[Federal Register Volume 78, Number 121 (Monday, June 24, 2013)]
[Proposed Rules]
[Pages 37885-37920]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-15022]



[[Page 37885]]

Vol. 78

Monday,

No. 121

June 24, 2013

Part II





 Nuclear Regulatory Commission





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10 CFR Part 50





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

Federal Register / Vol. 78, No. 121 / Monday, June 24, 2013 / 
Proposed Rules

[[Page 37886]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[NRC-2009-0359]
RIN 3150-AI72


Approval of American Society of Mechanical Engineers' Code Cases

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to 
amend its regulations to incorporate by reference the latest revisions 
of three regulatory guides (RGs) approving new and revised Code Cases 
published by the American Society of Mechanical Engineers (ASME). This 
proposed action would allow nuclear power plant licensees, and 
applicants for construction permits (CPs), operating licenses (OLs), 
combined licenses (COLs), standard design certifications, standard 
design approvals and manufacturing licenses, to use the Code Cases 
listed in these RGs as alternatives to engineering standards for the 
construction, inservice inspection (ISI), and inservice testing (IST) 
of nuclear power plant components.
    This rulemaking also includes consideration of a petition for 
rulemaking (PRM), PRM-50-89, submitted by Mr. Raymond West. This 
rulemaking also proposes resequencing NRC's requirements governing 
Codes and standards in order to comply with the Office of the Federal 
Register's (OFR) guidelines for incorporation by reference.

DATES: Submit comments by September 9, 2013. Comments received after 
this date will be considered if it is practical to do so, but the NRC 
is able to ensure consideration only of comments received on or before 
this date.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2009-0359. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this proposed rule.
     Email comments to: Rulemaking.Comments@nrc.gov. If you do 
not receive an automatic email reply confirming receipt, contact the 
NRC directly at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal 
workdays; telephone: 301-415-1677.
    You may submit comments on the information collections by the 
methods indicated in the Paperwork Reduction Act Statement.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Manash K. Bagchi, Office of Nuclear 
Reactor Regulation, telephone: 301-415-2905; email: 
Manash.Bagchi@nrc.gov; or Wallace Norris, Office of Nuclear Regulatory 
Research, telephone: 301-251-7506; email: Wallace.Norris@nrc.gov; U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION

Executive Summary

    The NRC is proposing to amend its regulations to incorporate by 
reference the latest revisions of three NRC RGs approving new and 
revised Code Cases published by the ASME. The three RGs that would be 
incorporated by reference are RG 1.84, ``Design, Fabrication, and 
Materials Code Case Acceptability, ASME Section III,'' Revision 36, 
(DG-1230 for this proposed rule); RG 1.147, ``Inservice Inspection Code 
Case Acceptability, ASME Section XI, Division 1,'' Revision 17, (DG-
1231 for this proposed rule); and RG 1.192, ``Operation and Maintenance 
[OM] Code Case Acceptability, ASME OM Code,'' Revision 1 (DG-1232 for 
this proposed rule). This proposed action would allow nuclear power 
plant licensees, and applicants for CPs, OLs, COLs, standard design 
certifications, standard design approvals, and manufacturing licenses, 
to use the Code Cases listed in these RGs as alternatives to 
engineering standards for the construction, ISI, and IST of nuclear 
power plant components.
    This rulemaking also includes consideration of PRM-50-89, submitted 
by Mr. Raymond West, requesting that the NRC amend its regulations to 
allow consideration of alternatives to the ASME Boiler and Pressure 
Vessel [BPV] and OM Code Cases. Lastly, this rulemaking proposes 
resequencing the order of NRC's requirements, governing Codes and 
standards in order to comply with the OFR guidelines for incorporating 
by reference.

