VT–3 examinations. 

(H) Metal containment examinations: 

Eighth provision. Containment bolted 

connections that are disassembled 

during the scheduled performance of 

the examinations in Item E1.11 of Table 

IWE–2500–1 must be examined using 

the VT–3 examination method. Flaws or 

degradation identified during the 

performance of a VT–3 examination 

must be examined in accordance with 

the VT–1 examination method. The 

criteria in the material specification or 

IWB–3517.1 must be used to evaluate 

containment bolting flaws or 

degradation. As an alternative to 

performing VT–3 examinations of 

containment bolted connections that are 

disassembled during the scheduled 

performance of Item E1.11, VT–3 

examinations of containment bolted 

connections may be conducted 

whenever containment bolted 

connections are disassembled for any 

reason. 

(I) Metal containment examinations: 

Ninth provision. The ultrasonic 

examination acceptance standard 

specified in IWE–3511.3 for Class MC 

pressure-retaining components must 

also be applied to metallic liners of 

Class CC pressure-retaining 

components. 

(J) Metal containment examinations: 

Tenth provision. In general, a repair/

replacement activity such as replacing a 

large containment penetration, cutting a 

large construction opening in the 

containment pressure boundary to 

replace steam generators, reactor vessel 

heads, pressurizers, or other major 

equipment; or other similar 

modification is considered a major 

containment modification. When 

applying IWE–5000 to Class MC 

pressure-retaining components, any 

major containment modification or 

repair/replacement must be followed by 

a Type A test to provide assurance of 

both containment structural integrity 

and leaktight integrity prior to returning 

to service, in accordance with 10 CFR 

part 50, Appendix J, Option A or Option 

B on which the applicant’s or licensee’s 

Containment Leak-Rate Testing Program 

is based. When applying IWE–5000, if a 

Type A, B, or C Test is performed, the 

test pressure and acceptance standard 

for the test must be in accordance with 

10 CFR part 50, Appendix J. 

(x) Section XI condition: Quality 

assurance. When applying Section XI 

editions and addenda later than the 

1989 Edition, the requirements of NQA– 

1, ‘‘Quality Assurance Requirements for 

Nuclear Facilities,’’ 1979 Addenda 

through the 1989 Edition, are acceptable 

as permitted by IWA–1400 of Section 

XI, if the licensee uses its 10 CFR part 

50, Appendix B, quality assurance 

program, in conjunction with Section XI 

requirements. Commitments contained 

in the licensee’s quality assurance 

program description that are more 

stringent than those contained in NQA– 

1 must govern Section XI activities. 

Further, where NQA–1 and Section XI 

do not address the commitments 

contained in the licensee’s Appendix B 

quality assurance program description, 

the commitments must be applied to 

Section XI activities. 

(xi) [Reserved] 

(xii) Section XI condition: Underwater 

welding. The provisions in IWA–4660, 

‘‘Underwater Welding,’’ of Section XI, 

1997 Addenda through the latest edition 

and addenda incorporated by reference 

in paragraph (a)(1)(ii) of this section, are 

not approved for use on irradiated 

material. 

(xiii) [Reserved] 

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(xiv) Section XI condition: Appendix 

VIII personnel qualification. All 

personnel qualified for performing 

ultrasonic examinations in accordance 

with Appendix VIII must receive 8 

hours of annual hands-on training on 

specimens that contain cracks. 

Licensees applying the 1999 Addenda 

through the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1)(ii) of this section may use the 

annual practice requirements in VII– 

4240 of Appendix VII of Section XI in 

place of the 8 hours of annual hands-on 

training provided that the supplemental 

practice is performed on material or 

welds that contain cracks, or by 

analyzing prerecorded data from 

material or welds that contain cracks. In 

either case, training must be completed 

no earlier than 6 months prior to 

performing ultrasonic examinations at a 

licensee’s facility. 

(xv) Section XI condition: Appendix 

VIII specimen set and qualification 

requirements. Licensees using 

Appendix VIII in the 1995 Edition 

through the 2001 Edition of the ASME 

Boiler and Pressure Vessel Code may 

elect to comply with all of the 

provisions in paragraphs (b)(2)(xv)(A) 

through (M) of this section, except for 

paragraph (b)(2)(xv)(F) of this section, 

which may be used at the licensee’s 

option. Licensees using editions and 

addenda after 2001 Edition through the 

2006 Addenda must use the 2001 

Edition of Appendix VIII and may elect 

to comply with all of the provisions in 

paragraphs (b)(2)(xv)(A) through (M) of 

this section, except for paragraph 

(b)(2)(xv)(F) of this section, which may 

be used at the licensee’s option. 

(A) Specimen set and qualification: 

First provision. When applying 

Supplements 2, 3, and 10 to Appendix 

VIII, the following examination coverage 

criteria requirements must be used: 

(1) Piping must be examined in two 

axial directions, and when examination 

in the circumferential direction is 

required, the circumferential 

examination must be performed in two 

directions, provided access is available. 

Dissimilar metal welds must be 

examined axially and circumferentially. 

(2) Where examination from both 

sides is not possible, full coverage credit 

may be claimed from a single side for 

ferritic welds. Where examination from 

both sides is not possible on austenitic 

welds or dissimilar metal welds, full 

coverage credit from a single side may 

be claimed only after completing a 

successful single-sided Appendix VIII 

demonstration using flaws on the 

opposite side of the weld. Dissimilar 

metal weld qualifications must be 

demonstrated from the austenitic side of 

the weld, and the qualification may be 

expanded for austenitic welds with no 

austenitic sides using a separate add-on 

performance demonstration. Dissimilar 

metal welds may be examined from 

either side of the weld. 

(B) Specimen set and qualification: 

Second provision. The following 

conditions must be used in addition to 

the requirements of Supplement 4 to 

Appendix VIII: 

(1) Paragraph 3.1, Detection 

acceptance criteria—Personnel are 

qualified for detection if the results of 

the performance demonstration satisfy 

the detection requirements of ASME 

Section XI, Appendix VIII, Table VIII– 

S4–1, and no flaw greater than 0.25 inch 

through-wall dimension is missed. 

(2) Paragraph 1.1(c), Detection test 

matrix—Flaws smaller than the 50 

percent of allowable flaw size, as 

defined in IWB–3500, need not be 

included as detection flaws. For 

procedures applied from the inside 

surface, use the minimum thickness 

specified in the scope of the procedure 

to calculate a/t. For procedures applied 

from the outside surface, the actual 

thickness of the test specimen is to be 

used to calculate a/t. 

(C) Specimen set and qualification: 

Third provision. When applying 

Supplement 4 to Appendix VIII, the 

following conditions must be used: 

(1) A depth sizing requirement of 0.15 

inch RMS must be used in lieu of the 

requirements in Subparagraphs 3.2(a) 

and 3.2(c), and a length sizing 

requirement of 0.75 inch RMS must be 

used in lieu of the requirement in 

Subparagraph 3.2(b). 

(2) In lieu of the location acceptance 

criteria requirements of Subparagraph 

2.1(b), a flaw will be considered 

detected when reported within 1.0 inch 

or 10 percent of the metal path to the 

flaw, whichever is greater, of its true 

location in the X and Y directions. 

(3) In lieu of the flaw type 

requirements of Subparagraph 1.1(e)(1), 

a minimum of 70 percent of the flaws 

in the detection and sizing tests must be 

cracks. Notches, if used, must be limited 

by the following: 

(i) Notches must be limited to the case 

where examinations are performed from 

the clad surface. 

(ii) Notches must be semielliptical 

with a tip width of less than or equal to 

0.010 inches. 

(iii) Notches must be perpendicular to 

the surface within ±2 degrees. 

(4) In lieu of the detection test matrix 

requirements in paragraphs 1.1(e)(2) and 

1.1(e)(3), personnel demonstration test 

sets must contain a representative 

distribution of flaw orientations, sizes, 

and locations. 

(D) Specimen set and qualification: 

Fourth provision. The following 

conditions must be used in addition to 

the requirements of Supplement 6 to 

Appendix VIII: 

(1) Paragraph 3.1, Detection 

Acceptance Criteria—Personnel are 

qualified for detection if: 

(i) No surface connected flaw greater 

than 0.25 inch through-wall has been 

missed. 

(ii) No embedded flaw greater than 

0.50 inch through-wall has been missed. 

(2) Paragraph 3.1, Detection 

Acceptance Criteria—For procedure 

qualification, all flaws within the scope 

of the procedure are detected. 

(3) Paragraph 1.1(b) for detection and 

sizing test flaws and locations—Flaws 

smaller than the 50 percent of allowable 

flaw size, as defined in IWB–3500, need 

not be included as detection flaws. 

Flaws that are less than the allowable 

flaw size, as defined in IWB–3500, may 

be used as detection and sizing flaws. 

(4) Notches are not permitted. 

(E) Specimen set and qualification: 

Fifth provision. When applying 

Supplement 6 to Appendix VIII, the 

following conditions must be used: 

(1) A depth sizing requirement of 0.25 

inch RMS must be used in lieu of the 

requirements of subparagraphs 3.2(a), 

3.2(c)(2), and 3.2(c)(3). 

(2) In lieu of the location acceptance 

criteria requirements in Subparagraph 

2.1(b), a flaw will be considered 

detected when reported within 1.0 inch 

or 10 percent of the metal path to the 

flaw, whichever is greater, of its true 

location in the X and Y directions. 

(3) In lieu of the length sizing criteria 

requirements of Subparagraph 3.2(b), a 

length sizing acceptance criteria of 0.75 

inch RMS must be used. 

(4) In lieu of the detection specimen 

requirements in Subparagraph 1.1(e)(1), 

a minimum of 55 percent of the flaws 

must be cracks. The remaining flaws 

may be cracks or fabrication type flaws, 

such as slag and lack of fusion. The use 

of notches is not allowed. 

(5) In lieu of paragraphs 1.1(e)(2) and 

1.1(e)(3) detection test matrix, personnel 

demonstration test sets must contain a 

representative distribution of flaw 

orientations, sizes, and locations. 

