[Federal Register Volume 82, Number 136 (Tuesday, July 18, 2017)]
[Rules and Regulations]
[Pages 32934-32986]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-14166]



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Vol. 82

Tuesday,

No. 136

July 18, 2017

Part II





 Nuclear Regulatory Commission





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 10 CFR Part 50





 Incorporation by Reference of American Society of Mechanical Engineers 
Codes and Code Cases; Final Rule

  Federal Register / Vol. 82 , No. 136 / Tuesday, July 18, 2017 / Rules 
and Regulations  

[[Page 32934]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[NRC-2011-0088]
RIN 3150-AI97


Incorporation by Reference of American Society of Mechanical 
Engineers Codes and 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 recent editions and addenda to 
the American Society of Mechanical Engineers (ASME) Codes for nuclear 
power plants and a standard for quality assurance. The NRC is also 
incorporating by reference six ASME Code Cases. This action is in 
accordance with the NRC's policy to periodically update the regulations 
to incorporate by reference new editions and addenda of the ASME Codes 
and is intended to maintain the safety of nuclear power plants and to 
make NRC activities more effective and efficient.

DATES: This final rule is effective on August 17, 2017. The 
incorporation by reference of certain publications listed in the 
regulation is approved by the Director of the Federal Register as of 
August 17, 2017.

ADDRESSES: Please refer to Docket ID NRC-2011-0088 when contacting the 
NRC about the availability of information for this action. You may 
obtain publicly-available information related to this action by any of 
the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2011-0088. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     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, instructions about obtaining materials 
referenced in this document are provided in the ``Availability of 
Documents'' section.
     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: Daniel I. Doyle, Office of Nuclear 
Reactor Regulation, telephone: 301-415-3748, email: 
[email protected]; or Keith Hoffman, Office of Nuclear Reactor 
Regulation, telephone: 301-415-1294, email: [email protected]. Both 
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.

SUPPLEMENTARY INFORMATION: 

Executive Summary

A. Need for the Regulatory Action

    The NRC is amending its regulations to incorporate by reference 
recent editions and addenda to the ASME Codes for nuclear power plants 
and an ASME standard for quality assurance. The NRC is also 
incorporating by reference six ASME Code Cases.
    This final rule is the latest in a series of rulemakings to amend 
the NRC's regulations to incorporate by reference revised and updated 
ASME Codes for nuclear power plants. The ASME is a voluntary consensus 
standards body, and the ASME Codes are voluntary consensus standards. 
The ASME periodically revises and updates its codes for nuclear power 
plants by issuing new editions and addenda. The NRC's use of the ASME 
Codes is consistent with applicable requirements of the National 
Technology Transfer and Advancement Act (NTTAA). This rulemaking is in 
accordance with the NRC's policy to update the regulations to 
incorporate by reference those new editions and addenda. The 
incorporation by reference of the new editions and addenda will 
maintain the safety of nuclear power plants, make NRC activities more 
effective and efficient, and allow nuclear power plant licensees and 
applicants to take advantage of the latest ASME Codes. Additional 
discussion of voluntary consensus standards and the NRC's compliance 
with the NTTAA is set forth in Section XIV of this document, 
``Voluntary Consensus Standards.''

B. Major Provisions

    Major provisions of this final rule include:
     Incorporation by reference of ASME Codes into the NRC's 
regulations and delineation of the NRC's requirements for the use of 
these codes, including conditions.
     Incorporation by reference of various versions of quality 
assurance standard NQA-1 into NRC regulations and approval for their 
use.
     Incorporation by reference of six ASME Code Cases.

C. Costs and Benefits

    The NRC prepared a regulatory analysis (ADAMS Accession No. 
ML16130A522) to identify the costs and benefits associated with this 
final rule. The regulatory analysis prepared for this rulemaking was 
used to determine if the rule is cost-effective, overall, and to help 
the NRC evaluate potentially costly conditions placed on specific 
provisions of the ASME Codes and Code Cases which are the subject of 
this rulemaking. Therefore, the regulatory analysis focuses on the 
marginal difference in benefits and costs for each provision of this 
final rule relative to the ``no action'' baseline alternative. The 
regulatory analysis identified costs and benefits in a quantitative 
fashion as well as in a qualitative fashion. An uncertainty analysis 
was performed to evaluate the effects of uncertainties in the 
quantitative estimation of both costs and benefits, and this analysis 
showed the rule alternative is cost effective with over 99 percent 
certainty. The standard deviation of the cost estimate net benefit is 
$4.1 million.

                      Table 1--Cost-Benefit Summary
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                                                          Alternative 2--
                                                              the rule
                                                            alternative
                                                           net benefits
                        Objective                             (costs)
                                                          (million) (Net
                                                          present value,
                                                            7% discount
                                                               rate)
------------------------------------------------------------------------
Industry................................................           $11.5
NRC.....................................................            3.28
Net benefit.............................................            14.7
------------------------------------------------------------------------

    Table 1 summarizes the costs and benefits for the alternative of 
proceeding with the final rule (Alternative 2) and shows that the final 
rule is quantitatively cost-beneficial with a net benefit of $14.7 
million to both the industry and the NRC when compared to the 
regulatory baseline (Alternative 1). The regulatory analysis shows that 
implementing the final rule is quantitatively cost-effective and an 
efficient use of NRC and Industry resources. Uncertainty analysis shows 
a standard deviation of $4.08 million, resulting in a net benefit range 
of $8.19 million to $21.6 million. Because the

[[Page 32935]]

rulemaking alternative is cost-effective, the rulemaking approach is 
recommended.
    There are several benefits associated with this final rule. The new 
motor-operated valve (MOV) provisions in this final rule result in over 
$25 million in averted costs (7-percent net present value) due to the 
removal of quarterly testing requirements and replacing those 
requirements with less frequent diagnostic and biannual testing 
requirements. Additionally, the provisions in this final rule will 
result in averted costs to the NRC and the industry from relief 
requests for the code cases in this final rule, in particular the ASME 
OMN-20 Code Case Time Period Extension provision, in excess of $5.1 
million (7-percent net present value).
    Qualitative factors which were considered include regulatory 
stability and predictability, regulatory efficiency, and consistency 
with the NTTAA. Table 50 in the regulatory analysis includes a 
discussion of the costs and benefits that were considered 
qualitatively. Considering non-quantified costs and benefits, the 
regulatory analysis shows that the rulemaking is justified because the 
number and significance of the non-quantified benefits outweigh the 
non-quantified costs. Certainly, if the qualitative benefits (including 
the safety benefit, regulatory efficiency, and other nonquantified 
benefits) are considered together with the quantified benefits, then 
the benefits would outweigh the identified quantitative and qualitative 
impacts. Therefore, integrating both quantified and non-quantified 
costs and benefits, the benefits of the final rule outweigh the 
identified quantitative and qualitative impacts attributable to the 
final rule.

Table of Contents

I. Background
II. Discussion
    A. ASME BPV Code, Section III
    B. ASME BPV Code, Section XI
    C. OM Code
    D. ASME Code Cases
III. Opportunities for Public Participation
IV. NRC Responses to Public Comments
V. Section-by-Section Analysis
VI. Generic Aging Lessons Learned Report
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Plain Writing
XI. Finding of No Significant Impact: Environmental Assessment
XII. Paperwork Reduction Act Statement
XIII. Congressional Review Act
XIV. Voluntary Consensus Standards
XV. Incorporation by Reference--Reasonable Availability to 
Interested Parties
XVI. Availability of Guidance
XVII. Availability of Documents

I. Background

    The ASME develops and publishes the ASME Boiler and Pressure Vessel 
Code (BPV Code), which contains requirements for the design, 
construction, and inservice inspection (ISI) of nuclear power plant 
components; and the OM Code,\1\ which contains requirements for 
inservice testing (IST) of nuclear power plant components. Until 2012, 
the ASME issued new editions of the ASME BPV Code every 3 years and 
addenda to the editions annually, except in years when a new edition 
was issued. Similarly, the ASME periodically published new editions and 
addenda of the OM Code. Starting in 2012, the ASME decided to issue 
editions of its BPV and OM Codes (no addenda) every 2 years. The new 
editions and addenda typically revise provisions of the ASME BPV and OM 
Codes (ASME Codes) to broaden their applicability, add specific 
elements to current provisions, delete specific provisions, and/or 
clarify them to narrow the applicability of the provision. The 
revisions to the editions and addenda of the ASME Codes do not 
significantly change philosophy or approach.
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    \1\ The editions and addenda of the ASME Code for Operation and 
Maintenance of Nuclear Power Plants have had different titles from 
2005 to 2012 and are referred to collectively in this rule as the 
``OM Code.''
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    It has been the NRC's practice to establish requirements for the 
design, construction, operation, ISI examination, and IST of nuclear 
power plants by approving the use of editions and addenda of the ASME 
Codes in Sec.  50.55a of title 10 of the Code of Federal Regulations 
(10 CFR), ``Codes and standards.'' The NRC approves and/or mandates the 
use of certain parts of editions and addenda of these ASME Codes in 
Sec.  50.55a through the rulemaking process of ``incorporation by 
reference.'' Upon incorporation by reference of the ASME Codes into 
Sec.  50.55a, the provisions of the ASME Codes are legally-binding NRC 
requirements as delineated in Sec.  50.55a and subject to the 
conditions on certain specific ASME Code provisions that are set forth 
in Sec.  50.55a. The editions and addenda of the ASME BPV and OM Codes 
were last incorporated by reference into the regulations in a final 
rule dated June 21, 2011 (76 FR 36232), subject to the NRC's 
conditions.
    The ASME Codes are consensus standards developed by participants 
with broad and varied interests, including the NRC and licensees of 
nuclear power plants. The ASME's adoption of new editions of, and 
addenda to, the ASME Codes does not mean that there is unanimity on 
every provision in the ASME Codes. There may be disagreement among the 
technical experts, including NRC representatives, on the ASME Code 
committees and subcommittees, regarding the acceptability or 
desirability of a particular Code provision included in an ASME-
approved Code edition or addenda. If the NRC believes that there is a 
significant technical or regulatory concern with a provision in an 
ASME-approved Code edition or addenda being considered for 
incorporation by reference, then the NRC will condition the use of that 
provision when it incorporates by reference that ASME Code edition or 
addenda. In some cases, the condition increases the level of safety 
afforded by the ASME Code provision or addresses a regulatory issue not 
considered by the ASME. In other instances, where research data or 
experience has shown that certain Code provisions are unnecessarily 
conservative, the condition may provide that the Code provision need 
not be complied with in some or all respects. The NRC's conditions are 
included in Sec.  50.55a, typically in paragraph (b) of that 
regulation. In a Staff Requirements Memorandum (SRM) dated September 
10, 1999, the Commission indicated that NRC rulemakings adopting 
(incorporating by reference) a voluntary consensus standard must 
identify and justify each part of the standard that is not adopted. For 
this rulemaking, the provisions of the 2009 Addenda, 2010 Edition, 2011 
Addenda, and 2013 Edition of Section III, Division 1; and the 2009 
Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section XI, 
Division 1, of the ASME BPV Code; and the 2009 Edition, 2011 Addenda, 
and 2012 Edition of the OM Code that the NRC is not adopting, or 
partially adopting, are identified in the Discussion, Regulatory 
Analysis, and Backfitting and Issue Finality sections of this document. 
The provisions of those specific editions and addenda and Code Cases 
that are the subject of this rulemaking that the NRC finds to be 
conditionally acceptable, together with the applicable conditions, are 
also identified in the Discussion, Regulatory Analysis, and Backfitting 
and Issue Finality sections of this document.
    The ASME Codes are voluntary consensus standards, and the NRC's 
incorporation by reference of these Codes is consistent with applicable 
requirements of the NTTAA. Additional discussion on NRC's compliance 
with the NTTAA is set forth in Section XIV

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of this document, ``Voluntary Consensus Standards.''
    This final rule reflects the NRC's redesignation of paragraphs 
within Sec.  50.55a set forth in a final rule dated November 5, 2014 
(79 FR 65776), as corrected on December 11, 2014 (79 FR 73461). The re-
designation of paragraphs was needed to address the Office of the 
Federal Register's requirements in 1 CFR part 51 for incorporation by 
reference. For additional information on the November 2014 final rule, 
please consult the statement of considerations (preamble) for that 
final rule.

II. Discussion

    The NRC regulations incorporate by reference ASME Codes for nuclear 
power plants. The ASME periodically revises and updates its codes for 
nuclear power plants. This final rule is the latest in a series of 
rulemakings to amend the NRC's regulations to incorporate by reference 
revised and updated ASME Codes for nuclear power plants. The proposed 
rule which led to this final rule was published on September 18, 2015 
(80 FR 56820). This rulemaking is intended to maintain the safety of 
nuclear power plants and make NRC activities more effective and 
efficient.
    The NRC follows a three-step process to determine acceptability of 
new provisions in new editions and addenda to the Codes and the need 
for conditions on the uses of these Codes. This process was employed in 
the review of the Codes that are the subject of this rule. First, the 
NRC staff actively participates with other ASME committee members with 
full involvement in discussions and technical debates in the 
development of new and revised Codes. This includes a technical 
justification of each new or revised Code. Second, the NRC committee 
representatives discuss the Codes and technical justifications with 
other cognizant NRC staff to ensure an adequate technical review. 
Third, the NRC position on each Code is reviewed and approved by NRC 
management as part of the rule amending Sec.  50.55a to incorporate by 
reference new editions and addenda of the ASME Codes and conditions on 
their use. This regulatory process, when considered together with the 
ASME's own process for developing and approving the ASME Codes, 
provides reasonable assurance that the NRC approves for use only those 
new and revised Code edition and addenda, with conditions as necessary, 
that provide reasonable assurance of adequate protection to public 
health and safety, and that do not have significant adverse impacts on 
the environment.
    The NRC is amending its regulations to incorporate by reference:
     The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 
Edition of the ASME BPV Code, Section III, Division 1 and Section XI, 
Division 1, with conditions on their use.
     The 2009 Edition, the 2011 Addenda, and the 2012 Edition 
of Division 1 of the OM Code, with conditions on their use.
     ASME Standard NQA-1, ``Quality Assurance Requirements for 
Nuclear Facility Applications,'' including several editions and addenda 
to NQA-1 from previous years with slightly varying titles as identified 
in Sec.  50.55a(a)(1)(v). More specifically, the NRC is incorporating 
by reference the 1983 Edition through the 1994 Edition, the 2008 
Edition, and the 2009-1a Addenda to the 2008 Edition of ASME NQA-1, 
with conditions on their use.
     ASME BPV Code Case N-513-3, ``Evaluation Criteria for 
Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping 
Section XI, Division 1,'' Mandatory Appendix I, ``Relations for Fm, Fb, 
and F for Through-Wall Flaws,'' Approval Date: January 26, 2009. This 
Code Case has already been approved for use by the NRC in Regulatory 
Guide (RG) 1.147 (75 FR 61321; October 5, 2010), but is now being 
incorporated by reference in order to adopt a condition on Nonmandatory 
Appendix U, which requires the use of this Code Case appendix.
     ASME BPV Code Case N-729-4, ``Alternative Examination 
Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having 
Pressure-Retaining Partial-Penetration Welds Section XI, Division 1,'' 
ASME approval date: June 22, 2012, with conditions on its use.
     ASME BPV Code Case N-770-2, ``Alternative Examination 
Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel 
Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler 
Material With or Without Application of Listed Mitigation Activities, 
Section XI, Division 1,'' ASME approval date: June 9, 2011, with 
conditions on its use.
     ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast 
Austenitic Piping Welds From the Outside Surface Section XI, Division 
1,'' ASME approval date: October 16, 2012.
     ASME BPV Code Case N-852, ``Application of the ASME NPT 
Stamp, Section III, Division 1; Section III, Division 2; Section III, 
Division 3; Section III, Division 5,'' Approval Date: February 9, 2015.
     OM Code Case OMN-20, ``Inservice Test Frequency.''
    The current regulations in Sec.  50.55a(a)(1)(ii) incorporate by 
reference ASME BPV Code, Section XI, 1970 Edition through the 1976 
Winter Addenda; and the 1977 Edition (Division 1) through the 2008 
Addenda (Division 1), subject to the existing conditions in Sec.  
50.55a(b)(2)(i) through (xxix). This amendment revises Sec.  
50.55a(a)(1)(ii) to incorporate by reference the 2009 Addenda (Division 
1) through the 2013 Edition (Division 1) of the ASME BPV Code, Section 
XI. It also clarifies the wording and adds, removes, or revises some of 
the conditions as explained in this document.
    The NRC is revising Sec.  50.55a(a)(1)(iv) to incorporate by 
reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division 
1 of the OM Code. Based on this revision, the NRC regulations will 
incorporate by reference in Sec.  50.55a the 1995 Edition through the 
2012 Edition of the OM Code.
    The NRC reviewed changes to the Codes in the editions and addenda 
of the Codes identified in this rulemaking, and published a proposed 
rule in the Federal Register setting forth the NRC's proposal to 
incorporate by reference the ASME Codes, together with proposed 
conditions on their use (80 FR 56820; September 18, 2015). After 
consideration of the public comments received on the proposed rule 
(public comments are discussed in Section IV of this document, ``NRC 
Responses to Public Comments''), the NRC concludes, in accordance with 
the process for review of changes to the Codes, that each of the 
editions and addenda of the Codes, and the 2008 Edition and the 2009-1a 
Addenda of NQA-1, are technically adequate, consistent with current NRC 
regulations, and approved for use with specified conditions set forth 
in this final rule. Each of the NRC conditions and the reasons for each 
condition are discussed in the following sections. The discussions are 
organized under the applicable ASME Code and Section.
    There is not a separate heading for ASME quality assurance standard 
NQA-1 because there are three separate discussions of NQA-1--one under 
the heading for ASME BPV Code, Section III, one under the heading for 
ASME BPV Code, Section XI, and one under the heading for OM Code--
because there are three conditions related to NQA-1, one in each of 
those areas (Sec.  50.55a(b)(1)(iv) for Section III, Sec.  
50.55a(b)(2)(x) for Section XI, and Sec.  50.55a(b)(3)(i) for the OM 
Code). In addition, administrative and editorial changes to various 
paragraphs of Sec.  50.55a are being adopted for accuracy, clarity, 
consistency, and general

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administrative convenience. These editorial changes are not further 
discussed in this heading, but are described in Section V of this 
document, ``Section-by-Section Analysis.''
    Four of the six ASME Code Cases being incorporated by reference in 
this rulemaking (N-729-4, N-770-2, N-824, and OMN-20) are discussed in 
Section II.D of this document, ``ASME Code Cases.'' A fifth ASME Code 
Case, N-852, is discussed in Section II.A, ``ASME BPV Code, Section 
III,'' because the NRC's approval of that Code Case relates to a 
provision of Section III, which is addressed in Sec.  50.55a(b)(1)(ix). 
The sixth ASME Code Case, N-513-3, is discussed in Section II.B, ``ASME 
BPV Code, Section XI,'' because the NRC's approval of that Code Case 
relates to a provision of Section XI, which is addressed in Sec.  
50.55a(b)(2)(xxxiv).

A. ASME BPV Code, Section III

10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section 
III
    The NRC is clarifying that Section III Nonmandatory Appendices are 
not incorporated by reference. This language was originally added in a 
final rule published on June 21, 2011 (76 FR 36232); however, it was 
omitted from the final rule published on November 5, 2014 (79 FR 
65776). The NRC is correcting the omission by inserting the 
parenthetical clause ``(excluding Nonmandatory Appendices)'' in Sec.  
50.55a(a)(1)(i).
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
    The NRC is identifying prohibited subparagraphs and notes for each 
ASME BPV Code edition and addenda in tabular form as opposed to the 
narrative form of the existing regulation. No substantive change to the 
requirements is intended by this revision. The NRC believes that 
presenting the information in tabular form will increase the clarity 
and understandability of the regulation.
    The existing condition in Sec.  50.55a(b)(1)(ii) prohibits, for 
welds with leg sizes less than 1.09 tn, the use of certain 
Code provisions in ASME BPV Code, Section III, Division 1. The Code 
provisions provide stress indices for welded joints used in the design 
of Class 2 and Class 3 piping. The use of these indices is prohibited 
for welds with leg sizes less than 1.09 tn, where 
tn is the nominal pipe thickness because this would result 
in a weld that would be weaker than the pipe to which it is adjoined 
under these dimensions. The location of the prohibited provisions vary 
in the Code editions and addenda from the 1989 Addenda through the 2013 
Edition, so in this final rule the NRC clearly identifies the 
prohibited code provisions in the editions and addenda in a tabular 
format.
    As an editorial matter, this final rule identifies the prohibited 
ASME BPV Code provisions as ``notes,'' which is the term used by the 
ASME, rather than ``footnotes.'' The NRC is using the terminology used 
by the ASME for clarity.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
    The NRC is approving for use the version of NQA-1 referenced in the 
2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code, 
Section III, Subsection NCA, Article 7000, which this rule is also 
incorporating by reference. This allows applicants and licensees to use 
the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2010 
and later editions and addenda of Section III.
    In the 2010 Edition of ASME BPV Code, Section III, Subsection NCA, 
Article NCA-4000, ``Quality Assurance,'' was updated to require N-Type 
Certificate Holders to comply with the requirements of Part 1 of the 
2008 Edition and the 2009-1a Addenda of ASME Standard NQA-1, ``Quality 
Assurance Requirements for Nuclear Facility Applications,'' as modified 
and supplemented in NCA-4120(b) and NCA-4134. In addition, NCA-4110(b) 
was revised to remove the reference to a specific edition and addenda 
of ASME NQA-1, and Table NCA-7100-2, ``Standards and Specifications 
Referenced in Division 1,'' was updated to require the 2008 Edition and 
2009-1a Addenda of NQA-1 when using the 2010 Edition of Section III. In 
light of these changes, the NRC reviewed the 2008 Edition and the 2009-
1a Addenda of NQA-1 and compared it to previously approved versions of 
NQA-1 and found that there were no significant differences. In 
addition, the NRC reviewed the changes to Subsection NCA that reference 
the 2008 Edition and 2009-1a Addenda of NQA-1, compared them to 
previously approved versions of Subsection NCA, and found that there 
were no significant differences. Therefore, the NRC has concluded that 
these editions and addenda of NQA-1 are acceptable for use.
    The NRC is revising Sec.  50.55a(b)(1)(iv) to clarify that an 
applicant's or licensee's commitments addressing those areas where NQA-
1 either does not address a requirement in appendix B to 10 CFR part 
50, ``Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants,'' or is less stringent than the comparable 
appendix B requirement govern the applicant's or licensee's Section III 
activities. The clarification is consistent with Sec.  50.55a(b)(2)(x) 
and (b)(3)(i). The NQA-1 provides the ASME's method for establishing 
and implementing a quality assurance (QA) program for the design and 
construction of nuclear power plants and fuel reprocessing plants. 
However, NQA-1, as modified and supplemented in NCA-4120(b) and NCA-
4134, does not address some of the requirements of appendix B to 10 CFR 
part 50. In some cases, the provisions of NQA-1 are less stringent than 
the comparable appendix B requirements. Therefore, in order to meet the 
requirements of appendix B, an applicant's or licensee's QA program 
description must contain commitments addressing those provisions of 
appendix B which are not covered by NQA-1, as well as provisions that 
supplement or replace the NQA-1 provisions where the appendix B 
requirement is more stringent.
    Finally, the NRC is removing the reference in Sec.  
50.55a(b)(1)(iv) to versions of NQA-1 older than the 1994 Edition 
because the NRC did not receive any adverse comments from any applicant 
or licensee about removing versions of NQA-1 older than the 1994 
Edition from the regulation. The NRC received only one comment 
regarding NQA-1. The comment expressed support for incorporation by 
reference of NQA-1 and did not respond to the NRC's request for comment 
regarding the removal of references to older versions of NQA-1.
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification 
and Demonstration of Function of Incompressible-Fluid Pressure-Relief 
Valves
    The NRC is revising Sec.  50.55a(b)(1)(vii) so that the existing 
condition prohibiting the use of paragraph NB-7742(a)(2) of the 2006 
Addenda through the 2007 Edition, up to and including the 2008 Addenda, 
is extended to include the editions and addenda up to the 2013 Edition, 
which are the subject of this rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME 
Certification Marks
    The NRC is adding Sec.  50.55a(b)(1)(viii) to allow licensees to 
use either the ASME BPV Code Symbol Stamps of editions and addenda 
earlier than the 2011 Addenda to the 2010 Edition of the ASME BPV Code 
or the ASME Certification Marks with the appropriate

[[Page 32938]]

certification designators and class designators as specified in the 
2013 Edition through the latest edition and addenda incorporated by 
reference in Sec.  50.55a.
    The ASME BPV Code requires, in certain instances, that components 
be stamped. The stamp signifies that the component has been designed, 
fabricated, examined and tested, as specified in the ASME BPV Code. The 
stamp also signifies that the required ASME BPV Code data report forms 
have been completed, and the authorized inspector has inspected the 
item and authorized the application of the ASME BPV Code Symbol Stamp.
    The ASME has instituted changes in the BPV Code to consolidate the 
different ASME BPV Code Symbol Stamps into a common ASME Certification 
Mark. This action was implemented in the 2011 Addenda to the 2010 
Edition of the ASME BPV Code. As of the end of 2012, ASME no longer 
utilizes the ASME BPV Code Symbol Stamp. Licensees, however, may not 
have updated to the edition or addenda that identifies the use of the 
ASME Certification Mark. Nevertheless, licensees are legally required 
to implement the ASME BPV Code Edition and Addenda identified as their 
current code of record. As ASME components are procured, these 
components may be received with the ASME Certification Mark, while the 
licensee's current code of record may require the component to have the 
ASME BPV Code Symbol Stamp. Installation of a component under such 
circumstances would not be in compliance with the regulations that the 
licensees are required to meet.
    Both the ASME Certification Mark and the ASME BPV Code Symbol Stamp 
are official ASME methods of certifying compliance with the Code. 
Although these ASME Certification Marks differ slightly in appearance, 
they serve the same purpose of certifying code compliance by the ASME 
Certificate Holder and continue to provide for the same level of 
quality assurance for the application of the ASME Certification Mark as 
was required for the application of the ASME BPV Code Symbol Stamp. The 
new ASME Certification Mark represents a small, non-safety significant 
modification of ASME's trademark. As such, it does not change the 
technical requirements of the Code. The ASME has confirmed that the 
Certification Mark with designator is equivalent to the corresponding 
BPV Code Symbol Stamp. Based on statements made by ASME in a letter 
dated August 17, 2012, the NRC has concluded that the ASME BPV Code 
Symbol Stamps and ASME Certification Mark with code-specific 
designators are equivalent with respect to their certification of 
compliance with the BPV Code. The NRC discussed this issue in 
Regulatory Issue Summary 2013-07, ``NRC Staff Position on the Use of 
American Society of Mechanical Engineers Certification Mark,'' dated 
May 28, 2013.
10 CFR 50.55a(b)(1)(ix) Section III Condition: NPT Code Symbol Stamps
    The NRC is adding Sec.  50.55a(b)(1)(ix) to allow licensees to use 
the NPT Code Symbol Stamp with the letters arranged horizontally as 
specified in ASME BPV Code Case N-852 for the service life of a 
component that had the NPT Code Symbol Stamp applied during the time 
period from January 1, 2005, through December 31, 2015.
    Public comments on the use of the NPT Code Symbol requested that 
the NRC accept the NPT Code Symbol Stamp having the NPT letters 
arranged horizontally as an acceptable NPT Stamp to certify Code 
compliance for fabricated items that have already been stamped prior to 
receiving a replacement NPT Code Symbol Stamp from the ASME. The 
comments requested that the NRC include acceptance of Code Case N-852 
in this final rule for this purpose. Within the context of its Code 
rules, ASME asserts that the NPT Code Symbol Stamp having the NPT 
letters arranged horizontally, although differing slightly in 
appearance from the NPT Code Symbol Stamp as illustrated in Section 
III, Table NCA-8100-1 of the ASME BPV Code, 2010 Edition and earlier 
editions and addenda, serves the same purpose of certifying Code 
compliance by the ASME NPT Certificate Holder with confirmation by the 
Authorized Nuclear Inspector and provides the same level of quality 
assurance. In addition, ASME indicated that on or after January 1, 
2016, the ASME will no longer authorize use of the NPT Code Symbol 
Stamp having the NPT letters arranged horizontally. Accordingly, on or 
after January 1, 2016, fabricated items will only be stamped with the 
NPT Code Symbol Stamp as illustrated in Section III, Table NCA-8100-1 
of the ASME BPV Code, 2010 Edition and earlier editions and addenda.
    The NRC agrees in general with this comment, in which the ASME 
asserts that the ASME NPT Code Symbol Stamp with the letters arranged 
horizontally to be equivalent to the ``N over PT'' ASME NPT Code Symbol 
Stamp. Therefore, using either Code Symbol Stamp serves the same 
purpose of certifying code compliance by the ASME Certificate Holder 
with confirmation by the Authorized Nuclear Inspector and provides the 
same level of quality assurance. The NRC also notes that the same 
administrative and technical requirements in the ASME Code still apply 
whether an ASME NPT Code Symbol Stamp with the letters arranged 
horizontally or an ``N over PT'' ASME NPT Code Symbol Stamp is applied. 
However, since this NPT Code Symbol Stamp having the NPT letters 
arranged horizontally will only be applied onto fabricated components 
from the time period of January 1, 2005, through December 31, 2015, the 
time period for when this NPT Code Symbol Stamp was applied to the 
component should be limited to these dates to prevent inadvertent 
fraudulent material. Therefore, the NRC agrees that the ASME BPV Code 
Case N-852 is acceptable for the service life of the component that had 
the NPT Code Symbol stamp applied from the time period of January 1, 
2005, through December 31, 2015. In response to this comment, the NRC 
added Sec.  50.55a(b)(1)(ix) to include a statement that licensees may 
use the NPT Code Symbol Stamp with the letters arranged horizontally as 
specified in ASME BPV Code Case N-852 for the service life of a 
component that had the NPT Code Symbol Stamp applied during the time 
period from January 1, 2005, through December 31, 2015. The NRC is 
incorporating by reference ASME BPV Code Case N-852 in Sec.  
50.55a(a)(1)(iii)(F) because it is referenced in Sec.  
50.55a(b)(1)(ix).
    Although the proposed rule did not include this Code Case, the NRC 
has determined that the incorporation by reference of this Code Case at 
the final rule stage is a logical outgrowth of the proposed rule. The 
NRC's intent to ensure that Sec.  50.55a identify all ASME-approved 
methods for labelling Code components is apparent from the statement of 
considerations for the proposed rule. See 80 FR 56820 (September 18, 
2015) at 56823-56824. The NRC did not entirely achieve that purpose, 
and this resulted in public comments seeking approval of this Code 
Case, which supports the proposition that the public had a reasonable 
opportunity to either propose the correction, with conditions as the 
commenter believes are necessary or desirable, or to indicate why the 
(anticipated) correction should not be made. Therefore, the NRC 
concludes that it may incorporate by reference ASME BPV Code Case N-
852.

[[Page 32939]]

B. ASME BPV Code, Section XI

10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section 
XI
    In the proposed rule, the NRC proposed a revision to Sec.  
50.55a(a)(1)(ii) that would have clarified that Section XI Nonmandatory 
Appendix U of the 2013 Edition of ASME BPV Code, Section XI was not 
incorporated by reference and therefore not approved for use. After 
considering public comments, the NRC has determined that it will not 
exclude Appendix U from the incorporation by reference because it is 
the integration of ASME BPV Code Cases N-513-3, ``Evaluation Criteria 
for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 
Piping Section XI, Division 1,'' and N-705, ``Evaluation Criteria for 
Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 
Vessels and Tanks Section XI, Division 1,'' into Section XI. The NRC 
has approved Code Cases N-513-3 and N-705 in RG 1.147. However, as 
described in the discussion for Sec.  50.55a(b)(2)(xxxiv) in Section 
II.B, ``ASME BPV Code Section XI,'' the NRC has found it necessary to 
adopt two new conditions to the use of Nonmandatory Appendix U.
    The NRC is adopting two conditions in the language of Sec.  
50.55a(a)(1)(ii)(C)(52) and (53) to address two inconsistencies that 
were identified between the NRC's position in a proposed rule regarding 
the acceptability of ASME Code Cases (81 FR 10780; March 2, 2016) (2016 
Code Case proposed rule) and the proposed rule for this rulemaking (80 
FR 56820; September 18, 2015). The first inconsistency is that the 
NRC's proposed conditions on ASME BPV Code Case N-799, ``Dissimilar 
Metal Welds Joining Vessel Nozzles to Components,'' in the 2016 Code 
Case proposed rule were not reflected in the 2015 proposed rule for 
this rulemaking, even though the technical content of ASME BPV Code 
Case N-799 has been incorporated into the 2011 Addenda and 2013 Edition 
of ASME BPV Code, Section XI. The second inconsistency is that the 
NRC's proposed disapproval of ASME BPV Code Case N-813, ``Alternative 
Requirements for Preservice Volumetric and Surface Examination,'' in 
the 2016 Code Case proposed rule was not reflected in the 2015 proposed 
rule for this rulemaking, even though the technical content of ASME BPV 
Code Case N-813 has been incorporated into the 2013 Edition of the ASME 
BPV Code, Section XI as IWB-3112(a)(3) and IWC-3112(a)(3). To address 
these two inconsistencies, the NRC is excluding these ASME BPV Code, 
Section XI items from incorporation by reference, as reflected in Sec.  
50.55a(a)(1)(ii)(C)(52) and (53) of the final rule. The NRC plans to 
complete the development of the regulatory approaches for examination 
of component-to-component welds for new construction plants and the 
acceptance of preservice flaws by analytical evaluation for operating 
plants and include them in a future rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and 
Addenda of Subsection IWE and Subsection IWL
    The NRC is revising Sec.  50.55a(b)(2)(vi) to expressly state that 
licensees that implemented the expedited examination of containment 
during the 5-year period from September 9, 1996, to September 9, 2001, 
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), 
when implementing the initial 120-month inspection interval for the 
containment ISI requirements of this section.
    The expedited examination involved the completion of the first set 
of examinations of the first or initial 120-month containment 
inspection interval. It is noted that all of the operating reactors in 
the previously stated class would have gone past their initial 120-
month inspection interval by 2011. The change removes the possibility 
of misinterpretation of the provision as requiring plants that do not 
fall in the previously stated class, such as reactors licensed after 
September 9, 2001, to use the 1992 Edition with 1992 Addenda or the 
1995 Edition with 1996 Addenda of Subsection IWE and Subsection IWL, 
Section XI for implementing the initial 120-month inspection interval 
of the containment ISI program. Applicants and licensees that do not 
fall in the previously stated class must use Code editions and addenda 
in accordance with Sec.  50.55a(g)(4)(i) and (ii), respectively, for 
the initial and successive 120-month containment ISI intervals.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment 
Examinations
    The NRC is revising Sec.  50.55a(b)(2)(viii) by removing the 
condition for using the 2007 Edition with 2009 Addenda through the 2013 
Edition of Subsection IWL requiring compliance with Sec.  
50.55a(b)(2)(viii)(E). To support the removal of the condition, the NRC 
is adding new requirements governing the performance and documentation 
of concrete containment examinations in Sec.  50.55a(b)(2)(viii)(H) and 
(I), which are discussed separately in the next two headings.
    Section 50.55a(b)(2)(viii)(E) is one of several conditions that 
apply to the inservice examination of concrete containments using 
Subsection IWL of various editions and addenda of the ASME BPV Code, 
Section XI, incorporated by reference in Sec.  50.55a(a)(1)(ii). The 
NRC is removing the condition in Sec.  50.55a(b)(2)(viii)(E) when 
applying the 2007 Edition with 2009 Addenda through the 2013 Edition of 
Subsection IWL because its intent has been incorporated into the Code 
in the new provision IWL-2512, ``Inaccessible Areas.''
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(viii)(H) to specify the 
information that must be provided in the ISI Summary Report required by 
IWA-6000, when inaccessible concrete surfaces are evaluated under the 
new Code provision IWL-2512. This new condition replaces the existing 
condition in Sec.  50.55a(b)(2)(viii)(E), when using the 2007 Edition 
with the 2009 Addenda through the 2013 Edition of Subsection IWL.
    The existing condition in Sec.  50.55a(b)(2)(viii)(E) of the 
current rule requires that, for Class CC applications, the licensee 
shall evaluate the acceptability of inaccessible areas when conditions 
exist in accessible areas that could indicate the presence of or result 
in degradation to such inaccessible areas, and provide the evaluation 
information required by Sec.  50.55a(b)(2)(viii)(E)(1), (2), and (3) in 
the IWA-6000 ISI Summary Report.
    In the 2009 Addenda Subsection IWL, the ASME revised existing 
provisions IWL-1220 and IWL-2510 and added the new provision IWL-2512 
intended to incorporate the condition in Sec.  50.55a(b)(2)(viii)(E) 
into Subsection IWL. The IWL-2510, ``Surface Examination,'' was 
restructured into new paragraphs in IWL-2511, ``Accessible Areas,'' 
with almost the same provisions as the previous IWL-2510 and IWL-2512, 
``Inaccessible Areas,'' to be specific to examinations required for 
accessible areas, and differentiate between those and the new 
requirements for inaccessible areas. The inaccessible areas addressed 
by the new IWL-2512 are: (1) Concrete surfaces obstructed by adjacent 
structures, parts or appurtenances (e.g., generally above-grade 
inaccessible areas); and (2)