I. Accessing Information and Submitting Comments
    A. Accessing Information
    B. Submitting Comments
II. Background
III. Discussion
    A. Code Cases Approved for Unconditional Use
    B. Code Case Approved for Use With Conditions
    Section III Code Cases (DG-1230/RG 1.84)
    Section XI Code Cases (DG-1231/RG 1.147)
    OM code Cases (DG-1232/RG 1.192)
    C. NRC Proposals for Code Cases on Which the NRC Received Public 
Comments in the 2009 Proposed ASME Code Case Rulemaking
    Section III Code Cases (DG-1230/RG 1.84)
    Section XI Code Cases (DG-1231/RG 1.147)
    D. ASME Code Cases Not Approved for Use
IV. Petition for Rulemaking (PRM-50-89)
V. Changes Addressing Office of the Federal Register Guidelines on 
Incorporation by Reference
VI. Addition of Headings to Paragraphs
VII. Paragraph-by-Paragraph Discussion
VIII. Plain Writing
IX. Availability of Documents
X. Voluntary Consensus Standards
XI. Finding of No Significant Environmental Impact: Environmental 
Assessment
XII. Paperwork Reduction Act Statement
XIII. Regulatory Analysis
XIV. Regulatory Flexibility Certification
XV. Backfitting and Issue Finality

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2009-0359 when contacting the NRC 
about the availability of information for this proposed rule. You may 
access information related to this proposed rule, which the NRC 
possesses and is publicly available, by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2009-0359.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents,'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to PDR.Resource@nrc.gov. The

[[Page 37887]]

ADAMS Accession Number for each document referenced in this proposed 
rule (if that document is available in ADAMS) is provided the first 
time that a document is referenced. In addition, for the convenience of 
the reader, the ADAMS Accession Numbers are provided in a table in 
Section IX, ``Availability of Documents,'' of this document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2009-0359 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

II. Background

    The ASME develops and publishes the ASME Boiler and Pressure Vessel 
Code (BPV Code), which contains requirements for the design, 
construction, and ISI of nuclear power plant components, and the ASME 
Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 
which contains requirements for IST of nuclear power plant components. 
In response to BPV and OM Code user requests, the ASME develops ASME 
Code Cases that provide alternatives to BPV and OM Code requirements 
under special circumstances.
    The NRC approves and/or mandates the use of the ASME BPV and OM 
Code in Sec.  50.55a of Title 10 of the Code of Federal Regulations (10 
CFR) through the process of incorporation by reference. As such, each 
provision of the ASME Codes incorporated by reference into, and 
mandated by, 10 CFR 50.55a, ``Codes and standards,'' constitutes a 
legally-binding NRC requirement imposed by rule. As noted previously, 
ASME Code Cases, for the most part, represent alternative approaches 
for complying with provisions of the ASME BPV and OM Codes.
    The NRC periodically amends 10 CFR 50.55a to incorporate by 
reference NRC RGs listing approved ASME Code Cases that may be used as 
alternatives to the BPV Code and the OM Code. See Federal Register 
notice (FRN), ``Incorporation by Reference of ASME BPV and OM Code 
Cases'' (68 FR 40469; July 8, 2003).
    This rulemaking is the latest in a series of rulemakings that 
incorporate by reference new versions of several RGs identifying new 
and revised \1\ unconditionally or conditionally acceptable ASME Code 
Cases that are approved for use. In developing these RGs, the NRC staff 
reviews ASME BPV and OM Code Cases, determines the acceptability of 
each Code Case, and publishes its findings in RGs. The RGs are revised 
periodically as new Code Cases are published by the ASME. The NRC 
incorporates by reference the RGs listing acceptable and conditionally 
acceptable ASME Code Cases into 10 CFR 50.55a. Currently, NRC RG 1.84, 
Revision 35, ``Design, Fabrication, and Materials Code Case 
Acceptability, ASME Section III''; RG 1.147, Revision 16, ``Inservice 
Inspection Code Case Acceptability, ASME Section XI, Division 1''; and 
RG 1.192, Revision 0, ``Operation and Maintenance Code Case 
Acceptability, ASME OM Code,'' are incorporated into the NRC's 
regulations at 10 CFR 50.55a. A request for comment on the draft RGs is 
published elsewhere in today's Federal Register (Docket ID NRC-2009-
0359).
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    \1\ ASME Code Cases can be categorized as one of two types: new 
or revised. A new Code Case provides for a new alternative to 
specific ASME Code provisions or addresses a new need. A revised 
Code Case is a revision (modification) to an existing Code Case to 
address, for example, technological advancements in examination 
techniques or to address NRC conditions imposed in one of the 
regulatory guides that have been incorporated by reference into 10 
CFR 50.55a.
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    This rulemaking also addresses PRM-50-89 that was submitted to the 
NRC on December 14, 2007, and revised on December 19, 2007, by Mr. 
Raymond West (ADAMS Accession No. ML073600974). The petition requests 
that the NRC amend 10 CFR 50.55a to allow NRC authorization of 
alternatives to NRC-approved ASME BPV and OM Code Cases. This 
rulemaking includes proposed provisions that address the PRM. A 
detailed discussion of the PRM is provided in Section IV, ``Petition 
for Rulemaking (PRM-50-89),'' of this document.

