[Federal Register Volume 79, Number 214 (Wednesday, November 5, 2014)]
[Rules and Regulations]
[Pages 65776-65814]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-25491]



[[Page 65775]]

Vol. 79

Wednesday,

No. 214

November 5, 2014

Part II





Nuclear Regulatory Commission





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10 CFR Part 50





Approval of American Society of Mechanical Engineers' Code Cases; Final 
Rule

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

[[Page 65776]]


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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.

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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: [email protected]. 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 publicly-available 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 [email protected]. 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: [email protected]; or 
Wallace Norris, Office of Nuclear Regulatory Research, telephone: 301-
251-7650; email: [email protected]; 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 NRC-approved 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 Sec.  
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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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 Sec.  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, Sec.  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 Sec.  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 Sec.  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.''
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    \1\ ASME 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.
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    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
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                                                               Comment submission ADAMS    Affected code cases/
           Commenter name                  Affiliation              Accession No.                 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          ML13259A040                  Proposed Rule.
                                      Institute.
                                                             ML13254A080 **

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Edward Colie.......................  South Carolina          ML13254A082                  Proposed Rule.
                                      Electric and Gas.
Patricia Campbell..................  GE Hitachi Nuclear      ML13259A038                  1332-6.
                                      Energy.
Devin Kelley.......................  AREVA.................  ML13259A039                  N-71-18.
David Helker.......................  Exelon Generation       ML13269A371                  N-60-5.
                                      Company, LLC.
                                                                                          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    ML13197A239                  N-805.
                                      Program Owners Group.
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** 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 Sec.  50.55a that reflects the 
commenter's view of the current regulatory process and suggested 
parsing the details of Sec.  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 Sec.  50.55a. (Culp-3)
    NRC Response: The main purpose of this rulemaking is to amend Sec.  
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 Sec.  50.55a.
    The NRC is not proposing a major restructuring or simplification of 
the requirements in Sec.  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 Sec.  
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 Sec.  
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 Sec.  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 plant-specific 
review process that covers more than the Codes and standards mentioned. 
It does not seem appropriate for Sec.  50.55a to reference Codes and 
standards that have been withdrawn (e.g., IEEE 279). The content of 
Sec.  50.55a represents an archive of once-upon-a-time 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

[[Page 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 Sec.  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 Sec.  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 Sec.  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 Sec.  50.55a in 
this rulemaking. The scope of this rulemaking is to: (1) Incorporate by 
reference the three Regulatory Guides identifying NRC-approved 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 Sec.  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 Sec.  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 Sec.  50.55a to aid the reader of this 
regulation. The hierarchy used in Sec.  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 Sec.  
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 Sec.  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 
Sec.  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 Sec.  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 Sec.  
50.55a into a suitable regulatory guide that can be referenced within 
the regulation. The NRC agrees that the format and organization of 
Sec.  50.55a could be improved, and the NRC may, in the future, conduct 
a rulemaking to restructure and simplify Sec.  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 Sec.  50.55a proposed rule.
    No change was made to the final rule as a result of this comment.
    Comment: The proposed revision to Sec.  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

[[Page 65780]]

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 Sec.  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 Sec.  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 NRC-
approved 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 Sec.  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 Sec.  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 Sec.  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 Sec.  50.55a: ``Final 
Reorganization of Paragraphs and Subparagraphs in 10 CFR 50.55a, `Codes 
and standards' '' (ADAMS Accession No. ML14015A191) and ``Cross-
Reference Tables'' (ADAMS Accession No. ML14211A050--package with two 
tables). The NRC is currently considering developing several 
alternatives to improve the format and organization of Sec.  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

[[Page 65781]]

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

[[Page 65782]]

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 Sec.  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

[[Page 65783]]

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[emsp14][deg]F (93 [deg]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[emsp14][deg]F (93 [deg]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[emsp14][deg]F 
(93 [deg]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 Sec.  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/doc-collections/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)


[[Page 65784]]


    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 
Sec.  50.55a(a)(3) [Sec.  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

[[Page 65785]]

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. (Comstock-2.4)
    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 Sec.  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.