[[Page 32940]]

concrete surfaces made inaccessible by foundation material or backfill 
(e.g., below-grade inaccessible areas).
    The revised IWL-2511(a) has a new requirement that states that, 
``If the Responsible Engineer determines that observed suspect 
conditions indicate the presence of, or could result in, degradation of 
inaccessible areas, the requirements of IWL-2512(a) shall be met.'' The 
new IWL-2512(a) requires the ``Responsible Engineer'' to evaluate 
suspect conditions and specify the type and extent of examinations, if 
any, required to be performed on inaccessible surface areas described 
in the previous paragraph. The acceptability of the evaluated 
inaccessible area would be determined either based on the evaluation or 
based on the additional examinations, if determined to be required. The 
new IWL-2512(b) further requires a periodic technical evaluation of 
below-grade inaccessible areas of concrete to be performed to determine 
and manage its susceptibility to degradation regardless of whether 
suspect conditions exist in accessible areas that would warrant an 
evaluation of inaccessible areas based on the condition in Sec.  
50.55a(b)(2)(viii)(E). Therefore, the revised IWL-2511(a) and new IWL-
2512 code provisions address the evaluation and acceptability of 
inaccessible areas consistent with the existing condition in Sec.  
50.55a(b)(2)(viii)(E), with one exception. The exception is that the 
new IWL-2512 provision does not explicitly require the information 
specified in Sec.  50.55a(b)(2)(viii)(E)(1), (2), and (3) of the 
existing condition to be provided in the IWA-6000 ISI Summary Report.
    For these reasons, the NRC is identifying the information that must 
be provided in the ISI Summary Report required by IWA-6000 when 
inaccessible concrete surfaces are evaluated under the new code 
provision IWL-2512. This new condition replaces the existing condition 
in Sec.  50.55a(b)(2)(viii)(E) when using the 2007 Edition with the 
2009 Addenda through the 2013 Edition of Subsection IWL. The 
information required by the new condition must be provided when 
inaccessible concrete areas are evaluated per IWL-2512(a) for 
degradation based on suspect conditions found in accessible areas, as 
well as when periodic technical evaluations of inaccessible below-grade 
concrete areas required by IWL-2512(b) are performed.
10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(viii)(I) to place a condition 
on the periodic technical evaluation requirements in the new IWL-
2512(b), for consistency with NUREG-1801, Revision 2, ``Generic Aging 
Lessons Learned (GALL) Report,'' with regard to aging management of 
below-grade containment concrete surfaces. The new IWL-2512(b) 
provision is applicable to inaccessible below-grade concrete surfaces 
exposed to foundation soil, backfill, or groundwater. This condition 
would apply only during the period of extended operation of a renewed 
license under 10 CFR part 54, when using IWL-2512(b) of the 2007 
Edition with 2009 Addenda through the 2013 Edition of Subsection IWL.
    In the 2009 Addenda of Subsection IWL, the ASME added new Code 
provisions, IWL-2512(b) and (c) as well as a new line item L1.13 in 
Table IWL-2500-1, intended to specifically address aging management 
concerns with potentially unidentified degradation of inaccessible 
below-grade containment concrete areas and to be responsive to actions 
outlined in the GALL Report related to aging management of inaccessible 
below-grade concrete surfaces. It is noted that these new Code 
provisions are an enhancement to the requirement of the existing 
condition in Sec.  50.55a(b)(2)(viii)(E) to specifically address aging 
management of inaccessible below-grade containment concrete areas and 
is generally acceptable to the NRC.
    The new IWL-2512(b) provides requirements for systematically 
performing a periodic technical evaluation of concrete surfaces exposed 
to foundation soil, backfill, or groundwater to determine 
susceptibility of the concrete to deterioration that could affect its 
ability to perform its intended design function under conditions 
anticipated through the service life of the structure. It requires the 
technical evaluation to be performed and documented at periodic 
intervals not to exceed 10 years regardless of whether conditions exist 
in accessible areas that would warrant an evaluation of inaccessible 
areas by the existing condition in Sec.  50.55a(b)(2)(viii)(E), which 
the NRC finds reasonable for the initial 40-year operating license 
period. The new IWL-2512(b) further provides the specific elements, 
including aging mechanisms considered, that the technical evaluation 
should include, as well as the definition of an aggressive below-grade 
environment. The new IWL-2512(c) requires that the evaluation results 
of IWL-2512(b) be used to define and document the condition monitoring 
program, if determined to be required, including required examinations 
and frequencies, to be implemented for the management of degradation 
and aging effects of the below-grade concrete surface areas. If it is 
determined that additional examinations are required, these 
examinations of inaccessible below-grade areas will be implemented in 
accordance with new line item L1.13 in Table IWL-2500-1 under 
Examination Category L-A, Concrete, with acceptance criteria based on 
IWL-3210. It should be noted that a technical evaluation approach, such 
as in IWL-2512(b), could be used, and is generally used, to determine 
acceptability of a below-grade inaccessible area to satisfy the 
condition in Sec.  50.55a(b)(2)(viii)(E).
    The technical evaluation requirements in IWL-2512(b) assist in 
determining the susceptibility to degradation and manage aging effects 
of inaccessible below-grade concrete surfaces, before the loss of 
intended function. The requirements are based on, and are generally 
consistent with, the guidance in the GALL Report, with the following 
two exceptions. The first exception is that IWL-2512(b) requires the 
technical evaluation to determine the susceptibility of the concrete to 
degradation and the ability to perform the intended design function 
through its service life at periodic intervals not to exceed 10 years. 
The aging management programs (AMPs) for safety-related structures 
(e.g., Structures Monitoring) in the GALL Report require such 
evaluation to be performed at intervals not to exceed 5 years, which is 
also consistent with applicant commitments during review of license 
renewal applications. The second exception is that IWL-2512(b) requires 
that examination of representative samples of below-grade concrete be 
performed if excavated for any reason when an aggressive below-grade 
environment is present. However, the NRC notes that the AMPs (X1.S6 
Structures Monitoring and X1.S7 Water Control Structures) in the GALL 
Report require the same examination even for a non-aggressive below-
grade environment.
    Based on these reasons, the NRC is adding Sec.  
50.55a(b)(2)(viii)(I) to place a condition on the periodic technical 
evaluation requirements in IWL-2512(b) for consistency with the GALL 
Report, when addressing the two exceptions previously described with 
respect to aging management of inaccessible below-grade concrete 
components of the

[[Page 32941]]

containment. The new condition requires that, during the period of 
extended operation of a renewed license, the technical evaluation under 
IWL-2512(b) of inaccessible below-grade concrete surfaces exposed to 
foundation soil, backfill, or groundwater be performed at periodic 
intervals not to exceed 5 years, as opposed to the 10-year interval in 
IWL-2512. In addition, the condition requires the examination of 
representative samples of the exposed portions of the below-grade 
concrete be performed when excavated for any reason as opposed to IWL-
2512, which limits the examination to excavations in aggressive, below-
grade environments. Since the GALL Report is the technical basis 
document for license renewal, this new condition applies only during 
the period of extended operation of a renewed license under 10 CFR part 
54, when using IWL-2512(b) of the 2007 Edition with 2009 Addenda 
through the 2013 Edition of Subsection IWL, Section XI.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment 
Examinations
    The NRC is extending the applicability of the existing conditions 
in Sec.  50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J), governing 
examinations of metal containments and the liners of concrete 
containments under Subsection IWE, to the ASME BPV Code editions and 
addenda which are the subject of this rulemaking (i.e., the 2007 
Edition with 2009 Addenda through the 2013 Edition). The last sentence 
of Sec.  50.55a(b)(2)(ix) prior to this final rule stated that the 
referenced conditions were applicable only to addenda, but not to 
editions, approved by the NRC after the 2007 Edition of the ASME BPV 
Code. To rectify this, the NRC is revising the last sentence of Sec.  
50.55a(b)(2)(ix) to refer to the latest ``edition and'' addenda after 
the 2007 Edition which are incorporated by reference into Sec.  50.55a.
    The NRC reviewed the Code changes in Subsection IWE of the 2009 
Addenda through the 2013 Edition of ASME BPV Code, Section XI, and 
noted that all of the changes were editorial or administrative with the 
intent to improve the clarity of the existing requirements or correct 
errors by errata. There were no changes to Subsection IWE in the Code 
editions and addenda that are the subject of this rulemaking that the 
NRC believes would require new regulatory conditions to ensure safety, 
nor do the changes to Subsection IWE address the NRC's reasons for 
adopting the conditions on the use of Subsection IWE. Accordingly, the 
NRC is extending the applicability of the existing conditions (by 
adding the words ``edition and'' to Sec.  50.55a(b)(2)(ix) as 
discussed) without any change to the provisions of the conditions.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
    The NRC is approving for use the version of NQA-1 referenced in the 
2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition of the 
ASME BPV Code, Section XI, Table IWA 1600-1, ``Referenced Standards and 
Specifications,'' which this rule is also incorporating by reference. 
This allows, but does not require, licensees to use the 1994 Edition or 
the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2009 
Addenda and later editions and addenda of Section XI.
    In the 2013 Edition of ASME BPV Code, Section XI, Table IWA 1600-1 
was updated to allow licensees to use the 1994 Edition or the 2008 
Edition with the 2009-1a Addenda of NQA-1 when using the 2013 Edition 
of Section XI. In the 2010 Edition of ASME BPV Code, Section XI, IWA-
1400, ``Owner's Responsibilities,'' Subparagraph (n)(2) was updated to 
reference the NQA-1 Part I, Basic Requirements and Supplementary 
Requirements for Nuclear Facilities. In the 2009 Addenda of the 2007 
Edition of ASME BPV Code, Section XI, Table IWA-1600-1, ``Referenced 
Standards and Specifications,'' was updated to allow licensees to use 
the 1994 Edition of NQA-1. The NRC reviewed the 2008 Edition and the 
2009-1a Addenda of NQA-1 and compared it to previously approved 
versions of NQA-1 and found that there were no significant differences. 
Therefore, the NRC has concluded that these editions and addenda of 
NQA-1 are acceptable for use.
    The NRC is amending Sec.  50.55a(b)(2)(x) to clarify that a 
licensee's commitments addressing those areas where NQA-1 either does 
not address a requirements in appendix B to 10 CFR part 50 or is less 
stringent than the comparable appendix B requirement govern the 
licensee's Section XI activities. The clarification is consistent with 
Sec.  50.55a(b)(1)(iv) and (b)(3)(i). The ASME's method for 
establishing and implementing a QA program for the design and 
construction of nuclear power plants and fuel reprocessing plants is 
described in NQA-1. However, NQA-1 does not address some of the 
requirements of appendix B to 10 CFR part 50. In some cases, the 
provisions of NQA-1 are less stringent than the comparable appendix B 
requirements. Therefore, in order to meet the requirements of appendix 
B, a licensee's QA program description must contain commitments 
addressing those provisions of appendix B which are not covered by NQA-
1, as well as provisions that supplement or replace the NQA-1 
provisions where the appendix B requirement is more stringent.
    Finally, the NRC is removing the reference in Sec.  50.55a(b)(2)(x) 
to versions of NQA-1 older than the 1994 Edition because the NRC did 
not receive any adverse comments from any applicant or licensee 
regarding concerns about removing versions of NQA-1 older than the 1994 
Edition from the regulation. The NRC received only one comment 
regarding NQA-1. The comment expressed support for incorporation by 
reference of NQA-1 and did not respond to the NRC's request for comment 
regarding the removal of references to older versions of NQA-1.
10 CFR 50.55a(b)(2)(xii) Section XI Condition: Underwater Welding
    The NRC is revising Sec.  50.55a(b)(2)(xii) to allow underwater 
welding on irradiated materials in accordance with IWA-4660, 
``Underwater Welding,'' of Section XI, 1997 Addenda through the latest 
edition and addenda incorporated by reference in Sec.  
50.55a(a)(1)(ii). The conditions for which underwater welding would be 
permitted without prior NRC approval are based on technical factors, 
such as neutron fluence and, for certain material classes, helium 
concentration.
    The existing condition in Sec.  50.55a(b)(2)(xii) does not allow 
underwater welding on irradiated materials by prohibiting the use of 
IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through 
the latest edition and addenda incorporated by reference in Sec.  
50.55a(a)(1)(ii) on materials that are irradiated; however, there are 
two problems with the restriction in Sec.  50.55a(b)(2)(xii). First, 
the neutron fluence threshold above which a material is considered to 
be irradiated is not defined in Sec.  50.55a(b)(2)(xii). Second, 
studies such as those documented in Boiling Water Reactor Vessel and 
Internals Project (BWRVIP) Report 1003020 (BWRVIP-97) have shown that 
reactor internals can tolerate some neutron irradiation without 
suffering damage to weldability, as long as the helium concentration in 
the material does not exceed a certain threshold. The NRC completed its 
Safety Evaluation of BWRVIP-97 in May 2008 and concluded that 
implementation of the guidelines in the BWRVIP-97 report, with some 
modifications as documented in the

[[Page 32942]]

NRC Safety Evaluation dated June 30, 2008, will provide an acceptable 
technical basis for the design of weld repairs based on the helium 
content of irradiated reactor vessel internals. The current version of 
Sec.  50.55a(b)(2)(xii) does not define a threshold of helium 
concentration below which the material is considered to be weldable.
    The most recent editions of the ASME BPV Code state in Article IWA-
4660 that underwater welding may not be performed on irradiated 
materials other than P-No. 8 materials containing less than 0.1 atomic 
parts per million (appm) measured or calculated helium content 
generated through irradiation. Some editions and addenda of the ASME 
BPV Code prior to 2010 state in Article IWA-4660 that underwater 
welding may only be performed in applications not predicted to exceed a 
thermal neutron fluence of 1 x 10\17\ n/cm\2\. Other editions and 
addenda of the ASME BPV Code prior to 2010 do not restrict the 
underwater welding of irradiated materials. Therefore, there is 
inconsistent treatment among the various editions and addenda of the 
ASME BPV Code on the underwater welding of irradiated materials.
    Current ASME BPV Code and Code Case requirements for welding on 
irradiated materials, other than the underwater welding requirements 
specified in IWA-4660, are inconsistent. Thresholds for weldability may 
be stated in terms of fast neutron fluence, thermal neutron fluence, or 
helium concentration. In some cases, thresholds are not defined and the 
Code or Code Case simply states that consideration must be given to 
irradiation effects when welding. The NRC believes that thresholds for 
welding on irradiated materials should be based on the current 
understanding of irradiation damage, as supported by technical studies 
(such as BWRVIP-97) which have been evaluated by the NRC. In addition, 
the NRC believes that these thresholds should be consistently applied 
for all Code and Code Case applications.
    During the public comment period for this rulemaking, a 
representative of ASME recommended that Sec.  50.55a(b)(2)(xii) be 
revised such that it applies only to those editions and addenda earlier 
than the 2010 Edition. The effect of such a revision would be to allow 
welding on P-No. 8 materials containing less than 0.1 appm measured or 
calculated helium content generated through irradiation. However, this 
proposed revision would not be consistent with other ASME BPV Code or 
Code Case requirements for welding on irradiated materials, and this 
proposed revision does not address standards for welding on material 
classes other than P-No. 8. Instead the NRC is adopting conditions that 
would apply to all materials and which can be consistently applied for 
all Code and Code Case applications. The first condition, Sec.  
50.55a(b)(2)(xii)(A), is based on fast neutron fluence and applies to 
ferritic materials. The second condition, Sec.  50.55a(b)(2)(xii)(B), 
is based on helium content and/or thermal fluence and applies to 
austenitic materials. For P-No. 8 austenitic materials, the evaluation 
of BWRVIP-97 supports a weldability threshold based on helium content 
and thermal fluence. For austenitic materials other than P-No. 8, there 
are insufficient data to support a weldability threshold based on 
helium content, and, therefore, the NRC is adopting a weldability 
threshold based on thermal fluence only.
    The conditions for which underwater welding are permitted, as 
stated in the revision of Sec.  50.55a(b)(2)(xii), were determined, in 
part, based on technical discussions in a Category 2 public meeting 
with industry representatives held on January 19, 2016. The NRC later 
presented the new conditions at a public meeting held on March 2, 2016. 
There were no comments on this change from the attendees at the March 
2, 2016, public meeting. Summaries of the January 19 and March 2, 2016, 
public meetings are available in ADAMS under Accession Nos. ML16050A383 
and ML16069A408, respectively.
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(xviii)(D) to prohibit 
applicants and licensees from using the ultrasonic examination 
nondestructive examination (NDE) personnel certification requirements 
in Section XI, Appendix VII and Subarticle VIII-2200 of the 2011 
Addenda and 2013 Edition of the ASME BPV Code. Paragraph (b)(2)(xviii) 
currently includes conditions on the certification of NDE personnel. In 
addition, the new paragraph will require applicants and licensees to 
use the 2010 Edition, Table VII-4110-1 training hour requirements for 
Levels I, II, and III ultrasonic examination personnel, and the 2010 
Edition, Subarticle VIII-2200 of Appendix VIII prerequisites for 
personnel requirements. In the 2011 Addenda and 2013 Edition, the ASME 
BPV Code added an accelerated Appendix VII training process for 
certification of ultrasonic examination personnel based on training and 
prior experience, and separated the Appendix VII training requirements 
from the Appendix VIII qualification requirements. These new ASME BPV 
Code provisions will provide personnel in training with less experience 
and exposure to representative flaws in representative materials and 
configurations common to operating nuclear power plants, and they would 
permit personnel with prior non-nuclear ultrasonic examination 
experience to qualify for examinations in nuclear power plants without 
exposure to the variety of defects, examination conditions, components, 
and regulations common to operating nuclear power plants.
    The impact of reduced training and nuclear power plant 
familiarization is unknown. The ASME BPV Code supplants training hours 
and field experience without a technical basis, minimum defined 
training criteria, process details, or standardization. For these 
reasons, the NRC is prohibiting the use of Appendix VII and Subarticle 
VIII-2200 of the 2011 Addenda and 2013 Edition. The NRC is requiring 
applicants and licensees using the 2011 Addenda and 2013 Edition to use 
the prerequisites for ultrasonic examination personnel certifications 
in Table VII-4110-1 and Subarticle VIII-2200, Appendix VIII in the 2010 
Edition.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements: 
First Provision
    The NRC is revising Sec.  50.55a(b)(2)(xxi)(A) to modify the 
standard for visual magnification resolution sensitivity and contrast 
for visual examinations performed on Examination Category B-D 
components instead of ultrasonic examinations, making the rule conform 
with ASME BPV Code, Section XI requirements for VT-1 examinations. The 
character recognition rules are used in ASME BPV Code, Section XI, 
Table IWA-2211-1 for VT-1 tests, and are the standard tests used for 
resolution and contrast checks of the VT-1 equipment. This revision 
essentially removed a requirement that was an addition to ASME BPV Code 
that required 1-mil wires to be used in licensees' Sensitivity, 
Resolution, and Contrast Standard targets. In 2004, the NRC published 
NUREG/CR-6860, ``An Assessment of Visual Testing,'' showing that a 
linear target, such as a wire, is not an effective method for testing 
the resolution of a video camera system. In addition, Boiling Water 
Reactor Vessel and Internals Project Report 105696 (BWRVIP-03) was 
changed to eliminate a \1/2\ mil wire from the Sensitivity, Resolution, 
and Contrast Standards due to similar concerns.

[[Page 32943]]

    Simple line detection can be a poor performance standard, allowing 
detection of a highly blurred image. This does not emulate sharpness 
quality recognition for evaluation of weld discontinuities. The 750 
[mu]m (30 mil) and the even smaller 25 [mu]m (1 mil) widths should not 
be used as performance standards because they do not determine image 
sharpness. This technique only measures the ``visible minimum'' for 
long linear indications, and does not measure a system's resolution or 
recognition limits. If the wire, or printed line, has a strong enough 
contrast against the background, then a linear feature well below the 
resolution of a system can be detected.
10 CFR 50.55a(b)(2)(xxiii) Section XI Condition: Evaluation of 
Thermally Cut Surfaces
    The NRC is revising Sec.  50.55a(b)(2)(xxiii) to clarify that this 
condition, prohibiting the ASME BPV Code provisions allowing 
elimination of mechanical processing of thermally cut surfaces under 
certain circumstances, only applies to the 2001 Edition through the 
2009 Addenda.
10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator 
Preservice Examinations
    In the proposed rule, the NRC proposed adding Sec.  
50.55a(b)(2)(xxx), with a condition regarding steam generator 
preservice examinations. The NRC received requests for clarification of 
the proposed condition, including elaboration on the kind of preservice 
examination that should be performed. The NRC agrees with the need for 
this clarification; however, during the development of the final rule, 
the NRC determined that additional time was needed to evaluate this 
proposed condition. Therefore, to ensure that this rulemaking is 
concluded as timely as possible, the NRC is not including this 
condition in this final rule and will address the need for a condition 
in a future rulemaking. The NRC has concluded that omitting this 
condition does not present a health or safety concern because licensees 
are currently performing appropriate steam generator preservice 
inspections under existing programs.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping 
Devices
    The NRC is adding Sec.  50.55a(b)(2)(xxxi) to require the use of 
Nonmandatory Appendix W when using a mechanical clamping device on an 
ASME BPV Code Class piping system. This condition, in part, clearly 
prohibits the use of mechanical clamping devices on small item Class 1 
piping and portions of piping systems that form the containment 
boundary. This condition also maintains the previously required design 
and testing requirements for the implementation of mechanical clamping 
devices on ASME BPV Code Class piping systems.
    In the 2010 Edition of the ASME BPV Code, a change was made to 
include mechanical clamping devices under the small items exclusion 
rules of IWA-4131. Currently in the 2007 Edition/2008 Addenda of 
Section XI under IWA-4133, ``Mechanical Clamping Devices Used as Piping 
Pressure Boundary,'' mechanical clamping devices may be used only if 
they meet the requirements of Mandatory Appendix IX of Section XI of 
the ASME BPV Code. Article IX-1000 (c) of Appendix IX prohibits the use 
of mechanical clamping devices on (1) Class 1 piping and (2) portions 
of a piping system that form the containment boundary.
    In the 2010 Edition, IWA-4133 was modified to allow use of IWA-
4131.1(c) for the installation of mechanical clamping devices. This 
change allowed the use of small items exclusion rules in the 
installation of mechanical clamping devices. Subparagraph IWA-4131.1(c) 
was added such that mechanical clamping devices installed on items 
classified as ``small items'' under IWA-4131, including Class 1 piping 
and portions of a piping system that form the containment boundary, 
would be allowed without a repair/replacement plan, pressure testing, 
services of an Authorized Inspection Agency, and completion of the NIS-
2 form. The NRC, in accordance with the previously approved IWA-4133 of 
the 2007 Edition/2008 Addenda of the ASME BPV Code, does not believe 
that the ASME has provided a sufficient technical basis to support the 
use of mechanical clamping devices on Class 1 piping or portions of a 
piping system that form the containment boundary as a permanent repair. 
Furthermore, the NRC finds that the ASME has not provided any basis for 
the small item exemption allowing the installation of mechanical clamps 
on these components. In the 2011 Addenda of the ASME BPV Code, IWA-
4131.1(c) was relocated to IWA-4131.1(d). To add clarity to the 
condition, the NRC has included statements such that the implementation 
of these paragraphs is now prohibited.
    In the 2013 Edition, Mandatory Appendix IX of Section XI of the 
ASME BPV Code was changed to Nonmandatory Appendix W of Section XI of 
the ASME BPV Code. The NRC found insufficient basis to make this 
change, removing the mandatory requirements for the use of mechanical 
clamping devices on ASME BPV Code Class piping systems. By taking this 
action, the ASME BPV Code now allows mechanical clamping devices to be 
installed in various methods through interpretations of the ASME BPV 
Code that do not maintain the requirements for design and testing of 
the formerly mandatory Appendix IX. Therefore, to clarify the 
requirement for the implementation of mechanical clamps in ASME BPV 
Code Class systems, the NRC requires the use of Appendix W of Section 
XI when using mechanical clamping devices, and prohibits the use of 
mechanical clamping devices on small item Class 1 piping and portions 
of a piping system that form the containment boundary, as would 
otherwise be permitted under IWA-4131.1(c) in the 2010 Edition and IWA-
4131.1(d) in the 2011 Addenda through 2013 Edition.
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report 
Submittal
    The NRC is adding Sec.  50.55a(b)(2)(xxxii) to require licensees 
using the 2010 Edition and later editions and addenda of Section XI to 
continue to submit Summary Reports as required in IWA-6240 of the 2009 
Addenda.
    Prior to the 2010 Edition, Section XI required the preservice 
summary report to be submitted prior to the date of placement of the 
unit into commercial service, and the inservice summary report to be 
submitted within 90 calendar days of the completion of each refueling 
outage. In the 2010 Edition, IWA-6240 was revised to state, ``Summary 
reports shall be submitted to the enforcement and regulatory 
authorities having jurisdiction at the plant site, if required by these 
authorities.'' This change in the 2010 Edition could lead to confusion 
as to whether or not the summary reports need to be submitted to the 
NRC, as well as the time for submitting the reports, if they were 
required. The NRC concludes that summary reports must continue to be 
submitted to the NRC in a timely manner because they provide valuable 
information regarding examinations performed, conditions noted, 
corrective actions taken, and the implementation status of preservice 
inspection and ISI programs. Therefore, the NRC is adding Sec.  
50.55a(b)(2)(xxxii) to ensure that preservice and inservice summary 
reports will continue to be submitted within the timeframes currently

[[Page 32944]]

established in Section XI editions and addenda prior to the 2010 
Edition.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed 
Allowable Pressure
    The NRC is adding Sec.  50.55a(b)(2)(xxxiii) to prohibit the use of 
Appendix G, Paragraph G-2216, in the 2011 Addenda and later editions 
and addenda of the ASME BPV Code, Section XI. The 2011 Addenda of the 
ASME BPV Code included, for the first time, a risk-informed methodology 
to compute allowable pressure as a function of inlet temperature for 
reactor heat-up and cool-down at rates not to exceed 100 degrees F/hr 
(56 degrees C/hr). This methodology was developed based upon 
probabilistic fracture mechanics (PFM) evaluations that investigated 
the likelihood of reactor pressure vessel (RPV) failure based on 
specific heat-up and cool-down scenarios.
    During the ASME's consideration of this change, the NRC staff noted 
that additional requirements would need to be placed on the use of this 
alternative. For example, the NRC staff indicated that it would be 
important for a licensee who wishes to utilize such a risk-informed 
methodology for determining plant-specific pressure-temperature limits 
to ensure that the material condition of its facility is consistent 
with assumptions made in the PFM evaluations that supported the 
development of the methodology. One aspect of this would be evaluating 
plant-specific ISI data to determine whether the facility's RPV flaw 
distribution was consistent with the flaw distribution assumed in the 
supporting PFM evaluations. This consideration is consistent with a 
similar requirement established by the NRC in Sec.  50.61a, 
``Alternative Fracture Toughness Requirements for Protection against 
Pressurized Thermal Shock Events.'' The PFM methodology that supports 
Sec.  50.61a is very similar to that which was used to support ASME BPV 
Code, Section XI, Appendix G, Paragraph G-2216. These concerns with the 
Paragraph G-2216 methodology for computing allowable pressure as a 
function of inlet temperature for reactor heat-up and cooldown were not 
addressed by the ASME. Accordingly, the NRC is prohibiting the use of 
Paragraph G-2216 in Appendix G of the 2010 Edition. The continued use 
of the deterministic methodology of Section XI, Appendix G to generate 
Pressure-Temperature (P-T) limits remains acceptable.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Nonmandatory Appendix 
U
    The NRC is adding Sec.  50.55a(b)(2)(xxxiv) to require that two 
conditions, (A) and (B), be satisfied when using Nonmandatory Appendix 
U of the 2013 Edition of the ASME BPV Code, Section XI. In the proposed 
rule, the NRC had proposed to exclude Nonmandatory Appendix U from the 
incorporation by reference and therefore not approve it for use. After 
considering public comments, the NRC has incorporated by reference 
Appendix U in this final rule because it integrates ASME BPV Code Cases 
N-513-3, ``Evaluation Criteria for Temporary Acceptance of Flaws in 
Moderate Energy Class 2 or 3 Piping Section XI, Division 1,'' and N-
705, ``Evaluation Criteria for Temporary Acceptance of Degradation in 
Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division 
1,'' into Section XI. The NRC has approved the use of ASME BPV Code 
Cases N-513-3 and N-705 in RG 1.147, which allows licensees to use 
these code cases without prior permission from the NRC.
    The first condition on the use of Appendix U is set forth in Sec.  
50.55a(b)(2)(xxxiv)(A) of this final rule and requires that an ASME BPV 
Code repair or replacement activity temporarily deferred under the 
provisions of Nonmandatory Appendix U to the 2013 Edition of the ASME 
BPV Code, Section XI, must be performed during the next scheduled 
outage. This condition is consistent with the NRC's condition on the 
use of ASME BPV Code Case N-513-3 in RG 1.147, Revision 17. Appendix U 
defines that the evaluation period is the operational time for which 
the temporary acceptance criteria are satisfied but not exceeding 26 
months from the initial discovery of the condition. Original versions 
of ASME BPV Code Case N-513 stated, in part, that certain flaws may be 
acceptable without performing a repair/replacement activity for a 
limited time, not to exceed the time to the next scheduled outage. The 
NRC staff found that the acceptance of ASME BPV Code Case N-513 was 
based on allowing continued plant operation with a monitored and 
evaluated low safety significant degraded condition for a limited time 
until plant shutdown. By allowing use of this Appendix, this option is 
allowed rather than requiring an unnecessary plant shutdown to repair 
the degradation. However, the NRC believes once the plant is shut down, 
the degraded piping must be repaired.
    The second condition on the use of Appendix U is set forth in Sec.  
50.55a(b)(2)(xxxiv)(B) of this final rule. This paragraph requires the 
use of the mandatory appendix in ASME BPV Code Case N-513-3 in lieu of 
the appendix referenced in paragraph U-S1-4.2.1(c) of Appendix U (which 
was inadvertently omitted from Appendix U). The NRC is incorporating by 
reference the mandatory appendix in ASME BPV Code Case N-513-3 in Sec.  
50.55a(a)(1)(iii)(A) because it is referenced in Sec.  
50.55a(b)(2)(xxxiv)(B).
    A proposed condition on Disposition of Flaws in Class 3 Components, 
which was located in Sec.  50.55a(b)(2)(xxxiv) of the proposed rule, is 
not included in this final rule based on public comments that the error 
has been corrected by ASME in published erratum.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 
in the KIa and KIc Equations
    The NRC is adding Sec.  50.55a(b)(2)(xxxv) to specify that when 
licensees use the 2013 Edition of the ASME BPV Code, Section XI, 
Appendix A, Paragraph A-4200, if T0 is available, then 
RTT0 may be used in place of RTNDT for 
applications using the KIc equation and the associated 
KIc curve, but not for applications using the KIa 
equation and the associated KIa curve.
    Nonmandatory Appendix A provides a procedure based on linear 
elastic fracture mechanics (LEFM) for determining the acceptability of 
flaws that have been detected during inservice inspections that exceed 
the allowable flaw indication standards of IWB-3500. Sub-article A-4200 
provides a procedure for determining fracture toughness of the material 
used in the LEFM analysis. The NRC staff's concern is related to the 
proposed insertion regarding an alternative based on the use of the 
Master Curve methodology to determine the nil-ductility transition 
reference temperature RTNDT, which is an important parameter 
in determining the fracture toughness of the material. Specifically, 
the insertion proposed to use the Master Curve reference temperature 
RTT0, which is defined as RTT0 = T0 + 
35 [deg]F, where T0 is a material-specific temperature value 
determined in accordance with ASTM E1921, ``Standard Test Method for 
Determination of Reference Temperature, T0, for Ferritic 
Steels in the Transition Range,'' to index (shift) the fracture 
toughness KIc curve, based on the lower bound of static 
initiation critical stress intensity factor, as well as the 
KIa curve, based on the lower bound

[[Page 32945]]

of crack arrest critical stress intensity factor.
    While use of RTT0 to index the KIc curve is 
acceptable, using RTT0 to index the KIa curve is 
questionable. This concern is based on the data analysis in ``A 
Physics-Based Model for the Crack Arrest Toughness of Ferritic 
Steels,'' written by NRC staff member Mark Kirk and published in 
``Fatigue and Fracture Mechanics, 33rd Volume, ASTM STP 1417'' which 
indicated that the crack arrest data does not support using 
RTT0 as RTNDT to index the KIa curve. 
This is also confirmed by industry data disclosed in a presentation, 
``Final Results from the CARINA Project on Crack Initiation and Arrest 
of Irradiated German RPV Steels for Neutron Fluences in the Upper 
Bound,'' by AREVA at the 26th Symposium on Effects of Radiation on 
Nuclear Materials (June 12-13, 2013, Indianapolis, Indiana, USA). The 
NRC staff recognized that the proposed insertion is consistent with 
ASME BPV Code Case N-629, ``Use of Fracture Toughness Test Data to 
Establish Reference Temperature for Pressure Retaining Materials,'' 
which was accepted by the NRC without conditions. In addition to the 
current NRC effort, the appropriate ASME BPV Code committee is in the 
process of correcting this issue in a future revision of Appendix A of 
Section XI.
    With this condition, users of Appendix A can avoid using an 
erroneous fracture toughness KIa value in their LEFM 
analysis for determining the acceptability of a detected flaw in 
applicable components. Therefore, the NRC is adding a condition which 
permits the use of RTT0 in place of RTNDT in 
applications using the KIc equation and the associated 
KIc curve, but does not permit the use of RTT0 in 
place of RTNDT in applications using the KIa 
equation and the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of 
Irradiated Materials
    The NRC is adding Sec.  50.55a(b)(2)(xxxvi) to require licensees 
using ASME BPV Code, Section XI, 2013 Edition, Appendix A, Paragraph A-
4400, to obtain NRC approval under Sec.  50.55a(z) before using 
irradiated T0 and the associated RTT0 in 
establishing fracture toughness of irradiated materials.
    Sub-article A-4400 provides guidance for considering irradiation 
effects on materials. The NRC staff's concern is related to use of 
RTT0 based on measured T0 of the irradiated 
materials. Specifically, the NRC staff has concerns over this sentence 
in the proposed insertion: ``Measurement of RTT0 of 
unirradiated or irradiated materials as defined in A-4200(b) is 
permitted, including use of the procedures given in ASTM E1921 to 
obtain direct measurement of irradiated T0.''
    Permission of measurement of RTT0 of irradiated 
materials, without providing guidelines regarding how to use the 
measured parameter in determining the fracture toughness of the 
irradiated materials, may mislead the users of Appendix A into adopting 
methodology that has not been accepted by the NRC. With this condition, 
users of Appendix A can avoid inappropriately using a fracture 
toughness KIc value based on the irradiated T0 
and the associated RTT0 in their LEFM analysis for 
determining the acceptability of a detected flaw in applicable 
components.
10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements
    The NRC is adding new paragraphs (g)(2)(i), (ii), and (iii) and 
revising current paragraphs (g) introductory text, (g)(2), (g)(3) 
introductory text, and (g)(3)(i), (ii), and (v) to distinguish the 
requirements for accessibility and preservice examination from those 
for inservice inspection in Sec.  50.55a(g). In addition, consistent 
with other paragraphs of this section, headings are added to the 
subordinate paragraphs of (g) in order to enhance readability of the 
regulation. No substantive change to the requirements are intended by 
these revisions.