(F) Specimen set and qualification: 

Sixth provision. The following 

conditions may be used for personnel 

qualification for combined Supplement 

4 to Appendix VIII and Supplement 6 to 

Appendix VIII qualification. Licensees 

choosing to apply this combined 

qualification must apply all of the 

provisions of Supplements 4 and 6 

including the following conditions: 

(1) For detection and sizing, the total 

number of flaws must be at least 10. A 

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minimum of 5 flaws must be from 

Supplement 4, and a minimum of 50 

percent of the flaws must be from 

Supplement 6. At least 50 percent of the 

flaws in any sizing must be cracks. 

Notches are not acceptable for 

Supplement 6. 

(2) Examination personnel are 

qualified for detection and length sizing 

when the results of any combined 

performance demonstration satisfy the 

acceptance criteria of Supplement 4 to 

Appendix VIII. 

(3) Examination personnel are 

qualified for depth sizing when 

Supplement 4 to Appendix VIII and 

Supplement 6 to Appendix VIII flaws 

are sized within the respective 

acceptance criteria of those 

supplements. 

(G) Specimen set and qualification: 

Seventh provision. When applying 

Supplement 4 to Appendix VIII, 

Supplement 6 to Appendix VIII, or 

combined Supplement 4 and 

Supplement 6 qualification, the 

following additional conditions must be 

used, and examination coverage must 

include: 

(1) The clad-to-base-metal-interface, 

including a minimum of 15 percent T 

(measured from the clad-to-base-metalinterface), must be examined from four 

orthogonal directions using procedures 

and personnel qualified in accordance 

with Supplement 4 to Appendix VIII. 

(2) If the clad-to-base-metal-interface 

procedure demonstrates detectability of 

flaws with a tilt angle relative to the 

weld centerline of at least 45 degrees, 

the remainder of the examination 

volume is considered fully examined if 

coverage is obtained in one parallel and 

one perpendicular direction. This must 

be accomplished using a procedure and 

personnel qualified for single-side 

examination in accordance with 

Supplement 6. Subsequent 

examinations of this volume may be 

performed using examination 

techniques qualified for a tilt angle of at 

least 10 degrees. 

(3) The examination volume not 

addressed by paragraph (b)(2)(xv)(G)(1) 

of this section is considered fully 

examined if coverage is obtained in one 

parallel and one perpendicular 

direction, using a procedure and 

personnel qualified for single sided 

examination when the conditions in 

paragraph (b)(2)(xv)(G)(2) are met. 

(H) Specimen set and qualification: 

Eighth provision. When applying 

Supplement 5 to Appendix VIII, at least 

50 percent of the flaws in the 

demonstration test set must be cracks 

and the maximum misorientation must 

be demonstrated with cracks. Flaws in 

nozzles with bore diameters equal to or 

less than 4 inches may be notches. 

(I) Specimen set and qualification: 

Ninth provision. When applying 

Supplement 5, Paragraph (a), to 

Appendix VIII, the number of false calls 

allowed must be D/10, with a maximum 

of 3, where D is the diameter of the 

nozzle. 

(J) [Reserved] 

(K) Specimen set and qualification: 

Eleventh provision. When performing 

nozzle-to-vessel weld examinations, the 

following conditions must be used 

when the requirements contained in 

Supplement 7 to Appendix VIII are 

applied for nozzle-to-vessel welds in 

conjunction with Supplement 4 to 

Appendix VIII, Supplement 6 to 

Appendix VIII, or combined 

Supplement 4 and Supplement 6 

qualification. 

(1) For examination of nozzle-tovessel welds conducted from the bore, 

the following conditions are required to 

qualify the procedures, equipment, and 

personnel: 

(i) For detection, a minimum of four 

flaws in one or more full-scale nozzle 

mock-ups must be added to the test set. 

The specimens must comply with 

Supplement 6, paragraph 1.1, to 

Appendix VIII, except for flaw locations 

specified in Table VIII S6–1. Flaws may 

be notches, fabrication flaws, or cracks. 

Seventy-five (75) percent of the flaws 

must be cracks or fabrication flaws. 

Flaw locations and orientations must be 

selected from the choices shown in 

paragraph (b)(2)(xv)(K)(4) of this 

section, Table VIII–S7–1—Modified, 

with the exception that flaws in the 

outer eighty-five (85) percent of the 

weld need not be perpendicular to the 

weld. There may be no more than two 

flaws from each category, and at least 

one subsurface flaw must be included. 

(ii) For length sizing, a minimum of 

four flaws as in paragraph 

(b)(2)(xv)(K)(1)(i) of this section must be 

included in the test set. The length 

sizing results must be added to the 

results of combined Supplement 4 to 

Appendix VIII and Supplement 6 to 

Appendix VIII. The combined results 

must meet the acceptance standards 

contained in paragraph (b)(2)(xv)(E)(3) 

of this section. 

(iii) For depth sizing, a minimum of 

four flaws as in paragraph 

(b)(2)(xv)(K)(1)(i) of this section must be 

included in the test set. Their depths 

must be distributed over the ranges of 

Supplement 4, Paragraph 1.1, to 

Appendix VIII, for the inner 15 percent 

of the wall thickness and Supplement 6, 

Paragraph 1.1, to Appendix VIII, for the 

remainder of the wall thickness. The 

depth sizing results must be combined 

with the sizing results from Supplement 

4 to Appendix VIII for the inner 15 

percent and to Supplement 6 to 

Appendix VIII for the remainder of the 

wall thickness. The combined results 

must meet the depth sizing acceptance 

criteria contained in paragraphs 

(b)(2)(xv)(C)(1), (b)(2)(xv)(E)(1), and 

(b)(2)(xv)(F)(3) of this section. 

(2) For examination of reactor 

pressure vessel nozzle-to-vessel welds 

conducted from the inside of the vessel, 

the following conditions are required: 

(i) The clad-to-base-metal-interface 

and the adjacent examination volume to 

a minimum depth of 15 percent T 

(measured from the clad-to-base-metalinterface) must be examined from four 

orthogonal directions using a procedure 

and personnel qualified in accordance 

with Supplement 4 to Appendix VIII as 

conditioned by paragraphs (b)(2)(xv)(B) 

and (C) of this section. 

(ii) When the examination volume 

defined in paragraph (b)(2)(xv)(K)(2)(i) 

of this section cannot be effectively 

examined in all four directions, the 

examination must be augmented by 

examination from the nozzle bore using 

a procedure and personnel qualified in 

accordance with paragraph 

(b)(2)(xv)(K)(1) of this section. 

(iii) The remainder of the examination 

volume not covered by paragraph 

(b)(2)(xv)(K)(2)(ii) of this section or a 

combination of paragraphs 

(b)(2)(xv)(K)(2)(i) and (ii) of this section, 

must be examined from the nozzle bore 

using a procedure and personnel 

qualified in accordance with paragraph 

(b)(2)(xv)(K)(1) of this section, or from 

the vessel shell using a procedure and 

personnel qualified for single sided 

examination in accordance with 

Supplement 6 to Appendix VIII, as 

conditioned by paragraphs (b)(2)(xv)(D) 

through (G) of this section. 

(3) For examination of reactor 

pressure vessel nozzle-to-shell welds 

conducted from the outside of the 

vessel, the following conditions are 

required: 

(i) The clad-to-base-metal-interface 

and the adjacent metal to a depth of 15 

percent T (measured from the clad-tobase-metal-interface) must be examined 

from one radial and two opposing 

circumferential directions using a 

procedure and personnel qualified in 

accordance with Supplement 4 to 

Appendix VIII, as conditioned by 

paragraphs (b)(2)(xv)(B) and (C) of this 

section, for examinations performed in 

the radial direction, and Supplement 5 

to Appendix VIII, as conditioned by 

paragraph (b)(2)(xv)(J) of this section, for 

examinations performed in the 

circumferential direction. 

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(ii) The examination volume not 

addressed by paragraph 

(b)(2)(xv)(K)(3)(i) of this section must be 

examined in a minimum of one radial 

direction using a procedure and 

personnel qualified for single sided 

examination in accordance with 

Supplement 6 to Appendix VIII, as 

conditioned by paragraphs (b)(2)(xv)(D) 

through (G) of this section. 

(4) Table VIII–S7–1, ‘‘Flaw Locations 

and Orientations,’’ Supplement 7 to 

Appendix VIII, is conditioned as 

follows: 

TABLE VIII—S7–1—MODIFIED

[Flaw locations and orientations] 

Parallel 

to weld 

Perpendicular 

to weld 

Inner 15 percent ...................................................................................................................... X X 

Outside Diameter Surface ....................................................................................................... X ........................................

Subsurface ............................................................................................................................... X ........................................

(L) Specimen set and qualification: 

Twelfth provision. As a condition to the 

requirements of Supplement 8, 

Subparagraph 1.1(c), to Appendix VIII, 

notches may be located within one 

diameter of each end of the bolt or stud. 

(M) Specimen set and qualification: 

Thirteenth provision. When 

implementing Supplement 12 to 

Appendix VIII, only the provisions 

related to the coordinated 

implementation of Supplement 3 to 

Supplement 2 performance 

demonstrations are to be applied. 

(xvi) Section XI condition: Appendix 

VIII single side ferritic vessel and piping 

and stainless steel piping examinations. 

When applying editions and addenda 

prior to the 2007 Edition of Section XI, 

the following conditions apply. 

(A) Ferritic and stainless steel piping 

examinations: First provision. 

Examinations performed from one side 

of a ferritic vessel weld must be 

conducted with equipment, procedures, 

and personnel that have demonstrated 

proficiency with single side 

examinations. To demonstrate 

equivalency to two sided examinations, 

the demonstration must be performed to 

the requirements of Appendix VIII, as 

conditioned by this paragraph and 

paragraphs (b)(2)(xv)(B) through (G) of 

this section, on specimens containing 

flaws with non-optimum sound energy 

reflecting characteristics or flaws similar 

to those in the vessel being examined. 

(B) Ferritic and stainless steel piping 

examinations: Second provision. 

Examinations performed from one side 

of a ferritic or stainless steel pipe weld 

must be conducted with equipment, 

procedures, and personnel that have 

demonstrated proficiency with single 

side examinations. To demonstrate 

equivalency to two sided examinations, 

the demonstration must be performed to 

the requirements of Appendix VIII, as 

conditioned by this paragraph and 

paragraph (b)(2)(xv)(A) of this section. 

(xvii) Section XI condition: 

Reconciliation of quality requirements. 