III. Discussion

    This proposed rule would incorporate by reference the latest 
revisions of the NRC regulatory guides that list ASME BPV and OM Code 
Cases the NRC finds to be acceptable or ``conditionally acceptable'' 
(i.e., NRC-specified conditions). Draft Regulatory Guide (DG)-1230, 
Regulatory Guide 1.84, Revision 36, (ADAMS Accession No. ML102590003) 
would supersede the incorporation by reference of Revision 35; DG-1231, 
RG 1.147, Revision 17, (ADAMS Accession No. ML102590004) would 
supersede the incorporation by reference of Revision 16; and DG-1232, 
RG 1.192, Revision 1, (ADAMS Accession No. ML102600001) would supersede 
the incorporation by reference of Revision 0.
    This proposed rule addresses two categories of ASME Code Cases. The 
first category of Code Cases are the new and revised Section III and 
Section XI Code Cases listed in Supplements 1 through 10 to the 2007 
Edition of the BPV Code, and the OM Code Cases published with the 2002 
Addenda through the 2006 Addenda. The second category is the Code Cases 
that were not addressed in the final rule published on October 5, 2010 
(75 FR 61321). The 2010 final rule addressed the new and revised 
Section III and Section XI Code Cases listed in Supplements 2 through 
11 to the 2004 Edition and Supplement 0 to the 2007 Edition of BPV 
Code. Public comments were received during the proposed rule stage 
(June 2, 2009; 74 FR 26303) requesting that the NRC include certain 
revised Code Cases in the final guides that were not listed in the 
draft guides. The NRC determined that the revised Code Cases 
represented changes significant enough to warrant broader public 
participation prior to the NRC making a final determination of them. 
Accordingly, the NRC is requesting comment on these Code Cases in this 
proposed rule.
    The latest editions and addenda of the ASME BPV and OM Codes that 
the NRC has approved for use are referenced in 10 CFR 50.55a. The ASME 
also publishes Code Cases that provide alternatives to existing Code 
requirements developed and approved by the ASME. The proposed rule 
would incorporate by reference RGs 1.84, 1.147, and 1.192. The NRC, by 
incorporating by reference these three RGs, would allow nuclear power 
plant licensees and applicants for standard

[[Page 37888]]