C. OM Code

10 CFR 50.55a(b)(3) Conditions on ASME OM Code
    The NRC is revising Sec.  50.55a(b)(3) to clarify that Subsections 
ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, 
III, and V; and Nonmandatory Appendices A through H and J through M of 
the OM Code are each incorporated by reference into Sec.  50.55a. The 
NRC is also clarifying that the OM Code Nonmandatory Appendices 
incorporated by reference into Sec.  50.55a are approved for use, but 
are not mandated. The Nonmandatory Appendices may be used by applicants 
and licensees of nuclear power plants, subject to the conditions in 
Sec.  50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
    The NRC is revising Sec.  50.55a(b)(3)(i) to allow use of the 1994 
Edition, 2008 Edition, and the 2009-1a Addenda of NQA-1, ``Quality 
Assurance Requirements for Nuclear Facility Applications.'' The NRC 
reviewed these editions and addenda, compared them to the previously 
approved versions of NQA-1, and found that there were no significant 
differences.
    The NRC is removing the reference in Sec.  50.55a(b)(3)(i) to 
versions of NQA-1 older than the 1994 Edition, inasmuch as these 
versions do not appear to be in use at any nuclear power plant. The NRC 
did not receive any adverse comments from any applicant or licensee 
regarding concerns about removing versions of NQA-1 older than the 1994 
Edition from the regulation. The NRC received one comment regarding 
NQA-1, supporting incorporation by reference of NQA-1 but not 
responding to the NRC's request for comment regarding the removal of 
references to older versions of NQA-1. Accordingly, the NRC concludes 
that removal of NQA-1 versions older than the 1994 Edition will not 
have any adverse effect on licensees, and the final rule removes these 
older versions from Sec.  50.55a(b)(3)(i).
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) 
Testing
    The NRC is revising Sec.  50.55a(b)(3)(ii) to reflect the new 
Appendix III, ``Preservice and Inservice Testing of Active Electric 
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,'' 
of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. Appendix 
III of the OM Code establishes provisions for periodic verification of 
the design-basis capability of MOVs within the scope of the IST 
program. Appendix III of the OM Code reflects the incorporation of OM 
Code Cases OMN-1, ``Alternative Rules for Preservice and Inservice 
Testing of Active Electric Motor-Operated Valve Assemblies in Light-
Water Reactor Power Plants,'' and OMN-11, ``Risk-Informed Testing for 
Motor-Operated Valves.'' The NRC is adding four new conditions on the 
use of Mandatory Appendix III in new Sec.  50.55a(b)(3)(ii)(A), (B), 
(C), and (D) to address periodic verification of MOV design-basis 
capability. These new conditions are discussed in the next four 
sections.
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval (First 
Condition on Use of Mandatory Appendix III)
    In the proposed rule, the NRC specified in Sec.  
50.55a(b)(3)(ii)(A) that licensees evaluate the adequacy of the 
diagnostic test interval for each MOV and adjust the interval as 
necessary, but not later than 5 years or three refueling outages 
(whichever is longer) from initial implementation of OM Code,

[[Page 32946]]

Appendix III. Paragraph III-3310(b) in Appendix III includes a 
provision stating that if insufficient data exist to determine the IST 
interval, then MOV inservice testing shall be conducted every two 
refueling outages or 3 years (whichever is longer) until sufficient 
data exist, from an applicable MOV or MOV group, to justify a longer 
IST interval. As discussed in a final rule published September 22, 1999 
(64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC 
considers it appropriate to include a modification requiring licensees 
to evaluate the information obtained for each MOV, during the first 5 
years or three refueling outages (whichever is longer) of the use of 
Appendix III to validate assumptions made in justifying a longer test 
interval.
    In response to public comments, the NRC revised Sec.  
50.55a(b)(3)(ii)(A) to clarify its intent for licensees to evaluate the 
test interval within 5 years or three refueling outages (whichever is 
longer) following implementation of Appendix III to the OM Code, rather 
than implying that every MOV must be tested within 5 years or three 
refueling outages of the initial implementation of Appendix III. For 
example, the condition allows grouping of MOVs to share test 
information in the evaluation of the MOV periodic verification 
intervals within 5 years or three refueling outages (whichever is 
longer) of the implementation of OM Code, Appendix III. Therefore, 
Sec.  50.55a(b)(3)(ii)(A) of this final rule states that licensees 
shall evaluate the adequacy of the diagnostic test intervals 
established for MOVs within the scope of OM Code, Mandatory Appendix 
III, not later than 5 years or three refueling outages (whichever is 
longer) from initial implementation of OM Code, Appendix III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk (Second Condition 
on Use of Mandatory Appendix III)
    The NRC is adding Sec.  50.55a(b)(3)(ii)(B) to require that when 
using Mandatory Appendix III, licensees ensure that the potential 
increase in core damage frequency (CDF) and large early release 
frequency (LERF) associated with the extension is acceptably small when 
extending exercise test intervals for high risk MOVs beyond a quarterly 
frequency. As discussed in a final rule published September 22, 1999 
(64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC 
considers it important for licensees to have sufficient information 
from the specific MOV, or similar MOVs, to demonstrate that exercising 
on a refueling outage frequency does not significantly affect component 
performance. The information may be obtained by grouping similar MOVs 
and establishing periodic exercising intervals of MOVs in the group 
over the refueling interval.
    Section 50.55a(b)(3)(ii)(B) requires that the increase in the 
overall plant CDF and LERF resulting from the extension be acceptably 
small. As presented in RG 1.174, ``An Approach for Using Probabilistic 
Risk Assessment [PRA] in Risk-Informed Decisions on Plant-Specific 
Changes to the Licensing Basis,'' the NRC considers acceptably small 
changes to be relative and to depend on the current plant CDF and LERF. 
For plants with total baseline CDF of 10-\4\ per year or 
less, acceptably small means CDF increases of up to 10-\5\ 
per year; and for plants with total baseline CDF greater than 
10-\4\ per year, acceptably small means CDF increases of up 
to 10-\6\ per year. For plants with total baseline LERF of 
10-\5\ per year or less, acceptably small LERF increases are 
considered to be up to 10-\6\ per year; and for plants with 
total baseline LERF greater than 10-\5\ per year, acceptably 
small LERF increases are considered to be up to 10-\7\ per 
year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization (Third Condition on 
Use of Mandatory Appendix III)
    The NRC is adding Sec.  50.55a(b)(3)(ii)(C) to require, when 
applying Mandatory Appendix III, that licensees categorize MOVs 
according to their safety significance using the methodology described 
in OM Code Case OMN-3, ``Requirements for Safety Significance 
Categorization of Components Using Risk Insights for Inservice Testing 
of LWR Power Plants,'' subject to the conditions discussed in RG 1.192, 
or using an MOV risk ranking methodology accepted by the NRC on a 
plant-specific or industry-wide basis in accordance with the conditions 
in the applicable safety evaluation. Paragraph III-3720 in Appendix III 
to the OM Code states that when applying risk insights, each MOV shall 
be evaluated and categorized using a documented risk ranking 
methodology. Further, Appendix III only addresses risk ranking 
methodologies that include two risk categories. In light of the 
potential extension of quarterly test intervals for high risk MOVs and 
the relaxation of IST activities for low risk MOVs based on risk 
insights, the NRC has determined that the rule should specify that 
plant-specific or industry-wide risk ranking methodologies must have 
been accepted by the NRC through RG 1.192 (which accepts OM Code Case 
OMN-3 with the specified conditions) or the issuance of safety 
evaluations. As noted in the response to public comments, the intent of 
this condition is to indicate that when applying Appendix III to the OM 
Code, licensees may use either a two-risk category approach (high or 
low) or a three-risk category approach (high, medium, and low), 
provided the risk ranking method has been accepted by the NRC.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time (Fourth Condition on Use of 
Mandatory Appendix III)
    The NRC is adding Sec.  50.55a(b)(3)(ii)(D) to require that when a 
licensee applies Paragraph III-3600, ``MOV Exercising Requirements,'' 
of Appendix III to the OM Code, the licensee verify that the stroke 
time of the MOV satisfies the assumptions in the plant's safety 
analyses. Previous editions and addenda of the OM Code specified that 
the licensee must perform quarterly MOV stroke time measurements that 
could be used to verify that the MOV stroke time satisfies the 
assumptions in the safety analyses consistent with plant TS. The need 
for verification of the MOV stroke time during periodic exercising is 
consistent with the NRC's lessons learned from the implementation of OM 
Code Case OMN-1. However, Paragraph III-3600 of Appendix III of the 
versions of the OM Code that will be incorporated by reference in this 
rulemaking no longer require the verification of MOV stroke time during 
periodic exercising. For this reason, the NRC is adopting this new 
condition, which will effectively retain the need to verify that the 
MOV stroke time during periodic exercising satisfies the assumptions in 
the plant's safety analyses.
    Based on the discussion during the public webinar on March 2, 2016, 
the NRC revised the condition to clarify that it applies to MOVs 
referenced in the plant TS. In particular, the NRC revised the 
condition to indicate that when a licensee applies Paragraph III-3600 
of Appendix III to the OM Code, the licensee shall verify that the 
stroke time of MOVs specified in plant technical specifications 
satisfies the assumptions in the plant's safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors
    The NRC is adding Sec.  50.55a(b)(3)(iii) to apply specific 
conditions for IST programs applicable to licensees of new nuclear 
power plants in addition to the provisions of the OM Code as 
incorporated by reference with conditions in Sec.  50.55a. Licensees of 
``new reactors'' are, as identified in the paragraph: (1) Holders of 
operating

[[Page 32947]]

licenses for nuclear power reactors that received construction permits 
under this part on or after the date 12 months after August 17, 2017, 
and (2) holders of combined licenses (COLs) issued under 10 CFR part 
52, whose initial fuel loading occurs on or after the date 12 months 
after August 17, 2017. This implementation schedule for new reactors is 
consistent with the NRC regulations governing inservice testing in 
Sec.  50.55a(f)(4)(i).
    Commission Papers SECY-90-016, ``Evolutionary Light Water Reactor 
(LWR) Certification Issues and Their Relationship to Current Regulatory 
Requirements;'' SECY-93-087, ``Policy, Technical, and Licensing Issues 
Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) 
Designs;'' SECY-94-084, ``Policy and Technical Issues Associated with 
the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant 
Designs;'' and SECY-95-132, ``Policy and Technical Issues Associated 
with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive 
Plant Designs (SECY-94-084),'' discuss IST programs for new reactors 
licensed under 10 CFR part 52.
    In recognition of new reactor designs and lessons learned from 
nuclear power plant operating experience, the ASME is updating the OM 
Code to incorporate improved IST provisions for components used in 
nuclear power plants that were issued (or will be issued) construction 
permits, or COLs, on or following January 1, 2000 (defined in the OM 
Code as post-2000 plants). The first phase of the ASME effort 
incorporated IST provisions that specify full flow pump testing and 
other clarifications for post-2000 plants in the OM Code beginning with 
the 2011 Addenda. The second phase of the ASME effort incorporated 
preservice and inservice inspection and surveillance provisions for 
pyrotechnic-actuated (squib) valves in the 2012 Edition of the OM Code. 
The ASME is considering further modifications to the OM Code to address 
additional lessons learned from valve operating experience and new 
reactor issues. As described in the following paragraphs, Sec.  
50.55a(b)(3)(iii) will include four specific conditions which are 
discussed in the following paragraphs.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves
    The NRC is adding Sec.  50.55a(b)(3)(iii)(A) to require that 
licensees within the scope of Sec.  50.55a(b)(3)(iii) periodically 
verify the capability of power-operated valves (POVs) to perform their 
design-basis safety functions. While Appendix III to the OM Code 
addresses this requirement for MOVs with the conditions specified in 
Sec.  50.55a, applicable applicants and licensees will need to develop 
programs to periodically verify the design-basis capability of other 
POVs. The NRC's Regulatory Issue Summary 2000-03, ``Resolution of 
Generic Issue 158: Performance of Safety-Related Power-Operated Valves 
Under Design Basis Conditions,'' provides attributes for a successful 
long-term periodic verification program for POVs by incorporating 
lessons learned from MOV performance at operating nuclear power plants 
and research programs. Implementation of Appendix III to the OM Code as 
accepted in Sec.  50.55a(b)(3)(ii) satisfies Sec.  50.55a(b)(3)(iii)(A) 
for MOVs.
    Section 50.55a(b)(3)(iii)(A) is consistent with the Commission 
policy for new reactors summarized in an NRC Staff Memorandum, 
``Consolidation of SECY-94-084 and SECY-95-132,'' dated July 24, 1995, 
that (a) the design capability of safety-related POVs should be 
demonstrated by a qualification test prior to installation; (b) prior 
to initial startup, POV capability under design-basis differential 
pressure and flow should be verified by a pre-operational test; and (c) 
during the operational phase, POV capability under design-basis 
differential pressure and flow should be verified periodically through 
a program similar to that developed for MOVs in Generic Letter 89-10, 
``Safety-Related Motor-Operated Valve Testing and Surveillance,'' dated 
June 28, 1989.\2\
---------------------------------------------------------------------------

    \2\ The NRC issued seven supplements to provide guidance for the 
implementation of the MOV testing program requested in Generic 
Letter 89-10. The supplements to Generic Letter 89-10 did not modify 
the substance of the MOV testing program requested in Generic Letter 
89-10 to provide reasonable assurance in the capability of safety-
related MOVs to perform their design-basis functions.
---------------------------------------------------------------------------

    The condition in Sec.  50.55a(b)(3)(iii)(A) specifies with the same 
level of detail as the condition in Sec.  50.55a(b)(3)(ii) that nuclear 
power plant licensees must establish a program to ensure the continued 
capability of MOVs in performing their design-basis safety functions. 
When establishing the MOV periodic verification condition, the NRC 
provided guidance in the final rule published September 22, 1999 (64 FR 
51370), for licensees to develop acceptable programs that would satisfy 
the MOV periodic verification condition. Similarly, the NRC staff is 
providing guidance herein for new reactor applicants and licensees to 
develop acceptable programs to periodically verify the capability of 
POVs to perform their design-basis safety functions.
    In NUREG-2124, ``Final Safety Evaluation Report [FSER] Related to 
the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and 
4,'' the NRC staff found the provisions established by the COL 
applicant for Vogtle Units 3 and 4 in its Final Safety Analysis Report 
(FSAR), Revision 5, Section 3.9.6.2.2, ``Valve Testing,'' to 
periodically verify the capability of POVs (such as air-operated valves 
(AOVs), solenoid-operated valves (SOVs), and hydraulic-operated valves 
(HOVs)) to perform their design-basis safety functions to be 
acceptable. In particular, the Vogtle Units 3 and 4 FSAR specifies 
that:

    Power-operated valves other than active MOVs are exercised 
quarterly in accordance with OM ISTC, unless justification is 
provided in the inservice testing program for testing these valves 
at other than Code mandated frequencies. Although the design basis 
capability of power-operated valves is verified as part of the 
design and qualification process, power-operated valves that perform 
an active safety function are tested again after installation in the 
plant, as required, to ensure valve setup is acceptable to perform 
their required functions, consistent with valve qualification. These 
tests, which are typically performed under static (no flow or 
pressure) conditions, also document the ``baseline'' performance of 
the valves to support maintenance and trending programs. During the 
testing, critical parameters needed to ensure proper valve setup are 
measured. Depending on the valve and actuator type, these parameters 
may include seat load, running torque or thrust, valve travel, 
actuator spring rate, bench set and regulator supply pressure. 
Uncertainties associated with performance of these tests and use of 
the test results (including those associated with measurement 
equipment and potential degradation mechanisms) are addressed 
appropriately. Uncertainties may be considered in the specification 
of acceptable valve setup parameters or in the interpretation of the 
test results (or a combination of both). Uncertainties affecting 
both valve function and structural limits are addressed. Additional 
testing is performed as part of the air-operated valve (AOV) 
program, which includes the key elements for an AOV Program as 
identified in the JOG AOV program document, Joint Owners Group Air 
Operated Valve Program Document, Revision 1, December 13, 2000 
(References 203 and 204) [JOG AOV Program Document, Revision 1, 
December 13, 2000 (ADAMS Accession No. ML010950310), and NRC comment 
letter dated October 8, 1999, to Nuclear Energy Institute (ADAMS 
Accession No. ML020360077)]. The AOV program incorporates the 
attributes for a successful power-operated valve long-term periodic 
verification program, as discussed in Regulatory Issue Summary 2000-
03, Resolution of Generic Safety Issue 158: Performance of Safety-
Related Power-Operated Valves Under Design Basis

[[Page 32948]]

Conditions, by incorporating lessons learned from previous nuclear 
power plant operations and research programs as they apply to the 
periodic testing of air- and other power-operated valves included in 
the IST program.
    For example, key lessons learned addressed in the AOV program 
include:
     Valves are categorized according to their safety 
significance and risk ranking.
     Setpoints for AOVs are defined based on current vendor 
information or valve qualification diagnostic testing, such that the 
valve is capable of performing its design-basis function(s).
     Periodic static testing is performed, at a minimum on 
high risk (high safety significance) valves, to identify potential 
degradation, unless those valves are periodically cycled during 
normal plant operation, under conditions that meet or exceed the 
worst case operating conditions within the licensing basis of the 
plant for the valve, which would provide adequate periodic 
demonstration of AOV capability. If required based on valve 
qualification or operating experience, periodic dynamic testing is 
performed to re-verify the capability of the valve to perform its 
required functions.
     Sufficient diagnostics are used to collect relevant 
data (e.g., valve stem thrust and torque, fluid pressure and 
temperature, stroke time, operating and/or control air pressure, 
etc.) to verify the valve meets the functional requirements of the 
qualification specification.
     Test frequency is specified, and is evaluated each 
refueling outage based on data trends as a result of testing. 
Frequency for periodic testing is in accordance with References 203 
and 204, with a minimum of 5 years (or 3 refueling cycles) of data 
collected and evaluated before extending test intervals.
     Post-maintenance procedures include appropriate 
instructions and criteria to ensure baseline testing is re-performed 
as necessary when maintenance on the valve, repair or replacement, 
have the potential to affect valve functional performance.
     Guidance is included to address lessons learned from 
other valve programs specific to the AOV program.
     Documentation from AOV testing, including maintenance 
records and records from the corrective action program are retained 
and periodically evaluated as a part of the AOV program.
* * * * *
    The attributes of the AOV testing program described above, to 
the extent that they apply to and can be implemented on other 
safety-related power-operated valves, such as electro-hydraulic 
operated valves, are applied to those other power-operated valves.'' 
(Vogtle Electric Generating Plant, Units 3 and 4, Updated Final 
Safety Analysis Report (UFSAR), Section 3.9.6, ``Inservice Testing 
of Pumps and Valves'')

    Applicable applicants and licensees may follow the method described 
in the Vogtle Units 3 and 4 FSAR in satisfying Sec.  
50.55a(b)(3)(iii)(A), or may establish a different method, subject to 
evaluation by the NRC during the licensing process or inspections.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
    The NRC is adding Sec.  50.55a(b)(3)(iii)(B) to require that 
licensees within the scope of Sec.  50.55a(b)(3)(iii) perform bi-
directional testing of check valves within the IST program where 
practicable. Nuclear power plant operating experience has revealed that 
testing check valves in only the flow direction can result in 
significant degradation, such as a missing valve disc, not being 
identified by the test. Nonmandatory Appendix M, ``Design Guidance for 
Nuclear Power Plant Systems and Component Testing,'' to OM Code, 2011 
Addenda and 2012 Edition, includes guidance for the design of new 
reactors to enable bi-directional testing of check valves. New reactor 
designs will provide the capability for licensees of new nuclear power 
plants to perform bi-directional testing of check valves within the IST 
program. Bi-directional testing of check valves in new reactors, as 
required by Sec.  50.55a(b)(3)(iii)(B), could be accomplished by valve-
specific testing or condition monitoring activities in accordance with 
Appendix II to the OM Code as accepted in Sec.  50.55a. The NRC is 
specifying this provision for bi-directional testing of check valves 
for new reactors in Sec.  50.55a(b)(3)(iii)(B) to emphasize that new 
reactors should include the capability for bi-directional testing of 
check valves as part of their initial design.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration
    In the proposed rule, the NRC proposed adding Sec.  
50.55a(b)(3)(iii)(C) to require that licensees subject to Sec.  
50.55a(b)(3)(iii) monitor flow-induced vibration (FIV) from 
hydrodynamic loads and acoustic resonance during preservice testing and 
inservice testing to identify potential adverse flow effects that might 
impact components within the scope of the IST program.
    Nuclear power plant operating experience has revealed the potential 
for adverse flow effects from vibration caused by hydrodynamic loads 
and acoustic resonance on components in the reactor coolant, steam, and 
feedwater systems. Therefore, the licensee will be required to address 
potential adverse flow effects on safety-related pumps, valves, and 
dynamic restraints within the IST program in the reactor coolant, 
steam, and feedwater systems from hydraulic loading and acoustic 
resonance during plant operation. In response to public comments, the 
NRC revised Sec.  50.55a(b)(3)(iii)(C) to clarify its intent that FIV 
monitoring of components may be conducted during preservice testing or 
inservice testing. This requirement will confirm that piping, 
components, restraints, and supports have been designed and installed 
to withstand the dynamic effects of steady-state FIV and anticipated 
operational transient conditions. As part of preservice testing 
activities, the initial test program may be used to verify that safety-
related piping and components are properly installed and supported such 
that vibrations caused by steady-state or dynamic effects do not result 
in excessive stress or fatigue in safety-related plant systems.
    In the Vogtle Units 3 and 4 FSER, the NRC staff found the 
provisions established by the COL applicant for Vogtle Units 3 and 4 in 
its FSAR, Revision 5, Section 3.9, ``Mechanical Systems and 
Components,'' Section 14.2.9, ``Preoperational Test Descriptions,'' and 
Section 14.2.10, ``Startup Test Procedures,'' with incorporation by 
reference of corresponding sections of the AP1000 Design Control 
Document (DCD), to monitor FIV from hydrodynamic loads and acoustic 
resonance during preservice testing or inservice testing to be 
acceptable. In particular, the NRC staff stated in the Vogtle Units 3 
and 4 FSER:

    AP1000 DCD Tier 2, Section 3.9.2, ``Dynamic Testing and 
Analysis,'' describes tests to confirm that piping, components, 
restraints, and supports have been designed to withstand the dynamic 
effects of steady-state FIV and anticipated operational transient 
conditions. Section 14.2.9.1.7, ``Expansion, Vibration and Dynamic 
Effects Testing,'' in AP1000 DCD Tier 2, Chapter 14, ``Initial Test 
Program,'' states that the purpose of the expansion, vibration and 
dynamic effects testing is to verify that safety-related, high 
energy piping and components are properly installed and supported 
such that, in addition to other factors, vibrations caused by 
steady-state or dynamic effects do not result in excessive stress or 
fatigue to safety-related plant systems. Nuclear power plant 
operating experience has revealed the potential for adverse flow 
effects from vibration caused by hydrodynamic loads and acoustic 
resonance on reactor coolant, steam, and feedwater systems. . . . In 
its response, SNC [Vogtle Units 3 and 4 COL applicant] stated that 
it intended to use the overall Initial Test Program to demonstrate 
that the plant has been constructed as designed and the systems 
perform consistent with design requirements. SNC referenced the 
provisions in the AP1000 DCD for vibration monitoring and testing to 
be implemented at VEGP. For example, the applicant notes that AP1000 
DCD Tier 2, Section 3.9.2.1, ``Piping Vibration, Thermal Expansion 
and Dynamic Effects,'' specifies that the preoperational test

[[Page 32949]]

program for ASME BPV Code, Section III, Class 1, 2, and 3 piping 
systems simulates actual operating modes to demonstrate that 
components comprising these systems meet functional design 
requirements and that piping vibrations are within acceptable 
levels. SNC indicates that the planned vibration testing program 
described in AP1000 DCD Tier 2, Sections 14.2.9 and 14.2.10, with 
the preservice and IST programs described in AP1000 DCD Tier 2, 
Sections 3.9.3.4.4 and 3.9.6, will confirm component installation in 
accordance with design requirements, and address the effects of 
steady-state (flow-induced) and transient vibration to ensure the 
operability of valves and dynamic restraints in the IST Program. The 
NRC staff considers the response by SNC clarifies its application of 
the provisions in the AP1000 DCD to ensure that potential adverse 
flow effects will be addressed at VEGP. Therefore, the NRC staff 
considers Standard Content Open Item 3.9-5 to be resolved for the 
VEGP COL application.'' (NUREG-2124, ``Final Safety Evaluation 
Report Related to the Combined Licenses for Vogtle Electric 
Generating Plant, Units 3 and 4,'' Section 3.9.6, ``Inservice 
Testing of Pumps and Valves (Related to RG 1.206, Section C.III.1, 
Chapter 3, C.I.3.9.6, `Functional Design, Qualification, and 
Inservice Testing Programs for Pumps, Valves, and Dynamic 
Restraints')'').

    As clarified in the final rule in response to public comments, a 
licensee may monitor components for adverse FIV effects during 
preservice testing or IST activities.
    Applicable applicants and licensees may either apply the methods 
described in the Vogtle Units 3 and 4 FSAR in satisfying Sec.  
50.55a(b)(3)(iii)(C) or develop their own plant-specific methods to 
satisfy Sec.  50.55a(b)(3)(iii)(C) for NRC review during the licensing 
process.
10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems
    The NRC is adding Sec.  50.55a(b)(3)(iii)(D) to require that 
licensees within the scope of Sec.  50.55a(b)(3)(iii) establish a 
program to assess the operational readiness of pumps, valves, and 
dynamic restraints within the scope of the Regulatory Treatment of Non-
Safety Systems (RTNSS) for applicable reactor designs. As of the time 
of this final rule, these are designs which have been certified in a 
design certification rule under 10 CFR part 52. In SECY-94-084 and 
SECY-95-132, the Commission discusses RTNSS policy and technical issues 
associated with passive plant designs. Some new nuclear power plants 
have advanced light-water reactor (ALWR) designs that use passive 
safety systems that rely on natural forces, such as density 
differences, gravity, and stored energy to supply safety injection 
water and to provide reactor core and containment cooling. Active 
systems in passive ALWR designs are categorized as non-safety systems 
with limited exceptions. Active systems in passive ALWR designs provide 
the first line of defense to reduce challenges to the passive systems 
in the event of a transient at the nuclear power plant. Active systems 
that provide a defense-in-depth function in passive ALWR designs need 
not meet all of the acceptance criteria for safety-related systems. 
However, there should be a high level of confidence that these active 
systems will be available and reliable when needed. The combined 
activities to provide confidence in the capability of these active 
systems in passive ALWR designs to perform their functions important to 
safety are referred to as the RTNSS program. In the NRC Staff 
Memorandum, ``Consolidation of SECY-94-084 and SECY-95-132,'' dated 
July 24, 1995, the NRC staff provided a consolidated list of the 
approved policy and technical positions associated with RTNSS equipment 
in passive plant designs discussed in SECY-94-084 and SECY-95-132. This 
new paragraph specifies the need for licensees to assess the 
operational readiness of RTNSS pumps, valves, and dynamic restraints.
    The July 24, 1995, staff memorandum summarizes the Commission 
policy positions related to inservice testing of RTNSS pumps and valves 
as follows:

    The staff also concluded that additional inservice testing 
requirements may be necessary for certain pumps and valves in 
passive plant designs. The unique passive plant design relies 
significantly on passive safety systems, but also depends on non-
safety systems (which are traditionally safety-related systems in 
current light-water reactors) to prevent challenges to passive 
systems. Therefore, the reliable performance of individual 
components is a very significant factor in enhancing the safety of 
passive plant design. The staff recommends that the following 
provisions be applied to passive ALWR plants to ensure reliable 
component performance.
    1. Important non-safety-related components are not required to 
meet criteria similar to safety-grade criteria. However, the non-
safety-related piping systems with functions that have been 
identified as being important by the RTNSS process should be 
designed to accommodate testing of pumps and valves to assure that 
the components meet their intended functions. Specific positions on 
the inservice testing requirements for those components will be 
determined as a part of the staff's review of plant-specific 
implementation of the regulatory treatment of non-safety systems for 
passive reactor designs.
    2. . . . The vendors for advanced passive reactors, for which 
the final designs are not complete, have sufficient time to include 
provisions in their piping system designs to allow testing at power. 
Quarterly testing is the base testing frequency in the Code and the 
original intent of the Code. Furthermore, the COL holder may need to 
test more frequently than during cold shutdowns or at every 
refueling outage to ensure that the reliable performance of 
components is commensurate with the importance of the safety 
functions to be performed and with system reliability goals. 
Therefore, to the extent practicable, the passive ALWR piping 
systems should be designed to accommodate the applicable Code 
requirements for the quarterly testing of valves. However, design 
configuration changes to accommodate Code-required quarterly testing 
should be done only if the benefits of the test outweigh the 
potential risk.
    3. The passive system designs should incorporate provisions (1) 
to permit all critical check valves to be tested for performance, to 
the extent practicable, in both forward- and reverse-flow 
directions, although the demonstration of a non-safety direction 
test need not be as rigorous as the corresponding safety direction 
test, and (2) to verify the movement of each check valve's obturator 
during inservice testing by observing a direct instrumentation 
indication of the valve position such as a position indicator or by 
using nonintrusive test methods.
    4. . . . Similarly, to the extent practicable, the design of 
non-safety-related piping systems with functions under design-basis 
condition that have been identified as being important by the RTNSS 
process should incorporate provisions to periodically test power-
operated valves in the system during operations to assure that the 
valves meet their intended functions under design-basis conditions.
    5. . . . Mispositioning may occur through actions taken locally 
(manual or electrical), at a motor control center, or in the control 
room, and includes deliberate changes of valve position to perform 
surveillance testing. The staff will determine if and the extent to 
which this concept should be applied to MOVs in important non-
safety-related systems when the staff reviews the implementation of 
the regulatory treatment of non-safety systems.'' (NRC Staff 
Memorandum, ``Consolidation of SECY-94-084 and SECY-95-132,'' July 
24, 1995, pages 26-28).

    Consistent with the Commission policy for RTNSS equipment, Sec.  
50.55a(b)(3)(iii)(D) specifies that new reactor licensees shall assess 
the operational readiness of pumps, valves, and dynamic restraints 
within the RTNSS scope. This regulatory requirement will allow 
licensees flexibility in developing programs to assess operational 
readiness of RTNSS components that satisfy the Commission policy. 
Guidance on the implementation of the Commission policy for RTNSS 
equipment is set forth in NRC Inspection Procedure 73758, ``Part 52, 
Functional Design and Qualification, and Preservice and Inservice 
Testing Programs for Pumps, Valves and

[[Page 32950]]

Dynamic Restraints,'' dated April 19, 2013.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
    The NRC is revising Sec.  50.55a(b)(3)(iv) to address Appendix II, 
``Check Valve Condition Monitoring Program,'' provided in the 2003 
Addenda through the 2012 Edition of the OM Code. In the proposed rule, 
the NRC proposed a condition in Sec.  50.55a(b)(3)(iv) to provide 
assurance that the valve or group of valves is capable of performing 
its intended function(s) over the entire interval. Public comments 
indicated that the proposed condition could be misinterpreted. 
Therefore, the NRC revised the proposed condition to clarify that the 
implementation of Appendix II must include periodic sampling of the 
check valves over the maximum interval allowed by Appendix II for the 
check valve condition monitoring program. A new table was added to the 
paragraph to specify the maximum intervals between check valve 
condition monitoring activities when applying interval extensions.
    The conditions currently specified for the use of Appendix II, 1995 
Edition with the 1996 and 1997 Addenda, and 1998 Edition through the 
2002 Addenda, of the OM Code remain unchanged by this final rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
    The NRC is adding a new condition, Sec.  50.55a(b)(3)(vii), to 
prohibit the use of Subsection ISTB, ``Inservice Testing of Pumps in 
Light-Water Reactor Nuclear Power Plants,'' in the 2011 Addenda of the 
OM Code. In the 2011 Addenda to the OM Code, the upper end of the 
``Acceptable Range'' and the ``Required Action Range'' for flow and 
differential or discharge pressure for comprehensive pump testing in 
Subsection ISTB was raised to higher values. The NRC staff on the OM 
Code committee accepted the proposed increase of the upper end of the 
``Acceptable Range'' and ``Required Action Range'' with the planned 
addition of a requirement for a pump periodic verification test program 
in the OM Code. However, the 2011 Addenda to the OM Code did not 
include the requirement for a pump periodic verification test program. 
Since then, the 2012 Edition of the OM Code has incorporated Mandatory 
Appendix V, ``Pump Periodic Verification Test Program,'' which supports 
the changes to the acceptable and required action ranges for 
comprehensive pump testing in Subsection ISTB. Therefore, the new Sec.  
50.55a(b)(3)(vii) prohibits the use of Subsection ISTB in the 2011 
Addenda of the OM Code. Licensees will be allowed to apply Subsection 
ISTB with the revised acceptable and required action ranges in the 2012 
Edition of the OM Code as incorporated by reference in Sec.  50.55a.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
    The NRC is adding Sec.  50.55a(b)(3)(viii) to specify that 
licensees who wish to implement Subsection ISTE, ``Risk-Informed 
Inservice Testing of Components in Light-Water Reactor Nuclear Power 
Plants,'' of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition, 
must request and obtain NRC approval in accordance with Sec.  50.55a(z) 
to apply Subsection ISTE on a plant-specific basis as a risk-informed 
alternative to the applicable IST requirements in the OM Code.
    In the 2009 Edition of the OM Code, the ASME included new 
Subsection ISTE that describes a voluntary risk-informed approach in 
developing an IST program for pumps and valves at nuclear power plants. 
If a licensee chooses to implement this risk-informed IST approach, 
Subsection ISTE indicates that all requirements in Subsection ISTA, 
``General Requirements,'' Subsection ISTB, and Subsection ISTC, 
``Inservice Testing of Valves in Light-Water Reactor Nuclear Power 
Plants,'' of the OM Code continue to apply, except those identified in 
Subsection ISTE. The ASME selected risk-informed guidance from OM Code 
Cases OMN-1, OMN-3, OMN-4, ``Requirements for Risk Insights for 
Inservice Testing of Check Valves at LWR Power Plants,'' OMN-7, 
``Alternative Requirements for Pump Testing,'' OMN-11, and OMN-12, 
``Alternative Requirements for Inservice Testing Using Risk Insights 
for Pneumatically and Hydraulically Operated Valve Assemblies in Light-
Water Reactor Power Plants,'' in preparing Subsection ISTE of the OM 
Code.
    During development of Subsection ISTE, the NRC staff participating 
on the OM Code committees indicated that the conditions specified in RG 
1.192 for the use of the applicable OM Code Cases need to be considered 
when evaluating the acceptability of the implementation of Subsection 
ISTE. In addition, the NRC staff noted that several aspects of 
Subsection ISTE will need to be addressed on a case-by-case basis when 
determining the acceptability of its implementation. Therefore, the new 
condition in Sec.  50.55a(b)(3)(viii) requires that licensees who wish 
to implement Subsection ISTE of the OM Code must request approval from 
the NRC to apply Subsection ISTE on a plant-specific basis as a risk-
informed alternative to the applicable IST requirements in the OM Code.
    Nuclear power plant applicants for construction permits under 10 
CFR part 50, or combined licenses for construction and operation under 
10 CFR part 52, may describe their proposed implementation of the risk-
informed IST approach specified in Subsection ISTE of the OM Code for 
NRC review in their applications.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
    The NRC is adding a condition on the use of Subsection ISTF in 
Sec.  50.55a(b)(3)(ix). First, the condition states that Subsection 
ISTF, 2011 Addenda, is prohibited for use. Second, the condition 
specifies that licensees applying Subsection ISTF, ``Inservice Testing 
of Pumps in Light-Water Reactor Nuclear Power Plants--Post-2000 
Plants,'' in the 2012 Edition of the OM Code shall satisfy the 
requirements of Mandatory Appendix V, ``Pump Periodic Verification Test 
Program,'' of the OM Code, 2012 Edition.
    As previously discussed regarding the new condition in Sec.  
50.55a(b)(3)(vii), the upper end of the ``Acceptable Range'' and the 
``Required Action Range'' for flow and differential or discharge 
pressure for comprehensive pump testing in Subsection ISTB in the OM 
Code was raised to higher values in combination with the incorporation 
of Mandatory Appendix V, ``Pump Periodic Verification Test Program.'' 
However, the 2011 Addenda of the OM Code does not include Appendix V. 
In addition, Subsection ISTF in the 2011 Addenda and 2012 Edition of 
the OM Code does not include a requirement for a pump periodic 
verification test program. Therefore, the new condition in Sec.  
50.55a(b)(3)(ix) requires that the provisions of Appendix V be applied 
when implementing Subsection ISTF of the 2012 Edition of the OM Code to 
support the application of the upper end of the Acceptable Range and 
the Required Action Range for flow and differential or discharge 
pressure for inservice pump testing in Subsection ISTF.
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
    The NRC is adding Sec.  50.55a(b)(3)(xi) to emphasize the 
provisions in OM Code, 2012 Edition, Subsection ISTC-3700, ``Position 
Verification Testing,'' to verify that valve obturator position is 
accurately indicated. Subsection ISTC-3700 of the OM Code requires that

[[Page 32951]]

valves with remote position indicators shall be observed locally at 
least once every 2 years to verify that valve operation is accurately 
indicated. Subsection ISTC-3700 states that where practicable, this 
local observation should be supplemented by other indications, such as 
the use of flow meters or other suitable instrumentation to verify 
obturator position. Subsection ISTC-3700 also states that where local 
observation is not possible, other indications shall be used for 
verification of valve operation. Nuclear power plant operating 
experience has revealed that reliance on indicating lights and stem 
travel are not sufficient to satisfy the requirement in ISTC-3700 to 
verify that valve operation is accurately indicated. Appendix A, 
``General Design Criteria for Nuclear Power Plants,'' to 10 CFR part 50 
requires that where generally recognized codes and standards are used, 
they shall be identified and evaluated to determine their 
applicability, adequacy, and sufficiency, and shall be supplemented or 
modified as necessary to assure a quality product in keeping with the 
required safety function. This new condition specifies that when 
implementing OM Code, Subsection ISTC-3700, licensees shall verify that 
valve operation is accurately indicated by supplementing valve position 
indicating lights with other indications, such as flow meters or other 
suitable instrumentation, to provide assurance of proper obturator 
position. The OM Code specifies obturator movement verification in 
order to detect certain internal valve failure modes consistent with 
the definition of `exercising' found in ISTA-2000, ``Definitions,'' 
(i.e., demonstration that the moving parts of a component function). 
Verification of the ability of an obturator to change or maintain 
position is an essential element of valve operational readiness 
determination, which is a fundamental aspect of the OM Code.
    The NRC initially emphasized the ASME OM Code requirement for valve 
position indication in 1995 in the original issuance of NUREG-1482, 
``Guidelines for Inservice Testing at Nuclear Power Plants,'' paragraph 
4.2.5. The NRC's position is further elaborated in NUREG-1482 (Revision 
2), ``Guidelines for Inservice Testing at Nuclear Power Plants: 
Inservice Testing of Pumps and Valves and Inservice Examination and 
Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants,'' 
paragraph 4.2.7. As discussed in NUREG-1482 (Revision 2), ISTC-3700 
allows flexibility to licensees in verifying that operation of valves 
with remote position indicators is accurately indicated. For example, 
NUREG-1482 refers to various methods to verify valve operation, such as 
nonintrusive techniques, flow initiation or absence of flow, leak 
testing, and pressure testing. The extent of verification necessary for 
valve operation to satisfy ISTC-3700 will depend on the type of valve, 
the sophistication of the diagnostic equipment used in testing the 
valve, possible failure modes of the valve, and the operating history 
of the valve and similar valve types. To satisfy ISTC-3700, the 
licensee is responsible for developing and implementing a method to 
provide reasonable assurance that valve operation is accurately 
indicated.
    The NRC is requiring this condition for the implementation of the 
2012 Edition of the OM Code for the 120-month IST interval in order to 
allow additional time for licensees to comply with this condition.
10 CFR 50.55a(f): Preservice and Inservice Testing Requirements
    The NRC is revising the introductory text of Sec.  50.55a(f) to 
indicate that systems and components must meet the requirements for 
``preservice and inservice testing'' in the applicable ASME Codes and 
that both activities are referred to as ``inservice testing'' in the 
remainder of paragraph (f). The change clarifies that the OM Code 
includes provisions for preservice testing of components as part of its 
overall provisions for IST programs. No expansion of IST program scope 
was intended by this clarification.
    In the proposed rule, the NRC included references to the OM Code in 
Sec.  50.55a(f)(3)(iii)(A), Class 1 Pumps and Valves: First Provision; 
Sec.  50.55a(f)(3)(iii)(B), Class 1 Pumps and Valves: Second Provision; 
Sec.  50.55a(f)(3)(iv)(A), Class 2 and 3 Pumps and Valves: First 
Provision; and Sec.  50.55a(f)(3)(iv)(B): Second Provision; to align 
the regulatory language with the current ASME OM Code used for IST 
programs. Because Sec.  50.55a(f)(3)(iii) and (iv) specifically 
reference Class 1, 2, or 3 pumps and valves, the proposed changes to 
these paragraphs referencing the OM Code are unnecessary and have not 
been adopted in this final rule.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for 
Operating Plants
    The NRC is revising Sec.  50.55a(f)(4) to clarify that this 
paragraph is applicable to pumps and valves that are within the scope 
of the OM Code. This revision aligns the scope of pumps and valves for 
inservice testing with the scope defined in the OM Code.
    Public comments on the alignment of the IST program scope in Sec.  
50.55a(f)(4) indicated that the nuclear industry is addressing the 
requirements in 10 CFR part 50, appendices A and B, to establish an IST 
program for safety-related pumps and valves that are not classified as 
ASME BPV Code Class 1, 2, or 3 components through either the OM Code 
provisions or augmented IST programs. For example, one public commenter 
indicated that generally, augmented IST programs are designed to meet 
the OM Code where practicable, but relief requests are not required 
when alternate testing is necessary. The NRC regulations in Sec.  
50.55a address the concept of augmented IST programs for pumps and 
valves at nuclear power plants. For example, Sec.  50.55a(f)(6)(ii), 
``Augmented IST requirements,'' indicates that the licensee may follow 
an augmented IST program for pumps and valves for which the NRC deems 
that added assurance of operational readiness is necessary. The NRC 
finds that an augmented IST program as addressed in Sec.  
50.55a(f)(6)(ii) is acceptable for safety-related pumps and valves that 
are not classified as ASME BPV Code Class 1, 2, or 3 components.
    Public commenters were concerned that the alignment of the scope of 
the OM Code and Sec.  50.55a would cause a potential paperwork burden 
for the submittal of relief or alternative requests for safety-related 
pumps and valves that are not classified as ASME BPV Code Class 1, 2, 
or 3 components. In response to these comments, the NRC included a 
provision in Sec.  50.55a(f)(4) that the IST requirements for pumps and 
valves that are within the scope of the OM Code but are not classified 
as ASME BPV Code Class 1, Class 2, or Class 3 may be satisfied as an 
augmented IST program in accordance with Sec.  50.55a(f)(6)(ii) without 
requesting relief under Sec.  50.55a(f)(5) or alternatives under Sec.  
50.55a(z). This use of an augmented IST program may be acceptable 
provided the basis for deviations from the OM Code, as incorporated by 
reference in this section, demonstrates an acceptable level of quality 
and safety, or that implementing the Code provisions would result in 
hardship or unusual difficulty without a compensating increase in the 
level of quality and safety, where documented and available for NRC 
review. This additional provision avoids the potential paperwork burden 
for the submittal of relief or alternative requests by allowing the 
licensee to maintain the documentation demonstrating an acceptable 
level of quality and safety on site for NRC review, as appropriate. The

[[Page 32952]]

documentation and availability of the basis for deviations from the OM 
Code for NRC review are acceptable for pumps and valves within the 
scope of the OM Code but not classified as ASME BPV Code Class 1, 2, or 
3, based on their lower safety significance in comparison to ASME BPV 
Code Class 1, 2, and 3 pumps and valves.
10 CFR 50.55a(g)(4) Inservice Inspection Standards Requirement for 
Operating Plants
    The NRC recognizes that updating an Appendix VIII program is a 
complex and time-consuming process. The NRC also recognizes that 
licensees would face the possibility of needing to maintain multiple 
Appendix VIII programs if units were to update their ISI programs on 
different dates. Maintaining certifications to multiple Appendix VIII 
programs would be very complicated, while not improving the 
effectiveness of the programs. Based on public comments, and to assist 
licensees in updating and coordinating their ISI programs, the NRC is 
adding two options to the regulations. First, the NRC is revising Sec.  
50.55a(g)(4)(i) and (ii) to clarify that a licensee whose ISI interval 
commences during the 12- to 18-month period after the approval date of 
this final rule, may delay the update of their Appendix VIII program by 
up to 18 months after the approval date of this final rule. This will 
provide licensees with enough time to incorporate the changes for the 
new Appendix VIII program. Second, the NRC is adding the option for 
licensees to update their ISI program to use the latest edition and 
addenda of Appendix VIII incorporated by reference in Sec.  
50.55a(a)(1) at any time in the licensee's ten-year interval. Licensees 
can normally update their ISI programs using all or portions of newer 
versions of ASME BPV Code Section XI under Sec.  50.55a(g)(4)(iv), 
subject to NRC review and approval. While some requests to use portions 
of ASME BPV Code Section XI require a detailed review by the NRC, a 
licensee asking to use the entire latest incorporated-by-reference 
version of Appendix VIII would certainly be approved by the NRC staff 
in this process. This provision will, therefore, allow licensees to use 
the latest incorporated version of Appendix VIII, as long as it is 
coupled with the same edition and addenda of Appendix I, without the 
NRC review and approval process. This will allow licensees to 
coordinate their ISI programs and use the latest approved version of 
Appendix VIII without the delay imposed by submitting a relief request 
under Sec.  50.55a (g)(4)(iv).