When purchasing replacement items, in 

addition to the reconciliation provisions 

of IWA–4200, 1995 Addenda through 

1998 Edition, the replacement items 

must be purchased, to the extent 

necessary, in accordance with the 

licensee’s quality assurance program 

description required by 10 CFR 

50.34(b)(6)(ii). 

(xviii) Section XI condition: NDE 

personnel certification. (A) NDE 

personnel certification: First provision. 

Level I and II nondestructive 

examination personnel must be 

recertified on a 3-year interval in lieu of 

the 5-year interval specified in the 1997 

Addenda and 1998 Edition of IWA– 

2314, and IWA–2314(a) and IWA– 

2314(b) of the 1999 Addenda through 

the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1)(ii) of this section. 

(B) NDE personnel certification: 

Second provision. When applying 

editions and addenda prior to the 2007 

Edition of Section XI, paragraph IWA– 

2316 may only be used to qualify 

personnel that observe leakage during 

system leakage and hydrostatic tests 

conducted in accordance with IWA 

5211(a) and (b). 

(C) NDE personnel certification: Third 

provision. When applying editions and 

addenda prior to the 2005 Addenda of 

Section XI, licensee’s qualifying visual 

examination personnel for VT–3 visual 

examination under paragraph IWA– 

2317 of Section XI must demonstrate the 

proficiency of the training by 

administering an initial qualification 

examination and administering 

subsequent examinations on a 3-year 

interval. 

(xix) Section XI condition: 

Substitution of alternative methods. The 

provisions for substituting alternative 

examination methods, a combination of 

methods, or newly developed 

techniques in the 1997 Addenda of 

IWA–2240 must be applied when using 

the 1998 Edition through the 2004 

Edition of Section XI of the ASME BPV 

Code. The provisions in IWA–4520(c), 

1997 Addenda through the 2004 

Edition, allowing the substitution of 

alternative methods, a combination of 

methods, or newly developed 

techniques for the methods specified in 

the Construction Code, are not approved 

for use. The provisions in IWA– 

4520(b)(2) and IWA–4521 of the 2008 

Addenda through the latest edition and 

addenda incorporated by reference in 

paragraph (a)(1)(ii) of this section, 

allowing the substitution of ultrasonic 

examination for radiographic 

examination specified in the 

Construction Code, are not approved for 

use. 

(xx) Section XI condition: System 

leakage tests—(A) System leakage tests: 

First provision. When performing 

system leakage tests in accordance with 

IWA–5213(a), 1997 through 2002 

Addenda, the licensee must maintain a 

10-minute hold time after test pressure 

has been reached for Class 2 and Class 

3 components that are not in use during 

normal operating conditions. No hold 

time is required for the remaining Class 

2 and Class 3 components provided that 

the system has been in operation for at 

least 4 hours for insulated components 

or 10 minutes for uninsulated 

components. 

(B) System leakage tests: Second 

provision. The NDE provision in IWA– 

4540(a)(2) of the 2002 Addenda of 

Section XI must be applied when 

performing system leakage tests after 

repair and replacement activities 

performed by welding or brazing on a 

pressure retaining boundary using the 

2003 Addenda through the latest edition 

and addenda incorporated by reference 

in paragraph (a)(1)(ii) of this section. 

(xxi) Section XI condition: Table IWB– 

2500–1 examination requirements. (A) 

Table IWB–2500–1 examination 

requirements: First provision. The 

provisions of Table IWB 2500–1, 

Examination Category B–D, Full 

Penetration Welded Nozzles in Vessels, 

Items B3.40 and B3.60 (Inspection 

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Program A) and Items B3.120 and 

B3.140 (Inspection Program B) of the 

1998 Edition must be applied when 

using the 1999 Addenda through the 

latest edition and addenda incorporated 

by reference in paragraph (a)(1)(ii) of 

this section. A visual examination with 

magnification that has a resolution 

sensitivity to detect a 1-mil width wire 

or crack, utilizing the allowable flaw 

length criteria in Table IWB–3512–1, 

1997 Addenda through the latest edition 

and addenda incorporated by reference 

in paragraph (a)(1)(ii) of this section, 

with a limiting assumption on the flaw 

aspect ratio (i.e., a/l = 0.5), may be 

performed instead of an ultrasonic 

examination. 

(B) [Reserved] 

(xxii) Section XI condition: Surface 

examination. The use of the provision 

in IWA–2220, ‘‘Surface Examination,’’ 

of Section XI, 2001 Edition through the 

latest edition and addenda incorporated 

by reference in paragraph (a)(1)(ii) of 

this section, that allows use of an 

ultrasonic examination method is 

prohibited. 

(xxiii) Section XI condition: 

Evaluation of thermally cut surfaces. 

The use of the provisions for 

eliminating mechanical processing of 

thermally cut surfaces in IWA–4461.4.2 

of Section XI, 2001 Edition through the 

latest edition and addenda incorporated 

by reference in paragraph (a)(1)(ii) of 

this section, is prohibited. 

(xxiv) Section XI condition: 

Incorporation of the performance 

demonstration initiative and addition of 

ultrasonic examination criteria. The use 

of Appendix VIII and the supplements 

to Appendix VIII and Article I–3000 of 

Section XI of the ASME BPV Code, 2002 

Addenda through the 2006 Addenda, is 

prohibited. 

(xxv) Section XI condition: Mitigation 

of defects by modification. The use of 

the provisions in IWA–4340, 

‘‘Mitigation of Defects by Modification,’’ 

Section XI, 2001 Edition through the 

latest edition and addenda incorporated 

by reference in paragraph (a)(1)(ii) of 

this section are prohibited. 

(xxvi) Section XI condition: Pressure 

testing Class 1, 2 and 3 mechanical 

joints. The repair and replacement 

activity provisions in IWA–4540(c) of 

the 1998 Edition of Section XI for 

pressure testing Class 1, 2, and 3 

mechanical joints must be applied when 

using the 2001 Edition through the 

latest edition and addenda incorporated 

by reference in paragraph (a)(1)(ii) of 

this section. 

(xxvii) Section XI condition: Removal 

of insulation. When performing visual 

examination in accordance with IWA– 

5242 of Section XI of the ASME BPV 

Code, 2003 Addenda through the 2006 

Addenda, or IWA–5241 of the 2007 

Edition through the latest edition and 

addenda incorporated by reference in 

paragraph (a)(1)(ii) of this section, 

insulation must be removed from 17–4 

PH or 410 stainless steel studs or bolts 

aged at a temperature below 1100 °F or 

having a Rockwell Method C hardness 

value above 30, and from A–286 

stainless steel studs or bolts preloaded 

to 100,000 pounds per square inch or 

higher. 

(xxviii) Section XI condition: Analysis 

of flaws. Licensees using ASME BPV 

Code, Section XI, Appendix A, must use 

the following conditions when 

implementing Equation (2) in A– 

4300(b)(1): 

For R < 0, DKI depends on the crack depth 

(a), and the flow stress (sf). The flow stress 

is defined by sf = 1/2(sys + sult), where sys

is the yield strength and sult is the ultimate 

tensile strength in units ksi (MPa) and (a) is 

in units in. (mm). For ¥2 ≤ R ≤ 0 and Kmax¥ 

Kmin ≤ 0.8 × 1.12 sf√(pa), S = 1 and DKI = 

Kmax. For R < ¥2 and Kmax¥ Kmin ≤ 0.8 × 1.12 

sf√(pa), S = 1 and DKI = (1 ¥ R) Kmax/3. For 

R < 0 and Kmax ¥ Kmin > 0.8 × 1.12 sf√(pa), 

S = 1 and DKI = Kmax¥Kmin. 

(xxix) Section XI condition: 

Nonmandatory Appendix R. 

Nonmandatory Appendix R, ‘‘RiskInformed Inspection Requirements for 

Piping,’’ of Section XI, 2005 Addenda 

through the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1)(ii) of this section, may not be 

implemented without prior NRC 

authorization of the proposed 

alternative in accordance with 

paragraph (z) of this section. 

(3) Conditions on ASME OM Code. As 

used in this section, references to the 

OM Code refer to the ASME Code for 

Operation and Maintenance of Nuclear 

Power Plants, Subsections ISTA, ISTB, 

ISTC, ISTD, Mandatory Appendices I 

and II, and Nonmandatory Appendices 

A through H and J, including the 1995 

Edition through the 2006 Addenda, 

subject to the following conditions: 

(i) OM condition: Quality assurance. 

When applying editions and addenda of 

the OM Code, the requirements of 

NQA–1, ‘‘Quality Assurance 

Requirements for Nuclear Facilities,’’ 

1979 Addenda, are acceptable as 

permitted by ISTA 1.4 of the 1995 

Edition through 1997 Addenda or 

ISTA–1500 of the 1998 Edition through 

the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1)(iv) of this section, provided the 

licensee uses its 10 CFR part 50, 

Appendix B, quality assurance program 

in conjunction with the OM Code 

requirements. Commitments contained 

in the licensee’s quality assurance 

program description that are more 

stringent than those contained in NQA– 

1 govern OM Code activities. If NQA– 

1 and the OM Code do not address the 

commitments contained in the 

licensee’s Appendix B quality assurance 

program description, the commitments 

must be applied to OM Code activities. 

(ii) OM condition: Motor-Operated 

Valve (MOV) testing. Licensees must 

comply with the provisions for MOV 

testing in OM Code ISTC 4.2, 1995 

Edition with the 1996 and 1997 

Addenda, or ISTC–3500, 1998 Edition 

through the latest edition and addenda 

incorporated by reference in paragraph 

(a)(1)(iv) of this section, and must 

establish a program to ensure that 

motor-operated valves continue to be 

capable of performing their design basis 

safety functions. 

(iii) [Reserved] 

(iv) OM condition: Check valves 

(Appendix II). Licensees applying 

Appendix II, ‘‘Check Valve Condition 

Monitoring Program,’’ of the OM Code, 

1995 Edition with the 1996 and 1997 

Addenda, must satisfy the requirements 

of (b)(3)(iv)(A) through (C) of this 

section. Licensees applying Appendix 

II, 1998 Edition through the 2002 

Addenda, must satisfy the requirements 

of (b)(3)(iv)(A), (B), and (D) of this 

section. 

(A) Check valves: First provision. 

Valve opening and closing functions 

must be demonstrated when flow testing 

or examination methods (nonintrusive, 

or disassembly and inspection) are used; 

(B) Check valves: Second provision. 