design certifications, standard design approvals, manufacturing 
licenses, applicants for Ols, CPs, and COLs under the regulations that 
govern license certifications, to use the Code Cases listed in these 
RGs as suitable alternatives to the ASME BPV and OM Codes for the 
construction, ISI, and IST of nuclear power plant components. This 
action would be consistent with the provisions of the National 
Technology Transfer and Advancement Act of 1995, Public Law 104-113, 
which encourages Federal regulatory agencies to consider adopting 
industry consensus standards as an alternative to de novo agency 
development of standards affecting an industry. This action would also 
be consistent with the NRC policy of evaluating the latest versions of 
consensus standards in terms of their suitability for endorsement by 
regulations or regulatory guides.
    The NRC follows a three-step process to determine the acceptability 
of new and revised Code Cases and the need for regulatory positions on 
the uses of these Code Cases. This process was employed in the review 
of the Code Cases in Supplements 1 through 10 to the 2007 Edition of 
the BPV Code and the 2002 Addenda through the 2006 Addenda of the OM 
Code. The Code Cases in these supplements are the subject of this 
proposed rule. First, the ASME develops Code Cases through a consensus 
development process, as administered by the American National Standards 
Institute (ANSI), which ensures that the various technical interests 
(e.g., utility, manufacturing, insurance, regulatory) are represented 
on standards development committees and that their viewpoints are 
addressed fairly. This process includes development of a technical 
justification in support of each new or revised Code Case. The ASME 
committee meetings are open to the public, and attendees are encouraged 
to participate. Task groups, working groups, and subgroups report to a 
standards committee. The standards committee is the decisive consensus 
committee and ensures that the development process fully complies with 
the ANSI consensus process. The NRC actively participates through full 
involvement in discussions and technical debates of the task groups, 
working groups, subgroups, and standards committee regarding the 
development of new and revised standards.
    Second, the standards committee transmits to its members a first 
consideration letter ballot requesting comment or approval of new and 
revised Code Cases. To be approved, Code Cases from the first 
consideration letter ballot must receive the following: (1) Approval 
votes from at least two thirds of the eligible consensus committee 
membership, (2) no disapprovals from the standards committee, and (3) 
no substantive comments from ASME oversight committees such as the 
Technical Oversight Management Committee (TOMC). The TOMC's duties, in 
part, are to oversee various standards committees to ensure technical 
adequacy and provide recommendations in the development of codes and 
standards, as required. The Code Cases that are disapproved or receive 
substantive comments from the first consideration ballot are reviewed 
by the working level group(s) responsible for their development to 
consider the comments received. These Code Cases may be approved by the 
standards committee on second consideration with an approval vote by at 
least two thirds of the eligible consensus committee membership, with 
no more than three disapprovals from the consensus committee.
    Third, the NRC reviews new and revised Code Cases to determine 
their acceptability for incorporation by reference in 10 CFR 50.55a 
through the subject RGs. This rulemaking process, when considered 
together with the ANSI process for developing and approving ASME codes 
and standards and ASME Code Cases, constitutes the NRC's basis that the 
Code Cases (with conditions as necessary) provide reasonable assurance 
of adequate protection to public health and safety.
    The NRC reviewed the new and revised Code Cases identified in this 
proposed rule and concluded, in accordance with the process previously 
described, that the Code Cases are technically adequate (with 
conditions as necessary) and consistent with current NRC regulations. 
Thus, the new and revised Code Cases listed in the subject RGs are 
approved for use subject to any specified conditions.

A. Code Cases Approved for Unconditional Use

    The NRC determined, in accordance with the process previously 
described for review of ASME Code Cases, that each ASME Code Case 
listed in Table I is appropriate for incorporation by reference without 
conditions into the NRC's regulations.

                                 Table I
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  Code Case No.        Supplement                    Title
------------------------------------------------------------------------
  Boiler and Pressure Vessel Code Section III (Addressed in DG-1230/RG
                             1.84, Table 1)
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N-4-13..........  5 (07 Edition).....  Special Type 403 Modified
                                        Forgings or Bars Class 1 and CS,
                                        Section III, Division 1.
N-570-2.........  7 (07 Edition).....  Alternative Rules for Linear
                                        Piping and Linear Standard
                                        Supports for Classes 1, 2, 3,
                                        and MC, Section III, Division 1.
N-580-2.........  4 (07 Edition).....  Use of Alloy 600 With Columbium
                                        Added, Section III, Division 1.
N-655-1.........  2 (07 Edition).....  Use of SA-738, Grade B, for Metal
                                        Containment Vessels, Class MC,
                                        Section III, Division 1.
N-708...........  2 (07 Edition).....  Use of JIS G-4303, Grades SUS304,
                                        SUS304L, SUS316, and SUS316L,
                                        Section III, Division 1.
N-759-2.........  4 (07 Edition).....  Alternative Rules for Determining
                                        Allowable External Pressure and
                                        Comprehensive Stress for
                                        Cylinders, Cones, Spheres, and
                                        Formed Heads, Section III,
                                        Division 1.
N-760-2.........  7 (07 Edition).....  Welding of Globe Valve Disks to
                                        Valve Stem Retainers, Classes 1,
                                        2, and 3, Section III, Division
                                        1.
N-767...........  4 (07 Edition).....  Use of 21 Cr-6Ni-9Mn (Alloy UNS
                                        S21904) Grade GXM-11 (Conforming
                                        to SA[dash]182/SA-182M and SA-
                                        336/SA-336M), Grade TPXM-11
                                        (Conforming to SA[dash]312/SA-
                                        312M) and Type XM-11 (Conforming
                                        to SA-666) Material, for Class 1
                                        Construction, Section III,
                                        Division 1.
N-774...........  7 (07 Edition).....  Use of 13Cr-4Ni (Alloy UNS
                                        S41500) Grade F6NM Forgings
                                        Weighing in Excess of 10,000 lb
                                        (4,540 kg) and Otherwise
                                        conforming to the Requirements
                                        of SA-336/SA-336M for Class 1,
                                        2, and 3 Construction, Section
                                        III, Division 1.
N-782...........  9 (07 Edition).....  Use of Editions, Addenda, and
                                        Cases, Section III, Division 1.