D. ASME Code Cases

Administrative Changes to References in Sec.  50.55a to NRC Regulatory 
Guides Identifying ASME Code Cases Approved for Use by the NRC
    The NRC is removing the revision number of the three RGs currently 
approved by the Office of the Federal Register for incorporation by 
reference throughout the substantive provisions of Sec.  50.55a 
addressing the ASME Code Cases, i.e., paragraphs (b) through (g). The 
revision numbers for the RGs approved for incorporation by reference 
(currently, RG 1.84, RG 1.147, and RG 1.192) will be retained in Sec.  
50.55a(a)(3)(i) through (iii), where the RGs are listed by full title, 
including revision number. These changes simplify the regulatory 
language containing cross-references to these RGs and reduce the 
possibility of NRC error in preparing future amendments to Sec.  50.55a 
with respect to these RGs. These changes are administrative in nature 
and do not change substantive requirements with respect to the RGs and 
the Code Cases listed in the RGs.
Administrative Changes To Comply With Requirements for Incorporation by 
Reference
    The NRC is revising Sec.  50.55a(a)(1)(iii) to maintain the ASME 
Code Cases in alphanumeric order.
Organization of NRC's Discussion of the Six ASME Code Cases 
Incorporated by Reference in This Final Rule
    The discussions under the following headings address four of the 
six ASME Code Cases being incorporated by reference in this rulemaking 
(N-729-4, N-770-2, N-824, and OMN-20). A fifth ASME Code Case, N-852, 
is discussed in Section II.A, ``ASME BPV Code, Section III,'' because 
the NRC's approval of that Code Case relates to a provision of Section 
III, which is addressed in Sec.  50.55a(b)(1)(ix). The sixth ASME Code 
Case, N-513-3, is discussed in Section II.B, ``ASME BPV Code, Section 
XI,'' because the NRC's approval of that Code Case relates to a 
provision of Section XI, which is addressed in Sec.  
50.55a(b)(2)(xxxiv).
ASME BPV Code Case N-729-4
    On September 10, 2008, the NRC issued a final rule to update Sec.  
50.55a to the 2004 Edition of the ASME BPV Code (73 FR 52730). As part 
of the final rule, Sec.  50.55a(g)(6)(ii)(D) implemented an augmented 
ISI program for the examination of pressurized water reactor RPV upper 
head penetration nozzles and associated partial penetration welds. The 
program required the implementation of ASME BPV Code Case N-729-1, with 
certain conditions.
    The application of ASME BPV Code Case N-729-1 was necessary because 
the inspections required by the 2004 Edition of the ASME BPV Code, 
Section XI were not written to address degradation of the RPV upper 
head penetration nozzles and associated welds by primary water stress 
corrosion cracking (PWSCC). The safety consequences of inadequate 
inspections can be significant. The NRC's determination that the ASME 
BPV Code required inspections are inadequate is based upon operating 
experience and analysis. The absence of an effective inspection regime 
could, over time, result in unacceptable circumferential cracking, or 
the degradation of the RPV upper head or other reactor coolant system 
components by leakage assisted corrosion. These degradation mechanisms 
increase the probability of a loss-of-coolant accident.
    Examination frequencies and methods for RPV upper head penetration 
nozzles and welds are provided in ASME BPV Code Case N-729-1. The use 
of code cases is voluntary, so these provisions were developed, in 
part, with the expectation that the NRC would incorporate the code case 
by reference into the CFR. Therefore, the NRC adopted rule language in 
Sec.  50.55a(g)(6)(ii)(D) requiring implementation of ASME BPV Code 
Case N-729-1, with conditions, in order to enhance the examination 
requirements in the ASME BPV Code, Section XI for RPV upper head 
penetration nozzles and welds. The examinations conducted in accordance 
with ASME BPV Code Case N-729-1 provide reasonable assurance that ASME 
BPV Code allowable limits will not be exceeded and that PWSCC will not 
lead to failure of the RPV upper head penetration nozzles or welds. 
However, the NRC concluded that certain conditions were needed in 
implementing the examinations in ASME BPV Code Case N-729-1. These 
conditions are set forth in Sec.  50.55a(g)(6)(ii)(D).
    On June 22, 2012, the ASME approved the fourth revision of ASME BPV 
Code Case N-729 (N-729-4). This revision changed certain requirements 
based on a consensus review of inspection techniques and frequencies. 
These changes were deemed necessary by the ASME to supersede the 
previous requirements under N-729-1 to establish an effective long-term 
inspection program for the RPV upper head penetration nozzles and 
associated welds in pressurized water reactors. The

[[Page 32953]]

major changes included incorporation of previous NRC conditions in the 
CFR. Minor changes were also made to address editorial issues, to 
correct figures or to add clarity.
    The NRC is updating the requirements of Sec.  50.55a(g)(6)(ii)(D) 
to require licensees to implement ASME BPV Code Case N-729-4, with 
conditions. One existing condition on ASME BPV Code Case N-729-1 has 
been modified, four existing conditions are being deleted in this final 
rule, one existing condition is being redesignated without substantive 
change, and two new conditions--in Sec.  50.55a(g)(6)(ii)(D)(3) and 
(4)--are adopted in this final rule in order to address the changes in 
ASME BPV Code Case N-729-4. The NRC's revisions to the conditions are 
discussed under the next three headings. As discussed earlier, this 
final rule incorporates by reference ASME BPV Code Case N-729-4 into 
Sec.  50.55a(a)(1)(iii)(C).
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
    The NRC is revising Sec.  50.55a(g)(6)(ii)(D)(1) to change the 
version of ASME BPV Code Case N-729 from N-729-1 to N-729-4 for the 
reasons previously set forth. Due to the incorporation of N-729-4, the 
date to establish applicability for licensed pressurized water reactors 
will be changed to the effective date of this final rule.
10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6) (Removed)
    The NRC is removing the existing conditions in Sec.  
50.55a(g)(6)(ii)(D)(2) through (5) and redesignating the condition 
currently in Sec.  50.55a(g)(6)(ii)(D)(6) as Sec.  
50.55a(g)(6)(ii)(D)(2) without any substantive change. The existing 
conditions in Sec.  50.55a(g)(6)(ii)(D)(2) through (5) have all been 
incorporated either verbatim or more conservatively in the revisions to 
ASME BPV Code Case N-729, up to version N-729-4. Therefore, there is no 
reason to retain these conditions in Sec.  50.55a.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency (New 
Condition)
    The NRC is adopting a new condition in Sec.  50.55a(g)(6)(ii)(D)(3) 
to modify the option in ASME BPV Code Case N-729-4 to extend bare metal 
visual inspections of the RPV upper head surface beyond the frequency 
listed in Table 1 of the Code Case. Previously, upper heads aged with 
less than eight effective degradation years were considered to have a 
low probability of initiating PWSCC, the cracking mechanism of concern. 
This ranking of effective degradation years was based on a simple time 
at temperature correlation. All of the upper heads within this 
category, with the exception of new heads using Alloy 600 penetration 
nozzles, were considered to have lower susceptibility to cracking due 
to the upper heads being at or near the cold leg operating temperature 
of the reactor coolant system. Therefore, these plants were referred to 
as having ``cold heads.'' All of the upper heads that had experienced 
cracking prior to 2006 were near the hot leg operating temperature of 
the reactor coolant system, which validated the time at temperature 
model.
    In 2006, one of the 21 ``cold head'' plants identified two 
indications within a penetration nozzle and the associated partial 
penetration weld. Then, between 2006 and 2013, five of the 21 ``cold 
head'' plants identified multiple indications within fifteen different 
penetration nozzles and the associated partial penetration welds. None 
of these indications caused leakage, and volumetric examination of the 
penetration nozzles showed that no flaws in the nozzle material had 
grown through-wall; however, this increasing trend creates a reasonable 
safety concern.
    Recent operational experience has shown that the volumetric 
inspection of penetration nozzles, at the current inspection frequency, 
is adequate to identify indications in the nozzle material prior to 
leakage; however, volumetric examinations cannot be performed on the 
partial penetration welds. Therefore, given the additional cracking 
identified at cold leg temperatures, the NRC staff has concerns about 
the adequacy of the partial penetration weld examinations.
    Leakage from a partial penetration weld into the annulus between 
the nozzle and head material can cause corrosion of the low alloy steel 
head. While initially limited in leak rate, due to limited surface area 
of the weld being in contact with the annulus region, corrosion of the 
vessel head material can expose more of the weld surface to the 
annulus, allowing a greater leak rate. Since an indication in the weld 
cannot be identified by a volumetric inspection, a postulated crack 
through the weld, just about to cause leakage, could exist as a plant 
performed its last volumetric and/or bare metal visual examination of 
the upper head material. This gives the crack years to breach the 
surface and leak prior to the next scheduled visual examination.
    Only a surface examination of the wetted surface of the partial 
penetration weld can reliably detect flaws in the weld. Unfortunately, 
this examination cannot size the flaws in the weld, and, if performed 
manually, requires significant radiological dose to examine all of the 
partial penetration welds on the upper head. As such, the available 
techniques are only able to detect a flaw after it has caused leakage. 
These techniques are a bare metal visual examination or a volumetric 
leak path assessment performed on the frequency of the volumetric 
examination.
    Volumetric leak path examinations are only done during outages when 
a volumetric examination of the nozzle is performed. Therefore, under 
the current requirements allowed by Note 4 of ASME BPV Code Case N-729-
4, leakage from a crack in the weld of a ``cold head'' plant could 
start and continue to grow for the 5 years between the required bare 
metal visual examinations to detect leakage through the partial 
penetration weld.
    Given the additional cracking identified at cold leg temperatures 
of upper head penetration nozzles and associated welds, the NRC finds 
limited basis to continue to categorize these ``cold head'' plants as 
having a low susceptibility to crack initiation. The NRC is increasing 
the frequency of the bare metal visual examinations of ``cold heads'' 
to identify potential leakage as soon as reasonably possible due to the 
volumetric examination limitations. Therefore, the NRC is conditioning 
Note 4 of ASME BPV Code Case N-729-4 to require a bare metal visual 
exam during each outage in which a volumetric exam is not performed. 
The NRC also will allow ``cold head'' plants to extend their bare metal 
visual inspection frequency from once each refueling outage, as stated 
in Table 1 of N-729-1, to once every 5 years, but only if the licensee 
performed a wetted surface examination of all of the partial 
penetration welds during the previous volumetric examination. Applying 
the conditioned bare metal visual inspection frequency or a volumetric 
examination each outage will allow licensees to identify any potential 
leakage through the partial penetration welds prior to significant 
degradation of the low alloy steel head material, thereby providing 
reasonable assurance of the structural integrity of the reactor coolant 
pressure boundary.
    These issues, including the operational experience, the fact that 
volumetric examination is not available to interrogate the partial 
penetration welds, and potential regulatory options, were discussed 
publicly at multiple ASME BPV Code meetings, at the annual Materials 
Programs Technical Information Exchange public meeting

[[Page 32954]]

held at the NRC Headquarters in June 2013, and at the 2013 NRC 
Regulatory Information Conference.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria (New 
Condition)
    The NRC is adopting a new condition in Sec.  50.55a(g)(6)(ii)(D)(4) 
to define surface examination acceptance criteria. Paragraph -3132(b) 
of ASME BPV Code Case N-729-4 sets forth the acceptance criteria for 
surface examinations. In general, throughout Section XI of the ASME BPV 
Code, the acceptance criteria for surface examinations default to 
Section III, Paragraph NB-5352, ``Acceptance Standards.'' Typically, 
for rounded indications, the indication was only unacceptable if it was 
greater than \3/16\-inch in size. The NRC requested that the code case 
authors include a requirement that any size rounded indication causing 
nozzle leakage is unacceptable due to operating experience identifying 
PWSCC under rounded indications less than \3/16\-inch in size.
    Recently, the ASME BPV Code Committee approved an interpretation of 
the language in Paragraph -3132(b), which implied that any size rounded 
indication is acceptable unless there is relevant indication of nozzle 
leakage, even those greater than \3/16\-inch. The NRC does not agree 
with the interpretation and maintains its original position on rounded 
indications that any size rounded indication is unacceptable if there 
is an indication of leakage. Since the adoption of ASME BPV Code Case 
N-729-1 into Sec.  50.55a(g)(6)(ii)(D), all licensees have used the 
NRC's position in implementing Paragraph -3132(b), even after the 
recent ASME BPV Code Committee interpretation approval over NRC 
objection.
    Therefore, in order to ensure compliance with the previous and 
ongoing requirement, the NRC is revising condition Sec.  
50.55a(g)(6)(ii)(D)(4) to include clarity within the acceptance 
criteria for surface examinations. The current edition requirements of 
NB-5352 of ASME BPV Code, Section III for the licensee's ongoing 10-
year inservice inspection interval shall be met.
ASME BPV Code Case N-770-2
    On June 21, 2011 (76 FR 36232), the NRC issued a final rule, which 
included Sec.  50.55a(g)(6)(ii)(F) that requires the implementation of 
ASME BPV Code Case N-770-1, ``Alternative Examination Requirements and 
Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt 
Welds Fabricated with UNS N06082 or UNS N86182 Weld Filler Material 
With or Without Application of Listed Mitigation Activities,'' with 
certain conditions.
    On June 9, 2011, the ASME approved the second revision of ASME BPV 
Code Case N-770 (N-770-2). The major changes from N-770-1 to N-770-2 
included establishing new ASME BPV Code Case, Table 1, inspection item 
classifications for optimized weld overlays and allowing alternatives 
when complete inspection coverage cannot be met. Minor changes were 
also made to address editorial issues, to correct figures, or to add 
clarity. The NRC found that the updates and improvements in N-770-2 are 
sufficient to update Sec.  50.55a(g)(6)(ii)(F).
    The NRC, therefore, is updating the requirements of Sec.  
50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code 
Case N-770-2, with conditions. The NRC conditions have been modified to 
address the changes in ASME BPV Code Case N-770-2 and to ensure that 
this regulatory framework will provide adequate protection of public 
health and safety. The following sections discuss each of the NRC's 
changes to the conditions on ASME BPV Code Case N-770-2. As discussed 
earlier, this final rule incorporates by reference ASME BPV Code Case 
N-770-2 into Sec.  50.55a(a)(1)(iii)(D).
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(1) to change the 
version of ASME BPV Code Case N-770 from N-770-1 to N-770-2 and to 
require its implementation, with conditions, to incorporate the updates 
and improvements contained in N-770-2. The NRC will allow licensees to 
begin using N-770-2 on the effective date of this rule.
10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(2) to provide 
clarification regarding categorization of each Alloy 82/182 butt weld, 
mitigated or not, under N-770-2. This paragraph also clarifies the 
NRC's position that Paragraph -1100(e) shall not be used to exempt 
welds that rely on Alloy 82/182 for structural integrity from more 
frequent ISI schedules until the NRC has reviewed and authorized an 
alternative categorization for the weld. Additionally, the NRC will 
change the inspection item categories for full structural weld overlays 
from C to C-1 and F to F-1 due to reclassification under ASME BPV Code 
Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(3) to clarify the 
baseline examination requirements by stating that previously-conducted 
examinations, in order to count as baseline examinations, must meet the 
requirements of ASME BPV Code Case N-770-2, as conditioned in this 
section. The 2011 rule required the use of ASME BPV Code Section XI 
Appendix VIII qualifications for baseline examinations, which is 
stricter than N-770-2 and does not provide requirements for optimized 
weld overlays. The revision also updates the deadline for baseline 
examination requirements, since the January 20, 2012, deadline from the 
previous rule has passed. Finally, upon implementation of this rule, if 
a licensee is currently in an outage, then the baseline inspection 
requirement can be met by performing the inspections in accordance with 
the previous regulatory requirements of Sec.  50.55a(g)(6)(ii)(F), in 
lieu of the examination requirements of Paragraphs -2500(a) or -2500(b) 
of ASME BPV Code Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(4) to define 
examination coverage for circumferential flaws and to prohibit the use 
of Paragraph -2500(d) of ASME BPV Code Case N-770-2 which, in some 
circumstances, allows unacceptably low examination coverage. Paragraph 
-2500(d) of N-770-2 would allow the reduction of circumferential 
volumetric examination coverage with analytical evaluation. Paragraph -
2500(c) was previously prohibited from use, and it continues to be 
prohibited. The NRC is establishing an essentially 100 percent 
volumetric examination coverage requirement, including greater than 90 
percent of the required volumetric examination coverage, for 
circumferential flaws to provide reasonable assurance of structural 
integrity of all ASME BPV Code Class 1 butt welds susceptible to PWSCC. 
Therefore, the NRC is adopting a condition prohibiting the use of 
Paragraphs -2500(c) and -2500(d). A licensee may request approval for 
use of these paragraphs under 10 CFR 50.55a(z).
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(5) to add the 
explanatory heading, ``Inlay/onlay

[[Page 32955]]

inspection frequency,'' and to make minor editorial corrections.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(6) to add the 
explanatory heading, ``Reporting requirements.''
10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(7) to add the 
explanatory heading, ``Defining `t'.''
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(8) to add the 
explanatory heading, ``Optimized weld overlay examination,'' and to 
maintain the requirement for the timing of the initial inservice 
examination of optimized weld overlays.
    Uncracked welds mitigated with optimized weld overlays were re-
categorized by ASME BPV Code Case N-770-2 from Inspection Item D to 
Inspection Item C-2; however, the initial inspection requirement was 
not incorporated into the Code Case for Inspection Item C-2. The NRC 
has determined that uncracked welds mitigated with an optimized weld 
overlay must have an initial inservice examination no sooner than the 
third refueling outage and no later than 10 years following the 
application of the weld overlay to identify unacceptable crack growth. 
Optimized weld overlays establish compressive stress on the inner half 
thickness of the weld, but the outer half thickness may also be under 
tensile stress. The requirement for an initial inservice examination no 
sooner than the third refueling outage and no later than 10 years 
following the application of the weld overlay is based on the design of 
optimized weld overlays, which require the outer quarter thickness of 
the susceptible material to provide structural integrity for the weld. 
Therefore, the NRC is continuing adoption of the condition, which 
requires the initial inservice examination of uncracked welds mitigated 
by optimized weld overlay (i.e., the welds which are subject to 
Inspection Item C-2 of ASME BPV Code Case N-770-2) within the specified 
timeframe.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(9) to add the 
explanatory heading, ``Deferral,'' and to address changes in ASME BPV 
Code Case N-770-2 which allow the deferral of the first inservice 
examination of uncracked welds mitigated with optimized weld overlays, 
Inspection Item C-2.
    Previously, under N-770-1, the initial inservice examination of 
these welds was not allowed to be deferred. Allowing deferral of the 
initial inservice examination in accordance with N-770-2 could, in 
certain circumstances, allow the initial inservice examination to be 
performed up to 20 years after installation. Therefore, the NRC is 
adopting a condition which would preclude the deferral of the initial 
inservice examination of uncracked welds mitigated by optimized weld 
overlays.
10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(10) to add the 
explanatory heading, ``Examination technique,'' and to address changes 
in ASME BPV Code Case N-770-2. Note 14(a) of Table 1 of ASME BPV Code 
Case N-770-2 provides the previously required full examination 
requirement for optimized weld overlays. The language of ASME BPV Code 
Case N-770-2, however, does not require the implementation of the full 
examination requirements of Note 14(a) of Table 1, if possible, before 
implementing the reduced examination coverage requirements of Note 
14(b) of Table 1 or Note (b) of Figure 5(a). The NRC agrees that 
reduced examination coverage is the best alternative if the full 
examination cannot be met; however, the full examination requirement 
should be implemented, if possible, before the option of reduced 
examination coverage is allowed. Therefore, the NRC is modifying the 
current condition in Sec.  50.55a(g)(6)(ii)(F)(10) to allow the use of 
Note 14(b) of Table 1 and Note (b) of Figure 5(a) of ASME BPV Code Case 
N-770-2 only after the determination that the requirements of Note 
14(a) of Table 1 of ASME BPV Code Case N-770-2 cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(11) to address 
examination requirements through cast stainless steel materials by 
requiring the use of Appendix VIII qualifications to meet the 
inspection requirements of Paragraph -2500(a) of ASME BPV Code Case N-
770-2. The requirements for volumetric examination of butt welds 
through cast stainless steel materials are currently being developed as 
Supplement 9 to the ASME BPV Code, Section XI, Appendix VIII. In 
accordance with Appendix VIII for supplements that have not been 
developed, the requirements of Appendix III apply. Appendix III 
requirements are not equivalent to Appendix VIII requirements. For the 
volumetric examination of ASME BPV Code Class 1 welds, the NRC has 
established the requirement for examination qualification under the 
Appendix VIII. Therefore, the NRC is adopting a condition requiring the 
use of Appendix VIII qualifications to meet the inspection requirements 
of Paragraph -2500(a) of ASME BPV Code Case N-770-2 by January 1, 2022.
    The development of a sufficient number of mockups would be required 
to establish an Appendix VIII program for examination of ASME BPV Code 
Class 1 piping and vessel nozzle butt welds through cast stainless 
steel materials. The NRC recognizes that significant time and resources 
are required to create mockups and to allow for qualification of 
equipment, procedures and personnel. Therefore, the NRC is requiring 
licensees to use these Appendix VIII qualifications no later than their 
first scheduled weld examinations involving cast stainless steel 
materials occurring after January 1, 2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(12) to clarify the 
examination coverage requirements allowed under Appendix I of ASME BPV 
Code Case N-770-2 for butt welds joining cast stainless steel material. 
Under current ASME BPV Code, Section XI, Appendix VIII requirements, 
the volumetric examination of butt welds through cast stainless steel 
materials is under Supplement 9. Supplement 9 rules are still being 
developed by the ASME BPV Code. Therefore, it is currently impossible 
to meet the requirement of Paragraph I.5.1 for butt welds joining cast 
stainless steel material.
    The material of concern is the weld material susceptible to PWSCC 
adjoining the cast stainless steel material. Appendix VIII qualified 
procedures are available to perform the inspection of the susceptible 
weld material, but they are not qualified to inspect the cast stainless 
steel materials. Therefore, the NRC is adopting a condition changing 
the inspection volume for stress-improved dissimilar metal welds with 
cast stainless steel from the ASME BPV Code Section XI requirements to 
``the maximum extent practical including 100 percent of the susceptible 
material volume.'' This will

[[Page 32956]]

remain applicable until an Appendix VIII qualified procedure for the 
inspection through cast stainless steel materials is available in 
accordance with the new condition in Sec.  50.55a(g)(6)(ii)(F)(11).
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(13) to require the 
encoding of ultrasonic volumetric examinations of Inspection Items A-1, 
A-2, B, E, F-2, J, and K in Table 1 of N-770-2. A human performance gap 
has been found between some ultrasonic testing procedures, as 
demonstrated during ASME BPV Code, Section XI, Appendix VIII 
qualification versus as applied in the field.
    The human factors that contributed to the licensee-performed 
examinations which failed to identify significant flaws at North Anna 
Power Station, Unit 1 in 2012 (Licensee Event Report 50-338/2012-001-
00) and at Diablo Canyon Nuclear Power Plant in 2013 (Relief Request 
REP-1 U2, Revision 2) can be avoided by the use of encoded ultrasonic 
examinations. Encoded ultrasonic examinations electronically store both 
the positional and ultrasonic information from the inspections. Encoded 
examinations allow for the inspector to evaluate the data and search 
for indications outside of a time limited environment to assure that 
the inspection was conducted properly and to allow for sufficient time 
to analyze the data. Additionally, the encoded examination would allow 
for an independent review of the data by other inspectors or an 
independent third party. Finally, the encoded examination could be 
compared to previous and/or future encoded examinations to determine if 
flaws are present and flaw growth rates. Therefore, the NRC is adopting 
a condition requiring the use of encoding for ultrasonic volumetric 
examinations of non-mitigated or cracked mitigated dissimilar metal 
butt welds in the reactor coolant pressure boundary which are within 
the scope of ASME BPV Code Case N-770-2.
ASME BPV Code Case N-824
10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii) to allow licensees to 
use the provisions of ASME BPV Code Case N-824, ``Ultrasonic 
Examination of Cast Austenitic Piping Welds From the Outside Surface 
Section XI, Division 1,'' subject to four conditions in Sec.  
50.55a(b)(2)(xxxvii)(A) through (D), when implementing inservice 
examinations in accordance with the ASME BPV Code, Section XI 
requirements.
    During the construction of nuclear power plants, it was recognized 
that the grain structure of cast austenitic stainless steel (CASS) 
could prevent effective ultrasonic inspections of piping welds where 
one or both sides of the welds were constructed of CASS. The high 
strength and toughness of CASS (prior to thermal embrittlement) made it 
desirable as a building material despite this known inspection issue. 
This choice of construction materials has rendered many pressure 
boundary components without a means to reliably inspect them 
volumetrically. While there is no operational experience of a CASS 
component failing, as part of the reactor pressure boundary, inservice 
volumetric inspection of these components is necessary to provide 
reasonable assurance of their structural integrity.
    The current regulatory requirements for the examination of CASS, 
provided in Sec.  50.55a, do not provide sufficient guidance to assure 
that the CASS components are being inspected adequately. To illustrate 
that ASME BPV Code does not provide adequate guidance, ASME BPV Code, 
Section XI, Appendix III, Supplement 1 states, ``Cast materials may 
preclude meaningful examinations because of geometry and attenuation 
variables.'' For this reason, over the past several decades, licensees 
have been unable to perform effective inspections of welds joining CASS 
components. To allow for continued operation of their plants, licensees 
submitted hundreds of requests for relief from the ASME BPV Code 
requirements for inservice inspection of CASS components to the NRC, 
resulting in a significant regulatory burden.
    The recent advances in inspection technology are driving renewed 
work at ASME BPV Code meetings to produce Section XI, Appendix VIII, 
Supplement 9 to resolve the CASS inspection issue, but it will be years 
before these code updates will be published, as well as additional time 
to qualify and approve procedures for use in the field. Until then, 
licensees would still use the requirements of ASME BPV Code Section XI, 
Appendix III, Supplement 1, which states that inspection of CASS 
materials meeting the ASME BPV Code requirements may not be meaningful. 
Consequently, less effective examinations would continue to be used in 
the field, and more relief requests would be generated between now and 
the implementation of Supplement 9.
    The NRC commissioned a research program to determine the 
effectiveness of the new technologies for inspections of CASS 
components in an effort to resolve some of the known inspection issues. 
The result of this work is published in NUREG/CR-6933, ``Assessment of 
Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using 
Advanced Low-Frequency Ultrasonic Methods'', March 2007, and NUREG/CR-
7122, ``An Evaluation of Ultrasonic Phased Array Testing for Cast 
Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,'' March 
2012. Based on the improvements in ultrasonic inspection technology and 
techniques for CASS components, the ASME approved BPV Code Case N-824 
(N-824) on October 16, 2012, which describes how to develop a procedure 
capable of meaningfully inspecting welds in CASS components.
    Effective examinations of CASS components require the use of lower 
frequencies and larger transducers than are typically used for 
ultrasonic inspections of piping welds and would require licensees to 
modify their inspection procedures. The NRC recognizes that requiring 
the use of spatial encoding will limit the full implementation of ASME 
BPV Code Case N-824, as spatial encoding is not practical for many weld 
configurations.
    At this time, the use of ASME BPV Code Case N-824, as conditioned, 
is the most effective known method for adequately examining welds with 
one or more CASS components. With the use of ASME BPV Code Case N-824, 
as conditioned, licensees will be able to take full credit for 
completion of the Sec.  50.55a required inservice volumetric inspection 
of welds involving CASS components. The implementation of ASME BPV Code 
Case N-824, as conditioned, will have the dual effect of improving the 
rigor of required volumetric inspections and reducing the number of 
uninspectable Class 1 and Class 2 pressure retaining welds.
    The NRC concludes that incorporation of ASME BPV Code Case N-824, 
subject to the four conditions in Sec.  50.55a(b)(2)(xxxvii)(A) through 
(D), will significantly improve the flaw detection capability of 
ultrasonic inspection of CASS components until Supplement 9 is 
implemented, thereby providing reasonable assurance of leak tightness 
and structural integrity. Additionally, it will reduce the regulatory 
burden on licensees and allow licensees to submit fewer relief requests 
for welds in CASS materials. The four conditions on the use of ASME BPV 
Code Case N-824,

[[Page 32957]]

Sec.  50.55a(b)(2)(xxxvii)(A) through (D), are discussed in the next 
four headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) (First Condition on Use of ASME BPV Code 
Case N-824)
    The NRC, based upon NUREG/CR-6933 and NUREG/CR-7122, has determined 
that inspections of CASS materials are very challenging, and sufficient 
technical basis exists to condition the code case to bring the code 
case into agreement with the NUREG/CR reports. The NUREG/CR reports 
also show that CASS materials produce high levels of coherent noise. 
The noise signals can be confusing and mask flaw indications. Use of 
encoded inspection data allows the inspector to mitigate this problem 
through the ability to electronically manipulate the data, which allows 
for discrimination between coherent noise and flaw indications. The NRC 
found that encoding CASS inspection data provides significant detection 
benefits. Therefore, the NRC is adding a condition in Sec.  
50.55a(b)(2)(xxxvii)(A) to require the use of encoded data when 
utilizing N-824 for the examination of CASS components.
10 CFR 50.55a(b)(2)(xxxvii)(B) (Second Condition on Use of ASME BPV 
Code Case N-824)
    The use of dual element phased-array search units showed the most 
promise in obtaining meaningful responses from flaws. For this reason, 
the NRC is adding a condition in Sec.  50.55a(b)(2)(xxxvii)(B) to 
require the use of dual, transmit-receive, refracted longitudinal wave, 
multi-element phased array search units when utilizing N-824 for the 
examination of CASS components.
10 CFR 50.55a(b)(2)(xxxvii)(C) (Third Condition on Use of ASME BPV Code 
Case N-824)
    The optimum inspection frequencies for examining CASS components of 
various thicknesses are described in NUREG/CR-6933 and NUREG/CR-7122. 
For this reason, the NRC is adding a condition in Sec.  
50.55a(b)(2)(xxxvii)(C) to require that ultrasonic examinations 
performed to implement ASME BPV Code Case N-824 on piping greater than 
1.6 inches (41 mm) thick shall use a phased array search unit with a 
center frequency of 500 kHz with a tolerance of + /- 20 percent.
10 CFR 50.55a(b)(2)(xxxvii)(D) (Fourth Condition on Use of ASME BPV 
Code Case N-824)
    NUREG/CR-6933 shows that the grain structure of CASS can reduce the 
effectiveness of some inspection angles. For this reason, the NRC is 
adding a condition in Sec.  50.55a(b)(2)(xxxvii)(D) to require that 
ultrasonic examinations performed to implement ASME BPV Code Case N-824 
shall use a phased array search unit which produces angles including, 
but not limited to, 30 to 55 degrees with a maximum increment of 5 
degrees.
OM Code Case OMN-20
10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20
    The NRC is adding Sec.  50.55a(b)(3)(x) to allow licensees to 
implement OM Code Case OMN-20, ``Inservice Test Frequency,'' in the OM 
Code, 2012 Edition, for the editions and addenda of the OM Code that 
are listed in Sec.  50.55a(a)(1)(iv) as being approved for 
incorporation by reference. As a conforming change, Sec.  
50.55a(a)(1)(iii)(G) is being added to incorporate by reference OM Code 
Case OMN-20 into Sec.  50.55a.
    Surveillance Requirement (SR) 3.0.3 from TS 5.5.6, ``Inservice 
Testing Program,'' allows licensees to apply a delay period before 
declaring the SR for TS equipment ``not met'' when the licensee 
inadvertently exceeds or misses the time limit for performing TS 
surveillance. Licensees have been applying SR 3.0.3 to inservice tests. 
The NRC has determined that licensees cannot use TS 5.5.6 to apply SR 
3.0.3 to inservice tests under Sec.  50.55a(f) that are not associated 
with a TS surveillance. To invoke SR 3.0.3, the licensee shall first 
discover that a TS surveillance was not performed at its specified 
frequency. Therefore, the delay period that SR 3.0.3 provides does not 
apply to non-TS support components tested under Sec.  50.55a(f). The OM 
Code does not provide for any inservice test frequency reductions or 
extensions. In order to provide inservice test frequency reductions or 
extensions that can no longer be provided by SR 3.0.3 from TS 5.5.6, 
the ASME has developed OM Code Case OMN-20. The NRC has reviewed OM 
Code Case OMN-20 and has found it acceptable for use. The NRC 
determined that OM Code Case OMN-20 may be safely used for all 
licensees using editions and addenda of the OM Code that are listed in 
Sec.  50.55a(a)(1)(iv). The NRC will include OM Code Case OMN-20 in the 
next revision of RG 1.192, at which time a conforming change will be 
made to delete both this paragraph and Sec.  50.55a(a)(1)(iii)(G).

III. Opportunities for Public Participation

    The proposed rule was published on September 18, 2015, for a 75-day 
comment period (80 FR 56820). The public comment period closed on 
December 2, 2015.
    After the close of the public comment period, the NRC held a public 
meeting on March 2, 2016, to discuss the proposed rule, to answer 
questions on specific provisions of the proposed rule, and to discuss 
public comments received on the proposed rule in order to enhance the 
NRC's understanding of the comments. The public meeting summary is 
available in ADAMS under Accession No. ML16069A408.

IV. NRC Responses to Public Comments

    The NRC received 27 letters and emails in response to the 
opportunity for public comment on the proposed rule. These comment 
submissions were submitted by the following commenters (listed in order 
of receipt):

1. Private citizen, Edward Cavey
2. Private citizen, Dale Matthews
3. Private citizen, Ron Clow
4. ASME
5. Iddeal Solutions, LLC
6. Electric Power Research Institute (EPRI)
7. Private citizen, William Taylor
8. ASME
9. Private citizen, Dan Nowakowski
10. Wolf Creek Nuclear Operating Corporation
11. Northern States Power Company--Minnesota
12. FirstEnergy Nuclear Operating Company
13. PSEG Nuclear
14. Dominion Resources Services, Inc.
15. Private citizen, Terence Chan
16. Nuclear Energy Institute
17. EPRI
18. Duke Energy
19. Private Citizen, William Taylor
20. Dominion Engineering, Inc.
21. Tennessee Valley Authority
22. Southern Nuclear Operating Company
23. Prairie Island Nuclear Plant
24. Inservice Test Owners Group
25. Exelon Generation Company
26. EPRI
27. EPRI

    In general, the comments:
     Suggested revising or rewording conditions to make them 
clearer.
     Supported incorporation of Code Cases N-729-4, N-770-2, N-
824, or OMN-20 into Sec.  50.55a.
     Supported the proposed changes to add or remove 
conditions.
     Opposed proposed conditions.
     Supplied additional information for NRC consideration.
     Proposed rewriting or renumbering of paragraphs.

[[Page 32958]]

     Asked questions or requested information from the NRC.
    Due to the large number of comments received and the length of the 
NRC's responses, this document summarizes the NRC's response to 
comments in areas of particular interest to stakeholders that prompted 
the NRC to make changes in this final rule from what was proposed. A 
discussion of all comments and complete NRC responses are presented in 
a separate document, ``2017 Final Rule (10 CFR 50.55a) American Society 
of Mechanical Engineers Codes and Code Cases: Analysis of Public 
Comments,'' (ADAMS Accession No. ML16130A531).
10 CFR 50.55a(a)(1)(ii), (b)(2); Nonmandatory Appendix U
    Public commenters were concerned that the NRC was proposing to 
exclude incorporating by reference Nonmandatory Appendix U because 
Nonmandatory Appendix U is the incorporation of the provisions of ASME 
BPV Code Cases N-513-3 and N-705, without any technical changes, into 
the Section XI Code. The NRC agrees with this comment, in that ASME BPV 
Code Cases N-513-3 and N-705 have been approved in RG 1.147. Based on 
these comments, the NRC has removed the proposed exclusion of 
Nonmandatory Appendix U from this final rule. However, the NRC has 
found it necessary to apply two new conditions in Sec.  
50.55a(b)(2)(xxxiv)(A) and (B) to Nonmandatory Appendix U. The first 
condition provides regulatory consistency with the approval of the code 
cases in RG 1.147. The second condition requires the use of an Appendix 
from ASME BPV Code Case N-513-3 that was unintentionally omitted from 
Appendix U. The NRC discussed these changes at the March 2, 2016, 
public meeting, and the NRC considered the public feedback from that 
meeting when developing this final rule.
10 CFR 50.55a(b)(2)(xii), Underwater Welding
    Public commenters were concerned that the proposed rule continued 
to prohibit the use of underwater welding in Sec.  50.55a(b)(2)(xii), 
when changes were made to address this condition in the 2010 Edition of 
Section XI. The NRC agrees that the condition should be modified to 
address the changes in the Code. After consideration of the public 
comments, the NRC noted other inconsistencies for addressing welding on 
irradiated materials that appear in the Code and in some Code Cases. 
Section 50.55a(b)(2)(xii) of this final rule reflects a change to 
include two conditions that provide consistency for welding of 
irradiated materials. The NRC discussed these changes at the March 2, 
2016, public meeting, and the NRC considered the public feedback from 
that meeting when developing this final rule.
10 CFR 50.55a(b)(2)(xxxi), Mechanical Clamping Devices
    Public commenters were concerned that the wording of the proposed 
condition in Sec.  50.55a(b)(2)(xxxi) was unclear and that citing the 
specific paragraphs of Section XI to which the NRC is taking exception 
would be clearer. The NRC agrees. To clarify the requirement for the 
implementation of mechanical clamps, the condition was changed to 
require the use of Appendix W of Section XI when using mechanical 
clamps. Additionally, use of IWA-4131.1(c) of the 2010 Edition of 
Section XI and IWA-4131.1(d) of the 2011 Addenda of the 2010 Edition 
and later versions of Section XI is prohibited. Identifying these 
specific subparagraphs was deemed necessary, as they may have caused 
confusion with the intended purpose of the original proposed condition 
in maintaining the previous regulatory requirements for mechanical 
clamping devices. Section 50.55a(b)(2)(xxxi) of this final rule 
reflects this change.
10 CFR 50.55a(b)(2)(xxxvii), ASME BPV Code Case N-824
    Public commenters had concerns with conditions proposed on ASME BPV 
Code Case N-824, ``Ultrasonic Examination of Cast Austenitic Piping 
Welds From the Outside Surface Section XI, Division 1,'' in Sec.  
50.55a(b)(2)(xxxvii)(A) through (E). There were concerns that the 
conditions would limit the use of Code Case N-824 and that some 
conditions did not have a sufficient technical basis. The NRC partially 
agreed with the comments requesting the removal and modification of 
some conditions in Sec.  50.55a(b)(2)(xxxvii) restricting the 
frequencies and angles usable on some cast austenitic welds. Based on 
the public comments, one condition was removed entirely and two others 
were modified. Section 50.55a(b)(2)(xxxvii)(A) through (D) of this 
final rule contain the modified and reduced conditions on the use of 
ASME BPV Code Case N-824. The NRC discussed these changes at the March 
2, 2016, public meeting, and the NRC considered the public feedback 
from that meeting when developing this final rule.
10 CFR 50.55a(b)(3)(xi), OM Condition: Valve Position Indication
    Public commenters raised concerns regarding the proposed condition 
in Sec.  50.55a(b)(3)(xi) to emphasize the OM Code provisions in 
Subsection ISTC-3700, ``Position Verification Testing,'' to verify that 
valve operation is accurately indicated. Public commenters indicated 
that because of the significance of implementing the condition, some 
licensees might need time to revise or create procedures to govern the 
implementation of this condition. Public commenters also suggested that 
the condition be limited to active valves. The NRC partially agrees and 
partially disagrees with these comments. The NRC agrees that additional 
time to implement the condition regarding valve position verification 
is appropriate. Therefore, the NRC has revised the condition to 
indicate that it will be effective with implementation of the 2012 
Edition of the OM Code. The NRC staff does not agree with the 
suggestion to limit the condition to active valves because the OM Code 
requires that passive valves undergo periodic verification of position 
indication.