[[Page 65786]]

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 Sec.  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 Sec.  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

[[Page 65787]]

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


[[Page 65788]]

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       The maximum yield strength of
                                                         Support Structures,     strain-hardened austenitic
                                                         Section III, Division   stainless steel shall not
                                                         I, Class 1.             exceed 90,000 psi in view of
                                                                                 the susceptibility of this
                                                                                 material to environmental
                                                                                 cracking.
N-208-2.......................  4.....................  Fatigue Analysis for    (1) In Figure A, the words ``No
                                                         Precipitation           mean stress'' shall be
                                                         Hardening Nickel        implemented with the
                                                         Alloy Bolting           understanding that it denotes
                                                         Material to             ``Maximum mean stress.''
                                                         Specification SB-637   (2) In Figure A, [sigma]y shall
                                                         N07718 for Class 1      be implemented with the
                                                         Construction, Section   understanding that it denotes
                                                         III, Division 1.        [sigma]max.
N-520-2.......................  4.....................  Alternative Rules for   The Code Case is considered
                                                         Renewal of Active or    acceptable with one
                                                         Expired N-type          clarification: an AIA is an
                                                         Certificates for        Authorized Inspection Agency
                                                         Plants Not in Active    and the AIA employs the
                                                         Construction, Section   Authorized Nuclear Inspector
                                                         III, Division 1.        (ANI).
N-757-1.......................  2.....................  Alternative Rules for   The design provisions of ASME
                                                         Acceptability for       Section III, Division 1,
                                                         Class 2 and 3 Valves    Appendix XIII, shall not be
                                                         (DN 25) and Smaller     used for Class 3 valves.
                                                         with Welded and
                                                         Nonwelded End
                                                         Connections Other
                                                         than Flanges, Section
                                                         III, Division 1.
----------------------------------------------------------------------------------------------------------------
                                         ASME BPV Code Case, Section XI
----------------------------------------------------------------------------------------------------------------
N-508-4.......................  8.....................  Rotation of Serviced    When Section XI requirements are
                                                         Snubbers and Pressure   used to govern the examination
                                                         Retaining Items for     and testing of snubbers and the
                                                         the Purpose of          ISI Code of Record is earlier
                                                         Testing, Section XI,    than Section XI, 2006 Addenda,
                                                         Division 1.             Footnote 1 shall not be
                                                                                 applied.
N-561-2.......................  1.....................  Alternative             (1) Paragraph 5(b): for repairs
                                                         Requirements for Wall   performed on a wet surface, the
                                                         Thickness Restoration   overlay is only acceptable
                                                         of Class 2 and High     until the next refueling
                                                         Energy Class 3 Carbon   outage.
                                                         Steel Piping, Section  (2) Paragraph 7(c): if the cause
                                                         XI, Division 1.         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             (1) Paragraph 5(b): for repairs
                                                         Requirements for Wall   performed on a wet surface, the
                                                         Thickness Restoration   overlay is only acceptable
                                                         of Class 3 Moderate     until the next refueling
                                                         Energy Carbon Steel     outage.
                                                         Piping, Section XI,    (2) Paragraph 7(c): if the cause
                                                         Division 1.             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     Requirements for        New condition (6): The
                                 Code Case. NRC had      Analytical Evaluation   evaluation criteria in Code
                                 proposed one new        of Pipe Wall            Case N-513-2 may be applied to
                                 condition in response   Thinning, Section XI,   Code Case N-597-2 for temporary
                                 to public comment on    Division 1.             acceptance of wall thinning
                                 last rulemaking.                                (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 [deg]F (93
                                                                                 [deg]C) and whose maximum
                                                                                 operating pressure does not
                                                                                 exceed 275 psig (1.9MPa). Code
                                                                                 Case N[dash]597-2 shall not be
                                                                                 used to evaluate through-wall
                                                                                 leakage conditions.
N-606-1.......................  Public comment          Similar and Dissimilar  Prior to welding, an examination
                                 received on             Metal Welding Using     or verification must be
                                 previously approved     Ambient Temperature     performed to ensure proper
                                 rule requesting         Machine GTAW Temper     preparation of the base metal,
                                 revision to             Bead Technique for      and that the surface is
                                 condition. Condition    BWR CRD Housing/Stub    properly contoured so that an
                                 was revised.            Tube Repairs, Section   acceptable weld can be
                                                         XI, Division 1.         produced. This verification is
                                                                                 to be required in the welding
                                                                                 procedures.
N-619.........................  Responding to comment   Alternative             In lieu of a UT examination,
                                 on previously           Requirements for        licensees may perform a VT-1
                                 approved Code Case.     Nozzle Inner Radius     examination in accordance with
                                                         Inspections for Class   the code of record for the
                                                         1 Pressurizer and       Inservice Inspection Program
                                                         Steam Generator         utilizing the allowable flaw
                                                         Nozzles, Section XI,    length criteria of Table IWB-
                                                         Division 1.             3512-1 with limiting
                                                                                 assumptions on the flaw aspect
                                                                                 ratio.
N-648-1.......................  Responding to comment   Alternative             In lieu of a UT examination,
                                 on previously           Requirements for        licensees may perform a VT-1
                                 approved Code Case.     Inner Radius            examination in accordance with
                                                         Inspections for Class   the code of record for the
                                                         1 Reactor Vessel        Inservice Inspection Program
                                                         Nozzles, Section XI,    utilizing the allowable flaw
                                                         Division 1.             length criteria of Table IWB-
                                                                                 3512-1 with limiting
                                                                                 assumptions on the flaw aspect
                                                                                 ratio.
N-661-2.......................  1.....................  Alternative             (1) Paragraph 5(b): for repairs
                                                         Requirements for Wall   performed on a wet surface, the
                                                         Thickness Restoration   overlay is only acceptable
                                                         of Classes 2 and 3      until the next refueling
                                                         Carbon Steel Piping     outage.
                                                         for Raw Water          (2) Paragraph 7(c): if the cause
                                                         Service, Section XI,    of the degradation has not been
                                                         Division 1.             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.