[[Page 37889]]

 
N-801...........  4 (10 Edition).....  Rules for Repair of N-Stamped
                                        Class 1, 2, and 3 Components by
                                        Organization Other Than the N
                                        Certificate Holder That
                                        Originally Stamped the Component
                                        Being Repaired, Section III,
                                        Division 1.
N-802...........  4 (10 Edition).....  Rules for Repair of Stamped
                                        Components by the N Certificate
                                        Holder That Originally Stamped
                                        the Component, Section III,
                                        Division 1.
------------------------------------------------------------------------
   Boiler and Pressure Vessel Code Section XI (Addressed in DG-1231/RG
                             1.147, Table 1)
------------------------------------------------------------------------
N-532-5.........  5 (10 Edition )....  Alternative Requirements to
                                        Repair and Replacement
                                        Documentation Requirements and
                                        Inservice Summary Report
                                        Preparation and Submission as
                                        Required by IWA-4000 and IWA-
                                        6000, Section XI, Division 1.
N-716-1.........  1 (13 Edition).....  Alternative Piping Classification
                                        and Examination Requirements,
                                        Section XI Division 1.
N-747...........  9 (04 Edition).....  Reactor Vessel Head-to-Flange
                                        Weld Examinations Section XI,
                                        Division 1.
N-762...........  1 (07 Edition).....  Temper Bead Procedure
                                        Qualification Requirements for
                                        Repair/Replacement Activities
                                        Without Postweld Heat Treatment,
                                        Section XI, Division 1.
N-765...........  8 (07 Edition).....  Alternative to Inspection
                                        Interval Scheduling Requirements
                                        of IWA-2430, Section XI,
                                        Division 1.
N-769...........  8 (07 Edition).....  Roll Expansion of Class 1 In-Core
                                        Housing Bottom Head Penetrations
                                        in BWR's, Section XI, Division
                                        1.
N-773...........  8 (07 Edition).....  Alternatives Qualification
                                        Criteria for Eddy Current
                                        Examinations of Piping Inside
                                        Surfaces, Section XI Division 1.
------------------------------------------------------------------------
Code for Operation and Maintenance (Addressed in DG-1232/RG 1.192, Table
                                   1).
------------------------------------------------------------------------
OMN-6...........  2006 Addenda.......  Alternate Rules for Digital
                                        Instruments.
OMN-8...........  2006 Addenda.......  Alternative Rules for Preservice
                                        and Inservice Testing of Power-
                                        Operated Valves That Are Used
                                        for System Control and Have a
                                        Safety Function per OM-10.
OMN-14..........  2004 Addenda.......  Alternative Rules for Valve
                                        Testing Operations and
                                        Maintenance, Appendix I, Boiling
                                        Water Reactor (BWR) Control Rod
                                        Drive Rupture Disk Exclusion.
OMN-16..........  2006 Addenda.......  Use of a Pump Curve for Testing.
------------------------------------------------------------------------