V. Section-by-Section Analysis

Administrative Changes

    The NRC is removing the revision number of the three RGs currently 
approved by the Office of the Federal Register for incorporation by 
reference throughout the substantive provisions of Sec.  50.55a 
addressing the ASME Code Cases, i.e., paragraphs (b) through (g). The 
revision numbers for the RGs approved for incorporation by reference 
(currently, RG 1.84, RG 1.147, and RG 1.192) will be retained in Sec.  
50.55a(a)(3)(i) through (iii), where the RGs are listed by full title, 
including revision number. That paragraph identifies the specific 
materials which the Office of the Federal Register has approved for 
incorporation by reference, as required by Office of the Federal 
Register requirements in 1 CFR 51.9. Readers would need to refer to 
Sec.  50.55a(a) to determine the specific revision of the relevant RG 
that is approved for incorporation by reference by the Office of the 
Federal Register. These changes are administrative in nature and do not 
change substantive requirements with respect to the RGs and the Code 
Cases listed in the RGs.
10 CFR 50.55a(a) Documents Approved for Incorporation by Reference
    The NRC is revising the incorporation by reference language to 
update the

[[Page 32959]]

contact information for the NRC Technical Library.
10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section 
III
    The NRC is revising Sec.  50.55a(a)(1)(i) to clarify that Section 
III Nonmandatory Appendices of the listed editions and addenda are 
excluded from the incorporation by reference. The exclusion was 
originally added in a final rule published on June 21, 2011 (76 FR 
36232); however, it was erroneously omitted from the final rule 
published on November 5, 2014 (79 FR 65776). The NRC is correcting the 
omission in this final rule by inserting ``(excluding Nonmandatory 
Appendices)'' in Sec.  50.55a(a)(1)(i). The NRC is relocating the 
definition of the term ``BPV Code,'' which is used throughout the 
section, from Sec.  50.55a(b) to Sec.  50.55a(a)(1)(i).
10 CFR 50.55a(a)(1)(i)(E) ``Rules for Construction of Nuclear Facility 
Components--Division 1''
    The NRC is revising Sec.  50.55a(a)(1)(i)(E) to add ASME BPV Code, 
Section III 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition.
10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section 
XI
    The NRC is revising Sec.  50.55a(a)(1)(ii) to include two minor 
editorial changes: to replace ``Boiler and Pressure Vessel Code'' with 
``BPV Code'' and to replace ``limited to'' with ``limited by.''
10 CFR 50.55a(a)(1)(ii)(C)(52) and (53) ``Rules for Inservice 
Inspection of Nuclear Power Plant Components--Division 1''
    The NRC is revising Sec.  50.55a(a)(1)(ii)(C)(52) and (53) to add 
ASME BPV Code, Section XI 2009 Addenda, 2010 Edition, 2011 Addenda, and 
2013 Edition. The examination requirements for Examination Category B-
F, Item Numbers B5.11 and B5.71, Nozzle-to-Component Butt Welds in the 
2011 Addenda and the 2013 Edition of ASME BPV Code, Section XI are 
expressly excluded from the incorporation by reference in Sec.  
50.55a(a)(1)(ii)(C)(52) and, therefore, not approved for use. 
Similarly, the requirements of IWB-3112(a)(3) and IWC-3112(a)(3) in the 
2013 Edition of ASME BPV Code, Section XI are expressly excluded from 
the incorporation by reference in Sec.  50.55a(a)(1)(ii)(C)(53) and are 
not approved for use.
10 CFR 50.55a(a)(1)(iii)(A) ASME BPV Code Case N-513-3 Mandatory 
Appendix I
    The NRC is revising Sec.  50.55a(a)(1)(iii)(A) to include 
information for a new standard that is being incorporated by reference, 
entitled, ``ASME BPV Code Case N-513-3 Mandatory Appendix I.''
10 CFR 50.55a(a)(1)(iii)(B) ASME BPV Code Case N-722-1
    The NRC is revising Sec.  50.55a(a)(1)(iii)(B) to maintain 
alphanumeric order for the ASME Code Cases listed in Sec.  
50.55a(a)(1)(iii). ASME BPV Code Case N-722-1 was previously approved 
for incorporation by reference.
10 CFR 50.55a(a)(1)(iii)(C) ASME BPV Code Case N-729-4
    The NRC is revising Sec.  50.55a(a)(1)(iii)(C) to add the title 
``ASME BPV Code Case N-729-4,'' and include information for the 
standard that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(D) ASME BPV Code Case N-770-2
    The NRC is adding Sec.  50.55a(a)(1)(iii)(D) to add the title 
``ASME BPV Code Case N-770-2,'' and include information for the 
standard that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(E) ASME BPV Code Case N-824
    The NRC is adding Sec.  50.55a(a)(1)(iii)(E) to include information 
for a new standard that is being incorporated by reference, entitled, 
``ASME BPV Code Case N-824.''
10 CFR 50.55a(a)(1)(iii)(F) ASME BPV Code Case N-852
    The NRC is adding Sec.  50.55a(a)(1)(iii)(F) to include information 
for a new standard that is being incorporated by reference, entitled, 
``ASME BPV Code Case N-852.''
10 CFR 50.55a(a)(1)(iii)(G) ASME OM Code Case OMN-20
    The NRC is adding Sec.  50.55a(a)(1)(iii)(G) to include information 
for a new standard that is being incorporated by reference, entitled, 
``ASME OM Code Case OMN-20.''
10 CFR 50.55a(a)(1)(iv) ASME Operation and Maintenance Code
    The NRC is revising Sec.  50.55a(a)(1)(iv) to correct the title of 
the OM Code and to relocate the definition of the term ``OM Code,'' 
which is used throughout the section, from Sec.  50.55a(b) to Sec.  
50.55a(a)(1)(iv).
10 CFR 50.55a(a)(1)(iv)(B) ``Operation and Maintenance of Nuclear Power 
Plants, Division 1: Section IST Rules for Inservice Testing of Light-
Water Reactor Power Plants''
    The NRC is adding new Sec.  50.55a(a)(1)(iv)(B) to include ASME OM 
Code 2009 Edition and 2011 Addenda.
10 CFR 50.55a(a)(1)(iv)(C) ``Operation and Maintenance of Nuclear Power 
Plants, Division 1: OM Code: Section IST''
    The NRC is adding new Sec.  50.55a(a)(1)(iv)(C) to include ASME OM 
Code 2012 Edition.
10 CFR 50.55a(a)(1)(v) ASME Quality Assurance Requirements
    The NRC is adding new Sec.  50.55a(a)(1)(v) to include information 
regarding NQA-1 standards and add the title ``ASME Quality Assurance 
Requirements'' for ASME NQA-1 Code as part of NRC titling convention.
10 CFR 50.55a(b) Use and Conditions on the Use of Standards
    The NRC is revising Sec.  50.55a(b) to correct the title of the 
ASME OM Code.
10 CFR 50.55a(b)(1) Conditions on ASME BPV Code Section III
    The NRC is revising Sec.  50.55a(b)(1) to reflect the latest 
edition incorporated by reference, the 2013 Edition.
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
    The NRC is revising Sec.  50.55a(b)(1)(ii) to clarify rule language 
and add Table I, which clarifies prohibited Section III provisions for 
welds with leg size less than 1.09 tn in tabular form.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
    The NRC is revising Sec.  50.55a(b)(1)(iv) to clarify that it 
allows, but does not require, applicants and licensees to use the 2008 
Edition through the 2009-1a Addenda of NQA-1 when applying the 2010 
Edition and later editions of the ASME BPV Code, Section III, up to the 
2013 Edition. Applicants and licensees are required to meet appendix B 
of 10 CFR part 50, and NQA-1 is one way of meeting portions of appendix 
B. An applicant or licensee may select any version of NQA-1 that has 
been approved for use in Sec.  50.55a, but they must also use the 
administrative, quality, and technical provisions contained in the 
version of NCA-4000 referencing that Edition or Addenda of

[[Page 32960]]

NQA-1 selected by the applicant or licensee.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1, as modified and supplemented 
by NCA-4000, does not meet all of the requirements of appendix B to 10 
CFR part 50.
    Section 50.55a(b)(1)(iv) clarifies that applicants and licensees 
using NQA-1 are also required to meet appendix B to 10 CFR part 50 and 
the commitments contained in their QA program descriptions. To meet the 
requirements of appendix B, when using NQA-1 during the design and 
construction phase, applicants and licensees must address, in their 
quality program description, those areas where NQA-1 is insufficient to 
meet appendix B. Additional guidance and regulatory positions on how to 
meet appendix B when using NQA-1 are provided in RG 1.28, ``Quality 
Assurance Program Criteria (Design and Construction).''
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification 
and Demonstration of Function of Incompressible-Fluid Pressure-Relief 
Valves
    The NRC is revising Sec.  50.55a(b)(1)(vii) to reflect the editions 
and addenda of the ASME BPV Code incorporated by reference in this 
rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME 
Certification Marks
    The NRC is adding Sec.  50.55a(b)(1)(viii) to allow licensees to 
use either the ASME BPV Code Symbol Stamp or ASME Certification Mark 
with the appropriate certification designator and class designator as 
specified in the 2013 Edition through the latest edition and addenda 
incorporated by reference in Sec.  50.55a.
10 CFR 50.55a(b)(1)(ix) Section III Condition: NPT Code Symbol Stamps
    The NRC is adding Sec.  50.55a(b)(1)(ix) to allow licensees to use 
the NPT Code Symbol Stamp with the letters arranged horizontally as 
specified in ASME BPV Code Case N-852 for the service life of a 
component that had the NPT Code Symbol Stamp applied during the time 
period from January 1, 2005, through December 31, 2015.
10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI
    The NRC is revising Sec.  50.55a(b)(2) to reflect the editions and 
addenda of the ASME BPV Code incorporated by reference in this 
rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and 
Addenda of Subsection IWE and Subsection IWL
    The NRC is revising Sec.  50.55a(b)(2)(vi) to clarify that the 
provision applies only to the class of licensees of operating reactors 
that were required by previous versions of Sec.  50.55a to develop and 
implement a containment ISI program in accordance with Subsection IWE 
and Subsection IWL, and complete an expedited examination of 
containment during the 5-year period from September 9, 1996 to 
September 9, 2001.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment 
Examinations
    The NRC is revising Sec.  50.55a(b)(2)(viii) by removing the 
condition for using the 2007 Edition with 2009 Addenda through the 2013 
Edition of Subsection IWL requiring compliance with Sec.  
50.55a(b)(2)(viii)(E). To support the removal of the condition, the NRC 
is adding new requirements governing the performance and documentation 
of concrete containment examinations in Sec.  50.55a(b)(2)(viii)(H) and 
(I), which are discussed separately in the next two headings.
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(viii)(H) to require licensees 
to provide the applicable information specified in paragraphs 
(b)(2)(viii)(E)(1), (2), and (3) of this section in the ISI Summary 
Report required by IWA-6000 for each inaccessible concrete surface area 
evaluated under the new code provision IWL-2512 of the 2009 Addenda up 
to and including the 2013 Edition.
10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(viii)(I) to provide a new 
condition requiring the technical evaluation required by IWL-2512(b) of 
the 2009 Addenda up to and including the 2013 Edition of inaccessible 
below-grade concrete surfaces exposed to foundation soil, backfill, or 
groundwater be performed at periodic intervals not to exceed 5 years. 
In addition, the licensee must examine representative samples of the 
exposed portions of the below-grade concrete, when such below-grade 
concrete is excavated for any reason. The condition applies only to 
holders of renewed licenses under 10 CFR part 54 during the period of 
extended operation (i.e., beyond the expiration date of the original 
40-year license) of a renewed license when using IWL-2512(b) of the 
2007 Edition with 2009 Addenda through the latest edition and addenda 
in Sec.  50.55a(a)(1)(ii)--the 2013 Edition under this final rule.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment 
Examinations
    The NRC is revising Sec.  50.55a(b)(2)(ix) to continue to apply the 
existing conditions in Sec.  50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) 
and (J) with respect to the metal containment examination requirements 
in Subsection IWE up to and including the 2013 Edition (and all future 
editions and addenda of the ASME BPV Code which the NRC incorporates by 
reference into Sec.  50.55a). The NRC is accomplishing this by adding 
the words ``edition and'' to the last sentence in Sec.  
50.55a(b)(2)(ix).
10 CFR 50.55a(b)(2)(ix)(D) Metal Containment Examinations: Fourth 
Provision
    The NRC is revising the rule text in Sec.  50.55a(b)(2)(ix)(D) to 
improve clarity. Section 50.55a(b)(2)(ix)(D) introductory text and 
(b)(2)(ix)(D)(1) are combined. The information required to be included 
in the ISI Summary report is now all on the same paragraph level. No 
substantive change to the requirements is intended by this revision.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
    The NRC is revising Sec.  50.55a(b)(2)(x) to clarify that it 
allows, but does not require, licensees to use the 1994 Edition or the 
2008 Edition through the 2009-1a Addenda of NQA-1 when applying the 
2009 Addenda and later editions and addenda of the ASME BPV Code, 
Section XI, up to the 2013 Edition. Licensees are required to meet 
appendix B of 10 CFR part 50, and NQA-1 is one way of meeting portions 
of appendix B. A licensee may select any version of NQA-1 that has been 
approved for use in Sec.  50.55a.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1 does not meet all of the 
requirements of appendix B to 10 CFR part 50. Section 50.55a(b)(2)(x) 
clarifies that licensees using NQA-1 are also required to meet appendix 
B to 10 CFR part 50 and the commitments contained in their QA program 
descriptions. To meet the

[[Page 32961]]

requirements of appendix B, when using NQA-1 during ISI phase, 
licensees must address, in their quality program description, those 
areas where NQA-1 is insufficient to meet appendix B. Additional 
guidance and regulatory positions on how to meet appendix B when using 
NQA-1 are provided in RG 1.28.
10 CFR 50.55a(b)(2)(xii) Section XI Condition: Underwater Welding
    The NRC is revising Sec.  50.55a(b)(2)(xii) to allow underwater 
welding on irradiated materials in accordance with IWA-4660 under 
certain conditions. Licensees are allowed to perform welding on 
irradiated materials if certain neutron fluence criteria and, for 
certain material classes, helium concentration criteria are not 
exceeded. If these criteria are exceeded, the licensee is prohibited 
from performing welding on irradiated materials unless the licensee 
obtains NRC approval in accordance with Sec.  50.55a(z).
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth 
Provision
    The NRC is adding Sec.  50.55a(b)(2)(xviii)(D) to provide a new 
condition prohibiting the use of Appendix VII and Subarticle VIII-2200 
of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV 
Code. Licensees are required to implement Appendix VII and Subarticle 
VIII-2200 of the 2010 Edition of Section XI.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements: 
First Provision
    The NRC is revising Sec.  50.55a(b)(2)(xxi)(A) to modify the 
standard for visual magnification resolution sensitivity and contrast 
for visual examinations performed on Examination Category B-D 
components instead of ultrasonic examinations. A visual examination 
with magnification that has a resolution sensitivity to resolve 0.044 
inch (1.1 mm) lower case characters without an ascender or descender 
(e.g., a, e, n, v), utilizing the allowable flaw length criteria in 
Table IWB-3512-1, 1997 Addenda through the latest edition and addenda 
incorporated by reference in Sec.  50.55a(a)(1)(ii), with a limiting 
assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed 
instead of an ultrasonic examination. This revision removes a 
requirement that was in addition to the ASME BPV Code that required 1-
mil wires to be used in licensees' Sensitivity, Resolution, and 
Contrast Standard targets.
10 CFR 50.55a(b)(2)(xxiii) Section XI Condition: Evaluation of 
Thermally Cut Surfaces
    The NRC is revising Sec.  50.55a(b)(2)(xxiii) to modify the 
applicability of the condition. The condition will only apply to the 
2001 Edition through the 2009 Addenda IWA-4461.4, which was revised in 
the 2010 Edition to remove paragraph IWA-4461.4.2, which permitted an 
application specific evaluation of thermally cut surfaces in lieu of a 
thermal metal removal process qualification.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping 
Devices
    The NRC is adding Sec.  50.55a(b)(2)(xxxi) to provide a new 
condition maintaining the requirement to use Appendix IX, now 
renumbered as Appendix W, when installing a mechanical clamping device 
on an ASME BPV Code Class piping system. Additionally, the condition 
prohibits the use of mechanical clamping devices in accordance with the 
changes made to IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d) in 
the 2011 Addenda through 2013 Edition on small item Class 1 piping and 
portions of a piping system that form the containment boundary.
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report 
Submittal
    The NRC is adding Sec.  50.55a(b)(2)(xxxii) to provide a new 
condition requiring licensees using the 2010 Edition or later editions 
and addenda of Section XI to follow the requirements of IWA-6240 of the 
2009 Addenda of Section XI for the submittal of Preservice and 
Inservice Summary Reports. The condition also describes the timing of 
the submission of the Summary Reports by referencing the specific 
Section XI paragraph IWA-6240(b) in the regulation.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed 
Allowable Pressure
    The NRC is adding Sec.  50.55a(b)(2)(xxxiii) to provide a new 
condition to prohibit the use of Appendix G, Paragraph G-2216, in the 
2011 Addenda and later editions and addenda of the ASME BPV Code, 
Section XI.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Nonmandatory Appendix 
U
    The NRC is adding Sec.  50.55a(b)(2)(xxxiv)(A) and (B) to require 
that two conditions be satisfied when using Nonmandatory Appendix U of 
the 2013 Edition of the ASME BPV Code, Section XI. Paragraph 
(b)(2)(xxxiv)(A) requires that an ASME BPV Code repair or replacement 
activity temporarily deferred under the provisions of Nonmandatory 
Appendix U to the 2013 Edition of the ASME BPV Code, Section XI, shall 
be performed during the next scheduled refueling outage. Paragraph 
(b)(2)(xxxiv)(B) requires the use of the mandatory appendix in ASME BPV 
Code Case N-513-3, in lieu of the appendix referenced in paragraph U-
S1-4.2.1(c) of Appendix U, which was inadvertently omitted from 
Appendix U.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0 
in the KIa and KIc Equations
    The NRC is adding Sec.  50.55a(b)(2)(xxxv) to provide a new 
condition to specify that when licensees use ASME BPV Code, Section XI, 
2013 Edition, Appendix A, paragraph A-4200, if T0 is 
available, then RTT0 may be used in place of 
RTNDT for applications using the KIc equation and 
the associated KIc curve, but not for applications using the 
KIa equation and the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of 
Irradiated Materials
    The NRC is adding Sec.  50.55a(b)(2)(xxxvi) to provide a new 
condition requiring licensees using ASME BPV Code, Section XI, 2013 
Edition, Appendix A, paragraph A-4400, to obtain NRC approval under 
Sec.  50.55a(z) before using irradiated T0 and the 
associated RTT0 in establishing fracture toughness of 
irradiated materials.
10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii) to provide a new 
provision that allows licensees to implement ASME BPV Code Case N-824, 
``Ultrasonic Examination of Cast Austenitic Piping Welds From the 
Outside Surface Section XI, Division 1,'' subject to four conditions in 
paragraphs (b)(2)(xxxvii)(A) through (D). Each of these paragraphs are 
discussed in the following headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) Section XI Condition: ASME BPV Code Case 
N-824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii)(A) to add a new 
condition that requires ultrasonic examinations performed to implement 
ASME BPV Code Case N-824 to be spatially encoded.

[[Page 32962]]

10 CFR 50.55a(b)(2)(xxxvii)(B) Section XI Condition: ASME BPV Code Case 
N-824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii)(B) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 shall use dual, transmit-receive, 
refracted longitudinal wave, multi-element phased array search units 
instead of the requirements of Paragraph 1(c)(1)(-a) of N-824.
10 CFR 50.55a(b)(2)(xxxvii)(C) Section XI Condition: ASME BPV Code Case 
N-824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii)(C) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 on piping greater than 1.6 inches 
(41 mm) thick shall use a phased array search unit with a center 
frequency of 500 kHz with a tolerance of + /- 20 percent instead of the 
requirements of Paragraph 1(c)(1)(-c)(-2).
10 CFR 50.55a(b)(2)(xxxvii)(D) Section XI Condition: ASME BPV Code Case 
N-824
    The NRC is adding Sec.  50.55a(b)(2)(xxxvii)(D) to add a new 
condition that requires that ultrasonic examinations performed to 
implement ASME BPV Code Case N-824 shall use a phased array search unit 
which produces angles including, but not limited to, 30 to 55 degrees 
with a maximum increment of 5 degrees instead of the requirements of 
Paragraph 1(c)(1)(-d).
10 CFR 50.55a(b)(3) Conditions on ASME OM Code
    The NRC is revising Sec.  50.55a(b)(3) to clarify that Subsections 
ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, 
III, and V; and Nonmandatory Appendices A through H and J through M of 
the OM Code are each incorporated by reference into Sec.  50.55a. The 
NRC is also clarifying that the OM Code Nonmandatory Appendices 
incorporated by reference into Sec.  50.55a are approved for use, but 
are not mandated. The Nonmandatory Appendices may be used by applicants 
and licensees of nuclear power plants, subject to the conditions in 
Sec.  50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
    The NRC is revising Sec.  50.55a(b)(3)(i) to allow licensees to use 
the 1994 Edition, 2008 Edition, and 2009-1a Addenda of NQA-1 when using 
the 1995 Edition through the 2012 Edition of the OM Code. Licensees are 
required to meet appendix B to 10 CFR part 50, and NQA-1 is one way of 
meeting portions of appendix B.
    NQA-1 provides a method for establishing and implementing a QA 
program for the design and construction of nuclear power plants and 
fuel reprocessing plants; however, NQA-1 does not meet all of the 
requirements of appendix B to 10 CFR part 50. Section 50.55a(b)(3)(i) 
clarifies that licensees using NQA-1 are also required to meet appendix 
B to 10 CFR part 50 and the commitments contained in their QA program 
descriptions. To meet the requirements of appendix B, licensees must 
address, in their quality program description, those areas where NQA-1 
is insufficient to meet appendix B. Additional guidance and regulatory 
positions on how to meet appendix B when using NQA-1 are provided in RG 
1.28.
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV) 
Testing
    The NRC is revising Sec.  50.55a(b)(3)(ii) to set forth four 
conditions on the use of mandatory Appendix III, ``Preservice and 
Inservice Testing of Active Electric Motor Operated Valve Assemblies in 
Light-Water Reactor Power Plants,'' in the OM Code, 2009 Edition, 2011 
Addenda, and 2012 Edition. The four conditions, which are set forth in 
paragraphs (b)(3)(ii)(A) through (D), are discussed in the next four 
headings.
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval
    The NRC is adding Sec.  50.55a(b)(3)(ii)(A) to require that 
licensees evaluate the adequacy of the diagnostic test intervals 
established for MOVs within the scope of OM Code, Appendix III, not 
later than 5 years or three refueling outages (whichever is longer) 
from initial implementation of OM Code, Appendix III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk
    The NRC is adding Sec.  50.55a(b)(3)(ii)(B) to require that 
licensees ensure that the potential increase in CDF and LERF associated 
with the extension is acceptably small when extending exercise test 
intervals for high risk MOVs beyond a quarterly frequency. As specified 
in RG 1.192, when extending exercise test intervals for high risk MOVs 
beyond a quarterly frequency, licensees must ensure that the potential 
increase in CDF and risk associated with the extension is small and 
consistent with the intent of the Commission's Safety Goal Policy 
Statement. As discussed earlier in Section II, the NRC provides 
guidance in RG 1.174 that acceptably small changes are relative and 
depend on the current plant CDF and LERF. For plants with total 
baseline CDF of 10-4 per year or less, acceptably small 
means CDF increases of up to 10-5 per year; and for plants 
with total baseline CDF greater than 10-4 per year, 
acceptably small means CDF increases of up to 10-6 per year. 
For plants with total baseline LERF of 10-5 per year or 
less, acceptably small LERF increases are considered to be up to 
10-6 per year; and for plants with total baseline LERF 
greater than 10-5 per year, acceptably small LERF increases 
are considered to be up to 10-7 per year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization
    The NRC is adding Sec.  50.55a(b)(3)(ii)(C) to require, when 
applying Appendix III to the OM Code, that licensees categorize MOVs 
according to their safety significance using the methodology described 
in OM Code Case OMN-3 subject to the conditions discussed in RG 1.192, 
or using an MOV risk ranking methodology accepted by the NRC on a 
plant-specific or industry-wide basis in accordance with the conditions 
in the applicable safety evaluation.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time
    The NRC is adding Sec.  50.55a(b)(3)(ii)(D) to require, when 
applying Paragraph III-3600, ``MOV Exercising Requirements,'' of 
Appendix III to the OM Code, licensees shall verify that the stroke 
time of MOVs specified in plant technical specifications satisfies the 
assumptions in the plant's safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors
    The NRC is adding Sec.  50.55a(b)(3)(iii) to specify that, in 
addition to complying with the provisions in the OM Code as required 
with the conditions specified in Sec.  50.55a(b)(3), holders of 
operating licenses for nuclear power reactors that received 
construction permits under this part on or after the date 12 months 
after August 17, 2017, and holders of COLs issued under 10 CFR part 52, 
whose initial fuel loading occurs on or after the date 12 months after 
August 17, 2017, shall also comply with four condition on power-
operated valves, check valves, flow-induced vibration, and operational 
readiness of high-risk non-safety systems, to the extent applicable. 
These four conditions, which are set forth in

[[Page 32963]]

Sec.  50.55a(b)(3)(iii)(A), (B), (C), and (D), are discussed in the 
next four headings.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves (First Condition on 
New Reactors)
    The NRC is adding Sec.  50.55a(b)(3)(iii)(A) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) periodically verify the 
capability of power-operated valves (POVs) to perform their design-
basis safety functions.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves (Second Condition on New 
Reactors)
    The NRC is adding Sec.  50.55a(b)(3)(iii)(B) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) perform bi-directional 
testing of check valves within the IST program where practicable.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration (Third Condition on 
New Reactors)
    The NRC is adding Sec.  50.55a(b)(3)(iii)(C) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) monitor flow-induced 
vibration (FIV) from hydrodynamic loads and acoustic resonance during 
preservice testing or inservice testing to identify potential adverse 
flow effects that might impact components within the scope of the IST 
program.
10 CFR 50.55a(b)(3)(iii)(D) High Risk Non-Safety Systems (Fourth 
Condition on New Reactors)
    The NRC is adding Sec.  50.55a(b)(3)(iii)(D) to require that 
licensees subject to Sec.  50.55a(b)(3)(iii) establish a program to 
assess the operational readiness of pumps, valves, and dynamic 
restraints within the scope of the Regulatory Treatment of Non-Safety 
Systems for applicable reactor designs. As of the time of this final 
rule, these are designs which have been certified in a design 
certification rule under 10 CFR part 52. This final rule refers to 
these RTNSS components using the term, ``high risk non-safety 
systems.''
    As noted by the public commenters, ASME is preparing guidance for 
new reactor licensees to use in developing programs for the treatment 
of RTNSS equipment. The NRC staff is participating on the OM Code 
committees to assist in developing guidance for the treatment of RTNSS 
equipment that is consistent with Commission policy. Guidance on the 
implementation of the Commission policy for RTNSS equipment is set 
forth in NRC Inspection Procedure 73758, ``Part 52, Functional Design 
and Qualification, and Preservice and Inservice Testing Programs for 
Pumps, Valves and Dynamic Restraints,'' dated April 19, 2013.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
    The NRC is revising Sec.  50.55a(b)(3)(iv) to extend the existing 
conditions on the use of Appendix II to the new Editions and Addenda 
which are the subject of this rulemaking. These conditions are that: 
(i) Trending and evaluation shall support the determination that the 
valve or group of valves is capable of performing its intended 
function(s) over the entire interval; and (ii) at least one of the 
Appendix II condition monitoring activities for a valve group shall be 
performed on each valve of the group at approximate equal intervals not 
to exceed the maximum interval shown in the following table:

       Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
                                           Maximum      Maximum interval
                                      interval between       between
                                        activities of     activities of
             Group size               member valves in    each valve in
                                         the groups         the group
                                           (years)           (years)
------------------------------------------------------------------------
>=4.................................               4.5                16
3...................................               4.5                12
2...................................                 6                12
1...................................    Not applicable                10
------------------------------------------------------------------------

    The conditions currently specified for the use of Appendix II, 1995 
Edition with the 1996 and 1997 Addenda, and 1998 Edition through the 
2002 Addenda, of the OM Code remain the same in this final rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
    The NRC is adding Sec.  50.55a(b)(3)(vii) to prohibit the use of 
Subsection ISTB in the 2011 Addenda to the OM Code.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
    The NRC is adding Sec.  50.55a(b)(3)(viii) to specify that 
licensees who wish to implement Subsection ISTE, ``Risk-Informed 
Inservice Testing of Components in Light-Water Reactor Nuclear Power 
Plants,'' of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition, 
must first request and obtain NRC approval in accordance with Sec.  
50.55a(z) to apply Subsection ISTE on a plant-specific basis as a risk-
informed alternative to the applicable IST requirements in the OM Code.
    The NRC will evaluate Sec.  50.55a(z) requests for approval to 
implement Subsection ISTE in accordance with the following 
considerations. These considerations are consistent with the guidance 
provided in RG 1.174.
1. Scope of Risk-Informed IST Program
    Subsection ISTE-1100, ``Applicability,'' establishes the component 
safety categorization methodology and process for dividing the 
population of pumps and valves, as identified in the IST Program Plan, 
into high safety significant component (HSSC) and low safety 
significant component (LSSC) categories. When establishing a risk-
informed IST program, the licensee should address a wide range of 
components important to safety at the nuclear power plant that includes 
both safety-related and nonsafety-related components. These components 
might extend beyond the scope of the OM Code.
2. Risk-Ranking Methodology
    The licensee should specify, in its request for authorization to 
implement a risk-informed IST program, the methodology to be applied in 
risk ranking its components. ISTE-4000, ``Specific Component 
Categorization Requirements,'' incorporates OM Code Case OMN-3 for the 
categorization of pumps and valves in developing a risk-informed IST 
program. The OMN-3 Code Case methodology for risk ranking uses two 
categories of safety

[[Page 32964]]

significance. The NRC staff has also accepted other methodologies for 
risk ranking that use three categories of safety significance.
3. Safety Significance Categorization
    The licensee should categorize components according to their safety 
significance based on the methodology described in Subsection ISTE with 
the applicable conditions on the use of OM Code Case OMN-3 specified in 
RG 1.192, or use other risk ranking methodologies accepted by the NRC 
on a plant-specific or industry-wide basis with applicable conditions 
specified by the NRC for their acceptance. The licensee should address 
the seven conditions in RG 1.192 for the use of OM Code Case OMN-3, as 
appropriate, in developing the risk-informed IST program described in 
Subsection ISTE. With respect to the provisions in Subsection ISTE, 
these conditions are:
    (a) The implementation of ISTE-1100 should include within the scope 
of a licensee's risk-informed IST program non-ASME OM Code pumps and 
valves categorized as HSSCs that might not currently be included in the 
IST program at the nuclear power plant.
    (b) The decision criteria discussed in ISTE-4410, ``Decision 
Criteria,'' and Nonmandatory Appendix L, ``Acceptance Guidelines,'' of 
the OM Code for evaluating the acceptability of aggregate risk effects 
(i.e., for CDF and LERF) should be consistent with the guidance 
provided in RG 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the 
Licensing Basis.''
    (c) The implementation of ISTE-4440, ``Defense in Depth,'' should 
be consistent with the guidance contained in Section 2.2.1, ``Defense-
in-Depth Evaluation,'' and Section 2.2.2, ``Safety Margin Evaluation,'' 
of RG 1.175, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Inservice Testing.''
    (d) The implementation of ISTE-4500, ``Inservice Testing Program,'' 
and ISTE-6100, ``Performance Monitoring,'' should be consistent with 
the guidance contained in Section 3.2, ``Program Implementation,'' and 
Section 3.3, ``Performance Monitoring,'' of RG 1.175.
    (e) The implementation of ISTE-3210, ``Plant-Specific PRA,'' should 
be consistent with the guidance that the Owner is responsible for 
demonstrating and justifying the technical adequacy of the PRA analyses 
used as the basis to perform component risk ranking and for estimating 
the aggregate risk impact. For example, RG 1.200, ``An Approach for 
Determining the Technical Adequacy of Probabilistic Risk Assessment 
Results for Risk-Informed Activities,'' and RG 1.201, ``Guidelines for 
Categorizing Structures, Systems, and Components in Nuclear Power 
Plants According to their Safety Significance,'' provide guidance for 
PRA technical adequacy and component risk ranking.
    (f) The implementation of ISTE-4240, ``Reconciliation,'' should 
specify that 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 LSSC.
    (g) The implementation of ISTE-3220, ``Living PRA,'' should be 
consistent with the following: (i) 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; (ii) 
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 (iii) the review of the 
plant changes must be performed in a timely manner and must be 
performed once every two refueling outages, or as required by Sec.  
50.71(h)(2) for COL holders.
4. Pump Testing
    Subsection ISTE-5100, ``Pumps,'' incorporates OM Code Case OMN-7 
for risk-informed testing of pumps categorized as LSSCs. Subsection 
ISTE-5100 allows the interval for Group A and Group B testing of LSSC 
pumps specified in Subsection ISTB of the OM Code to be extended from 
the current 3-month interval to intervals of 6 months or 2 years. 
Subsection ISTE-5100 eliminates the requirement in Subsection ISTB to 
perform comprehensive pump testing for LSSC pumps. Table ISTE-5121-1, 
``LSSC Pump Testing,'' specifies that pump operation may be required 
more frequently than the specified test frequency (6 months) to meet 
vendor recommendations. Subsection ISTE-4500, ``Inservice Testing 
Program,'' specifies in ISTE-4510, ``Maximum Testing Interval,'' that 
the maximum testing interval shall be based on the more limiting of (a) 
the results of the aggregate risk, or (b) the performance history of 
the component. ISTE-5130, ``Maximum Test Interval--Pre-2000 Plants,'' 
specifies that the most limiting interval for LSSC pump testing shall 
be determined from ISTE-4510 and ISTE-5120, ``Low Safety Significant 
Pump Testing.'' The ASME developed the comprehensive pump test 
requirements in the OM Code to address weaknesses in the Code 
requirements to assess the operational readiness of pumps to perform 
their design-basis safety function. Therefore, the licensee should 
ensure that testing under Subsection ISTE will provide assurance of the 
operational readiness of pumps in each safety significant 
categorization to perform their design-basis safety function as 
described in RGs 1.174 and 1.175.
5. Motor-Operated Valve Testing
    Subsection ISTE-5300, ``Motor Operated Valve Assemblies,'' provides 
a risk-informed IST approach instead of the IST requirements for MOVs 
in Mandatory Appendix III to the OM Code. The ASME prepared Appendix 
III to the OM Code to replace the requirement for quarterly stroke-time 
testing of MOVs with a program of periodic exercising and diagnostic 
testing to address lessons learned from nuclear power plant operating 
experience and industry and regulatory research programs for MOV 
performance. Subsection ISTC of the OM Code specifies the 
implementation of Appendix III for periodic exercising and diagnostic 
testing of MOVs to replace quarterly stroke-time testing previously 
required for MOVs. Appendix III incorporates provisions that allow a 
risk-informed IST approach for MOVs as described in OM Code Cases OMN-1 
and OMN-11. Subsection ISTE-5300 is not consistent with the provisions 
for the risk-informed IST program for MOVs specified in Appendix III to 
the OM Code (and Code Cases OMN-1 and 11). Therefore, licensees who 
wish to implement Subsection ISTE should address the provisions in 
paragraph III-3700, ``Risk-Informed MOV Inservice Testing,'' of 
Appendix III to the OM Code as incorporated by reference in Sec.  
50.55a, with the applicable conditions, instead of ISTE-5300.
6. Pneumatically and Hydraulically Operated Valve Testing
    Subsection ISTE-5400, ``Pneumatically and Hydraulically Operated 
Valves,'' specifies that licensees test their AOVs and HOVs in 
accordance with Appendix IV to the OM Code. Subsection ISTE-5400 
indicates that Appendix IV is in the course of preparation. The NRC 
staff will need to review Appendix IV prior to accepting its use as 
part of Subsection ISTE. Therefore, licensees who wish to implement 
Subsection ISTE should describe the planned IST provisions for AOVs and 
HOVs in its request for approval to implement Subsection ISTE.