[[Page 65789]]

 
N-702.........................  Responding to comment   Alternative             The technical basis supporting
                                 on previously           Requirements for        the implementation of this Code
                                 approved Code Case.     Boiling Water Reactor   Case is addressed by BWRVIP-
                                                         (BWR) Nozzle Inner      108: BWR Vessel and Internals
                                                         Radius and Nozzle-to-   Project, ``Technical Basis for
                                                         Shell Welds, Section    the Reduction of Inspection
                                                         XI, Division 1.         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   Licensees may use Code Case OMN-
                                                         Preservice and          1, ``Alternative Rules for
                                                         Inservice Testing of    Preservice and Inservice
                                                         Active Electric Motor-  Testing of Certain Electric
                                                         Operated Valve          Motor-Operated Valve Assemblies
                                                         Assemblies in Light-    in Light-Water Reactor Power
                                                         Water Reactor Power     Plants,'' in lieu of the
                                                         Plants.                 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.

[[Page 65790]]

 
OMN-3.........................  2004 Edition..........  Requirements for        In addition to those components
                                                         Safety Significance     identified in ASME IST Program
                                                         Categorization of       Plan, implementation of Section
                                                         Components Using Risk   1, ``Applicability,'' of the
                                                         Insights for            Code Case must include within
                                                         Inservice Testing of    the scope of a licensee's risk-
                                                         LWR Power Plants.       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,
                                                                                 Risk-Informed 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   (1) Valve opening and closing
                                                         Insights for            functions must be demonstrated
                                                         Inservice Testing of    when flow testing or
                                                         Check Valves at LWR     examination methods
                                                         Power Plants.           (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     (1) When a reference curve may
                                                         for Testing.            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.

[[Page 65791]]

 
OMN-12........................  2004 Edition..........  Alternative             (1) Paragraph 4.2, ``Inservice
                                                         Requirements for        Test Requirements,'' of OMN-12
                                                         Inservice Testing       specifies inservice test
                                                         Using Risk Insights     requirements for pneumatically
                                                         for Pneumatically and   and hydraulically operated
                                                         Hydraulically           valve assemblies categorized as
                                                         Operated Valve          high safety significant within
                                                         Assemblies in Light-    the scope of the Code Case. The
                                                         Water Reactor Power     inservice testing program must
                                                         Plants (OM-Code 1998,   include a mix of static and
                                                         Subsection ISTC).       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 Sec.  50.55a. Regulatory 
Guide 1.193 is prepared by the NRC as a resource for stakeholders, 
allowing

[[Page 65792]]

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 Sec.  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 Sec.  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 Sec.  50.55a(z) (previously 
Sec.  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 Sec.  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 Sec. Sec.  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 Sec.  50.55a. For the least disruption 
to the existing structure of the section, the NRC incorporated by 
reference the multiple standards into Sec.  50.55a(a), the first 
paragraph of the section. Each national consensus standard that is 
being incorporated by reference in Sec.  50.55a has been listed 
separately. Accordingly, the regulatory language of Sec. Sec.  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 Sec.  
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-by-Paragraph Discussion,'' of 
this document. The regulatory text of Sec.  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 Sec.  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 Sec.  50.55a to the new regulations.
    Some alternatives the NRC considered were a new regulation adjacent 
to Sec.  50.55a (e.g., Sec. Sec.  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 Sec.  50.200, and continuing with 
Sec. Sec.  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 re-designate paragraphs in Sec.  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 Sec.  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 Sec.  50.55a with public input. The current 
reorganization of this