B. Code Cases Approved for Use With Conditions

    The NRC has determined that certain Code Cases, as issued by the 
ASME, are generally acceptable for use, but that the alternative 
requirements specified in those Code Cases must be supplemented to 
provide an acceptable level of quality and safety. Accordingly, the NRC 
proposes to impose conditions on the use of these Code Cases to modify, 
limit or clarify their requirements. For each applicable Code Case, the 
conditions would specify the additional activities that must be 
performed, the limits on the activities specified in the Code Case, 
and/or the supplemental information needed to provide clarity. These 
ASME Code Cases are included in Table 2 of the following: DG-1230 (RG 
1.84), DG-1231 (RG 1.147), and DG-1232 (RG 1.192). The NRC's evaluation 
of the Code Cases and the reasons for the NRC's proposed conditions are 
discussed in the following paragraphs. Notations have been made to 
indicate the conditions duplicated from previous versions of the RGs.
    The NRC requests public comment on these Code Cases and the 
proposed conditions. It should also be noted that the following 
paragraphs only address those Code Cases for which the NRC proposes to 
impose a condition or conditions that are listed in the RG for the 
first time (e.g., the conditions on OMN-4, 2004 are identical to those 
listed in Revision 0 to RG 1.192 on OMN-4, 1999 Addenda).
Section III Code Cases (DG-1230/RG 1.84)
    NRC-proposed changes to Tables 1 and 2 of DG-1230/RG 1.84 for Code 
Cases N-520-2, N-655-1, N-757-1, N-759-2, and N-782, are discussed in 
this notice under the heading, NRC Proposals for Code Cases on which 
NRC Received Public Comments in the 2009 Proposed ASME Code Case 
Rulemaking.
Code Case N-60-5
    Type: Revised
    Title: Material for Core Support Structures, Section III, Division 
1
    Published: Supplement 12, 2001 Edition
    The NRC proposes to reinstate a condition on the use of ASME Code 
Case N-60-5, which in a previous publication was inadvertently deleted. 
Code Case N-60-5 was originally listed in RG 1.85, ``Materials Code 
Case Acceptability, ASME Section III, Division 1.'' Two conditions were 
listed in RG 1.85 for Code Case N-60-5: 1) welding of age-hardenable 
Alloy SA-453 Grade 660 and SA-637 Grade 688 should be performed when 
the material is in the solution-treated condition, and 2) the maximum 
yield strength of strain-hardened austenitic stainless steel should not 
exceed 90,000 psi in view of the susceptibility of this material to 
environmentally assisted cracking. Revision 31 of RG 1.85 was last 
published in May 1999. In June 18, 2004 (69 FR 34202), RG 1.85 was 
merged into RG 1.84. The combined RG 1.84 now lists all Section III 
Code Cases, and RG 1.85 is no longer published. When RG 1.85 was merged 
into RG.1.84, the NRC inadvertently dropped the two conditions 
applicable to Code Case N-60-5. The NRC is now proposing to reinstate 
the second of the two conditions by reinstating Code Case N-60-5 in DG-
1230/RG 1.84, Table 2, ``Conditionally Acceptable Section III Code 
Cases.''
    The NRC has determined that the first condition, regarding age-
hardenable Alloy SA-453 Grade 660 and SA-637 Grade 688, is no longer 
needed. These alloy materials are used for bolting and pins that are 
not typically subjected to welding.
    The second condition was instituted because operating experience 
and laboratory testing showed that strain hardened (also known as cold-
worked), austenitic stainless steel in excess of 90,000 psi yield 
strength, is susceptible to environmentally induced cracking. The 
caution regarding the limit on the maximum yield strength of strain-