[[Page 32965]]

7. Pump Periodic Verification Test
    Subsection ISTE does not include a requirement to implement the 
pump periodic verification test program specified in Mandatory Appendix 
V to the OM Code, 2012 Edition. Therefore, licensee should address the 
consideration of a pump periodic verification test program in its risk-
informed IST program, proposed as part of the authorization request to 
implement Subsection ISTE.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
    The NRC is adding Sec.  50.55a(b)(3)(ix) to specify that licensees 
applying Subsection ISTF, ``Inservice Testing of Pumps in Light-Water 
Reactor Nuclear Power Plants--Post-2000 Plants,'' in the 2012 Edition 
of the OM Code shall satisfy the requirements of Mandatory Appendix V, 
``Pump Periodic Verification Test Program,'' of the OM Code, 2012 
Edition. The paragraph also states that Subsection ISTF, 2011 Addenda, 
is not acceptable for use.
10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20
    The NRC is adding Sec.  50.55a(b)(3)(x) to allow licensees to 
implement OM Code Case OMN-20, ``Inservice Test Frequency,'' in the OM 
Code, 2012 Edition, for the editions and addenda of the OM Code that 
are listed in Sec.  50.55a(a)(1)(iv).
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
    The NRC is adding Sec.  50.55a(b)(3)(xi) to emphasize the 
provisions in the OM Code, 2012 Edition, Subsection ISTC-3700, 
``Position Verification Testing,'' to verify that valve obturator 
position is accurately indicated. The OM Code, Subsection ISTC-3700 
requires valves with remote position indicators shall be observed 
locally at least once every 2 years to verify that valve operation is 
accurately indicated. Licensees will be required to implement the 
condition when adopting the 2012 Edition of the OM Code as their Code 
of Record for the applicable 120-month IST interval.
10 CFR 50.55a(f) Preservice and Inservice Testing Requirements
    The NRC is revising the heading for Sec.  50.55a(f) and clarifying 
that the OM Code includes provisions for preservice testing of 
components as part of its overall provisions for IST programs.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for 
Operating Plants
    The NRC is revising Sec.  50.55a(f)(4) to ensure that the paragraph 
is applicable to pumps and valves that are within the scope of the OM 
Code. The NRC is also including an additional provision in Sec.  
50.55a(f)(4) stating that the IST requirements for pumps and valves 
that are within the scope of the OM Code but are not classified as ASME 
BPV Code Class 1, Class 2, or Class 3 may be satisfied as an augmented 
IST program, in accordance with Sec.  50.55a(f)(6)(ii), without 
requesting relief under Sec.  50.55a(f)(5) or alternatives under Sec.  
50.55a(z). This use of an augmented IST program may be acceptable 
provided the basis for deviations from the OM Code, as incorporated by 
reference in this section, demonstrates an acceptable level of quality 
and safety, or that implementing the Code provisions would result in 
hardship or unusual difficulty without a compensating increase in the 
level of quality and safety, where documented and available for NRC 
review. These changes align the scope of pumps and valves for inservice 
testing with the scope defined in the OM Code without imposing an 
unnecessary paperwork burden on nuclear power plant licensees for the 
submittal of relief and alternative requests for pumps and valves 
within the scope of the OM Code but not classified as ASME BPV Code 
Class 1, Class 2, or Class 3 components.
10 CFR 50.55a(g) Preservice and Inservice Inspection Requirements
    The NRC is revising the heading in Sec.  50.55a(g), adding new 
paragraphs (g)(2)(i), (ii), and (iii), and revising current paragraphs 
(g) introductory text, (g)(2), (g)(3) introductory text, and (g)(3)(i), 
(ii), and (v) to distinguish the requirements for accessibility, 
preservice examination, and inservice inspection. No substantive change 
to the requirements is intended by these revisions.
10 CFR 50.55a(g)(4) Inservice Inspection Standards Requirement for 
Operating Plants
    The NRC is revising Sec.  50.55a(g)(4)(ii) to add an implementation 
period of 18-months for licensees whose ISI interval commences during 
the 12 through 18-month period after the publication of this final 
rule. The NRC is also revising Sec.  50.55a(g)(4)(i) and (ii) to add a 
provision allowing licensees to adopt the latest version of Appendix 
VIII of the ASME BPV Code edition or addenda listed in Sec.  
50.55a(a)(1) at any time in the licensee's 120-month ISI interval.
10 CFR 50.55a(g)(6)(ii)(D) Augmented ISI Requirements: Reactor Vessel 
Head Inspections
    The NRC is revising Sec.  50.55a(g)(6)(ii)(D) to reflect the NRC's 
approval of ASME BPV Code Case N-729-4, which supersedes the NRC's 
earlier approval of ASME BPV Code Case N-729-1. The revisions include 
changes to the conditions governing the use of the Code Case to reflect 
the change from N-729-1 to N-729-4. The effect of these changes is to 
require licensees to implement an augmented ISI program for the 
examination of the pressurized water reactor RPV upper head 
penetrations. The following discussions provide a more detailed 
discussion of the revisions to Sec.  50.55a(g)(6)(ii)(D).
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
    The NRC is revising Sec.  50.55a(g)(6)(ii)(D)(1) to require 
licensees to implement an augmented ISI program for the examination of 
the pressurized water reactor RPV upper head penetrations meeting ASME 
BPV Code Case N-729-4 instead of the previously approved requirements 
to use ASME BPV Code Case N-729-1, as conditioned by the NRC.
Removal of Existing Conditions in 10 CFR 50.55a(g)(6)(ii)(D)(2) Through 
(5)
    The NRC is removing the existing conditions in Sec.  
50.55a(g)(6)(ii)(D)(2) through (5) and redesignating the existing 
condition in Sec.  50.55a(g)(6)(ii)(D)(6) as Sec.  
50.55a(g)(6)(ii)(D)(2).
10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I Use
    The NRC is revising the existing condition in Sec.  
50.55a(g)(6)(ii)(D)(6), which is redesignated as Sec.  
50.55a(g)(6)(ii)(D)(2) in this final rule, to require NRC approval 
prior to implementing Appendix I of ASME BPV Code Case N-729-4.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency
    The NRC is adding a new condition in Sec.  50.55a(g)(6)(ii)(D)(3) 
which requires cold head plants with less than eight effective 
degradation years (EDY<8) without PWSCC flaws to perform a bare metal 
visual examination (VE) each outage a volumetric exam is not performed 
and allows these plants to extend the bare metal visual inspection 
frequency from once each refueling outage, as stated in Table 1 of N-
729-4, to once every 5 years, only if the licensee performed a wetted 
surface examination of all of the partial

[[Page 32966]]

penetration welds during the previous volumetric examination. In 
addition, this new condition clarifies that a bare metal visual 
examination is not required during refueling outages when a volumetric 
or surface examination is performed of the partial penetration welds.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria
    The NRC is adding a new condition in Sec.  50.55a(g)(6)(ii)(D)(4) 
clarifying that rounded indications found by surface examinations of 
the partial-penetration or associated fillet welds in accordance with 
N-729-4 must meet the acceptance criteria for surface examinations of 
paragraph NB-5352 of ASME 2013 Edition of Section III for the 
licensee's ongoing 10-year ISI interval.
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(1) to require 
licensees to implement an augmented ISI program for the examination of 
ASME Class 1 piping and nozzle butt welds meeting ASME BPV Code Case N-
770-2 instead of the previously approved ASME BPV Code Case N-770-1.
    Furthermore, the NRC is revising Sec.  50.55a(g)(6)(ii)(F)(1) to 
update the date of applicability for pressurized water reactors, to 
note the change to implement ASME BPV Code Case N-770-2 instead of N-
770-1, and to reflect the number of conditions which must be applied.
10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(2) to clarify the 
requirements for licensees to establish the initial categorization of 
each weld and modify the wording to reflect the ASME BPV Code Case N-
770-2 change in the inspection item category for full structural weld 
overlays (C to C-1 and F to F-1). Additionally, the NRC is adding a 
sentence which clarifies the NRC position that Paragraph -1100(e) of 
ASME BPV Code Case N-770-2 shall not be used to exempt welds that rely 
on Alloy 82/182 for structural integrity from any requirement of Sec.  
50.55a(g)(6)(ii)(F).
10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(3) to clarify the 
current requirement in this paragraph to complete baseline examinations 
by stating that previously-conducted examinations, in order to count as 
baseline examinations, must meet the requirements of ASME BPV Code Case 
N-770-2, as conditioned in this section. Additionally, this condition 
clarifies that the examination coverage requirements, for a licensee to 
count previous inspections as baseline examinations, must meet the 
examination coverage requirements described in Paragraphs -2500(a) or -
2500(b) of ASME BPV Code Case N-770-2, as conditioned by the NRC in 
this section. Upon implementation of this rule, if a licensee is 
currently in an outage, then the baseline inspection requirement can be 
met by performing the inspections in accordance with the previous 
regulatory requirements of Sec.  50.55a(g)(6)(ii)(F), in lieu of the 
examination requirements of Paragraphs -2500(a) or -2500(b) of ASME BPV 
Code Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(4) to clarify that 
when licensees are implementing paragraph -2500(a) of ASME BPV Code 
Case N-770-2, essentially 100 percent of the required volumetric 
examination coverage shall be obtained, including greater than 90 
percent volumetric examination coverage is obtained for circumferential 
flaws, to continue the restriction on the licensee's use of Paragraph -
2500(c) and to continue the restriction that the use of new Paragraph -
2500(d) of ASME BPV Code Case N-770-2 is not allowed without prior NRC 
review and approval in accordance with Sec.  50.55a(z), as it would 
permit a reduction in volumetric examination coverage for 
circumferential flaws. However, a licensee may request approval for use 
of these paragraphs under Sec.  50.55a(z), and the NRC may approve the 
request if technically justified.
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(5) to add an 
explanatory heading, ``Inlay/onlay inspection frequency,'' and to make 
minor editorial corrections without substantive changes in the 
requirement.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(6) to add an 
explanatory heading, ``Reporting requirements.''
10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(7) to add an 
explanatory heading, ``Defining `t'.''
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(8) to add an 
explanatory heading, ``Optimized weld overlay examination,'' and to 
continue the current condition located in Sec.  50.55a(g)(6)(ii)(F)(9) 
which requires that the initial examination of optimized weld overlays 
(i.e., Inspection Item C-2 of ASME BPV Code Case N-770-2) be performed 
between the third refueling outage and no later than 10 years after 
application of the overlay and delete the other current examination 
requirements for optimized weld overlay examination frequency, as these 
requirements were included in the revision from N-770-1 to N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(9) to add an 
explanatory heading, ``Deferral,'' and to modify the current condition 
to continue denial of the deferral of the initial inservice examination 
of uncracked welds mitigated by optimized weld overlays. These welds 
shall continue to have their initial inservice examinations as 
prescribed in N-770-1 within 10 years of the application of the 
optimized weld overlay and not allow deferral of this initial 
examination. Subsequent inservice examinations may be deferred as 
allowed by N-770-2. Additionally, the modified condition will delete 
the current condition on examination requirements for the deferral of 
welds mitigated by inlay, onlay, stress improvement and optimized weld 
overlay, as these requirements were, with one exception (i.e., 
optimized weld overlay), included in the revision from N-770-1 to N-
770-2.
10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique
    The NRC is revising Sec.  50.55a(g)(6)(ii)(F)(10) to add an 
explanatory heading, ``Examination technique,'' and to modify the 
current condition to allow the previously prohibited alternate 
examination requirements of Note (b) of Figure 5(a) of ASME BPV Code 
Cases N-770-1 and N-770-2 and the same requirements in Note 14(b) of 
Table 1 of ASME BPV Code Case N-770-2 for optimized weld overlays only 
if the full examination requirements of Note 14(a) of Table 1 of

[[Page 32967]]

ASME BPV Code Case N-770-2 cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(11) to provide a new 
condition requiring licensees to establish a Section XI, Appendix VIII, 
qualification requirement for ultrasonic inspection of cast stainless 
steel and through cast stainless steel to meet the examination 
requirements of Paragraph -2500(a) of ASME BPV Code Case N-770-2 by 
January 1, 2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(12) to provide a new 
condition that would allow licensees to implement a stress improvement 
mitigation technique for items containing cast stainless steel that 
would meet the requirements of Appendix I of ASME BPV Code Case N-770-
2, if the required examination volume can be examined by Appendix VIII 
procedures to the maximum extent practical including 100 percent of the 
susceptible material volume.
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination
    The NRC is adding Sec.  50.55a(g)(6)(ii)(F)(13) to provide a new 
condition requiring licensees to perform encoded examinations of 100 
percent of the required inspection volume when required to perform 
volumetric examinations of all non-mitigated and cracked mitigated butt 
welds in the reactor coolant pressure boundary in accordance with ASME 
BPV Code Case N-770-2.

VI. Generic Aging Lessons Learned Report

Background

    In December 2010, the NRC issued NUREG-1801, Revision 2, for 
applicants to use in preparing their license renewal applications. The 
GALL Report provides aging management programs (AMPs) that the NRC 
staff has concluded are sufficient for aging management in accordance 
with the license renewal rule, as required in Sec.  54.21(a)(3). In 
addition, NUREG-1800, Revision 2, ``Standard Review Plan for Review of 
License Renewal Applications for Nuclear Power Plants,'' was issued in 
December 2010 to ensure the quality and uniformity of NRC staff reviews 
of license renewal applications and to present a well-defined basis on 
which the NRC staff evaluates the applicant's aging management programs 
and activities. In April 2011, the NRC issued NUREG-1950, ``Disposition 
of Public Comments and Technical Bases for Changes in the License 
Renewal Guidance Documents NUREG-1801 and NUREG-1800,'' which describes 
the technical bases for the changes in Revision 2 of the GALL Report 
and Revision 2 of the SRP for review of license renewal applications. 
Revision 2 of the GALL Report, in Sections XI.M1, XI.S1, XI.S2, and 
XI.S3, describes the evaluation and technical bases for determining the 
sufficiency of ASME BPV Code Subsections IWB, IWC, IWD, IWE, IWF, and 
IWL for managing aging during the period of extended operation. In 
addition, many other AMPs in the GALL Report rely, in part but to a 
lesser degree, on the requirements specified in the ASME BPV Code, 
Section XI. Revision 2 of the GALL Report also states that the 1995 
Edition through the 2004 Edition of the ASME BPV Code, Section XI, 
Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as modified and limited 
by Sec.  50.55a, were found to be acceptable editions and addenda for 
complying with the requirements of Sec.  54.21(a)(3), unless 
specifically noted in certain sections of the GALL Report. The GALL 
Report further states that the future Federal Register notices that 
amend Sec.  50.55a will discuss the acceptability of editions and 
addenda more recent than the 2004 edition for their applicability to 
license renewal.
    In a final rule issued on June 21, 2011 (76 FR 36232), subsequent 
to Revision 2 of the GALL Report, the NRC found that the 2004 Edition 
with the 2005 Addenda through the 2007 Edition with the 2008 Addenda of 
Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, IWE, IWF, 
and IWL, as subject to the conditions in Sec.  50.55a, are acceptable 
for the AMPs in the GALL Report and the conclusions of the GALL Report 
remain valid with the augmentations specifically noted in the GALL 
Report.

Evaluation With Respect to Aging Management

    As part of this rulemaking, the NRC evaluated whether those AMPs in 
Revision 2 of the GALL Report which rely upon Subsections IWB, IWC, 
IWD, IWE, IWF, and IWL of Section XI in the editions and addenda of the 
ASME BPV Code incorporated by reference into Sec.  50.55a, continue to 
be acceptable if the AMP relies upon the versions of these Subsections 
in the 2007 Edition with the 2009 Addenda through the 2013 Edition. The 
NRC finds that the 2007 Edition with the 2009 Addenda through the 2013 
Edition of Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, 
IWE, IWF, and IWL, as subject to the conditions of this rule, are 
acceptable for the AMPs in the GALL Report and the conclusions of the 
GALL Report remain valid with the augmentations specifically noted in 
the GALL Report. Accordingly, an applicant for license renewal may use, 
in its plant-specific license renewal application, Subsections IWB, 
IWC, IWD, IWE, IWF, and IWL of Section XI of the 2007 Edition with the 
2009 Addenda through the 2013 Edition of the ASME BPV Code, as subject 
to the conditions in this rule, without additional justification.
    Similarly, a licensee approved for license renewal that relied on 
the GALL AMPs may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of 
Section XI of the 2007 Edition with the 2009 Addenda through the 2013 
Edition of the ASME BPV Code. However, a licensee must assess and 
follow applicable NRC requirements with regard to changes to its 
licensing basis. Some of the AMPs in the GALL Report recommend 
augmentation of certain Code requirements in order to ensure adequate 
aging management for license renewal. The technical and regulatory 
aspects of the AMPs for which augmentations are recommended also apply 
if the editions or addenda from the 2007 Edition with the 2009 Addenda 
through the 2013 Edition of Section XI of the ASME BPV Code are used to 
meet the requirements of Sec.  54.21(a)(3). The NRC staff evaluated the 
changes in the 2007 Edition with the 2009 Addenda through the 2013 
Edition of Section XI of the ASME BPV Code to determine if the 
augmentations described in the GALL Report remain necessary. The NRC 
staff's evaluation has concluded that the augmentations described in 
the GALL Report are necessary to ensure adequate aging management. For 
example, Table IWB-2500-1, in the 2007 Edition with the 2009 Addenda of 
ASME BPV Code, Section XI, Subsection IWB, requires surface examination 
of ASME BPV Code Class 1 branch pipe connection welds less than nominal 
pipe size (NPS) 4 under Examination Category B-J. However, the NRC 
staff finds that volumetric or opportunistic destructive examination, 
rather than surface examination, is necessary to adequately detect and 
manage the aging effect due to stress corrosion cracking or thermal, 
mechanical and vibratory loadings in the components for the period of 
extended operation. Therefore, GALL Report Section XI.M35, ``One-Time 
Inspection of ASME BPV Code Class 1 Small-Bore Piping,'' includes the 
augmentation of the requirements in ASME BPV Code, Section XI,

[[Page 32968]]

Subsection IWB to perform a one-time inspection of a sample of ASME BPV 
Code Class 1 piping less than NPS 4 and greater than or equal to NPS 1 
using volumetric or opportunistic destructive examination. The GALL 
Report addresses this augmentation to confirm that there is no need to 
manage age-related degradation through periodic volumetric inspections 
or that an existing AMP (for example, Water Chemistry AMP) is effective 
to manage the aging effect due to stress corrosion cracking or thermal, 
mechanical and vibratory loadings for the period of extended operation. 
A license renewal applicant may either augment its AMPs as described in 
the GALL Report, or propose alternatives for the NRC to review as part 
of the applicant's plant-specific justification for its AMPs.

VII. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC 
certifies that this rule does not have a significant economic impact on 
a substantial number of small entities. This final rule affects only 
the licensing and operation of nuclear power plants. The companies that 
own these plants do not fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
size standards established by the NRC (Sec.  2.810).

VIII. Regulatory Analysis

    The NRC has prepared a final regulatory analysis on this 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by the NRC. The regulatory analysis is 
available as indicated in the ``Availability of Documents'' section of 
this document.

IX. Backfitting and Issue Finality

Introduction

    The NRC's Backfit Rule in Sec.  50.109 states that the NRC shall 
require the backfitting of a facility only when it finds the action to 
be justified under specific standards stated in the rule. Section 
50.109(a)(1) defines backfitting as the modification of or addition to 
systems, structures, components, or design of a facility; the design 
approval or manufacturing license for a facility; or the procedures or 
organization required to design, construct, or operate a facility. Any 
of these modifications or additions may result from a new or amended 
provision in the NRC's rules or the imposition of a regulatory position 
interpreting the NRC's rules that is either new or different from a 
previously applicable NRC position after issuance of the construction 
permit or the operating license or the design approval.
    Section 50.55a requires nuclear power plant licensees to:
     Construct ASME BPV Code Class 1, 2, and 3 components in 
accordance with the rules provided in Section III, Division 1, of the 
ASME BPV Code (``Section III'').
     Inspect Class 1, 2, 3, Class MC, and Class CC components 
in accordance with the rules provided in Section XI, Division 1, of the 
ASME BPV Code (``Section XI'').
     Test Class 1, 2, and 3 pumps, valves, and dynamic 
restraints (snubbers) in accordance with the rules provided in the OM 
Code.
    This final rule is incorporating by reference the 2009 Addenda, 
2010 Edition, 2011 Addenda, and the 2013 Edition of the ASME BPV Code, 
Section III, Division 1 and ASME BPV Code, Section XI, Division 1, 
including NQA-1 (with conditions on its use), as well as the 2009 
Edition and 2011 Addenda and 2012 Edition of the OM Code and Code Cases 
N-770-2 and N-729-4.
    The ASME BPV and OM Codes are national consensus standards 
developed by participants with broad and varied interests, in which all 
interested parties (including the NRC and utilities) participate. A 
consensus process involving a wide range of stakeholders is consistent 
with the NTTAA, inasmuch as the NRC has determined that there are sound 
regulatory reasons for establishing regulatory requirements for design, 
maintenance, ISI, and IST by rulemaking. The process also facilitates 
early stakeholder consideration of backfitting issues. Therefore, the 
NRC believes that the NRC need not address backfitting with respect to 
the NRC's general practice of incorporating by reference updated ASME 
Codes.

Overall Backfitting Considerations: Section III of the ASME BPV Code

    Incorporation by reference of more recent editions and addenda of 
Section III of the ASME BPV Code does not affect a plant that has 
received a construction permit or an operating license or a design that 
has been approved. This is because the edition and addenda to be used 
in constructing a plant are, under Sec.  50.55a, determined based on 
the date of the construction permit, and are not changed thereafter, 
except voluntarily by the licensee. The incorporation by reference of 
more recent editions and addenda of Section III ordinarily applies only 
to applicants after the effective date of a final rule incorporating 
these new editions and addenda. Therefore, incorporation by reference 
of a more recent edition and addenda of Section III does not constitute 
``backfitting'' as defined in Sec.  50.109(a)(1).

Overall Backfitting Considerations: Section XI of the ASME BPV Code and 
the OM Code

    Incorporation by reference of more recent editions and addenda of 
Section XI of the ASME BPV Code and the OM Code affects the ISI and IST 
programs of operating reactors. However, the Backfit Rule generally 
does not apply to incorporation by reference of later editions and 
addenda of the ASME BPV Code (Section XI) and OM Code. As previously 
mentioned, the NRC's longstanding regulatory practice has been to 
incorporate later versions of the ASME Codes into Sec.  50.55a. Under 
Sec.  50.55a, licensees shall revise their ISI and IST programs every 
120 months to the latest edition and addenda of Section XI of the ASME 
BPV Code and the OM Code incorporated by reference into Sec.  50.55a 12 
months before the start of a new 120-month ISI and IST interval. 
Therefore, when the NRC approves and requires the use of a later 
version of the Code for ISI and IST, it is implementing this 
longstanding regulatory practice and requirement.
    Other circumstances where the NRC does not apply the Backfit Rule 
to the approval and requirement to use later Code editions and addenda 
are as follows:
    1. When the NRC takes exception to a later ASME BPV Code or OM Code 
provision but merely retains the current existing requirement, 
prohibits the use of the later Code provision, limits the use of the 
later Code provision, or supplements the provisions in a later Code. 
The Backfit Rule does not apply because the NRC is not imposing new 
requirements. However, the NRC explains any such exceptions to the Code 
in the statement of considerations and regulatory analysis for the 
rule.
    2. When an NRC exception relaxes an existing ASME BPV Code or OM 
Code provision but does not prohibit a licensee from using the existing 
Code provision. The Backfit Rule does not apply because the NRC is not 
imposing new requirements.
    3. The NRC's consideration of backfitting for modifications and 
limitations imposed during previous routine updates of Sec.  50.55a 
have established a precedent for determining the kinds of modifications 
or limitations which should be considered backfitting, or require a 
backfit analysis (e.g., final rule dated September 10, 2008 (73 FR 
52730), and a correction dated October 2, 2008 (73 FR 57235)). The 
consideration of backfitting and issue

[[Page 32969]]

finality with respect to the modifications and limitations in this 
rulemaking are consistent with the consideration and application of 
backfitting and issue finality requirements to analogous modifications 
and limitations in previous Sec.  50.55a rulemakings.
    The incorporation by reference and adoption of a requirement 
mandating the use of a later ASME BPV Code or OM Code may constitute 
backfitting in some circumstances. In these cases, the NRC would 
perform a backfit analysis or documented evaluation in accordance with 
Sec.  50.109. These include the following:
    1. When the NRC endorses a later provision of the ASME BPV Code or 
OM Code that takes a substantially different direction from the 
existing requirements, the action is treated as a backfit (e.g., 61 FR 
41303 (August 8, 1996)).
    2. When the NRC requires implementation of a later ASME BPV Code or 
OM Code provision on an expedited basis, the action is treated as a 
backfit. This applies when implementation is required sooner than it 
would be required if the NRC simply endorsed the Code without any 
expedited language (e.g., 64 FR 51370 (September 22, 1999)).
    3. When the NRC takes an exception to an ASME BPV Code or OM Code 
provision and imposes a requirement that is substantially different 
from the existing requirement as well as substantially different from 
the later Code (e.g., 67 FR 60529 (September 26, 2002)).

Detailed Backfitting Discussion: Changes Beyond Those Necessary To 
Incorporate by Reference the New ASME BPV and OM Code Provisions

    This section discusses the backfitting considerations for all the 
changes to Sec.  50.55a that go beyond the minimum changes necessary 
and required to adopt the new ASME Code Addenda into Sec.  50.55a.

ASME BPV Code, Section III

    1. Revise Sec.  50.55a(b)(1)(ii), ``Weld leg dimensions,'' to 
clarify rule language and add Table I, which clarifies prohibited 
Section III provisions for welds with leg sizes less than 1.09 
tn in tabular form. This change does not alter the original 
intent of this requirement and, therefore, does not impose a new 
requirement. Therefore, this change is not a backfit.
    2. Revise Sec.  50.55a(b)(1)(iv), ``Quality assurance,'' to require 
that when applying editions and addenda later than the 1989 Edition of 
Section III, the requirements of NQA-1, 1994 Edition, 2008 Edition, and 
the 2009-1a Addenda are acceptable for use, provided that the edition 
and addenda of NQA-1 specified in either NCA-4000 or NCA-7000 is used 
in conjunction with the administrative, quality, and technical 
provisions contained in the edition and addenda of Section III being 
used. This revision clarifies the current requirements, and is 
considered to be consistent with the meaning and intent of the current 
requirements, and therefore is not considered to result in a change in 
requirements. Therefore, this change is not a backfit.
    3. Add a new condition as Sec.  50.55a(b)(1)(viii), ``Use of ASME 
Certification Marks,'' to allow licensees to use either the ASME BPV 
Code Symbol Stamp or ASME Certification Mark with the appropriate 
certification designator and class designator as specified in the 2013 
Edition through the latest edition and addenda incorporated by 
reference in Sec.  50.55a. This condition does not result in a change 
in requirements previously approved in the Code and, therefore, is not 
a backfit.

ASME BPV Code, Section XI

    1. Revise Sec.  50.55a(b)(2)(vi), ``Effective edition and addenda 
of Subsection IWE and Subsection IWL,'' to clarify that the provision 
applies only to the class of licensees of operating reactors that were 
required by previous versions of Sec.  50.55a to develop, implement a 
containment ISI program in accordance with Subsection IWE and 
Subsection IWL, and complete an expedited examination of containment 
during the 5-year period from September 9, 1996, to September 9, 2001. 
This revision clarifies the current requirements, is considered to be 
consistent with the meaning and intent of the current requirements, and 
is not considered to result in a change in requirements. Therefore, 
this change is not a backfit.
    2. Revise Sec.  50.55a(b)(2)(viii), ``Concrete containment 
examinations,'' so that when using the 2007 Edition with 2009 Addenda 
through the 2013 Edition of Subsection IWL, the conditions in Sec.  
50.55a(b)(2)(viii)(E) do not apply, but the new conditions in Sec.  
50.55a(b)(2)(viii)(H) and (I) do apply. This revision does not require 
Sec.  50.55a(b)(2)(viii)(E) to be used when following the 2007 Edition 
with 2009 Addenda through the 2013 Edition of Subsection IWL because 
most of its requirements have been included in IWL-2512, ``Inaccessible 
Areas.'' Therefore, this change is not a backfit because the 
requirements have not changed. The revision to add the condition in 
Sec.  50.55a(b)(2)(viii)(H) captures the reporting requirements of the 
current Sec.  50.55a(b)(2)(viii)(E) which were not included in IWL-
2512. Therefore, this change is not a backfit because the requirements 
have not changed. The revision to add the condition in Sec.  
50.55a(b)(2)(viii)(I) addresses a new code provision in IWL-2512(b) for 
evaluation of below-grade concrete surfaces during the period of 
extended operation of a renewed license. The condition assures 
consistency with the GALL Report, Revision 2, and applies to plants 
going forward using the 2007 Edition with 2009 Addenda through the 2013 
Edition of Subsection IWL. The requirements remain unchanged from the 
recommendations in the GALL Report and, therefore, this change is not a 
backfit.
    3. Revise Sec.  50.55a(b)(2)(ix), ``Metal containment 
examinations,'' to extend the applicability of the existing conditions 
in Sec.  50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) to the 2007 
Edition with 2009 Addenda through the 2013 Edition of Subsection IWE. 
This condition does not result in a change to current requirements, and 
is therefore not a backfit.
    4. Revise Sec.  50.55a(b)(2)(x), ``Quality assurance,'' to require 
that when applying the editions and addenda later than the 1989 Edition 
of ASME BPV Code, Section XI, the requirements of NQA-1, 1994 Edition, 
the 2008 Edition, and the 2009-1a Addenda specified in either IWA-1400 
or Table IWA 1600-1, ``Referenced Standards and Specifications,'' of 
that edition and addenda of Section XI are acceptable for use, provided 
the licensee uses its appendix B to 10 CFR part 50 QA program in 
conjunction with Section XI requirements. This revision clarifies the 
current requirements, which the NRC considers to be consistent with the 
meaning and intent of the current requirements. Therefore, the NRC does 
not consider the clarification to be a change in requirements. 
Therefore, this change is not a backfit.
    5. Revise Sec.  50.55a(b)(2)(xii), ``Underwater welding,'' to allow 
underwater welding on irradiated materials under certain conditions. 
The revision eliminates the prohibition on welding on irradiated 
materials. Therefore, this change is not a backfit.
    6. Add a new condition as Sec.  50.55a(b)(2)(xviii)(D), ``NDE 
personnel certification: Fourth provision,'' to prohibit the use of 
Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013 
Edition of Section XI of the ASME BPV Code. Licensees are required to 
implement Appendix VII

[[Page 32970]]

and Subarticle VIII-2200 of the 2010 Edition of Section XI. This 
condition does not constitute a change in NRC position because the use 
of the subject provisions is not currently allowed by Sec.  50.55a. 
Therefore, the addition of this new condition is not a backfit.
    7. Revise Sec.  50.55a(b)(2)(xxi)(A), ``Table IWB-2500-1 
examination requirements: First provision,'' to modify the standard for 
visual magnification resolution sensitivity and contrast for visual 
examinations of Examination Category B-D components, making the rule 
conform with ASME BPV Code, Section XI requirements for VT-1 
examinations. This revision removes a condition that was in addition to 
the ASME BPV Code requirements and does not impose a new requirement. 
Therefore, this change is not a backfit.
    8. Add a new condition as Sec.  50.55a(b)(2)(xxxi), ``Mechanical 
clamping devices;'' to prohibit the use of mechanical clamping devices 
in accordance with IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d) 
in the 2011 Addenda through 2013 Edition on small item Class 1 piping 
and portions of a piping system that forms the containment boundary. 
This condition does not constitute a change in NRC position and does 
not affect licensees because the use of the subject provisions is not 
currently allowed by Sec.  50.55a. Therefore, the addition of this new 
condition is not a backfit.
    9. Add a new condition as Sec.  50.55a(b)(2)(xxxii), ``Summary 
report submittal,'' to clarify that licensees using the 2010 Edition or 
later editions and addenda of Section XI must continue to submit to the 
NRC the Preservice and Inservice Summary Reports required by IWA-6240 
of the 2009 Addenda of Section XI. This condition does not result in a 
change in the NRC's requirements insomuch as these reports have been 
required in the 2009 Addenda of Section XI and all previous editions 
and addenda. Therefore, the addition of this new condition is not a 
backfit.
    10. Add a new condition as Sec.  50.55a(b)(2)(xxxiii), ``Risk-
Informed allowable pressure,'' to prohibit the use of ASME BPV Code, 
Section XI, Appendix G, Paragraph G-2216. The use of Paragraph G-2216 
is not currently allowed by Sec.  50.55a. Therefore, the condition does 
not constitute a new or changed NRC position on the lack of 
acceptability of Paragraph G-2216. Therefore, the addition of this new 
condition is not a backfit.
    11. Add a new condition as Sec.  50.55a(b)(2)(xxxiv), 
``Nonmandatory Appendix U.'' Paragraph (b)(2)(xxxiv)(A) requires that 
repair or replacement activities temporarily deferred under the 
provisions of Nonmandatory Appendix U shall be performed during the 
next scheduled refueling outage. This condition is imposed to ensure 
that repairs/replacements are performed on degraded components when a 
unit is shutdown for refueling. This change is consistent with the 
condition previously placed on ASME BPV Code Case N-513-3 and, 
therefore, does not impose a new requirement. This change is not a 
backfit. Paragraph (b)(2)(xxxiv)(B) requires that the mandatory 
appendix in ASME BPV Code Case N-513-3 be used in lieu of the appendix 
referenced in Paragraph U-S1-4.2.1(c) of Appendix U. This change is 
required because the appendix referenced in Appendix U was 
unintentionally omitted. This change is not a backfit.
    12. Add a new condition as Sec.  50.55a(b)(2)(xxxv), ``Use of 
RTT0 in the KIa and KIc equations,'' 
to specify that when licensees use ASME BPV Code, Section XI 2013 
Edition Nonmandatory Appendix A, Paragraph A-4200, if T0 is 
available, then RTT0 may be used in place of 
RTNDT for applications using the KIc equation and 
the associated KIc curve, but not for applications using the 
KIa equation and the associated KIa curve. 
Conditions on the use of ASME BPV Code, Section XI, Nonmandatory 
Appendices do not constitute backfitting inasmuch as those provisions 
apply to voluntary actions initiated by the licensee to use the 
``nonmandatory compliance'' provisions in these Appendices of the rule.
    13. Add a new condition as Sec.  50.55a(b)(2)(xxxvi), ``Fracture 
toughness of irradiated materials,'' to require licensees using ASME 
BPV Code, Section XI 2013 Edition Nonmandatory Appendix A, Paragraph A-
4400, to obtain NRC approval before using irradiated T0 and 
the associated RTT0 in establishing fracture toughness of 
irradiated materials. Conditions on the use of ASME BPV Code, Section 
XI, Nonmandatory Appendices do not constitute backfitting inasmuch as 
those provisions apply to voluntary actions initiated by the licensee 
to use the ``nonmandatory compliance'' provisions in these Appendices 
of the rule.
    14. Add a new condition as Sec.  50.55a(b)(2)(xxxvii), ``ASME BPV 
Code Case N-824,'' to allow the use of the code case as conditioned. 
Conditions on the use of ASME BPV Code Case N-824 do not constitute 
backfitting, inasmuch as the use of this code case is not required by 
the NRC but instead is an alternative which may be voluntarily used by 
the licensee (i.e., a ``voluntary alternative'').