The initial interval for tests and 

associated examinations may not exceed 

two fuel cycles or 3 years, whichever is 

longer; any extension of this interval 

may not exceed one fuel cycle per 

extension with the maximum interval 

not to exceed 10 years. Trending and 

evaluation of existing data must be used 

to reduce or extend the time interval 

between tests. 

(C) Check valves: Third provision. If 

the Appendix II condition monitoring 

program is discontinued, then the 

requirements of ISTC 4.5.1 through 4.5.4 

must be implemented. 

(D) Check valves: Fourth provision. 

The applicable provisions of subsection 

ISTC must be implemented if the 

Appendix II condition monitoring 

program is discontinued. 

(v) OM condition: Snubbers ISTD. 

Article IWF–5000, ‘‘Inservice Inspection 

Requirements for Snubbers,’’ of the 

ASME BPV Code, Section XI, must be 

used when performing inservice 

inspection examinations and tests of 

snubbers at nuclear power plants, 

except as conditioned in paragraphs 

(b)(3)(v)(A) and (B) of this section. 

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(A) Snubbers: First provision. 

Licensees may use Subsection ISTD, 

‘‘Preservice and Inservice Examination 

and Testing of Dynamic Restraints 

(Snubbers) in Light-Water Reactor 

Power Plants,’’ ASME OM Code, 1995 

Edition through the latest edition and 

addenda incorporated by reference in 

paragraph (a)(1)(iv) of this section, in 

place of the requirements for snubbers 

in the editions and addenda up to the 

2005 Addenda of the ASME BPV Code, 

Section XI, IWF–5200(a) and (b) and 

IWF–5300(a) and (b), by making 

appropriate changes to their technical 

specifications or licensee-controlled 

documents. Preservice and inservice 

examinations must be performed using 

the VT–3 visual examination method 

described in IWA–2213. 

(B) Snubbers: Second provision. 

Licensees must comply with the 

provisions for examining and testing 

snubbers in Subsection ISTD of the 

ASME OM Code and make appropriate 

changes to their technical specifications 

or licensee-controlled documents when 

using the 2006 Addenda and later 

editions and addenda of Section XI of 

the ASME BPV Code. 

(vi) OM condition: Exercise interval 

for manual valves. Manual valves must 

be exercised on a 2-year interval rather 

than the 5-year interval specified in 

paragraph ISTC–3540 of the 1999 

through the 2005 Addenda of the ASME 

OM Code, provided that adverse 

conditions do not require more frequent 

testing. 

(4) Conditions on Design, Fabrication, 

and Materials Code Cases. Each 

manufacturing license, standard design 

approval, and design certification 

application under part 52 of this chapter 

is subject to the following conditions. 

Licensees may apply the ASME BPV 

Code Cases listed in NRC Regulatory 

Guide 1.84, Revision 36, without prior 

NRC approval, subject to the following 

conditions: 

(i) Design, Fabrication, and Materials 

Code Case condition: Applying Code 

Cases. When an applicant or licensee 

initially applies a listed Code Case, the 

applicant or licensee must apply the 

most recent version of that Code Case 

incorporated by reference in paragraph 

(a) of this section. 

(ii) Design, Fabrication, and Materials 

Code Case condition: Applying different 

revisions of Code Cases. If an applicant 

or licensee has previously applied a 

Code Case and a later version of the 

Code Case is incorporated by reference 

in paragraph (a) of this section, the 

applicant or licensee may continue to 

apply the previous version of the Code 

Case as authorized or may apply the 

later version of the Code Case, including 

any NRC-specified conditions placed on 

its use, until it updates its Code of 

Record for the component being 

constructed. 

(iii) Design, Fabrication, and 

Materials Code Case condition: 

Applying annulled Code Cases. 

Application of an annulled Code Case is 

prohibited unless an applicant or 

licensee applied the listed Code Case 

prior to it being listed as annulled in 

Regulatory Guide 1.84. If an applicant or 

licensee has applied a listed Code Case 

that is later listed as annulled in 

Regulatory Guide 1.84, the applicant or 

licensee may continue to apply the Code 

Case until it updates its Code of Record 

for the component being constructed. 

(5) Conditions on inservice inspection 

Code Cases. Licensees may apply the 

ASME BPV Code Cases listed in 

Regulatory Guide 1.147, Revision 17, 

without prior NRC approval, subject to 

the following: 

(i) ISI Code Case condition: Applying 

Code Cases. When a licensee initially 

applies a listed Code Case, the licensee 

must apply the most recent version of 

that Code Case incorporated by 

reference in paragraph (a) of this 

section. 

(ii) ISI Code Case condition: Applying 

different revisions of Code Cases. If a 

licensee has previously applied a Code 

Case and a later version of the Code 

Case is incorporated by reference in 

paragraph (a) of this section, the 

licensee may continue to apply, to the 

end of the current 120-month interval, 

the previous version of the Code Case, 

as authorized, or may apply the later 

version of the Code Case, including any 

NRC-specified conditions placed on its 

use. Licensees who choose to continue 

use of the Code Case during subsequent 

120-month ISI program intervals will be 

required to implement the latest version 

incorporated by reference into 10 CFR 

50.55a as listed in Tables 1 and 2 of 

Regulatory Guide 1.147, Revision 17. 

(iii) ISI Code Case condition: 

Applying annulled Code Cases. 

Application of an annulled Code Case is 

prohibited unless a licensee previously 

applied the listed Code Case prior to it 

being listed as annulled in Regulatory 

Guide 1.147. If a licensee has applied a 

listed Code Case that is later listed as 

annulled in Regulatory Guide 1.147, the 

licensee may continue to apply the Code 

Case to the end of the current 120- 

month interval. 

(6) Conditions on Operation and 

Maintenance of Nuclear Power Plants 

Code Cases. Licensees may apply the 

ASME Operation and Maintenance Code 

Cases listed in Regulatory Guide 1.192, 

Revision 1, without prior NRC approval, 

subject to the following: 

(i) OM Code Case condition: Applying 

Code Cases. When a licensee initially 

applies a listed Code Case, the licensee 

must apply the most recent version of 

that Code Case incorporated by 

reference in paragraph (a) of this 

section. 

(ii) OM Code Case condition: 

Applying different revisions of Code 

Cases. If a licensee has previously 

applied a Code Case and a later version 

of the Code Case is incorporated by 

reference in paragraph (a) of this 

section, the licensee may continue to 

apply, to the end of the current 120- 

month interval, the previous version of 

the Code Case, as authorized, or may 

apply the later version of the Code Case, 

including any NRC-specified conditions 

placed on its use. Licensees who choose 

to continue use of the Code Case during 

subsequent 120-month ISI program 

intervals will be required to implement 

the latest version incorporated by 

reference into 10 CFR 50.55a as listed in 

Tables 1 and 2 of Regulatory Guide 

1.192, Revision 1. 

(iii) OM Code Case condition: 

Applying annulled Code Cases. 

Application of an annulled Code Case is 

prohibited unless a licensee previously 

applied the listed Code Case prior to it 

being listed as annulled in Regulatory 

Guide 1.192. If a licensee has applied a 

listed Code Case that is later listed as 

annulled in Regulatory Guide 1.192, the 

licensee may continue to apply the Code 

Case to the end of the current 120- 

month interval. 

(c) Reactor coolant pressure 

boundary. Systems and components of 

boiling and pressurized water-cooled 

nuclear power reactors must meet the 

requirements of the ASME BPV Code as 

specified in this paragraph. Each 

manufacturing license, standard design 

approval, and design certification 

application under part 52 of this chapter 

and each combined license for a 

utilization facility is subject to the 

following conditions: 

(1) Standards requirement for reactor 

coolant pressure boundary components. 

Components that are part of the reactor 

coolant pressure boundary must meet 

the requirements for Class 1 

components in Section III 1,4 of the 

ASME BPV Code, except as provided in 

paragraphs (c)(2) through (4) of this 

section. 

(2) Exceptions to reactor coolant 

pressure boundary standards 

requirement. Components that are 

connected to the reactor coolant system 

and are part of the reactor coolant 

pressure boundary as defined in § 50.2 

need not meet the requirements of 

paragraph (c)(1) of this section, 

provided that: 

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(i) Exceptions: Shutdown and cooling 

capability. In the event of postulated 

failure of the component during normal 

reactor operation, the reactor can be 

shut down and cooled down in an 

orderly manner, assuming makeup is 

provided by the reactor coolant makeup 

system; or 

(ii) Exceptions: Isolation capability. 

The component is or can be isolated 

from the reactor coolant system by two 

valves in series (both closed, both open, 

or one closed and the other open). Each 

open valve must be capable of automatic 

actuation and, assuming the other valve 

is open, its closure time must be such 

that, in the event of postulated failure of 

the component during normal reactor 

operation, each valve remains operable 

and the reactor can be shut down and 

cooled down in an orderly manner, 

assuming makeup is provided by the 

reactor coolant makeup system only. 

(3) Applicable Code and Code Cases 

and conditions on their use. The Code 

edition, addenda, and optional ASME 

Code Cases to be applied to components 

of the reactor coolant pressure boundary 

must be determined by the provisions of 

paragraph NCA–1140, Subsection NCA 

of Section III of the ASME BPV Code, 

subject to the following conditions: 

(i) Reactor coolant pressure boundary 

condition: Code edition and addenda. 

The edition and addenda applied to a 

component must be those that are 

incorporated by reference in paragraph 

(a)(1)(i) of this section; 

(ii) Reactor coolant pressure boundary 

condition: Earliest edition and addenda 

for pressure vessel. The ASME Code 

provisions applied to the pressure 

vessel may be dated no earlier than the 

summer 1972 Addenda of the 1971 

Edition; 

(iii) Reactor coolant pressure 

boundary condition: Earliest edition and 

addenda for piping, pumps, and valves. 

The ASME Code provisions applied to 

piping, pumps, and valves may be dated 

no earlier than the Winter 1972 

Addenda of the 1971 Edition; and 

(iv) Reactor coolant pressure 

boundary condition: Use of Code Cases. 

The optional Code Cases applied to a 

component must be those listed in NRC 

Regulatory Guide 1.84 that is 

incorporated by reference in paragraph 

(a)(3)(i) of this section. 