[[Page 65793]]

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 Sec.  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 Sec.  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 first-level 
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 Sec.  
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 Sec. Sec.  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 Sec.  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 Sec.  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, Sec.  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 Sec.  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., Sec.  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 Sec.  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 Sec.  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 Sec.  50.55a(a)(3). If a Code Case is incorporated by reference 
into Sec.  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 Sec.  
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

[[Page 65794]]

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 Sec.  50.55a(z). If a Code Case is 
incorporated by reference into Sec.  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 Sec.  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 Sec. Sec.  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 Sec.  50.54, the introductory statement has been revised to 
include a reference to Sec.  50.55a. This revision clarifies that 
nuclear power plant licensees, as described in the introductory 
paragraph of Sec.  50.54, also are subject to the applicable 
requirements delineated in Sec.  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 Sec.  50.55a(a)(1), and no 
change to the requirement is intended by the transfer of this 
requirement from Sec.  50.55a(a)(1) to Sec.  50.54(jj), except for 
clarification of its applicability.

Section 50.55

    In Sec.  50.55, the introductory text has been revised to include 
references to existing Sec.  50.55a, and paragraphs (g) and (h) have 
been added and reserved for future use. Further, existing Sec.  
50.55a(a)(1) has been moved to a newly created Sec.  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 Sec.  
50.55 has been revised to maintain the existing applicability of the 
requirement in the newly created Sec.  50.55(i) to construction permits 
for utilization facilities.

Section 50.55a

    The introductory text to Sec.  50.55a was relocated to several 
other locations. There is no introductory text to Sec.  50.55a in the 
new rule. The first sentence in the previous introductory text was 
relocated to the first sentence in Sec.  50.55. The remaining sentences 
were relocated to Sec.  50.55a(b) (second sentence), Sec.  50.55a(b)(1) 
(first sentence), Sec.  50.55a(b)(4) (first sentence), Sec.  50.55a(c) 
(second sentence), Sec.  50.55a(d) (second sentence), Sec.  50.55a(e) 
(second sentence), Sec.  50.55a(f) (second and third sentences), Sec.  
50.55a(g) (second and third sentences), and Sec.  50.55a(h) (second 
sentence).
    In addition to moving existing paragraphs, creating new paragraphs, 
and revising introductory and regulatory texts, the footnotes in Sec.  
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 Sec.  50.55a 
to incorporate by reference the multiple standards currently identified 
in existing Sec.  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

[[Page 65795]]

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 Sec.  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 Sec.  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).

[[Page 65796]]

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 non-radiological 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 [email protected]; 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 government-unique 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 ASME-
approved 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 Sec.  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 Sec.  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 Sec.  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

[[Page 65797]]

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 Sec.  50.55a(a)(3), requests for relief under Sec.  50.55a(f) 
and (g), or requests for exemptions under Sec. Sec.  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     ML12289A121
 and Subparagraphs.
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    ML14015A191
 Subparagraphs.
Cross-Reference Tables (package)........  ML14211A050
RG 1.84, ``Design, Fabrication, and       ML13339A515
 Materials Code Case Acceptability, ASME
 Section III,'' Revision 36.
RG 1.147, ``Inservice Inspection Code     ML13339A689
 Case Acceptability, ASME Section XI,
 Division 1,'' Revision 17.
RG 1.192, ``Operation and Maintenance     ML13340A034
 Code Case Acceptability, ASME OM
 Code,'' Revision 1.
RG 1.193, ``ASME Code Cases Not Approved  ML13350A001
 for Use,'' Revision 4.
RG 1.200, ``An Approach for Determining   ML090410014
 the Technical Adequacy of Probabilistic
 Risk Assessment Results for Risk-
 informed Activities,'' Revision 2.
RG 1.201, ``Guidelines for Categorizing   ML061090627
 Structures, Systems, and Components in
 Nuclear Power Plants According to Their
 Safety Significance,'' Revision 1.
2007/12/19--``SECY--Petition for          ML073600974
 Rulemaking to amend 10 CFR 50.55a--
 Rev.1'' submitted by Ray West.
Hatch Plant Report--``Hatch, Units 1 &    ML033280037
 2, Farley, Units 1 & 2, Vogtle, Units 1
 & 2, Safety Evaluation Re. Request to
 Use ASME Code Case N-661''.