OM Code

    1. Add a new condition as Sec.  50.55a(b)(3)(ii)(A), ``MOV 
diagnostic test interval,'' to require that licensees evaluate the 
adequacy of the diagnostic test intervals established for MOVs within 
the scope of OM Code, Appendix III, not later than 5 years or three 
refueling outages (whichever is longer) from initial implementation of 
Appendix III of the OM Code. This condition represents an exception to 
a later OM Code provision but merely retains the current NRC condition 
on ASME OM Code Case OMN-1, and is therefore not a backfit because the 
NRC is not imposing a new requirement.
    2. Add a new condition as Sec.  50.55a(b)(3)(ii)(B), ``MOV testing 
impact on risk,'' to require that licensees ensure that the potential 
increase in core damage frequency and large early release frequency 
associated with the extension is acceptably small when extending 
exercise test intervals for high risk MOVs beyond a quarterly 
frequency. This condition represents an exception to a later OM Code 
provision but merely retains the current NRC condition on ASME OM Code 
Case OMN-1, and is therefore not a backfit because the NRC is not 
imposing a new requirement.
    3. Add a new condition as Sec.  50.55a(b)(3)(ii)(C), ``MOV risk 
categorization,'' to require, when applying Appendix III to the OM 
Code, that licensees categorize MOVs according to their safety 
significance using the methodology described in OM Code Case OMN-3 
subject to the conditions discussed in RG 1.192, or using an MOV risk 
ranking methodology accepted by the NRC on a plant-specific or 
industry-wide basis in accordance with the conditions in the applicable 
safety evaluation. This condition represents an exception to a later OM 
Code provision but merely retains the current NRC condition on ASME OM 
Code Case OMN-1, and is therefore not a backfit because the NRC is not 
imposing a new requirement.
    4. Add a new condition as Sec.  50.55a(b)(3)(ii)(D), ``MOV stroke 
time,'' to require that, when applying Paragraph III-3600, ``MOV 
Exercising Requirements,'' of Appendix III to the OM Code, licensees 
shall verify that the stroke time of the MOVs specified in plant 
technical specifications satisfies the assumptions in the plant's 
safety analyses. This condition retains the MOV stroke time requirement 
for a smaller set of MOVs than was specified in previous editions and 
addenda of the

[[Page 32971]]

OM Code. The retention of this requirement is not a backfit.
    5. Add new conditions as Sec.  50.55a(b)(3)(iii)(A) through (D), 
``New reactors,'' to apply specific conditions for IST programs 
applicable to licensees of new nuclear power plants in addition to the 
provisions of the OM Code as incorporated by reference with conditions 
in Sec.  50.55a. Licensees of ``new reactors'' are, as identified in 
the paragraph: (1) Holders of operating licenses for nuclear power 
reactors that received construction permits under this part on or after 
the date 12 months after August 17, 2017, and (2) holders of COLs 
issued under 10 CFR part 52, whose initial fuel loading occurs on or 
after the date 12 months after August 17, 2017. This implementation 
schedule for new reactors is consistent with the NRC regulations in 
Sec.  50.55a(f)(4)(i). These conditions represent an exception to a 
later OM Code provision but merely retain a current NRC requirement, 
and are therefore not a backfit because the NRC is not imposing a new 
requirement.
    6. Revise Sec.  50.55a(b)(3)(iv), ``Check valves (Appendix II),'' 
to specify that Appendix II, ``Check Valve Condition Monitoring 
Program,'' of the OM Code, 2003 Addenda through the 2012 Edition, is 
acceptable for use with the following clarification: Trending and 
evaluation shall support the determination that the valve or group of 
valves is capable of performing its intended function(s) over the 
entire interval. At least one of the Appendix II condition monitoring 
activities for a valve group shall be performed on each valve of the 
group at approximate equal intervals not to exceed the maximum interval 
shown in the following table:

       Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
                                                             Maximum
                                           Maximum          interval
                                      interval between       between
             Group size                 activities of     activities of
                                        member valves    each valve  in
                                        in the groups       the group
                                           (years)           (years)
------------------------------------------------------------------------
>=4.................................               4.5                16
3...................................               4.5                12
2...................................                 6                12
1...................................    Not applicable                10
------------------------------------------------------------------------

    The regulation is being revised to extend the applicability of this 
existing NRC condition on the OM Code to the 2012 Edition of the OM 
Code and to update the clarification for the use of Appendix II. This 
does not represent a change in the NRC's position that the condition is 
needed with respect to the OM Code. Therefore, this condition is not a 
backfit.
    7. Add a new condition as Sec.  50.55a(b)(3)(vii), ``Subsection 
ISTB,'' to prohibit the use of Subsection ISTB in the 2011 Addenda to 
the OM Code because the complete set of planned Code modifications to 
support the changes to the comprehensive pump test acceptance criteria 
was not made in that addenda. This condition represents an exception to 
a later OM Code provision but merely limits the use of the later Code 
provision, and is therefore not a backfit because the NRC is not 
imposing a new requirement.
    8. Add a new condition as Sec.  50.55a(b)(3)(viii), ``Subsection 
ISTE,'' to allow licensees to implement Subsection ISTE, ``Risk-
Informed Inservice Testing of Components in Light-Water Reactor Nuclear 
Power Plants,'' in the OM Code, 2009 Edition, 2011 Addenda and 2012 
Edition, where the licensee has obtained authorization to implement 
Subsection ISTE as an alternative to the applicable IST requirements in 
the OM Code on a case-by-case basis in accordance with Sec.  50.55a(z). 
This condition represents an exception to a later OM Code provision but 
merely limits the use of the later Code provision, and is therefore not 
a backfit because the NRC is not imposing a new requirement.
    9. Add a new condition as Sec.  50.55a(b)(3)(ix), ``Subsection 
ISTF,'' to specify that licensees applying Subsection ISTF, 2012 
Edition, shall satisfy the requirements of Mandatory Appendix V, ``Pump 
Periodic Verification Test Program,'' of the OM Code, 2012 Edition. The 
condition also specifies that Subsection ISTF, 2011 Addenda, is not 
acceptable for use. This condition represents an exception to a later 
OM Code provision but merely limits the use of the later Code 
provision, and is therefore not a backfit because the NRC is not 
imposing a new requirement.
    10. Add a new condition as Sec.  50.55a(b)(3)(x), ``ASME OM Code 
Case OMN-20,'' to allow licensees to implement OM Code Case OMN-20, 
``Inservice Test Frequency,'' in the OM Code, 2012 Edition. This 
condition allows voluntary action initiated by the licensee to use the 
code case and is, therefore, not a backfit.
    11. Add a new condition as Sec.  50.55a(b)(3)(xi), ``Valve Position 
Indication,'' to emphasize, when implementing OM Code (2012 Edition), 
Subsection ISTC-3700, ``Position Verification Testing,'' licensees 
shall implement the OM Code provisions to verify that valve operation 
is accurately indicated. This condition emphasizes the OM Code 
requirements for valve position indication and is not a change to those 
requirements. As such, this condition is not a backfit.
    12. Revise Sec.  50.55a(f), ``Preservice and inservice testing 
requirements,'' to clarify that the OM Code includes provisions for 
preservice testing of components as part of its overall provisions for 
IST programs. No expansion of IST program scope is intended by this 
clarification. This condition does not result in a change in 
requirements previously approved in the Code and is, therefore, not a 
backfit.
    13. Revise Sec.  50.55a(f)(4), ``Inservice testing standards for 
operating plants,'' to state that the paragraph is applicable to pumps 
and valves that are within the scope of the OM Code. Also, revise Sec.  
50.55a(f)(4) to state that the IST requirements for pumps and valves 
that are within the scope of the OM Code but are not classified as ASME 
BPV Code Class 1, Class 2, or Class 3 may be satisfied as an augmented 
IST program in accordance with Sec.  50.55a(f)(6)(ii) without 
requesting relief under Sec.  50.55a(f)(5) or alternatives under Sec.  
50.55a(z). This use of an augmented IST program may be acceptable 
provided the basis for deviations from the OM Code as incorporated by 
reference in this section demonstrates an acceptable level of quality 
and safety, or that implementing the Code provisions would result in 
hardship or

[[Page 32972]]

unusual difficulty without a compensating increase in the level of 
quality and safety, where documented and available for NRC review. 
These changes align the scope of pumps and valves for inservice testing 
with the scope defined in the OM Code. These changes do not result in a 
change in requirements previously approved in the Code, and is 
therefore not a backfit.

ASME BPV Code Case N-729-4

    Revise Sec.  50.55a(g)(6)(ii)(D), ``Reactor vessel head 
inspections.''
    On June 22, 2012, the ASME approved the fourth revision of ASME BPV 
Code Case N-729 (N-729-4). The NRC proposed to update the requirements 
of Sec.  50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV 
Code Case N-729-4, with conditions. The ASME BPV Code Case N-729-4 
contains similar requirements as N-729-1; however, N-729-4 also 
contains new requirements to address previous NRC conditions, including 
changes to inspection frequency and qualifications. The new NRC 
conditions on the use of ASME BPV Code Case N-729-4 address operational 
experience, clarification of implementation, and the use of 
alternatives to the code case.
    The current regulatory requirements for the examination of 
pressurized water reactor upper RPV heads that use nickel-alloy 
materials are provided in Sec.  50.55a(g)(6)(ii)(D). This section was 
first created by rulemaking, dated September 10, 2008 (73 FR 52730), to 
require licensees to implement ASME BPV Code Case N-729-1, with 
conditions, instead of the inspections previously required by the ASME 
BPV Code, Section XI. The action did constitute a backfit; however, the 
NRC concluded that imposition of ASME BPV Code Case N-729-1, as 
conditioned, constituted an adequate protection backfit.
    The General Design Criteria (GDC) for nuclear power plants 
(appendix A to 10 CFR part 50) or, as appropriate, similar requirements 
in the licensing basis for a reactor facility, provide bases and 
requirements for NRC assessment of the potential for, and consequences 
of, degradation of the reactor coolant pressure boundary (RCPB). The 
applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC 
31 (Fracture Prevention of Reactor Coolant Pressure Boundary), and GDC 
32 (Inspection of Reactor Coolant Pressure Boundary). General Design 
Criterion 14 specifies that the RCPB be designed, fabricated, erected, 
and tested so as to have an extremely low probability of abnormal 
leakage, of rapidly propagating failure, and of gross rupture. General 
Design Criterion 31 specifies that the probability of rapidly 
propagating fracture of the RCPB be minimized. General Design Criterion 
32 specifies that components that are part of the RCPB have the 
capability of being periodically inspected to assess their structural 
and leak-tight integrity.
    The NRC concludes that ASME BPV Code Case N-729-4, as conditioned, 
shall be mandatory in order to ensure that the requirements of the GDC 
are satisfied. Imposition of ASME BPV Code Case N-729-4, with 
conditions, ensures that the ASME BPV Code-allowable limits will not be 
exceeded, leakage will likely not occur, and potential flaws will be 
detected before they challenge the structural or leak-tight integrity 
of the RPV upper head within current nondestructive examination 
limitations. The NRC concludes that the regulatory framework for 
providing adequate protection of public health and safety is 
accomplished by the incorporation of ASME BPV Code Case N-729-4 into 
Sec.  50.55a, as conditioned. All current licensees of U.S. pressurized 
water reactors will be required to implement ASME BPV Code Case N-729-
4, as conditioned. The Code Case provisions on examination requirements 
for RPV upper heads are essentially the same as those established under 
ASME BPV Code Case N-729-1, as conditioned. One exception is the 
condition in Sec.  50.55a(g)(6)(ii)(D)(3), which will require, for 
upper heads with Alloy 600 penetration nozzles, that bare metal visual 
examinations be performed each outage in accordance with Table 1 of 
ASME BPV Code Case N-729-4. Accordingly, the NRC imposition of the ASME 
BPV Code Case N-729-4, as conditioned, may be deemed to be a 
modification of the procedures to operate a facility resulting from the 
imposition of the new regulation, and as such, this rulemaking 
provision may be considered backfitting under Sec.  50.109(a)(1).
    The NRC continues to find that inspections of RPV upper heads, 
their penetration nozzles, and associated partial penetration welds are 
necessary for adequate protection of public health and safety and that 
the requirements of ASME BPV Code Case N-729-4, as conditioned, 
represent an acceptable approach, developed, in part, by a voluntary 
consensus standards body for performing future inspections. The NRC 
concludes that approval of ASME BPV Code Case N-729-4, as conditioned, 
by incorporation by reference of the Code Case into Sec.  50.55a, is 
necessary to ensure that the facility provides adequate protection to 
the health and safety of the public and constitutes a redefinition of 
the requirements necessary to provide reasonable assurance of adequate 
protection of public health and safety. Therefore, a backfit analysis 
need not be prepared for this portion of the rule in accordance with 
Sec.  50.109(a)(4)(ii) and (iii).

ASME BPV Code Case N-770-2

    Revise Sec.  50.55a(g)(6)(ii)(F), ``Examination requirements for 
Class 1 piping and nozzle dissimilar metal butt welds.''
    On June 9, 2011, the ASME approved the second revision of ASME BPV 
Code Case N-770 (N-770-2). The NRC is updating the requirements of 
Sec.  50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV 
Code Case N-770-2, with conditions. The ASME BPV Code Case N-770-2 
contains similar baseline and ISI requirements for unmitigated nickel-
alloy butt welds, and preservice and ISI requirements for mitigated 
butt welds as N-770-1. However, N-770-2 also contains new requirements 
for optimized weld overlays, a specific mitigation technique and 
volumetric inspection coverage. Further, the NRC conditions on the use 
of ASME BPV Code Case N-770-2 have been modified to address the changes 
in the code case, clarify inspection coverage requirements and require 
the development of inspection qualifications to allow complete weld 
inspection coverage in the future.
    The current regulatory requirements for the examination of ASME 
Class 1 piping and nozzle dissimilar metal butt welds that use nickel-
alloy materials is provided in Sec.  50.55a(g)(6)(ii)(F). This section 
was first created by rulemaking, dated June 21, 2011 (76 FR 36232), to 
require licensees to implement ASME BPV Code Case N-770-1, with 
conditions. The NRC added Sec.  50.55a(g)(6)(ii)(F) to require 
licensees to implement ASME BPV Code Case N-770-1, with conditions, 
instead of the inspections previously required by the ASME BPV Code, 
Section XI. The action did constitute a backfit; however, the NRC 
concluded that imposition of ASME BPV Code Case N-770-1, as 
conditioned, constituted an adequate protection backfit.
    The GDC for nuclear power plants (appendix A to 10 CFR part 50) or, 
as appropriate, similar requirements in the licensing basis for a 
reactor facility, provide bases and requirements for NRC assessment of 
the potential for, and consequences of, degradation of the RCPB. The 
applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC 
31 (Fracture Prevention of Reactor Coolant Pressure Boundary) and GDC 
32 (Inspection of Reactor Coolant Pressure

[[Page 32973]]

Boundary). General Design Criterion 14 specifies that the RCPB be 
designed, fabricated, erected, and tested so as to have an extremely 
low probability of abnormal leakage, of rapidly propagating failure, 
and of gross rupture. General Design Criterion 31 specifies that the 
probability of rapidly propagating fracture of the RCPB be minimized. 
General Design Criterion 32 specifies that components that are part of 
the RCPB have the capability of being periodically inspected to assess 
their structural and leak-tight integrity.
    The NRC concludes that ASME BPV Code Case N-770-2, as conditioned, 
must be imposed in order to ensure that the requirements of the GDC are 
satisfied. Imposition of ASME BPV Code Case N-770-2, with conditions, 
ensures that the requirements of the GDC are met for all mitigation 
techniques currently in use for Alloy 82/182 butt welds because ASME 
BPV Code-allowable limits will not be exceeded, leakage would likely 
not occur and potential flaws will be detected before they challenge 
the structural or leak-tight integrity of piping welds. All current 
licensees of U.S. pressurized water reactors will be required to 
implement ASME BPV Code Case N-770-2, as conditioned. The Code Case 
provisions on examination requirements for ASME Class 1 piping and 
nozzle nickel-alloy dissimilar metal butt welds are somewhat different 
from those established under ASME BPV Code Case N-770-1, as 
conditioned, and will require a licensee to modify its procedures for 
inspection of ASME Class 1 nickel-alloy welds to meet these 
requirements. Accordingly, the NRC imposition of the ASME BPV Code Case 
N-770-2, as conditioned, may be deemed to be a modification of the 
procedures to operate a facility resulting from the imposition of the 
new regulation, and as such, this rulemaking provision may be 
considered backfitting under Sec.  50.109(a)(1).
    The NRC continues to find that ASME Class 1 nickel-alloy dissimilar 
metal weld inspections are necessary for adequate protection of public 
health and safety, and that the requirements of ASME BPV Code Case N-
770-2, as conditioned, represent an acceptable approach developed by a 
voluntary consensus standards body for performing future ASME Class 1 
nickel-alloy dissimilar metal weld inspections. The NRC concludes that 
approval of ASME BPV Code Case N-770-2, as conditioned, by 
incorporation by reference of the Code Case into Sec.  50.55a, is 
necessary to ensure that the facility provides adequate protection to 
the health and safety of the public and constitutes a redefinition of 
the requirements necessary to provide reasonable assurance of adequate 
protection of public health and safety. Therefore, a backfit analysis 
need not be prepared for this portion of the rule in accordance with 
Sec.  50.109(a)(4)(ii) and (iii).

Conclusion

    The NRC finds that incorporation by reference into Sec.  50.55a of 
the 2009 Addenda through 2013 Edition of Section III, Division 1, of 
the ASME BPV Code, subject to the identified conditions; the 2009 
Addenda through 2013 Edition of Section XI, Division 1, of the ASME BPV 
Code, subject to the identified conditions; and the 2009 Edition 
through the 2012 Edition of the OM Code, subject to the identified 
conditions, does not constitute backfitting or represent an 
inconsistency with any issue finality provisions in 10 CFR part 52.
    The NRC finds that the incorporation by reference of Code Cases N-
824 and OMN-20 does not constitute backfitting or represent an 
inconsistency with any issue finality provisions in 10 CFR part 52.
    The NRC finds that the inclusion of a new condition on Code Case N-
729-4 and a new condition on Code Case N-770-2 constitutes backfitting 
necessary for adequate protection.

X. 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).

XI. Finding of No Significant Environmental Impact: Environmental 
Assessment

    This final rule is in accordance with the NRC's policy to 
incorporate by reference in Sec.  50.55a new editions and addenda of 
the ASME BPV and OM Codes to provide updated rules for constructing and 
inspecting components and testing pumps, valves, and dynamic restraints 
(snubbers) in light-water nuclear power plants. The ASME Codes are 
national voluntary consensus standards and are required by the NTTAA to 
be used by government agencies unless the use of such a standard is 
inconsistent with applicable law or otherwise impractical. The National 
Environmental Policy Act (NEPA) requires Federal 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 proposed action and alternatives to 
the proposed action (42 U.S.C. 4332(C); NEPA Sec. 102(C)).
    The NRC has determined under NEPA, as amended, and the NRC's 
regulations in subpart A of 10 CFR part 51, that this rule is not a 
major Federal action significantly affecting the quality of the human 
environment and, therefore, an environmental impact statement is not 
required. The rulemaking does not significantly increase the 
probability or consequences of accidents, no changes are being made in 
the types of effluents that may be released off-site, and there is no 
significant increase in public radiation exposure. The NRC estimates 
the radiological dose to plant personnel performing the inspections 
required by ASME BPV Code Case N-770-2 would be about 3 rem per plant 
over a 10-year interval, and a one-time exposure for mitigating welds 
of about 30 rem per plant. The NRC estimates the radiological dose to 
plant personnel performing the inspections required by ASME BPV Code 
Case N-729-4 would be about 3 rem per plant over a 10-year interval and 
a one-time exposure for mitigating welds of about 30 rem per plant. As 
required by 10 CFR part 20, and in accordance with current plant 
procedures and radiation protection programs, plant radiation 
protection staff will continue monitoring dose rates and would make 
adjustments in shielding, access requirements, decontamination methods, 
and procedures as necessary to minimize the dose to workers. The 
increased occupational dose to individual workers stemming from the 
ASME BPV Code Case N-770-2 and N-729-4 inspections must be maintained 
within the limits of 10 CFR part 20 and as low as reasonably 
achievable. Therefore, the NRC concludes that the increase in 
occupational exposure would not be significant. This final rule does 
not involve non-radiological plant effluents and has no other 
environmental impacts. Therefore, no significant non-radiological 
impacts are associated with this action. The determination of this 
environmental assessment is that there will be no significant off-site 
impact to the public from this action.

XII. Paperwork Reduction Act Statement

    This final rule amends collections of information subject to the 
Paperwork Reduction Act of 1995 (44 U.S.C. 3501

[[Page 32974]]

et seq.). The collections of information were approved by the Office of 
Management and Budget (OMB), approval number 3150-0011.
    Because the rule will reduce the burden for existing information 
collections, the public burden for the information collections is 
expected to be decreased by 58.5 hours per response. This reduction 
includes the time for reviewing instructions, searching existing data 
sources, gathering and maintaining the data needed, and completing and 
reviewing the information collection.
    The information collection is being conducted to document the plans 
for and the results of ISI and IST programs. The records are generally 
historical in nature and provide data on which future activities can be 
based. The practical utility of the information collection for the NRC 
is that appropriate records are available for auditing by NRC personnel 
to determine if ASME BPV and OM Code provisions for construction, 
inservice inspection, repairs, and inservice testing are being properly 
implemented in accordance with Sec.  50.55a, or whether specific 
enforcement actions are necessary. Responses to this collection of 
information are generally mandatory under 10 CFR 50.55a.
    You may submit comments on any aspect of the information 
collection(s), including suggestions for reducing the burden, by the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2011-0088.
     Mail comments to: Information Services Branch, Office of 
the Chief Information Officer, Mail Stop: T-2F43, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001 or to Aaron Szabo, 
Desk Officer, Office of Information and Regulatory Affairs (3150-0011), 
NEOB-10202, Office of Management and Budget, Washington, DC 20503; 
telephone: 202-395-3621, email: [email protected].

Public Protection Notification

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

XIII. Congressional Review Act

    This final rule is a rule as defined in the Congressional Review 
Act (5 U.S.C. 801-808). However, OMB has not found it to be a major 
rule as defined in the Congressional Review Act.

XIV. Voluntary Consensus Standards

    Section 12(d)(3) of the National Technology Transfer and 
Advancement Act of 1995, Public Law 104-113 (NTTAA), and implementing 
guidance in OMB Circular A-119 (February 10, 1998), requires 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 which are not 
acceptable to the agency.
    In this final rule, the NRC is continuing its existing practice of 
establishing requirements for the design, construction, operation, ISI 
(examination), and IST of nuclear power plants by approving the use of 
the latest editions and addenda of the ASME Codes in Sec.  50.55a. The 
ASME Codes are voluntary consensus standards, developed by participants 
with broad and varied interests, in which all interested parties 
(including the NRC and licensees of nuclear power plants) participate. 
Therefore, the NRC's incorporation by reference of the ASME Codes is 
consistent with the overall objectives of the NTTAA and OMB Circular A-
119.
    In this final rule, the NRC is also 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 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 in this final rule is consistent with the overall objectives of 
the NTTAA and OMB Circular A-119.
    As discussed in Section II of this document, ``Discussion,'' the 
NRC is conditioning the use of certain provisions of the 2009 Addenda, 
2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code 
Section III, Division 1 and Section XI, Division 1. The NRC is also 
conditioning the use of certain provisions of the 2009 Edition, the 
2011 Addenda, and the 2012 Edition of the OM Code, Division 1. This 
final rule also includes various versions of quality assurance standard 
NQA-1 and Code Cases N-729-4, N-770-2, N-824, OMN-20, N-513-3 Mandatory 
Appendix I, and N-852. In addition, this final rule does not adopt 
(``excludes'') certain provisions of the ASME Codes, as discussed in 
this statement of considerations and in the regulatory analysis for 
this rulemaking. The NRC staff's position is that this final rule 
complies with the NTTAA and OMB Circular A-119 despite these conditions 
and ``exclusions.''
    If the NRC did not conditionally accept ASME editions, addenda, and 
code cases, the NRC would disapprove these entirely. The effect would 
be that licensees and applicants would submit a larger number of 
requests for use of alternatives under Sec.  50.55a(z), requests for 
relief under Sec.  50.55a(f) and (g), or requests for exemptions under 
Sec.  50.12 and/or Sec.  52.7. These requests would likely include 
broad scope requests for approval to issue the full scope of the ASME 
Code editions and addenda which would otherwise be approved in this 
final rule (i.e., the request would not be simply for approval of a 
specific ASME Code provision with conditions). These requests would be 
an unnecessary additional burden for both the licensee and the NRC, 
inasmuch as the NRC has already determined that the ASME Codes and Code 
Cases which are the subject of this final rule are acceptable for use 
(in some cases with conditions). For these reasons, the NRC concludes 
that this final rule's treatment of ASME Code editions and addenda, and 
code cases and any conditions placed on them 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 U.S. voluntary consensus standards bodies for use within 
the United States, which the NRC could incorporate by reference instead 
of the ASME Codes. The NRC also did not

[[Page 32975]]

identify any voluntary consensus standards, developed by multinational 
voluntary consensus standards bodies for use on a multinational basis, 
which the NRC could incorporate by reference instead of the ASME Codes. 
The NRC identified codes addressing the same subject as the ASME Codes 
for use in individual countries. At least one country, Korea, directly 
translated the ASME Code for use in that country. In other countries 
(e.g., Japan), ASME Codes were the basis for development of the 
country's codes, but the ASME Codes were substantially modified to 
accommodate that country's regulatory system and reactor designs. 
Finally, there are countries (e.g., the Russian Federation) where that 
country's code was developed without regard to the ASME Code. However, 
some of these codes may not meet the definition of a voluntary 
consensus standard because they were developed by the state rather than 
a voluntary consensus standards body. The NRC's evaluation of other 
countries' codes to determine whether each code provides a comparable 
or enhanced level of safety, when compared against the level of safety 
provided under the ASME Codes, would require a significant expenditure 
of agency resources. This expenditure does not seem justified, given 
that substituting another country's code for the U.S. voluntary 
consensus standard does not appear to substantially further the 
apparent underlying objectives of the NTTAA.
    In summary, this final rule satisfies the requirements of Section 
12(d)(3) of the NTTAA and OMB Circular A-119.

XV. Incorporation by Reference--Reasonable Availability to Interested 
Parties

    The NRC is incorporating by reference recent editions and addenda 
to the ASME Codes for nuclear power plants and a standard for quality 
assurance. The NRC is also incorporating by reference six ASME Code 
Cases. As described in the ``Background'' and ``Discussion'' sections 
of this document, these materials provide rules for safety governing 
the design, fabrication, and inspection of nuclear power plant 
components.
    The NRC is required by law to obtain approval for incorporation by 
reference from the Office of the Federal Register (OFR). The OFR's 
requirements for incorporation by reference are set forth in 1 CFR part 
51. On November 7, 2014, the OFR adopted changes to its regulations 
governing incorporation by reference (79 FR 66267). The OFR regulations 
require an agency to include in a final rule a discussion of the ways 
that the materials the agency incorporates by reference are reasonably 
available to interested parties and how interested parties can obtain 
the materials. The discussion in this section complies with the 
requirement for final rules as set forth in Sec.  51.5(b).
    The NRC considers ``interested parties'' to include all potential 
NRC stakeholders, not only the individuals and entities regulated or 
otherwise subject to the NRC's regulatory oversight. These NRC 
stakeholders are not a homogenous group, so the considerations for 
determining ``reasonable availability'' vary by class of interested 
parties. The NRC identifies six classes of interested parties with 
regard to the material to be incorporated by reference in an NRC rule:
     Individuals and small entities regulated or otherwise 
subject to the NRC's regulatory oversight who are subject to the 
material to be incorporated by reference by rulemaking. This class also 
includes applicants and potential applicants for licenses and other NRC 
regulatory approvals. In this context, ``small entities'' has the same 
meaning as a ``small entity'' under Sec.  2.810.
     Large entities otherwise subject to the NRC's regulatory 
oversight who are subject to the material to be incorporated by 
reference by rulemaking. This class also includes applicants and 
potential applicants for licenses and other NRC regulatory approvals. 
In this context, ``large entities'' are those which do not qualify as a 
``small entity'' under Sec.  2.810.
     Non-governmental organizations with institutional 
interests in the matters regulated by the NRC.
     Other Federal agencies, states, local governmental bodies 
(within the meaning of Sec.  2.315(c)).
     Federally-recognized and State-recognized \3\ Indian 
tribes.
---------------------------------------------------------------------------

    \3\ State-recognized Indian tribes are not within the scope of 
Sec.  2.315(c). However, for purposes of the NRC's compliance with 1 
CFR 51.5, ``interested parties'' includes a broad set of 
stakeholders including State-recognized Indian tribes.
---------------------------------------------------------------------------

     Members of the general public (i.e., individual, 
unaffiliated members of the public who are not regulated or otherwise 
subject to the NRC's regulatory oversight) who may wish to gain access 
to the materials that the NRC proposes to incorporate by reference in 
order to participate in the rulemaking.
    The NRC makes the materials to be incorporated by reference 
available for inspection to all interested parties, by appointment, at 
the NRC Technical Library, which is located at Two White Flint North, 
11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-
7000; email: [email protected].
    Interested parties may purchase a copy of the materials from ASME 
at Three Park Avenue, New York, NY 10016, or at the ASME Web site 
https://www.asme.org/shop/standards. The materials are also accessible 
through third-party subscription services such as IHS (15 Inverness Way 
East, Englewood, CO 80112; https://global.ihs.com) and Thomson Reuters 
Techstreet (3916 Ranchero Dr., Ann Arbor, MI 48108; http://www.techstreet.com). The purchase prices for individual documents range 
from $225 to $720 and the cost to purchase all documents is 
approximately $9,000.
    For the class of interested parties constituting members of the 
general public who wish to gain access to the materials to be 
incorporated by reference in order to participate in the rulemaking, 
the NRC recognizes that the $9,000 cost may be so high that the 
materials could be regarded as not reasonably available for purposes of 
commenting on this rulemaking, despite the NRC's actions to make the 
materials available at the NRC's PDR.
    Accordingly, the NRC sent a letter to the ASME on April 9, 2015, 
requesting that they consider enhancing public access to these 
materials during the public comment period. In an April 21, 2015, 
letter to the NRC, the ASME agreed to make the materials available 
online in a read-only electronic access format during the public 
comment period.
    During the public comment period, the ASME made publicly-available 
the editions and addenda to the ASME Codes for nuclear power plants, 
the ASME standard for quality assurance, and the ASME Code Cases which 
the NRC proposed to incorporate by reference. The ASME made the 
materials publicly-available in read-only format at the ASME Web site 
http://go.asme.org/NRC.
    The materials are available to all interested parties in multiple 
ways and in a manner consistent with their interest in this rulemaking. 
Therefore, the NRC concludes that the materials the NRC is 
incorporating by reference in this rulemaking are reasonably available 
to all interested parties.

XVI. Availability of Guidance

    The NRC will not be issuing guidance for this rulemaking. The ASME 
BPV Code and OM Code provide direction for the performance of 
activities to satisfy the Code requirements for design, inservice 
inspection, and inservice testing of nuclear power plant SSCs. In 
addition, the NRC provides

[[Page 32976]]

guidance in this Federal Register notice for the implementation of the 
new conditions on the ASME BPV Code and OM Code, as necessary. The NRC 
has a number of standard review plans (SRPs), which provide guidance to 
NRC reviewers and make communication and understanding of NRC review 
processes available to members of the public and the nuclear power 
industry. NUREG-0800, ``Review of Safety Analysis Reports for Nuclear 
Power Plants,'' has numerous sections which discuss implementation of 
various aspects of the ASME BPV Code and OM Code (e.g., Sections 3.2.2, 
3.8.1, 3.8.2, 3.9.3, 3.9.6, 3.9.7, 3.9.8, 3.13, 5.2.1.1, 5.2.1.2, 
5.2.4, and 6.6). The NRC also publishes Regulatory Guides and Generic 
Communications (i.e., Regulatory Issue Summaries, Information Notices) 
to communicate and clarify NRC technical or policy positions on 
regulatory matters which may contain guidance relative to this 
rulemaking.
    Revision 2 of NUREG-1482, ``Guidelines for Inservice Testing at 
Nuclear Power Plants,'' provides guidance for the development and 
implementation of IST programs at nuclear power plants. With direction 
provided in the ASME BPV and OM Codes, and guidance in this Federal 
Register notice, the NRC has determined that preparation of a separate 
guidance document is not necessary for this update to Sec.  50.55a. 
However, the NRC will consider preparation of a revision to NUREG-1482 
in the future to address the latest edition of the ASME OM Code 
incorporated by reference in Sec.  50.55a.

XVII. Availability of Documents

    The NRC is making the documents identified in Table 2 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 2--Availability of Documents
------------------------------------------------------------------------
                                                   ADAMS Accession No./
                    Document                         Federal Register
                                                     citation/Web link
------------------------------------------------------------------------
Proposed Rule Documents:
    Proposed Rule--Federal Register Notice......  80 FR 56820 (September
                                                   18, 2015).
    Draft Regulatory Analysis...................  ML14170B104.
Final Rule Documents:
    Final Regulatory Analysis...................  ML16130A522.
    2017 Final Rule (10 CFR 50.55a) American      ML16130A531.
     Society of Mechanical Engineers Codes and
     Code Cases: Analysis of Public Comments.
Related Documents:
    Fatigue and Fracture Mechanics: 33rd Volume,  https://www.astm.org/
     ASTM STP 1417, W.G. Reuter and R.S.           DIGITAL_LIBRARY/STP/
     Piascik, Eds., ASTM International, West       SOURCE_PAGES/
     Conshohocken, PA, 2002.                       STP1417.htm.
    Final Results from the CARINA Project on      http://www.astm.org/
     Crack Initiation and Arrest of Irradiated     DIGITAL_LIBRARY/STP/
     German RPV Steels for Neutron Fluences in     PAGES/
     the Upper Bound, H. Hein et al., ASTM         STP157220130113.htm.
     International, West Conshohocken, PA, June
     2014.
    Letter from Brian Thomas, NRC, to Michael     ML15085A206.
     Merker, ASME, ``Public Access to Material
     the NRC Seeks to Incorporate by Reference
     into its Regulations,'' April 9, 2015.
    Letter from Mark Maxin, NRC, to Rick Libra,   ML081680730.
     BWRVIP Chairman, ``Safety Evaluation for
     Electric Power Research Institute (EPRI)
     Boiling Water Reactor (BWR) Vessel and
     Internals Project (BWRVIP) Report 1003020
     (BWRVIP-97), `BWR Vessel and Internals
     Project, Guidelines for Performing Weld
     Repairs to Irradiated BWR Internals' (TAC
     No. MC3948),'' June 30, 2008.
    Letter from Michael Merker, ASME, to Brian    ML15112A064.
     Thomas, NRC; April 21, 2015.
    Licensee Event Report 50-338/2012-001-00....  ML12151A441.
    NUREG-0800, ``Standard Review Plan for the    ML070660036.
     Review of Safety Analysis Reports for
     Nuclear Power Plants, LWR Edition''.
    NUREG-0800, Section 3.9.6, Revision 3,        ML070720041.
     ``Functional Design, Qualification, and
     Inservice Testing Programs for Pumps,
     Valves, and Dynamic Restraints,'' March
     2007.
    NUREG-1482, Revision 2, ``Guidelines for      ML13295A020.
     Inservice Testing at Nuclear Power Plants:
     Inservice Testing of Pumps and Valves and
     Inservice Examination and Testing of
     Dynamic Restraints (Snubbers) at Nuclear
     Power Plants,'' October 2013.
    NUREG-1800, Revision 2, ``Standard Review     ML103490036.
     Plan for Review of License Renewal
     Applications for Nuclear Power Plants,''
     December 2010.
    NUREG-1801, Revision 2, ``Generic Aging       ML103490041.
     Lessons Learned (GALL) Report,'' December
     2010.
    NUREG-1950, ``Disposition of Public Comments  ML11116A062.
     and Technical Bases for Changes in the
     License Renewal Guidance Documents NUREG-
     1801 and NUREG-1800,'' April 2011.
    NUREG-2124, ``Final Safety Evaluation Report  ML12271A045.
     Related to the Combined Licenses for Vogtle
     Electric Generating Plant, Units 3 and 4,''
     Section 3.9.6, ``Inservice Testing of Pumps
     and Valves (Related to RG 1.206, Section
     C.III.1, Chapter 3, C.I.3.9.6, `Functional
     Design, Qualification, and Inservice
     Testing Programs for Pumps, Valves, and
     Dynamic Restraints')''.
    NUREG/CR-6860, ``An Assessment of Visual      ML043630040.
     Testing,'' November 2004.
    NUREG/CR-6933, ``Assessment of Crack          ML071020410 and
     Detection in Heavy-Walled Cast Stainless      ML071020414.
     Steel Piping Welds Using Advanced Low-
     Frequency Ultrasonic Methods,'' March 2007.
    NUREG/CR-7122, ``An Evaluation of Ultrasonic  ML12087A004.
     Phased Array Testing for Cast Austenitic
     Stainless Steel Pressurizer Surge Line
     Piping Welds,'' March 2012.
    NRC Generic Letter 89-10, ``Safety-Related    ML031150300.
     Motor-Operated Valve Testing and
     Surveillance,'' June 1989.
    NRC Generic Letter 90-05, ``Guidance for      ML031140590.
     Performing Temporary Non-Code Repair of
     ASME Code Class 1, 2, and 3 Piping (Generic
     Letter 90-05),'' June 1990.
    NRC Meeting Summary of June 5-7, 2013,        ML14003A230.
     Annual Materials Programs Technical
     Information Exchange Public Meeting.
    NRC Meeting Summary of January 19, 2016,      ML16050A383.
     Category 2 Public Meeting with Industry
     Representatives to Discuss Welding on
     Neutron Irradiated Ferritic and Austenitic
     Materials.

[[Page 32977]]

 
    NRC Meeting Summary of March 2, 2016, Public  ML16069A408.
     Meeting on Stakeholder Comments on the
     Proposed Rule.
    NRC Staff Memorandum, ``Consolidation of      ML003708048.
     SECY-94-084 and SECY-95-132,'' July 24,
     1995.
    NRC Memorandum, ``Staff Requirements--        ML003755050.
     Affirmation Session, 11:30 a.m., Friday,
     September 10, 1999, Commissioners'
     Conference Room, One White Flint North,
     Rockville, Maryland (Open to Public
     Attendance),'' September 10, 1999.
    NRC Regulatory Guide 1.28, Revision 4,        ML100160003.
     ``Quality Assurance Program Criteria
     (Design and Construction),'' June 2010.
    NRC Regulatory Guide 1.147, Revision 17,      ML13339A689.
     ``Inservice Inspection Code Case
     Acceptability, ASME Section XI, Division
     1,'' August 2014.
    NRC Regulatory Guide 1.174, Revision 2, ``An  ML100910006.
     Approach for Using Probabilistic Risk
     Assessment in Risk-Informed Decisions on
     Plant-Specific Changes to the Licensing
     Basis,'' May 2011.
    NRC Regulatory Guide 1.175, ``An Approach     ML003740149.
     for Plant-Specific, Risk-Informed
     Decisionmaking: Inservice Testing,'' August
     1998.
    NRC Regulatory Guide 1.192, Revision 1,       ML13340A034.
     ``Operation and Maintenance Code Case
     Acceptability, ASME OM Code,'' August 2014.
    NRC Regulatory Guide 1.200, Revision 2, ``An  ML090410014.
     Approach for Determining the Technical
     Adequacy of Probabilistic Risk Assessment
     Results for Risk-Informed Activities,''
     March 2009.
    NRC Regulatory Guide 1.201, Revision 1,       ML061090627.
     ``Guidelines for Categorizing Structures,
     Systems, and Components in Nuclear Power
     Plants According to Their Safety
     Significance,'' May 2006.
    NRC Regulatory Information Conference,        http://www.nrc.gov/
     Recent Operating Reactors Materials Issues,   public-involve/
     Presentation Materials, 2013.                 conference-symposia/
                                                   ric/past/2013/docs/
                                                   abstracts/
                                                   sessionabstract-
                                                   19.html.
    NRC Regulatory Issue Summary 2013-07, ``NRC   ML13003A207.
     Staff Position on the Use of American
     Society of Mechanical Engineers
     Certification Mark,'' May 28, 2013.
    Relief Request REP-1 U2, Revision 2.........  ML13232A308.
    SECY-90-016, ``Evolutionary Light Water       ML003707849.
     Reactor (LWR) Certification Issues and
     Their Relationship to Current Regulatory
     Requirements''.
    SECY-93-087, ``Policy, Technical, and         ML003708021.
     Licensing Issues Pertaining to Evolutionary
     and Advanced Light-Water Reactor (ALWR)
     Designs''.
    SECY-94-084, ``Policy and Technical Issues    ML003708068.
     Associated with the Regulatory Treatment of
     Non-Safety Systems in Passive Plant
     Designs''.
    SECY-95-132, ``Policy and Technical Issues    ML003708005.
     Associated with the Regulatory Treatment of
     Non-Safety Systems (RTNSS) in Passive Plant
     Designs (SECY-94-084)''.
    Vogtle Electric Generating Plant, Units 3     ML14183B276.
     and 4, Updated Final Safety Analysis
     Report, Revision 3, Chapter 3, Section 3.9,
     Mechanical Systems and Components.
------------------------------------------------------------------------

List of Subjects in 10 CFR Part 50

    Administrative practice and procedure, Antitrust, Classified 
information, Criminal penalties, Education, Fire prevention, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Penalties, Radiation protection, 
Reactor siting criteria, Reporting and recordkeeping requirements, 
Whistleblowing.

    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 continues to read as follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste 
Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National 
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 
note; Sec. 109, Pub. L. 96-295, 94 Stat. 783.