(4) Standards requirement for 

components in older plants. For a 

nuclear power plant whose construction 

permit was issued prior to May 14, 

1984, the applicable Code edition and 

addenda for a component of the reactor 

coolant pressure boundary continue to 

be that Code edition and addenda that 

were required by Commission 

regulations for such a component at the 

time of issuance of the construction 

permit. 

(d) Quality Group B components. 

Systems and components of boiling and 

pressurized water-cooled nuclear power 

reactors must meet the requirements of 

the ASME BPV Code as specified in this 

paragraph. Each manufacturing license, 

standard design approval, and design 

certification application under part 52 

of this chapter, and each combined 

license for a utilization facility is subject 

to the following conditions: 

(1) Standards requirement for Quality 

Group B components. 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, 

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 B7 must meet the requirements 

for Class 2 Components in Section III of 

the ASME BPV Code. 

(2) Quality Group B: Applicable Code 

and Code Cases and conditions on their 

use. The Code edition, addenda, and 

optional ASME Code Cases to be 

applied to the systems and components 

identified in paragraph (d)(1) of this 

section must be determined by the rules 

of paragraph NCA–1140, Subsection 

NCA of Section III of the ASME BPV 

Code, subject to the following 

conditions: 

(i) Quality Group B condition: Code 

edition and addenda. The edition and 

addenda must be those that are 

incorporated by reference in paragraph 

(a)(1)(i) of this section; 

(ii) Quality Group B condition: 

Earliest edition and addenda for 

components. The ASME Code 

provisions applied to the systems and 

components may be dated no earlier 

than the 1980 Edition; and 

(iii) Quality Group B condition: Use of 

Code Cases. The optional Code Cases 

must be those listed in NRC Regulatory 

Guide 1.84 that is incorporated by 

reference in paragraph (a)(3)(i) of this 

section. 

(e) Quality Group C components. 

Systems and components of boiling and 

pressurized water-cooled nuclear power 

reactors must meet the requirements of 

the ASME BPV Code as specified in this 

paragraph. Each manufacturing license, 

standard design approval, and design 

certification application under part 52 

of this chapter and each combined 

license for a utilization facility is subject 

to the following conditions. 

(1) Standards requirement for Quality 

Group C components. 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, 

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 BPV Code. 

(2) Quality Group C applicable Code 

and Code Cases and conditions on their 

use. The Code edition, addenda, and 

optional ASME Code Cases to be 

applied to the systems and components 

identified in paragraph (e)(1) of this 

section must be determined by the rules 

of paragraph NCA–1140, subsection 

NCA of Section III of the ASME BPV 

Code, subject to the following 

conditions: 

(i) Quality Group C condition: Code 

edition and addenda. The edition and 

addenda must be those incorporated by 

reference in paragraph (a)(1)(i) of this 

section; 

(ii) Quality Group C condition: 

Earliest edition and addenda for 

components. The ASME Code 

provisions applied to the systems and 

components may be dated no earlier 

than the 1980 Edition; and 

(iii) Quality Group C condition: Use of 

Code Cases. The optional Code Cases 

must be those listed in NRC Regulatory 

Guide 1.84 that is incorporated by 

reference in paragraph (a)(3)(i) of this 

section. 

(f) Inservice testing requirements. 

Systems and components of boiling and 

pressurized water-cooled nuclear power 

reactors must meet the requirements of 

the ASME BPV Code and ASME Code 

for Operation and Maintenance of 

Nuclear Power Plants as specified in 

this paragraph. Each operating license 

for a boiling or pressurized water-cooled 

nuclear facility is subject to the 

following conditions. Each combined 

license for a boiling or pressurized 

water-cooled nuclear facility is subject 

to the following conditions, but the 

conditions in paragraphs (f)(4) through 

(6) of this section must be met only after 

the Commission makes the finding 

under § 52.103(g) of this chapter. 

Requirements for inservice inspection of 

Class 1, Class 2, Class 3, Class MC, and 

Class CC components (including their 

supports) are located in § 50.55a(g). 

(1) Inservice testing requirements for 

older plants (pre-1971 CPs). For a 

boiling or pressurized water-cooled 

nuclear power facility whose 

construction permit was issued prior to 

January 1, 1971, pumps and valves must 

meet the test requirements of paragraphs 

(f)(4) and (5) of this section to the extent 

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practical. Pumps and valves that are 

part of the reactor coolant pressure 

boundary must meet the requirements 

applicable to components that are 

classified as ASME Code Class 1. Other 

pumps and valves that perform a 

function to shut down the reactor or 

maintain the reactor in a safe shutdown 

condition, mitigate the consequences of 

an accident, or provide overpressure 

protection for safety-related systems (in 

meeting the requirements of the 1986 

Edition, or later, of the BPV or OM 

Code) must meet the test requirements 

applicable to components that are 

classified as ASME Code Class 2 or 

Class 3. 

(2) Design and accessibility 

requirements for performing inservice 

testing in plants with CPs issued 

between 1971 and 1974. For a boiling or 

pressurized water-cooled nuclear power 

facility whose construction permit was 

issued on or after January 1, 1971, but 

before July 1, 1974, pumps and valves 

that are classified as ASME Code Class 

1 and Class 2 must be designed and 

provided with access to enable the 

performance of inservice tests for 

operational readiness set forth in 

editions and addenda of Section XI of 

the ASME BPV incorporated by 

reference in paragraph (a)(1)(ii) of this 

section (or the optional ASME Code 

Cases listed in NRC Regulatory Guide 

1.147, Revision 17, or Regulatory Guide 

1.192, Revision 1, that are incorporated 

by reference in paragraphs (a)(3)(ii) and 

(iii) of this section, respectively) in 

effect 6 months before the date of 

issuance of the construction permit. The 

pumps and valves may meet the 

inservice test requirements set forth in 

subsequent editions of this Code and 

addenda that are incorporated by 

reference in paragraph (a)(1)(ii) of this 

section (or the optional ASME Code 

Cases listed in NRC Regulatory Guide 

1.147, Revision 17; or Regulatory Guide 

1.192, Revision 1, that are incorporated 

by reference in paragraphs (a)(3)(ii) and 

(iii) of this section, respectively), subject 

to the applicable conditions listed 

therein. 

(3) Design and accessibility 

requirements for performing inservice 

testing in plants with CPs issued after 

1974. 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: 

(i)–(ii) [Reserved] 

(iii) IST design and accessibility 

requirements: Class 1 pumps and 

valves. (A) Class 1 pumps and valves: 

First provision. In facilities whose 

construction permit was issued before 

November 22, 1999, pumps and valves 

that are classified as ASME Code Class 

1 must be designed and 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 BPV 

Code incorporated by reference in 

paragraph (a)(1)(ii) of this section (or the 

optional ASME Code Cases listed in 

NRC Regulatory Guide 1.147, Revision 

17, or Regulatory Guide 1.192, Revision 

1, that are incorporated by reference in 

paragraphs (a)(3)(ii) and (iii) of this 

section, respectively) applied to the 

construction of the particular pump or 

valve or the summer 1973 Addenda, 

whichever is later. 

(B) Class 1 pumps and valves: Second 

provision. In facilities whose 

construction permit under this part, or 

design certification, design approval, 

combined license, or manufacturing 

license under part 52 of this chapter, 

issued on or after November 22, 1999, 

pumps and valves that are classified as 

ASME Code Class 1 must be designed 

and 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 NRC Regulatory Guide 1.192, 

Revision 1, that are incorporated by 

reference in paragraph (a)(3)(iii) of this 

section), incorporated by reference in 

paragraph (a)(1)(iv) of this section at the 

time the construction permit, combined 

license, manufacturing license, design 

certification, or design approval is 

issued. 

(iv) IST design and accessibility 

requirements: Class 2 and 3 pumps and 

valves. (A) Class 2 and 3 pumps and 

valves: First provision. In facilities 

whose construction permit was issued 

before November 22, 1999, pumps and 

valves that are classified as ASME Code 

Class 2 and Class 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 the 

editions and addenda of Section XI of 

the ASME BPV Code incorporated by 

reference in paragraph (a)(1)(ii) of this 

section (or the optional ASME Code 

Cases listed in NRC Regulatory Guide 

1.147, Revision 17, that are incorporated 

by reference in paragraph (a)(3)(ii) of 

this section) applied to the construction 

of the particular pump or valve or the 

Summer 1973 Addenda, whichever is 

later. 

(B) Class 2 and 3 pumps and valves: 

Second provision. In facilities whose 

construction permit under this part, or 

design certification, design approval, 

combined license, or manufacturing 

license under part 52 of this chapter, 

issued on or after November 22, 1999, 

pumps and valves that are classified as 

ASME Code Class 2 and 3 must be 

designed and 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 OM 

Code Cases listed in NRC Regulatory 

Guide 1.192, Revision 1, that are 

incorporated by reference in paragraph 

(a)(3)(iii) of this section), incorporated 

by reference in paragraph (a)(1)(iv) of 

this section at the time the construction 

permit, combined license, or design 

certification is issued. 

(v) IST design and accessibility 

requirements: Meeting later IST 

requirements. All pumps and valves 

may meet the test requirements set forth 

in subsequent editions of codes and 

addenda or portions thereof that are 

incorporated by reference in paragraph 

(a) of this section, subject to the 

conditions listed in paragraph (b) of this 

section. 

(4) Inservice testing standards 

requirement for operating plants. 

Throughout the service life of a boiling 

or pressurized water-cooled nuclear 

power facility, pumps and valves that 

are classified as ASME Code Class 1, 

Class 2, and Class 3 must meet the 

inservice test requirements (except 

design and access provisions) set forth 

in the ASME OM Code and addenda 

that become effective subsequent to 

editions and addenda specified in 

paragraphs (f)(2) and (3) of this section 

and that are incorporated by reference 

in paragraph (a)(1)(iv) of this section, to 

the extent practical within the 

limitations of design, geometry, and 

materials of construction of the 

components. 

(i) Applicable IST Code: Initial 120- 

month interval. 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 OM Code incorporated 

by reference in paragraph (a)(1)(iv) 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 1.192, Revision 

1, that is incorporated by reference in 

paragraph (a)(3)(iii) of this section, 

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subject to the conditions listed in 

paragraph (b) of this section). 