[[Page 65798]]

 
EPRI Technical Report--Project No. 704--  ML023330203
 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.
Safety Evaluation of Proprietary EPRI     ML073600374
 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).
Comment Letter--Comment (4) of Bryan A.   ML092190138
 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.
SRM-COMNJD-03-0002--Stabilizing the PRA   ML033520457
 Quality Expectations and Requirements.
SECY-04-0118--Plan for the                ML041470505
 Implementation of the Commission's
 Phased Approach to Probabilistic Risk
 Assessment Quality.
SRM-SECY-04-0118--Plan for the            ML042800369
 Implementation of the Commission's
 Phased Approach to Probabilistic Risk
 Assessment Quality.
NUREG-0800--Chapter 4, Section 4.5.1,     ML070230007
 Revision 3, Control Rod Drive
 Structural Materials, dated March 2007.
NUREG-0800--Chapter 5, Section 5.2.3,     ML063190006
 Revision 3, Reactor Coolant Pressure
 Boundary Materials, dated March 2007.
NUREG/CR-6943--A Study of Remote Visual   ML073110060
 Methods to Detect Cracking in Reactor
 Components.
------------------------------------------------------------------------

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

0
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).


0
2. In Sec.  50.54, revise the introductory text, add reserved paragraph 
(ii), and add paragraph (jj) to read as follows:


Sec.  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 Sec.  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.

0
3. In Sec.  50.55, revise the introductory text, add reserved 
paragraphs (g) and (h), and add paragraph (i) to read as follows:


Sec.  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 
Sec.  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 Sec.  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 Sec.  50.55a until the date that the Commission makes 
the finding under Sec.  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.

0
4. Revise Sec.  50.55a to read as follows:


Sec.  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:

[[Page 65799]]


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

[[Page 65800]]

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: [email protected]; http://www.nrc.gov/reading-rm/doc-collections/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[emsp14][deg]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

[[Page 65801]]

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: Pressure-retaining 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

[[Page 65802]]

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 ``owner-defined'' 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]

[[Page 65803]]

    (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

[[Page 65804]]

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-metal-interface), 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-to-vessel 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-metal-interface) 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-to-base-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.

[[Page 65805]]

    (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

[[Page 65806]]

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[emsp14][deg]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, [Delta]KI depends on the crack depth (a), 
and the flow stress ([sigma]f). The flow stress is 
defined by [sigma]f = 1/2([sigma]ys + 
[sigma]ult), where [sigma]ys is the yield 
strength and [sigma]ult 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 x 1.12 
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI = 
Kmax. For R < -2 and Kmax- Kmin <= 
0.8 x 1.12 [sigma]f[radic]([pi]a), S = 1 and 
[Delta]KI = (1 - R) Kmax/3. For R < 0 and 
Kmax - Kmin > 0.8 x 1.12 
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI = 
Kmax-Kmin.

    (xxix) Section XI condition: Nonmandatory Appendix R. Nonmandatory 
Appendix R, ``Risk-Informed 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.

[[Page 65807]]

    (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 Sec.  50.2 need not meet the requirements of paragraph (c)(1) of 
this section, provided that:

[[Page 65808]]

    (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 B \7\ 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 C \9\ 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 Sec.  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 Sec.  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

[[Page 65809]]

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 water-cooled nuclear power facility whose construction 
permit under this part or design approval, design certification, 
combined license, or manufacturing license under part 52 of this 
chapter was issued on or after July 1, 1974:
    (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,

[[Page 65810]]

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 Sec.  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 Sec.  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 Sec.  52.103(g) of this chapter. 
Requirements for inservice testing of Class 1, Class 2, and Class 3 
pumps and valves are located in Sec.  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

[[Page 65811]]

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 post-tensioning 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

[[Page 65812]]

for amendment of the technical specifications to conform the technical 
specifications to the revised program. The licensee must submit this 
application, as specified in Sec.  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 Sec.  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 through-wall 
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,

[[Page 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 
(MSIP\TM\) 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.

[[Page 65814]]

    (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 Sec.  50.55a:

    \1\ USAS 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].
    \4\ For 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 
Sec.  50.34(a) and Sec.  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].
    \10\ For 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