0
2. In Sec.  50.55a:
0
a. Revise paragraphs (a) introductory text, (a)(1)(i) introductory text 
and (a)(1)(i)(E)(12) and (13) and add paragraphs (a)(1)(i)(E)(14) 
through (17);
0
b. Revise paragraphs (a)(1)(ii) introductory text and (a)(1)(ii)(C)(48) 
and (49) and add paragraphs (a)(1)(ii)(C)(50) through (53);
0
c. Revise paragraphs (a)(1)(iii)(A) through (C) and add paragraphs 
(a)(1)(iii)(D) through (G);
0
d. Revise paragraph (a)(1)(iv) introductory text and add paragraphs 
(a)(1)(iv)(B) and (C);
0
e. Add paragraph (a)(1)(v);
0
f. Revise paragraphs (b) introductory text, (b)(1) introductory text 
and (b)(1)(ii), (iv), and (vii) and add paragraphs (b)(1)(viii) and 
(ix);
0
g. Revise paragraphs (b)(2) introductory text, (b)(2)(vi), and 
(b)(2)(viii) introductory text, add paragraphs (b)(2)(viii)(H) and (I), 
revise paragraphs (b)(2)(ix) introductory text, (b)(2)(ix)(D), and 
(b)(2)(x) and (xii), add paragraph (b)(2)(xviii)(D), revise paragraphs 
(b)(2)(xxi)(A) and (b)(2)(xxiii), add and reserve paragraph 
(b)(2)(xxx), and add paragraphs (b)(2)(xxxi) through (xxxvii);
0
h. Revise paragraphs (b)(3) introductory text and (b)(3)(i) and (ii), 
add paragraph (b)(3)(iii), revise paragraph (b)(3)(iv) introductory 
text, and add paragraphs (b)(3)(vii) through (xi);
0
i. Revise paragraphs (b)(4) introductory text and (b)(5) and (6);
0
j. Revise paragraphs (f) heading and introductory text, (f)(2), 
(f)(3)(iii)(A) and (B), (f)(3)(iv)(A) and (B), (f)(4) introductory 
text, and (f)(4)(i) and (ii); and
0
k. Revise paragraphs (g) heading and introductory text, (g)(2), and 
(g)(3)

[[Page 32978]]

heading, remove paragraph (g)(3) introductory text, revise paragraphs 
(g)(3)(i), (ii), and (v), (g)(4)(i) and (ii), and (g)(6)(ii)(D)(1) 
through (4), remove paragraphs (g)(6)(ii)(D)(5) and (6), revise 
paragraphs (g)(6)(ii)(F)(1) through (10), and add paragraphs 
(g)(6)(ii)(F)(11) through (13).
    The revisions and additions read as follows:


Sec.  50.55a  Codes and standards.

    (a) Documents approved for incorporation by reference. The 
standards listed in this paragraph (a) 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, by appointment, at the NRC Technical Library, 
which is located at Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland 20852; telephone: 301-415-7000; email: 
[email protected]; 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) * * *
    (i) ASME Boiler and Pressure Vessel Code, Section III. The editions 
and addenda for Section III of the ASME Boiler and Pressure Vessel Code 
(excluding Nonmandatory Appendices) (referred to herein as ASME BPV 
Code) are listed in this paragraph (a)(1)(i), but limited by those 
provisions identified in paragraph (b)(1) of this section.
* * * * *
    (E) * * *
    (12) 2007 Edition,
    (13) 2008 Addenda,
    (14) 2009b Addenda (including Subsection NCA; and Division 1 
subsections NB through NH and Appendices),
    (15) 2010 Edition (including Subsection NCA; and Division 1 
subsections NB through NH and Appendices),
    (16) 2011a Addenda (including Subsection NCA; and Division 1 
subsections NB through NH and Appendices), and
    (17) 2013 Edition (including Subsection NCA; and Division 1 
subsections NB through NH and Appendices).
    (ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions 
and addenda for Section XI of the ASME BPV Code are listed in this 
paragraph (a)(1)(ii), but limited by those provisions identified in 
paragraph (b)(2) of this section.
* * * * *
    (C) * * *
    (48) 2007 Edition,
    (49) 2008 Addenda,
    (50) 2009b Addenda,
    (51) 2010 Edition,
    (52) 2011a Addenda (Excluding Article IWB-2000: IWB-2500 
``Examination and Inspection: Examination and Pressure Test 
Requirements,'' Table IWB-2500-1 ``Examination Categories,'' Item 
numbers B5.11 and B5.71), and
    (53) 2013 Edition (Excluding Article IWB-2000: IWB-2500 
``Examination and Inspection: Examination and Pressure Test 
Requirements,'' Table IWB-2500-1 (B-F) ``Examination Category B-F, 
Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles,'' Item 
numbers B5.11 and B5.71; Article IWB-3000 ``Acceptance Standards,'' 
IWB-3100 ``Evaluation of Examination Results,'' IWB-3110 ``Preservice 
Volumetric and Surface Examinations,'' IWB-3112 ``Acceptance,'' 
paragraph (a)(3); and Article IWC-3000 ``Acceptance Standards,'' IWC-
3100 ``Evaluation of Examination Results,'' IWC-3110 ``Preservice 
Volumetric and Surface Examinations,'' IWC-3112 ``Acceptance,'' 
paragraph (a)(3)).
    (iii) * * *
    (A) ASME BPV Code Case N-513-3 Mandatory Appendix I. ASME BPV Code 
Case N-513-3, ``Evaluation Criteria for Temporary Acceptance of Flaws 
in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,'' 
Mandatory Appendix I, ``Relations for Fm, Fb, and F for Through-Wall 
Flaws'' (Approval Date: January 26, 2009). ASME BPV Code Case N-513-3 
Mandatory Appendix I is referenced in paragraph (b)(2)(xxxiv)(B) of 
this section.
    (B) ASME BPV Code Case N-722-1. ASME BPV 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.
    (C) ASME BPV Code Case N-729-4. ASME BPV Code Case N-729-4, 
``Alternative Examination Requirements for PWR Reactor Vessel Upper 
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds 
Section XI, Division 1'' (Approval Date: June 22, 2012), with the 
conditions in paragraph (g)(6)(ii)(D) of this section.
    (D) ASME BPV Code Case N-770-2. ASME BPV Code Case N-770-2, 
``Alternative Examination Requirements and Acceptance Standards for 
Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS 
N06082 or UNS W86182 Weld Filler Material With or Without Application 
of Listed Mitigation Activities Section XI, Division 1'' (Approval 
Date: June 9, 2011), with the conditions in paragraph (g)(6)(ii)(F) of 
this section.
    (E) ASME BPV Code Case N-824. ASME BPV Code Case N-824, 
``Ultrasonic Examination of Cast Austenitic Piping Welds From the 
Outside Surface Section XI, Division 1'' (Approval Date: October 16, 
2012), with the conditions in paragraphs (b)(2)(xxxvii)(A) through (D) 
of this section.
    (F) ASME BPV Code Case N-852. ASME BPV Code Case N-852, 
``Application of the ASME NPT Stamp, Section III, Division 1; Section 
III, Division 2; Section III, Division 3; Section III, Division 5'' 
(Approval Date: February 9, 2015). ASME BPV Code Case N-852 is 
referenced in paragraph (b)(1)(ix) of this section.
    (G) ASME OM Code Case OMN-20. ASME OM Code Case OMN-20, ``Inservice 
Test Frequency,'' in the 2012 Edition of the ASME OM Code. OMN-20 is 
referenced in paragraph (b)(3)(x) of this section.
    (iv) ASME Operation and Maintenance Code. The editions and addenda 
for the ASME Operation and Maintenance of Nuclear Power Plants (various 
edition titles referred to herein as ASME OM Code) are listed in this 
paragraph (a)(1)(iv), but limited by those provisions identified in 
paragraph (b)(3) of this section.
* * * * *
    (B) ``Operation and Maintenance of Nuclear Power Plants, Division 
1: Section IST Rules for Inservice Testing of Light-Water Reactor Power 
Plants:''
    (1) 2009 Edition; and
    (2) 2011 Addenda.
    (C) ``Operation and Maintenance of Nuclear Power Plants, Division 
1: OM Code: Section IST:''
    (1) 2012 Edition.
    (2) [Reserved]
    (v) ASME Quality Assurance Requirements. (A) ASME NQA-1, ``Quality 
Assurance Program Requirements for Nuclear Facilities:''
    (1) NQA-1--1983 Edition;
    (2) NQA-1a--1983 Addenda;
    (3) NQA-1b--1984 Addenda;
    (4) NQA-1c--1985 Addenda;
    (5) NQA-1--1986 Edition;
    (6) NQA-1a--1986 Addenda;
    (7) NQA-1b--1987 Addenda;
    (8) NQA-1c--1988 Addenda;
    (9) NQA-1--1989 Edition;

[[Page 32979]]

    (10) NQA-1a--1989 Addenda;
    (11) NQA-1b--1991 Addenda; and
    (12) NQA-1c--1992 Addenda.
    (B) ASME NQA-1, ``Quality Assurance Requirements for Nuclear 
Facility Applications:''
    (1) NQA-1--1994 Edition;
    (2) NQA-1--2008 Edition; and
    (3) NQA-1a--2009 Addenda.
* * * * *
    (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 BPV Code and the ASME 
OM Code as specified in this paragraph (b). 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 10 
CFR part 52 is subject to the following conditions. As used in this 
section, references to Section III refer to Section III of the ASME BPV 
Code and include the 1963 Edition through 1973 Winter Addenda and the 
1974 Edition (Division 1) through the 2013 Edition (Division 1), 
subject to the following conditions:
* * * * *
    (ii) Section III condition: Weld leg dimensions. When applying the 
1989 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1) of this section, applicants and licensees 
may not apply the Section III provisions identified in Table I of this 
section for welds with leg size less than 1.09 tn:

                   Table I--Prohibited Code Provisions
------------------------------------------------------------------------
         Editions and addenda                    Code provision
------------------------------------------------------------------------
1989 Addenda through 2013 Edition.....  Subparagraph NB-3683.4(c)(1);
                                         Subparagraph NB-3683.4(c)(2).
1989 Addenda through 2003 Addenda.....  Note 11 to Figure NC-3673.2(b)-
                                         1; Note 11 to Figure ND-
                                         3673.2(b)-1.
2004 Edition through 2010 Edition.....  Note 13 to Figure NC-3673.2(b)-
                                         1; Note 13 to Figure ND-
                                         3673.2(b)-1.
2011 Addenda through 2013 Edition.....  Note 11 to Table NC-3673.2(b)-1;
                                         Note 11 to Table ND-3673.2(b)-
                                         1.
------------------------------------------------------------------------

* * * * *
    (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 
Facility Applications,'' 1994 Edition, 2008 Edition, and the 2009-1a 
Addenda specified in either NCA-4000 or NCA-7000 of that edition and 
addenda of Section III may be used by an applicant or licensee, 
provided that the administrative, quality, and technical provisions 
contained in that edition and addenda of Section III are used in 
conjunction with the applicant's or licensee's appendix B to this part 
quality assurance program; and that the applicant's or licensee's 
Section III activities comply with those commitments contained in the 
applicant's or licensee's quality assurance program description. Where 
NQA-1 and Section III do not address the commitments contained in the 
applicant's or licensee's appendix B quality assurance program 
description, those licensee commitments must be applied to Section III 
activities.
* * * * *
    (vii) Section III condition: Capacity certification and 
demonstration of function of incompressible-fluid pressure-relief 
valves. When applying the 2006 Addenda through the 2013 Edition, 
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 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.
    (viii) Section III condition: Use of ASME certification marks. When 
applying editions and addenda earlier than the 2011 Addenda to the 2010 
Edition, licensees may use either the ASME BPV Code Symbol Stamps or 
the ASME Certification Marks with the appropriate certification 
designators and class designators as specified in the 2013 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1) of this section.
    (ix) Section III Condition: NPT Code Symbol Stamps. Licensees may 
use the NPT Code Symbol Stamp with the letters arranged horizontally as 
specified in ASME BPV Code Case N-852 for the service life of a 
component that had the NPT Code Symbol Stamp applied during the time 
period from January 1, 2005, through December 31, 2015.
    (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 BPV Code, and include the 1970 Edition through the 1976 Winter 
Addenda and the 1977 Edition through the 2013 Edition, subject to the 
following conditions:
* * * * *
    (vi) Section XI condition: Effective edition and addenda of 
Subsection IWE and Subsection IWL. Licensees that implemented the 
expedited examination of containment, in accordance with Subsection IWE 
and Subsection IWL, during the period from September 9, 1996, to 
September 9, 2001, 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.
* * * * *
    (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 up to and including the 
2008 Addenda must apply paragraph (b)(2)(viii)(E) of this section. 
Applicants or licensees applying Subsection IWL, 2007 Edition with the 
2009 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section, must apply 
paragraphs (b)(2)(viii)(H) and (I) of this section.
* * * * *

[[Page 32980]]

    (H) Concrete containment examinations: Eighth provision. For each 
inaccessible area of concrete identified for evaluation under IWL-
2512(a), or identified as susceptible to deterioration under IWL-
2512(b), the licensee must provide the applicable information specified 
in paragraphs (b)(2)(viii)(E)(1), (2), and (3) of this section in the 
ISI Summary Report required by IWA-6000.
    (I) Concrete containment examinations: Ninth provision. During the 
period of extended operation of a renewed license under part 54 of this 
chapter, the licensee must perform the technical evaluation under IWL-
2512(b) of inaccessible below-grade concrete surfaces exposed to 
foundation soil, backfill, or groundwater at periodic intervals not to 
exceed 5 years. In addition, the licensee must examine representative 
samples of the exposed portions of the below-grade concrete, when such 
below-grade concrete is excavated for any reason.
    (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 (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 (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 or licensees applying Subsection IWE, 2007 Edition through 
the latest edition and 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.
* * * * *
    (D) Metal containment examinations: Fourth provision. This 
paragraph (b)(2)(ix)(D) may be used as an alternative to the 
requirements of IWE-2430. 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:
    (1) A description of each flaw or area, including the extent of 
degradation, and the conditions that led to the degradation;
    (2) The acceptability of each flaw or area and the need for 
additional examinations to verify that similar degradation does not 
exist in similar components;
    (3) A description of necessary corrective actions; and
    (4) The number and type of additional examinations to ensure 
detection of similar degradation in similar components.
* * * * *
    (x) Section XI condition: Quality assurance. When applying the 
editions and addenda later than the 1989 Edition of ASME BPV Code, 
Section XI, the edition and addenda of NQA-1, ``Quality Assurance 
Requirements for Nuclear Facility Applications,'' 1994 Edition, the 
2008 Edition, and the 2009-1a Addenda specified in either IWA-1400 or 
Table IWA 1600-1 of that edition and addenda of Section XI, may be used 
by a licensee provided that the licensee uses its appendix B to this 
part quality assurance program in conjunction with Section XI 
requirements and the commitments contained in the licensee's quality 
assurance program description. Where NQA-1 and Section XI do not 
address the commitments contained in the licensee's appendix B quality 
assurance program description, those licensee commitments must be 
applied to Section XI activities.
* * * * *
    (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 approved for use on irradiated material 
with the following conditions:
    (A) Underwater welding: First provision. Licensees must obtain NRC 
approval in accordance with paragraph (z) of this section regarding the 
welding technique to be used prior to performing welding on ferritic 
material exposed to fast neutron fluence greater than 1 x 10\17\ n/
cm\2\ (E > 1 MeV).
    (B) Underwater welding: Second provision. Licensees must obtain NRC 
approval in accordance with paragraph (z) of this section regarding the 
welding technique to be used prior to performing welding on austenitic 
material other than P-No. 8 material exposed to thermal neutron fluence 
greater than 1 x 10\17\ n/cm\2\ (E < 0.5 eV). Licensees must obtain NRC 
approval in accordance with paragraph (z) regarding the welding 
technique to be used prior to performing welding on P-No. 8 austenitic 
material exposed to thermal neutron fluence greater than 1 x 10\17\ n/
cm\2\ (E < 0.5 eV) and measured or calculated helium concentration of 
the material greater than 0.1 atomic parts per million.
* * * * *
    (xviii) * * *
    (D) NDE personnel certification: Fourth provision. The use of 
Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013 
Edition of Section XI of the ASME BPV Code is prohibited. When using 
ASME BPV Code, Section XI editions and addenda later than the 2010 
Edition, licensees and applicants must use the prerequisites for 
ultrasonic examination personnel certifications in Table VII-4110-1 and 
Subarticle VIII-2200, Appendix VIII in the 2010 Edition.
* * * * *
    (xxi) * * *
    (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 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 resolve 0.044 inch 
(1.1 mm) lower case characters without an ascender or descender (e.g., 
a, e, n, v), 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.
* * * * *
    (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 2009 Addenda, is prohibited.
* * * * *
    (xxx) [Reserved]
    (xxxi) Section XI condition: Mechanical clamping devices. When

[[Page 32981]]

installing a mechanical clamping device on an ASME BPV Code class 
piping system, Appendix W of Section XI shall be treated as a mandatory 
appendix and all of the provisions of Appendix W shall be met for the 
mechanical clamping device being installed. Additionally, use of IWA-
4131.1(c) of the 2010 Edition of Section XI and IWA-4131.1(d) of the 
2011 Addenda of the 2010 Edition and later versions of Section XI is 
prohibited on small item Class 1 piping and portions of a piping system 
that form the containment boundary.
    (xxxii) Section XI condition: Summary report submittal. When using 
ASME BPV Code, Section XI, 2010 Edition through the latest edition and 
addenda incorporated by reference in paragraph (a)(1)(ii) of this 
section, Summary Reports described in IWA-6000 must be submitted to the 
NRC as described in IWA-6240(a) and IWA-6240(b). Preservice inspection 
summary reports shall be submitted prior to the date of placement of 
the unit into commercial service and inservice inspection summary 
reports shall be submitted within 90 calendar days of the completion of 
each refueling outage.
    (xxxiii) Section XI condition: Risk-Informed allowable pressure. 
The use of Paragraph G-2216 in Appendix G in the 2011 Addenda and later 
editions and addenda of the ASME BPV Code, Section XI is prohibited.
    (xxxiv) Section XI condition: Nonmandatory Appendix U. When using 
Nonmandatory Appendix U of the 2013 Edition of the ASME BPV Code, 
Section XI the following conditions apply:
    (A) The repair or replacement activities temporarily deferred under 
the provisions of Nonmandatory Appendix U must be performed during the 
next scheduled refueling outage.
    (B) In lieu of the appendix referenced in paragraph U-S1-4.2.1(c) 
of Appendix U the mandatory appendix in ASME BPV Code Case N-513-3 must 
be used.
    (xxxv) Section XI condition: Use of RTT0 in the KIa and KIc 
equations. When using the 2013 Edition of the ASME BPV Code, Section 
XI, Appendix A, paragraph A-4200, if T0 is available, then 
RTT0 may be used in place of RTNDT for 
applications using the KIc equation and the associated 
KIc curve, but not for applications using the KIa 
equation and the associated KIa curve.
    (xxxvi) Section XI condition: Fracture toughness of irradiated 
materials. When using the 2013 Edition of the ASME BPV Code, Section 
XI, Appendix A paragraph A-4400, the licensee shall obtain NRC approval 
under paragraph (z) of this section before using irradiated 
T0 and the associated RTT0 in establishing 
fracture toughness of irradiated materials.
    (xxxvii) Section XI condition: ASME BPV Code Case N-824. Licensees 
may use the provisions of ASME BPV Code Case N-824, ``Ultrasonic 
Examination of Cast Austenitic Piping Welds From the Outside Surface 
Section XI, Division 1,'' subject to the following conditions.
    (A) Ultrasonic examinations must be spatially encoded.
    (B) Instead of Paragraph 1(c)(1)(-a), licensees shall use dual, 
transmit-receive, refracted longitudinal wave, multi-element phased 
array search units.
    (C) Instead of Paragraph 1(c)(1)(-c)(-2), licensees shall use a 
phased array search unit with a center frequency of 500 kHz with a 
tolerance of  20 percent.
    (D) Instead of Paragraph 1(c)(1)(-d), the phased array search unit 
must produce angles including, but not limited to, 30 to 55 degrees 
with a maximum increment of 5 degrees.
    (3) Conditions on ASME OM Code. As used in this section, references 
to the ASME OM Code are to the ASME OM Code, Subsections ISTA, ISTB, 
ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and 
Nonmandatory Appendices A through H and J through M, in the 1995 
Edition through the 2012 Edition, as specified in paragraph (a)(1)(iv) 
of this section. Mandatory appendices must be used if required by the 
OM Code; nonmandatory appendices are approved for use by the NRC but 
need not be used. The following conditions are applicable when 
implementing the ASME OM Code:
    (i) OM condition: Quality assurance. When applying editions and 
addenda of the ASME OM Code, the requirements of ASME Standard NQA-1, 
``Quality Assurance Requirements for Nuclear Facility Applications,'' 
1994 Edition, 2008 Edition, and 2009-1a Addenda, are acceptable as 
permitted by either ISTA 1.4 of the 1995 Edition through 1997 Addenda 
or ISTA-1500 of the 1998 Edition through the latest edition and addenda 
of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) 
of this section, provided the licensee uses its appendix B to this part 
quality assurance program in conjunction with the ASME OM Code 
requirements and the commitments contained in the licensee's quality 
assurance program description. Where NQA-1 and the ASME OM Code do not 
address the commitments contained in the licensee's appendix B quality 
assurance program description, the commitments must be applied to ASME 
OM Code activities.
    (ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees 
must comply with the provisions for testing MOVs in ASME 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 MOVs continue to be capable of performing their 
design basis safety functions. Licensees implementing ASME OM Code, 
Mandatory Appendix III, ``Preservice and Inservice Testing of Active 
Electric Motor Operated Valve Assemblies in Light-Water Reactor Power 
Plants,'' of the 2009 Edition, 2011 Addenda, and 2012 Edition shall 
comply with the following conditions:
    (A) MOV diagnostic test interval. Licensees shall evaluate the 
adequacy of the diagnostic test intervals established for MOVs within 
the scope of ASME OM Code, Appendix III, not later than 5 years or 
three refueling outages (whichever is longer) from initial 
implementation of ASME OM Code, Appendix III.
    (B) MOV testing impact on risk. Licensees shall ensure that the 
potential increase in core damage frequency and large early release 
frequency associated with the extension is acceptably small when 
extending exercise test intervals for high risk MOVs beyond a quarterly 
frequency.
    (C) MOV risk categorization. When applying Appendix III to the ASME 
OM Code, licensees shall categorize MOVs according to their safety 
significance using the methodology described in ASME OM Code Case OMN-
3, ``Requirements for Safety Significance Categorization of Components 
Using Risk Insights for Inservice Testing of LWR Power Plants,'' 
subject to the conditions applicable to OMN-3 which are set forth in 
Regulatory Guide 1.192, or using an MOV risk ranking methodology 
accepted by the NRC on a plant-specific or industry-wide basis in 
accordance with the conditions in the applicable safety evaluation.
    (D) MOV stroke time. When applying Paragraph III-3600, ``MOV 
Exercising Requirements,'' of Appendix III to the ASME OM Code, 
licensees shall verify that the stroke time of MOVs specified in plant 
technical specifications satisfies the assumptions in the plant's 
safety analyses.
    (iii) OM condition: New reactors. In addition to complying with the 
provisions in the ASME OM Code with the conditions specified in 
paragraph (b)(3) of this section, holders of

[[Page 32982]]

operating licenses for nuclear power reactors that received 
construction permits under this part on or after the date 12 months 
after August 17, 2017, and holders of combined licenses issued under 10 
CFR part 52, whose initial fuel loading occurs on or after the date 12 
months after August 17, 2017, shall also comply with the following 
conditions, as applicable:
    (A) Power-operated valves. Licensees shall periodically verify the 
capability of power-operated valves to perform their design-basis 
safety functions.
    (B) Check valves. Licensees must perform bi-directional testing of 
check valves within the IST program where practicable.
    (C) Flow-induced vibration. Licensees shall monitor flow-induced 
vibration from hydrodynamic loads and acoustic resonance during 
preservice testing or inservice testing to identify potential adverse 
flow effects on components within the scope of the IST program.
    (D) High risk non-safety systems. Licensees shall assess the 
operational readiness of pumps, valves, and dynamic restraints within 
the scope of the Regulatory Treatment of Non-Safety Systems for 
applicable reactor designs.
    (iv) OM condition: Check valves (Appendix II). Licensees applying 
Appendix II, ``Check Valve Condition Monitoring Program,'' of the ASME 
OM Code, 1995 Edition with the 1996 and 1997 Addenda, shall satisfy the 
requirements of paragraphs (b)(3)(iv)(A) through (C) of this section. 
Licensees applying Appendix II, 1998 Edition through the 2012 Edition, 
shall satisfy the requirements of paragraphs (b)(3)(iv)(A), (B), and 
(D) of this section. Appendix II of the ASME OM Code, 2003 Addenda 
through the 2012 Edition, is acceptable for use with the following 
requirements. Trending and evaluation shall support the determination 
that the valve or group of valves is capable of performing its intended 
function(s) over the entire interval. At least one of the Appendix II 
condition monitoring activities for a valve group shall be performed on 
each valve of the group at approximate equal intervals not to exceed 
the maximum interval shown in the following table:

  Table II--Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
                                           Maximum           Maximum
                                      interval between  interval between
                                        activities of     activities of
             Group size                 member valves    each valve  in
                                        in the groups       the group
                                           (years)           (years)
------------------------------------------------------------------------
>=4.................................               4.5                16
3...................................               4.5                12
2...................................                 6                12
1...................................    Not applicable                10
------------------------------------------------------------------------

* * * * *
    (vii) OM condition: Subsection ISTB. Subsection ISTB, 2011 Addenda, 
is prohibited for use.
    (viii) OM condition: Subsection ISTE. Licensees may not implement 
the risk-informed approach for inservice testing (IST) of pumps and 
valves specified in Subsection ISTE, ``Risk-Informed Inservice Testing 
of Components in Light-Water Reactor Nuclear Power Plants,'' in the 
ASME OM Code, 2009 Edition, 2011 Addenda, or 2012 Edition, without 
first obtaining NRC authorization to use Subsection ISTE as an 
alternative to the applicable IST requirements in the ASME OM Code, 
pursuant to paragraph (z) of this section.
    (ix) OM condition: Subsection ISTF. Licensees applying Subsection 
ISTF, 2012 Edition, shall satisfy the requirements of Mandatory 
Appendix V, ``Pump Periodic Verification Test Program,'' of the ASME OM 
Code, 2012 Edition. Subsection ISTF, 2011 Addenda, is prohibited for 
use.
    (x) OM condition: ASME OM Code Case OMN-20. Licensees may implement 
ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' which is 
incorporated by reference in paragraph (a)(1)(iii)(G) of this section, 
for editions and addenda of the ASME OM Code listed in paragraph 
(a)(1)(iv) of this section.
    (xi) OM condition: Valve Position Indication. When implementing 
ASME OM Code, 2012 Edition, Subsection ISTC-3700, ``Position 
Verification Testing,'' licensees shall verify that valve operation is 
accurately indicated by supplementing valve position indicating lights 
with other indications, such as flow meters or other suitable 
instrumentation, to provide assurance of proper obturator position.
    (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, as incorporated by reference in 
paragraph (a)(3)(i) of this section, without prior NRC approval, 
subject to the following conditions:
* * * * *
    (5) Conditions on inservice inspection Code Cases. Licensees may 
apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section, 
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 this section as listed in Tables 1 and 2 of NRC 
Regulatory Guide 1.147, as incorporated by reference in paragraph 
(a)(3)(ii) of this section.
    (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 NRC

[[Page 32983]]

Regulatory Guide 1.147. If a licensee has applied a listed Code Case 
that is later listed as annulled in NRC 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 ASME OM Code Cases. Licensees may apply the ASME 
OM Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by 
reference in paragraph (a)(3)(iii) of this section, 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 this section as listed in Tables 1 and 2 of NRC 
Regulatory Guide 1.192, as incorporated by reference in paragraph 
(a)(3)(iii) of this section.
    (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 NRC Regulatory Guide 1.192. If a licensee has applied a 
listed Code Case that is later listed as annulled in NRC Regulatory 
Guide 1.192, the licensee may continue to apply the Code Case to the 
end of the current 120-month interval.
* * * * *
    (f) Preservice and inservice testing requirements. Systems and 
components of boiling and pressurized water-cooled nuclear power 
reactors must meet the requirements for preservice and inservice 
testing (referred to in this paragraph (f) collectively as inservice 
testing) of the ASME BPV Code and ASME OM Code as specified in this 
paragraph (f). 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 paragraph (g) of this section.
* * * * *
    (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 BPV 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 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 or NRC Regulatory Guide 1.192, as 
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 or NRC Regulatory Guide 1.192, as incorporated by reference 
in paragraphs (a)(3)(ii) and (iii) of this section, respectively), 
subject to the applicable conditions listed therein.
* * * * *
    (3) * * *
    (iii) * * *
    (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 BPV 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 or 
NRC Regulatory Guide 1.192, as 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 BPV 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 OM Code 
Cases listed in NRC Regulatory Guide 1.192, as 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) * * *
    (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 BPV 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 BPV Code Cases listed in NRC Regulatory 
Guide 1.147, as 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 BPV 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, as 
incorporated by reference in paragraph (a)(3)(iii) of this section), 
incorporated by reference in paragraph (a)(1)(iv) of this section at 
the

[[Page 32984]]

time the construction permit, combined license, or design certification 
is issued.
* * * * *
    (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 within the scope of 
the ASME OM Code 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. The inservice test requirements for 
pumps and valves that are within the scope of the ASME OM Code but are 
not classified as ASME BPV Code Class 1, Class 2, or Class 3 may be 
satisfied as an augmented IST program in accordance with paragraph 
(f)(6)(ii) of this section without requesting relief under paragraph 
(f)(5) of this section or alternatives under paragraph (z) of this 
section. This use of an augmented IST program may be acceptable 
provided the basis for deviations from the ASME OM Code, as 
incorporated by reference in this section, demonstrates an acceptable 
level of quality and safety, or that implementing the Code provisions 
would result in hardship or unusual difficulty without a compensating 
increase in the level of quality and safety, where documented and 
available for NRC review.
    (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 ASME 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 OM Code Cases 
listed in NRC Regulatory Guide 1.192, as incorporated by reference in 
paragraph (a)(3)(iii) of this section, 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 ASME 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 or NRC Regulatory Guide 1.192 as 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.
* * * * *
    (g) Preservice and 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 paragraph (f) of this section.
* * * * *
    (2) Accessibility requirements--(i) Accessibility requirements for 
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 that are classified as ASME BPV Code Class 1 and Class 2 and 
supports for components that are classified as ASME BPV Code Class 1 
and Class 2 must be designed and be provided with the access necessary 
to perform the required preservice and inservice examinations 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 BPV Code Cases listed in NRC Regulatory Guide 1.147, 
as incorporated by reference in paragraph (a)(3)(ii) of this section) 
in effect 6 months before the date of issuance of the construction 
permit.
    (ii) Accessibility requirements for 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, components 
that are classified as ASME BPV Code Class 1, Class 2, and Class 3 and 
supports for components that are classified as ASME BPV Code Class 1, 
Class 2, and Class 3 must be designed and provided with the access 
necessary to perform the required preservice and inservice examinations 
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 BPV Code Cases listed in NRC Regulatory 
Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of 
this section) applied to the construction of the particular component.
    (iii) Accessibility requirements: Meeting later Code 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.
    (3) Preservice examination requirements--(i) Preservice examination 
requirements for 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 that are classified as ASME BPV Code Class 1 
and Class 2 and supports for components that are classified as ASME BPV 
Code Class 1 and Class 2 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 BPV Code Cases listed in 
NRC Regulatory Guide 1.147, as incorporated by reference in paragraph 
(a)(3)(ii) of this section) in effect 6 months before the date of 
issuance of the construction permit.
    (ii) Preservice examination requirements for 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, components that are classified as ASME BPV Code Class 1, Class 2, 
and Class 3 and supports for components that are classified as ASME BPV 
Code Class 1, Class 2, and Class 3 must meet the preservice examination 
requirements set forth in the editions

[[Page 32985]]

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 BPV Code Cases listed in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section) 
applied to the construction of the particular component.
* * * * *
    (v) Preservice examination requirements: Meeting later Code 
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) * * *
    (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 ASME 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, when 
using ASME BPV Code, Section XI, or NRC Regulatory Guide 1.192, when 
using the ASME OM Code, as 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. Licensees may, at 
any time in their 120-month ISI interval, elect to use the Appendix 
VIII in the latest edition and addenda of the ASME BPV Code 
incorporated by reference in paragraph (a) of this section, subject to 
any applicable conditions listed in paragraph (b) of this section. 
Licensees using this option must also use the same edition and addenda 
of Appendix I as Appendix VIII, including any applicable 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 ASME 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, when using ASME 
BPV Code, Section XI, or NRC Regulatory Guide 1.192, when using the 
ASME OM Code, as 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 August 
17, 2017, may delay the update of their Appendix VIII program by up to 
18 months after August 17, 2017. Alternatively, licensees may, at any 
time in their 120-month ISI interval, elect to use the Appendix VIII in 
the latest edition and addenda of the ASME BPV Code incorporated by 
reference in paragraph (a) of this section, subject to any applicable 
conditions listed in paragraph (b) of this section. Licensees using 
this option must also use the same Edition and Addenda of Appendix I as 
Appendix VIII, including any applicable conditions listed in paragraph 
(b) of this section.
* * * * *
    (6) * * *
    (ii) * * *
    (D) * * *
    (1) Implementation. Holders of operating licenses or combined 
licenses for pressurized-water reactors as of or after August 17, 2017 
shall implement the requirements of ASME BPV Code Case N-729-4 instead 
of ASME BPV Code Case N-729-1, subject to the conditions specified in 
paragraphs (g)(6)(ii)(D)(2) through (4) of this section, by the first 
refueling outage starting after August 17, 2017.
    (2) Appendix I use. Appendix I of ASME BPV Code Case N-729-4 shall 
not be implemented without prior NRC approval.
    (3) Bare metal visual frequency. Instead of Note 4 of ASME BPV Code 
Case N-729-4, the following shall be implemented. If effective 
degradation years (EDY) < 8 and if no flaws are found that are 
attributed to primary water stress corrosion cracking:
    (i) A bare metal visual examination is not required during 
refueling outages when a volumetric or surface examination is 
performed; and
    (ii) If a wetted surface examination has been performed of all of 
the partial penetration welds during the previous non-visual 
examination, the reexamination frequency may be extended to every third 
refueling outage or 5 calendar years, whichever is less, provided an 
IWA-2212 VT-2 visual examination of the head is performed under the 
insulation through multiple access points in outages that the VE is not 
completed. This IWA-2212 VT-2 visual examination may be performed with 
the reactor vessel depressurized.
    (4) Surface exam acceptance criteria. In addition to the 
requirements of Paragraph -3132.1(b) of ASME BPV Code Case N-729-4, a 
component whose surface examination detects rounded indications greater 
than allowed in Paragraph NB-5352 in size on the partial-penetration or 
associated fillet weld shall be classified as having an unacceptable 
indication and corrected in accordance with the provisions of 
paragraph-3132.2 of ASME BPV Code Case N-729-4.
* * * * *
    (F) * * *
    (1) Implementation. Holders of operating licenses or combined 
licenses for pressurized-water reactors as of or after August 17, 2017, 
shall implement the requirements of ASME BPV Code Case N-770-2 instead 
of ASME BPV Code Case N-770-1, subject to the conditions specified in 
paragraphs (g)(6)(ii)(F)(2) through (13) of this section, by the first 
refueling outage starting after August 17, 2017.
    (2) Categorization. Full structural weld overlays, authorized by 
the NRC staff in accordance with the alternatives approval process of 
this section, may be categorized as Inspection Items C-1 or F-1, 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 the purpose of determining ISI frequencies, all 
other butt welds that rely on Alloy 82/182 for structural integrity 
shall 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 an alternative code 
case Inspection Item is used based on conformance with an ASME 
mitigation code case endorsed in NRC Regulatory Guide 1.147 with any 
applying conditions specified in NRC Regulatory Guide 1.147, as 
incorporated by reference in paragraph (a)(3)(ii) of this section. 
Paragraph -1100(e) of ASME BPV Code Case N-770-2 shall not be used to 
exempt welds that rely on Alloy 82/182 for structural integrity from 
any requirement of paragraph (g)(6)(ii)(F) of this section.
    (3) Baseline examinations. Baseline examinations for welds in Table 
1 of ASME BPV Code Case N-770-2, Inspection Items A-1, A-2, and B, if 
not

[[Page 32986]]

previously performed or currently scheduled to be performed in an 
ongoing refueling outage as of August 17, 2017, in accordance with 
paragraph (g)(6)(ii)(F) of this section, shall be completed by the end 
of the next refueling outage. Previous examinations of these welds can 
be credited for baseline examinations only if they were performed 
within the re-inspection period for the weld item in Table 1 of ASME 
BPV Code Case N-770-2 and the examination of each weld meets the 
examination requirements of paragraphs -2500(a) or -2500(b) of ASME BPV 
Code Case N-770-2 as conditioned in this section. Other previous 
examinations that do not meet these requirements can be used to meet 
the baseline examination requirement, provided NRC approval in 
accordance with paragraph (z)(1) or (2) of this section, is granted 
prior to the end of the next refueling outage.
    (4) Examination coverage. When implementing Paragraph -2500(a) of 
ASME BPV Code Case N-770-2, essentially 100 percent of the required 
volumetric examination coverage shall be obtained, including greater 
than 90 percent of the volumetric examination coverage for 
circumferential flaws. Licensees are prohibited from using Paragraphs -
2500(c) and -2500(d) of ASME BPV Code Case N-770-2 to meet examination 
requirements.
    (5) Inlay/onlay inspection frequency. All hot-leg operating 
temperature welds in Inspection Items G, H, J, and K shall be inspected 
each inspection interval. A 25 percent sample of Inspection Items G, H, 
J, and K cold-leg operating temperature welds shall be inspected 
whenever the core barrel is removed (unless it has already been 
inspected within the past 10 years) or within 20 years, whichever is 
less.
    (6) Reporting requirements. 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) Defining ``t''. 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) Optimized weld overlay examination. Initial inservice 
examination of Inspection Item C-2 welds shall be performed between the 
third refueling outage and no later than 10 years after application of 
the overlay.
    (9) Deferral. Note (11)(b)(1) in ASME BPV Code Case N-770-2 shall 
not be used to defer the initial inservice examination of optimized 
weld overlays (i.e., Inspection Item C-2 of ASME BPV Code Case N-770-
2).
    (10) Examination technique. Note 14(b) of Table 1 and Note (b) of 
Figure 5(a) of ASME BPV Code Case N-770-2 may only be implemented if 
the requirements of Note 14(a) of Table 1 of ASME BPV Code Case N-770-2 
cannot be met.
    (11) Cast stainless steel. Examination of ASME BPV Code Class 1 
piping and vessel nozzle butt welds involving cast stainless steel 
materials, shall be performed with Appendix VIII, Supplement 9 
qualifications, or qualifications similar to Appendix VIII, Supplement 
2 or 10 using cast stainless steel mockups no later than the next 
scheduled weld examination after January 1, 2022, in accordance with 
the requirements of Paragraph -2500(a).
    (12) Stress improvement inspection coverage. Under Paragraph I.5.1, 
for cast stainless steel items, the required examination volume shall 
be examined by Appendix VIII procedures to the maximum extent practical 
including 100 percent of the susceptible material volume.
    (13) Encoded ultrasonic examination. Ultrasonic examinations of 
non-mitigated or cracked mitigated dissimilar metal butt welds in the 
reactor coolant pressure boundary must be performed in accordance with 
the requirements of Table 1 for Inspection Item A-1, A-2, B, E, F-2, J, 
and K for 100 percent of the required inspection volume using an 
encoded method.
* * * * *

    Dated at Rockville, Maryland, this 30th day of June 2017.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2017-14166 Filed 7-17-17; 8:45 am]
 BILLING CODE 7590-01-P