(ii) Applicable IST Code: Successive 

120-month intervals. Inservice tests to 

verify operational readiness of pumps 

and valves, whose function is required 

for safety, conducted during successive 

120-month intervals must comply with 

the requirements of the latest edition 

and addenda of the OM Code 

incorporated by reference in paragraph 

(a)(1)(iv) of this section 12 months 

before the start of the 120-month 

interval (or the optional ASME Code 

Cases listed in NRC Regulatory Guide 

1.147, Revision 17, or Regulatory Guide 

1.192, Revision 1, that are incorporated 

by reference in paragraphs (a)(3)(ii) and 

(iii) of this section, respectively), subject 

to the conditions listed in paragraph (b) 

of this section. 

(iii) [Reserved] 

(iv) Applicable IST Code: Use of later 

Code editions and addenda. Inservice 

tests of pumps and valves may meet the 

requirements set forth in subsequent 

editions and addenda that are 

incorporated by reference in paragraph 

(a)(1)(iv) of this section, subject to the 

conditions listed in paragraph (b) of this 

section, and subject to NRC approval. 

Portions of editions or addenda may be 

used, provided that all related 

requirements of the respective editions 

or addenda are met. 

(5) Requirements for updating IST 

programs—(i) IST program update: 

Applicable IST Code editions and 

addenda. The inservice test program for 

a boiling or pressurized water-cooled 

nuclear power facility must be revised 

by the licensee, as necessary, to meet 

the requirements of paragraph (f)(4) of 

this section. 

(ii) IST program update: Conflicting 

IST Code requirements with technical 

specifications. If a revised inservice test 

program for a facility conflicts with the 

technical specifications for the facility, 

the licensee must apply to the 

Commission for amendment of the 

technical specifications to conform the 

technical specifications to the revised 

program. The licensee must submit this 

application, as specified in § 50.4, at 

least 6 months before the start of the 

period during which the provisions 

become applicable, as determined by 

paragraph (f)(4) of this section. 

(iii) IST program update: Notification 

of impractical IST Code requirements. If 

the licensee has determined that 

conformance with certain Code 

requirements is impractical for its 

facility, the licensee must notify the 

Commission and submit, as specified in 

§ 50.4, information to support the 

determination. 

(iv) IST program update: Schedule for 

completing impracticality 

determinations. Where a pump or valve 

test requirement by the Code or addenda 

is determined to be impractical by the 

licensee and is not included in the 

revised inservice test program (as 

permitted by paragraph (f)(4) of this 

section), the basis for this determination 

must be submitted for NRC review and 

approval not later than 12 months after 

the expiration of the initial 120-month 

interval of operation from the start of 

facility commercial operation and each 

subsequent 120-month interval of 

operation during which the test is 

determined to be impractical. 

(6) Actions by the Commission for 

evaluating impractical and augmented 

IST Code requirements—(i) Impractical 

IST requirements: Granting of relief. The 

Commission will evaluate 

determinations under paragraph (f)(5) of 

this section that code requirements are 

impractical. The Commission may grant 

relief and may impose such alternative 

requirements as it determines are 

authorized by law, will not endanger 

life or property or the common defense 

and security, and are otherwise in the 

public interest, giving due consideration 

to the burden upon the licensee that 

could result if the requirements were 

imposed on the facility. 

(ii) Augmented IST requirements. The 

Commission may require the licensee to 

follow an augmented inservice test 

program for pumps and valves for 

which the Commission deems that 

added assurance of operational 

readiness is necessary. 

(g) Inservice inspection requirements. 

Systems and components of boiling and 

pressurized water-cooled nuclear power 

reactors must meet the requirements of 

the ASME BPV Code as specified in this 

paragraph. Each operating license for a 

boiling or pressurized water-cooled 

nuclear facility is subject to the 

following conditions. Each combined 

license for a boiling or pressurized 

water-cooled nuclear facility is subject 

to the following conditions, but the 

conditions in paragraphs (g)(4) through 

(6) of this section must be met only after 

the Commission makes the finding 

under § 52.103(g) of this chapter. 

Requirements for inservice testing of 

Class 1, Class 2, and Class 3 pumps and 

valves are located in § 50.55a(f). 

(1) Inservice inspection requirements 

for older plants (pre-1971 CPs). For a 

boiling or pressurized water-cooled 

nuclear power facility whose 

construction permit was issued before 

January 1, 1971, components (including 

supports) must meet the requirements of 

paragraphs (g)(4) and (g)(5) of this 

section to the extent practical. 

Components that are part of the reactor 

coolant pressure boundary and their 

supports must meet the requirements 

applicable to components that are 

classified as ASME Code Class 1. Other 

safety-related pressure vessels, piping, 

pumps and valves, and their supports 

must meet the requirements applicable 

to components that are classified as 

ASME Code Class 2 or Class 3. 

(2) Design and accessibility 

requirements for performing inservice 

inspection in plants with CPs issued 

between 1971 and 1974. For a boiling or 

pressurized water-cooled nuclear power 

facility whose construction permit was 

issued on or after January 1, 1971, but 

before July 1, 1974, components 

(including supports) that are classified 

as ASME Code Class 1 and Class 2 must 

be designed and be provided with 

access to enable the performance of 

inservice examination of such 

components (including supports) and 

must meet the preservice examination 

requirements set forth in editions and 

addenda of Section III or Section XI of 

the ASME BPV Code incorporated by 

reference in paragraph (a)(1) of this 

section (or the optional ASME Code 

Cases listed in NRC Regulatory Guide 

1.147, Revision 17, that are incorporated 

by reference in paragraph (a)(3)(ii) of 

this section) in effect 6 months before 

the date of issuance of the construction 

permit. The components (including 

supports) may meet the requirements set 

forth in subsequent editions and 

addenda of this Code that are 

incorporated by reference in paragraph 

(a) of this section (or the optional ASME 

Code Cases listed in NRC Regulatory 

Guide 1.147, Revision 17, that are 

incorporated by reference in paragraph 

(a)(3)(ii) of this section), subject to the 

applicable limitations and 

modifications. 

(3) Design and accessibility 

requirements for performing inservice 

inspection in plants with CPs issued 

after 1974. For a boiling or pressurized 

water-cooled nuclear power facility, 

whose construction permit under this 

part, or design certification, design 

approval, combined license, or 

manufacturing license under part 52 of 

this chapter, was issued on or after July 

1, 1974, the following are required: 

(i) ISI design and accessibility 

requirements: Class 1 components and 

supports. Components (including 

supports) that are classified as ASME 

Code Class 1 must be designed and be 

provided with access to enable the 

performance of inservice examination of 

these components and must meet the 

preservice examination requirements set 

forth in the editions and addenda of 

Section III or Section XI of the ASME 

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BPV Code incorporated by reference in 

paragraph (a)(1) of this section (or the 

optional ASME Code Cases listed in 

NRC Regulatory Guide 1.147, Revision 

17, that are incorporated by reference in 

paragraph (a)(3)(ii) of this section) 

applied to the construction of the 

particular component. 

(ii) ISI design and accessibility 

requirements: Class 2 and 3 components 

and supports. Components that are 

classified as ASME Code Class 2 and 

Class 3 and supports for components 

that are classified as ASME Code Class 

1, Class 2, and Class 3 must be designed 

and provided with access to enable the 

performance of inservice examination of 

these components and must meet the 

preservice examination requirements set 

forth in the editions and addenda of 

Section XI of the ASME BPV Code 

incorporated by reference in paragraph 

(a)(1)(ii) of this section (or the optional 

ASME Code Cases listed in NRC 

Regulatory Guide 1.147, Revision 17, 

that are incorporated by reference in 

paragraph (a)(3)(ii) of this section) 

applied to the construction of the 

particular component. 

(iii)–(iv) [Reserved] 

(v) ISI design and accessibility 

requirements: Meeting later ISI 

requirements. All components 

(including supports) may meet the 

requirements set forth in subsequent 

editions of codes and addenda or 

portions thereof that are incorporated by 

reference in paragraph (a) of this 

section, subject to the conditions listed 

therein. 

(4) Inservice inspection standards 

requirement for operating plants. 

Throughout the service life of a boiling 

or pressurized water-cooled nuclear 

power facility, components (including 

supports) that are classified as ASME 

Code Class 1, Class 2, and Class 3 must 

meet the requirements, except design 

and access provisions and preservice 

examination requirements, set forth in 

Section XI of editions and addenda of 

the ASME BPV Code (or ASME OM 

Code for snubber examination and 

testing) that become effective 

subsequent to editions specified in 

paragraphs (g)(2) and (3) of this section 

and that are incorporated by reference 

in paragraph (a)(1)(ii) or (iv) for snubber 

examination and testing of this section, 

to the extent practical within the 

limitations of design, geometry, and 

materials of construction of the 

components. Components that are 

classified as Class MC pressure retaining 

components and their integral 

attachments, and components that are 

classified as Class CC pressure retaining 

components and their integral 

attachments, must meet the 

requirements, except design and access 

provisions and preservice examination 

requirements, set forth in Section XI of 

the ASME BPV Code and addenda that 

are incorporated by reference in 

paragraph (a)(1)(ii) of this section, 

subject to the condition listed in 

paragraph (b)(2)(vi) of this section and 

the conditions listed in paragraphs 

(b)(2)(viii) and (ix) of this section, to the 

extent practical within the limitation of 

design, geometry, and materials of 

construction of the components. 

(i) Applicable ISI Code: Initial 120- 

month interval. Inservice examination 

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 

(a) 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, Revision 

17, when using Section XI, or 

Regulatory Guide 1.192, Revision 1, 

when using the OM Code, that are 

incorporated by reference in paragraphs 

(a)(3)(ii) and (iii) of this section, 

respectively), subject to the conditions 

listed in paragraph (b) of this section. 

(ii) Applicable ISI Code: Successive 

120-month intervals. Inservice 

examination of components and system 

pressure tests conducted during 

successive 120-month inspection 

intervals must comply with the 

requirements of the latest edition and 

addenda of the Code incorporated by 

reference in paragraph (a) of this section 

12 months before the start of the 120- 

month inspection interval (or the 

optional ASME Code Cases listed in 

NRC Regulatory Guide 1.147, Revision 

17, when using Section XI, or 

Regulatory Guide 1.192, Revision 1, 

when using the OM Code, that are 

incorporated by reference in paragraphs 

(a)(3)(ii) and (iii) of this section), subject 

to the conditions listed in paragraph (b) 

of this section. However, a licensee 

whose inservice inspection interval 

commences during the 12 through 18- 

month period after July 21, 2011, may 

delay the update of their Appendix VIII 

program by up to 18 months after July 

21, 2011. 

(iii) Applicable ISI Code: Optional 

surface examination requirement. When 

applying editions and addenda prior to 

the 2003 Addenda of Section XI of the 

ASME BPV Code, licensees may, but are 

not required to, perform the surface 

examinations of high-pressure safety 

injection systems specified in Table 

IWB–2500–1, Examination Category B– 

J, Item Numbers B9.20, B9.21, and 

B9.22. 

(iv) Applicable ISI Code: Use of 

subsequent Code editions and addenda. 

Inservice examination of components 

and system pressure tests may meet the 

requirements set forth in subsequent 

editions and addenda that are 

incorporated by reference in paragraph 

(a) of this section, subject to the 

conditions listed in paragraph (b) of this 

section, and subject to Commission 

approval. Portions of editions or 

addenda may be used, provided that all 

related requirements of the respective 

editions or addenda are met. 

(v) Applicable ISI Code: Metal and 

concrete containments. For a boiling or 

pressurized water-cooled nuclear power 

facility whose construction permit 

under this part or combined license 

under part 52 of this chapter was issued 

after January 1, 1956, the following are 

required: 

(A) Metal and concrete containments: 

First provision. Metal containment 

pressure retaining components and their 

integral attachments must meet the 

inservice inspection, repair, and 

replacement requirements applicable to 

components that are classified as ASME 

Code Class MC; 

(B) Metal and concrete containments: 

Second provision. Metallic shell and 

penetration liners that are pressure 

retaining components and their integral 

attachments in concrete containments 

must meet the inservice inspection, 

repair, and replacement requirements 

applicable to components that are 

classified as ASME Code Class MC; and 

(C) Metal and concrete containments: 

Third provision. Concrete containment 

pressure retaining components and their 

integral attachments, and the posttensioning systems of concrete 

containments, must meet the inservice 

inspections, repair, and replacement 

requirements applicable to components 

that are classified as ASME Code Class 

CC. 

(5) Requirements for updating ISI 

programs—(i) ISI program update: 

Applicable ISI Code editions and 

addenda. The inservice inspection 

program for a boiling or pressurized 

water-cooled nuclear power facility 

must be revised by the licensee, as 

necessary, to meet the requirements of 

paragraph (g)(4) of this section. 

(ii) ISI program update: Conflicting 

ISI Code requirements with technical 

specifications. If a revised inservice 

inspection program for a facility 

conflicts with the technical 

specifications for the facility, the 

licensee must apply to the Commission 

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for amendment of the technical 

specifications to conform the technical 

specifications to the revised program. 

The licensee must submit this 

application, as specified in § 50.4, at 

least six months before the start of the 

period during which the provisions 

become applicable, as determined by 

paragraph (g)(4) of this section. 

(iii) ISI program update: Notification 

of impractical ISI Code requirements. If 

the licensee has determined that 

conformance with a Code requirement is 

impractical for its facility the licensee 

must notify the NRC and submit, as 

specified in § 50.4, information to 

support the determinations. 

Determinations of impracticality in 

accordance with this section must be 

based on the demonstrated limitations 

experienced when attempting to comply 

with the Code requirements during the 

inservice inspection interval for which 

the request is being submitted. Requests 

for relief made in accordance with this 

section must be submitted to the NRC 

no later than 12 months after the 

expiration of the initial or subsequent 

120-month inspection interval for which 

relief is sought. 

(iv) ISI program update: Schedule for 

completing impracticality 

determinations. Where the licensee 

determines that an examination 

required by Code edition or addenda is 

impractical, the basis for this 

determination must be submitted for 

NRC review and approval not later than 

12 months after the expiration of the 

initial or subsequent 120-month 

inspection interval for which relief is 

sought. 

(6) Actions by the Commission for 

evaluating impractical and augmented 

ISI Code requirements—(i) Impractical 

ISI requirements: Granting of relief. The 

Commission will evaluate 

determinations under paragraph (g)(5) of 

this section that code requirements are 

impractical. The Commission may grant 

such relief and may impose such 

alternative requirements as it 

determines are authorized by law, will 

not endanger life or property or the 

common defense and security, and are 

otherwise in the public interest giving 

due consideration to the burden upon 

the licensee that could result if the 

requirements were imposed on the 

facility. 

(ii) Augmented ISI program. The 

Commission may require the licensee to 

follow an augmented inservice 

inspection program for systems and 

components for which the Commission 

deems that added assurance of 

structural reliability is necessary. 

(A) [Reserved] 

(B) Augmented ISI requirements: 

Submitting containment ISI programs. 

Licensees do not have to submit to the 

NRC for approval of their containment 

inservice inspection programs that were 

developed to satisfy the requirements of 

Subsection IWE and Subsection IWL 

with specified conditions. The program 

elements and the required 

documentation must be maintained on 

site for audit. 

(C) Augmented ISI requirements: 

Implementation of Appendix VIII to 

Section XI. (1) Appendix VIII and the 

supplements to Appendix VIII to 

Section XI, Division 1, 1995 Edition 

with the 1996 Addenda of the ASME 

BPV Code must be implemented in 

accordance with the following schedule: 

Appendix VIII and Supplements 1, 2, 3, 

and 8—May 22, 2000; Supplements 4 

and 6—November 22, 2000; Supplement 

11—November 22, 2001; and 

Supplements 5, 7, and 10—November 

22, 2002. 

(2) Licensees implementing the 1989 

Edition and earlier editions and 

addenda of IWA–2232 of Section XI, 

Division 1, of the ASME BPV Code must 

implement the 1995 Edition with the 

1996 Addenda of Appendix VIII and the 

supplements to Appendix VIII of 

Section XI, Division 1, of the ASME 

BPV Code. 

(D) Augmented ISI requirements: 

Reactor vessel head inspections—(1) All 

licensees of pressurized water reactors 

must augment their inservice inspection 

program with ASME Code Case N–729– 

1, subject to the conditions specified in 

paragraphs (g)(6)(ii)(D)(2) through (6) of 

this section. Licensees of existing 

operating reactors as of September 10, 

2008, must implement their augmented 

inservice inspection program by 

December 31, 2008. Once a licensee 

implements this requirement, the First 

Revised NRC Order EA–03–009 no 

longer applies to that licensee and shall 

be deemed to be withdrawn. 

(2) Note 9 of ASME Code Case N– 

729–1 must not be implemented. 

(3) Instead of the specified 

‘‘examination method’’ requirements for 

volumetric and surface examinations in 

Note 6 of Table 1 of Code Case N–729– 

1, the licensee must perform volumetric 

and/or surface examination of 

essentially 100 percent of the required 

volume or equivalent surfaces of the 

nozzle tube, as identified by Figure 2 of 

ASME Code Case N–729–1. A 

demonstrated volumetric or surface leak 

path assessment through all J-groove 

welds must be performed. If a surface 

examination is being substituted for a 

volumetric examination on a portion of 

a penetration nozzle that is below the 

toe of the J-groove weld [Point E on 

Figure 2 of ASME Code Case N–729–1], 

the surface examination must be of the 

inside and outside wetted surface of the 

penetration nozzle not examined 

volumetrically. 

(4) By September 1, 2009, ultrasonic 

examinations must be performed using 

personnel, procedures, and equipment 

that have been qualified by blind 

demonstration on representative 

mockups using a methodology that 

meets the conditions specified in 

paragraphs (g)(6)(ii)(D)(4)(i) through (iv), 

instead of the qualification requirements 

of Paragraph –2500 of ASME Code Case 

N–729–1. References herein to Section 

XI, Appendix VIII, must be to the 2004 

Edition with no addenda of the ASME 

BPV Code. 

(i) The specimen set must have an 

applicable thickness qualification range 

of +25 percent to ¥40 percent for 

nominal depth through-wall thickness. 

The specimen set must include 

geometric and material conditions that 

normally require discrimination from 

primary water stress corrosion cracking 

(PWSCC) flaws. 

(ii) The specimen set must have a 

minimum of ten (10) flaws that provide 

an acoustic response similar to PWSCC 

indications. All flaws must be greater 

than 10 percent of the nominal pipe 

wall thickness. A minimum of 20 

percent of the total flaws must initiate 

from the inside surface and 20 percent 

from the outside surface. At least 20 

percent of the flaws must be in the 

depth ranges of 10–30 percent throughwall thickness and at least 20 percent 

within a depth range of 31–50 percent 

through-wall thickness. At least 20 

percent and no more than 60 percent of 

the flaws must be oriented axially. 

(iii) Procedures must identify the 

equipment and essential variables and 

settings used for the qualification, in 

accordance with Subarticle VIII–2100 of 

Section XI, Appendix VIII. The 

procedure must be requalified when an 

essential variable is changed outside the 

demonstration range as defined by 

Subarticle VIII–3130 of Section XI, 

Appendix VIII, and as allowed by 

Articles VIII–4100, VIII–4200, and VIII– 

4300 of Section XI, Appendix VIII. 

Procedure qualification must include 

the equivalent of at least three personnel 

performance demonstration test sets. 

Procedure qualification requires at least 

one successful personnel performance 

demonstration. 

(iv) Personnel performance 

demonstration test acceptance criteria 

must meet the personnel performance 

demonstration detection test acceptance 

criteria of Table VIII—S10–1 of Section 

XI, Appendix VIII, Supplement 10. 

Examination procedures, equipment, 

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Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 65813 

and personnel are qualified for depth 

sizing and length sizing when the RMS 

error, as defined by Subarticle VIII–3120 

of Section XI, Appendix VIII, of the flaw 

depth measurements, as compared to 

the true flaw depths, do not exceed 1⁄8

inch (3 mm) and the root mean square 

(RMS) error of the flaw length 

measurements, as compared to the true 

flaw lengths, do not exceed 3⁄8 inch (10 

mm), respectively. 

(5) If flaws attributed to PWSCC have 

been identified, whether acceptable or 

not for continued service under 

Paragraphs –3130 or –3140 of ASME 

Code Case N–729–1, the re-inspection 

interval must be each refueling outage 

instead of the re-inspection intervals 

required by Table 1, Note (8), of ASME 

Code Case N–729–1. 

(6) Appendix I of ASME Code Case 

N–729–1 must not be implemented 

without prior NRC approval. 

(E) Augmented ISI requirements: 

Reactor coolant pressure boundary 

visual inspections 10—(1) All licensees 

of pressurized water reactors must 

augment their inservice inspection 

program by implementing ASME Code 

Case N–722–1, subject to the conditions 

specified in paragraphs (g)(6)(ii)(E)(2) 

through (4) of this section. The 

inspection requirements of ASME Code 

Case N–722–1 do not apply to 

components with pressure retaining 

welds fabricated with Alloy 600/82/182 

materials that have been mitigated by 

weld overlay or stress improvement. 

(2) If a visual examination determines 

that leakage is occurring from a specific 

item listed in Table 1 of ASME Code 

Case N–722–1 that is not exempted by 

the ASME Code, Section XI, IWB– 

1220(b)(1), additional actions must be 

performed to characterize the location, 

orientation, and length of a crack or 

cracks in Alloy 600 nozzle wrought 

material and location, orientation, and 

length of a crack or cracks in Alloy 82/ 

182 butt welds. Alternatively, licensees 

may replace the Alloy 600/82/182 

materials in all the components under 

the item number of the leaking 

component. 

(3) If the actions in paragraph 

(g)(6)(ii)(E)(2) of this section determine 

that a flaw is circumferentially oriented 

and potentially a result of primary water 

stress corrosion cracking, licensees must 

perform non-visual NDE inspections of 

components that fall under that ASME 

Code Case N–722–1 item number. The 

number of components inspected must 

equal or exceed the number of 

components found to be leaking under 

that item number. If circumferential 

cracking is identified in the sample, 

non-visual NDE must be performed in 

the remaining components under that 

item number. 

(4) If ultrasonic examinations of butt 

welds are used to meet the NDE 

requirements in paragraphs 

(g)(6)(ii)(E)(2) or (3) of this section, they 

must be performed using the 

appropriate supplement of Section XI, 

Appendix VIII, of the ASME BPV Code. 

(F) Augmented ISI requirements: 

Examination requirements for Class 1 

piping and nozzle dissimilar-metal butt 

welds—(1) Licensees of existing, 

operating pressurized-water reactors as 

of July 21, 2011, must implement the 

requirements of ASME Code Case N– 

770–1, subject to the conditions 

specified in paragraphs (g)(6)(ii)(F)(2) 

through (10) of this section, by the first 

refueling outage after August 22, 2011. 

(2) Full structural weld overlays 

authorized by the NRC staff may be 

categorized as Inspection Items C or F, 

as appropriate. Welds that have been 

mitigated by the Mechanical Stress 

Improvement Process (MSIPTM) may be 

categorized as Inspection Items D or E, 

as appropriate, provided the criteria in 

Appendix I of the Code Case have been 

met. For ISI frequencies, all other butt 

welds that rely on Alloy 82/182 for 

structural integrity must be categorized 

as Inspection Items A–1, A–2 or B until 

the NRC staff has reviewed the 

mitigation and authorized an alternative 

Code Case Inspection Item for the 

mitigated weld, or until an alternative 

Code Case Inspection Item is used based 

on conformance with an ASME 

mitigation Code Case endorsed in 

Regulatory Guide 1.147 with conditions, 

if applicable, and incorporated by 

reference in this section. 

(3) Baseline examinations for welds in 

Table 1, Inspection Items A–1, A–2, and 

B, must be completed by the end of the 

next refueling outage after January 20, 

2012. Previous examinations of these 

welds can be credited for baseline 

examinations if they were performed 

within the re-inspection period for the 

weld item in Table 1 using Section XI, 

Appendix VIII, requirements and met 

the Code required examination volume 

of essentially 100 percent. Other 

previous examinations that do not meet 

these requirements can be used to meet 

the baseline examination requirement, 

provided NRC approval of alternative 

inspection requirements in accordance 

with paragraphs (z)(1) or (2) of this 

section is granted prior to the end of the 

next refueling outage after January 20, 

2012. 

(4) The axial examination coverage 

requirements of Paragraph—2500(c) 

may not be considered to be satisfied 

unless essentially 100 percent coverage 

is achieved. 

(5) All hot-leg operating temperature 

welds in Inspection Items G, H, J, and 

K must be inspected each inspection 

interval. A 25 percent sample of 

Inspection Items G, H, J, and K cold-leg 

operating temperature welds must be 

inspected whenever the core barrel is 

removed (unless it has already been 

inspected within the past 10 years) or 20 

years, whichever is less. 

(6) For any mitigated weld whose 

volumetric examination detects growth 

of existing flaws in the required 

examination volume that exceed the 

previous IWB–3600 flaw evaluations or 

new flaws, a report summarizing the 

evaluation, along with inputs, 

methodologies, assumptions, and causes 

of the new flaw or flaw growth is to be 

provided to the NRC prior to the weld 

being placed in service other than 

modes 5 or 6. 

(7) For Inspection Items G, H, J, and 

K, when applying the acceptance 

standards of ASME BPV Code, Section 

XI, IWB–3514, for planar flaws 

contained within the inlay or onlay, the 

thickness ‘‘t’’ in IWB–3514 is the 

thickness of the inlay or onlay. For 

planar flaws in the balance of the 

dissimilar metal weld examination 

volume, the thickness ‘‘t’’ in IWB–3514 

is the combined thickness of the inlay 

or onlay and the dissimilar metal weld. 

(8) Welds mitigated by optimized 

weld overlays in Inspection Items D and 

E are not permitted to be placed into a 

population to be examined on a sample 

basis and must be examined once each 

inspection interval. 

(9) Replace the first two sentences of 

Extent and Frequency of Examination 

for Inspection Item D in Table 1 of Code 

Case N–770–1 with, ‘‘Examine all welds 

no sooner than the third refueling 

outage and no later than 10 years 

following stress improvement 

application.’’ Replace the first two 

sentences of Note (11)(b)(2) in Code 

Case N–770–1 with, ‘‘The first 

examination following weld inlay, 

onlay, weld overlay, or stress 

improvement for Inspection Items D 

through K must be performed as 

specified.’’ 

(10) General Note (b) to Figure 5(a) of 

Code Case N–770–1 pertaining to 

alternative examination volume for 

optimized weld overlays may not be 

applied unless NRC approval is 

authorized under paragraphs (z)(1) or (2) 

of this section. 

(h) Protection and safety systems. 

Protection systems of nuclear power 

reactors of all types must meet the 

requirements specified in this 

paragraph. Each combined license for a 

utilization facility is subject to the 

following conditions. 

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65814 Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations 

(1) [Reserved] 

(2) Protection systems. For nuclear 

power plants with construction permits 

issued after January 1, 1971, but before 

May 13, 1999, protection systems must 

meet the requirements stated in either 

IEEE Std. 279, ‘‘Criteria for Protection 

Systems for Nuclear Power Generating 

Stations,’’ or in IEEE Std. 603–1991, 

‘‘Criteria for Safety Systems for Nuclear 

Power Generating Stations,’’ and the 

correction sheet dated January 30, 1995. 

For nuclear power plants with 

construction permits issued before 

January 1, 1971, protection systems 

must be consistent with their licensing 

basis or may meet the requirements of 

IEEE Std. 603–1991 and the correction 

sheet dated January 30, 1995. 

(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. 

(i)–(y) [Reserved] 

(z) Alternatives to codes and 

standards requirements. Alternatives to 

the requirements of paragraphs (b) 

through (h) of this section or portions 

thereof may be used when authorized by 

the Director, Office of Nuclear Reactor 

Regulation, or Director, Office of New 

Reactors, as appropriate. A proposed 

alternative must be submitted and 

authorized prior to implementation. The 

applicant or licensee must demonstrate 

that: 

(1) Acceptable level of quality and 

safety. The proposed alternative would 

provide an acceptable level of quality 

and safety; or 

(2) Hardship without a compensating 

increase in quality and safety. 

Compliance with the specified 

requirements of this section would 

result in hardship or unusual difficulty 

without a compensating increase in the 

level of quality and safety. Footnotes to 

§ 50.55a: 

1USAS and ASME Code addenda issued 

prior to the winter 1977 Addenda are 

considered to be ‘‘in effect’’ or ‘‘effective’’ 6 

months after their date of issuance and after 

they are incorporated by reference in 

paragraph (a) of this section. Addenda to the 

ASME Code issued after the summer 1977 

Addenda are considered to be ‘‘in effect’’ or 

‘‘effective’’ after the date of publication of the 

addenda and after they are incorporated by 

reference in paragraph (a) of this section. 2–3 [Reserved]. 4For ASME Code editions and addenda 

issued prior to the winter 1977 Addenda, the 

Code edition and addenda applicable to the 

component is governed by the order or 

contract date for the component, not the 

contract date for the nuclear energy system. 

For the winter 1977 Addenda and subsequent 

editions and addenda the method for 

determining the applicable Code editions and 

addenda is contained in Paragraph NCA 1140 

of Section III of the ASME Code. 5–6 [Reserved]. 7 Guidance for quality group classifications 

of components that are to be included in the 

safety analysis reports pursuant to § 50.34(a) 

and § 50.34(b) may be found in Regulatory 

Guide 1.26, ‘‘Quality Group Classifications 

and Standards for Water-, Steam-, and 

Radiological-Waste-Containing Components 

of Nuclear Power Plants,’’ and in Section 

3.2.2 of NUREG–0800, ‘‘Standard Review 

Plan for Review of Safety Analysis Reports 

for Nuclear Power Plants.’’ 8–9 [Reserved]. 10For inspections to be conducted once per 

interval, the inspections must be performed 

in accordance with the schedule in Section 

XI, paragraph IWB–2400, except for plants 

with inservice inspection programs based on 

a Section XI edition or addenda prior to the 

1994 Addenda. For plants with inservice 

inspection programs based on a Section XI 

edition or addenda prior to the 1994 

Addenda, the inspection must be performed 

in accordance with the schedule in Section 

XI, paragraph IWB–2400, of the 1994 

Addenda. 

Dated at Rockville, Maryland, this 11th day 

of August 2014. 

For the Nuclear Regulatory Commission. 

Daniel H. Dorman, 

Acting Director, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2014–25491 Filed 11–4–14; 8:45 am] 

BILLING CODE 7590–01–P 

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