[[Page 37890]]

hardened austenitic stainless steels has been addressed in the Standard 
Review Plan (SRP) for over 30 years and has been used as guidance by 
the NRC staff in its review of reactor coolant pressure boundary 
materials in all new reactors since the condition was inadvertently 
dropped in RG 1.84. Specifically, the limit is addressed in NUREG-0800, 
``Standard Review Plan for the Review of Safety Analysis Reports for 
Nuclear Power Plants SRP Section 4.5.1, Control Rod Drive Structural 
Materials, and Section 5.2.3, Reactor Coolant Pressure Boundary 
Materials. In Section II, SRP Acceptance Criteria state the need for 
such a limitation: ``Laboratory stress corrosion tests and service 
experience provide the basis for the criterion that cold-worked 
austenitic stainless steels used in the reactor coolant pressure 
boundary should have an upper limit on the yield strength of 620 MPa 
(90,000 psi).''
    Thus, the technical basis for the condition is well-established and 
continues to be valid because these materials are used in current 
reactor designs and may be used in future reactor designs. Accordingly, 
the NRC proposes to reinstate this condition on Code Case N-60-5 in 
Table 2 of DG-1230/RG 1.84. A licensee that implemented Code Case N-60-
5 after RG 1.84 and RG 1.85 were combined (i.e., Code Case N-60-5 
unconditionally approved) would not have to comply with the reinstated 
condition limiting the maximum yield strength. Two of the five new 
reactor designs, Economic Simplified Boiling Water Reactor [ESBWR] and 
US-EPR, specified the use of Code Case N-60-5 during the time period 
that no conditions were listed in RG 1.84. These new reactor design 
certifications were reviewed by the NRC staff for conformance with this 
condition using the guidelines of the SRP. The condition is included in 
the Design Control Document for each of these two designs. Operating 
reactor licensees, who specified Code Case N-60-5 during the time that 
it was unconditionally approved, are required to meet the ISI 
examinations in ASME Code Section XI, to ensure detection of 
environmentally assisted cracking that might result from using strain 
hardened austenitic stainless steels with yield strength in excess of 
90,000 psi.
    Reinstatement of this condition would not impact combined license 
applications that are currently under review by the NRC or have been 
approved. The condition would only apply to those applicants or 
licensees in the future that implement Code Case N-60-5 in accordance 
with Revision 36 (or later) of the final RG 1.84.
Code Case N-208-2
    Type: Revised
    Title: Fatigue Analysis for Precipitation Hardening Nickel Alloy 
Bolting Material to Specification SB-637 N07718 for Class 1 
Construction, Section III, Division 1
    Published: Supplement 4, January 4, 2008
    Figure A, ``Design Fatigue Curve for Nickel-Chromium Alloy 718,'' 
Code Case N-208-2, presents maximum mean stress curves. The upper-most 
curve is labeled ``No mean stress or [sigma]max < 100 ksi.'' 
The words ``No mean stress'' may be confusing to users and should be 
implemented with the condition that this means ``Maximum mean stress.'' 
In addition, the lower-most curve is labeled as 
``[sigma]y,'' which may also be confusing to users. The 
[sigma]y should be implemented with the condition that it 
means [sigma]max. Therefore, the NRC proposes to add two 
conditions to Code Case N-208-2 in Table 2 of DG-1230/RG 1.84 that 
would provide definitions for ``no mean stress'' and 
``[sigma]max'' with respect to Figure A.
Section XI Code Cases (DG-1231/RG 1.147)
    NRC-proposed changes to Tables 1 and 2 of DG-1231/RG 1.147 for Code 
Cases N-508-4, N-597-2, N-619, N-648-1, and N-702, are discussed in 
this notice under the heading, NRC Proposals for Code Cases on which 
NRC Received Public Comments in the 2009 Proposed ASME Code Case 
Rulemaking.
Code Case N-561-2 [Supplement 1]
    Type: Revised
    Title: Alternative Requirements for Wall Thickness Restoration of 
Class 2 and High Energy Class 3 Carbon Steel Piping, Section XI, 
Division 1
    Published: Supplement 1, 2007 Edition
    The original version and first version of this Code Case were not 
approved by the NRC for use. The NRC's basis for not approving the use 
of this Code Case was that: 1) no criteria for determining the rate or 
extent of degradation of the repair of the wall thickness restoration 
or the surrounding base metal were provided, and 2) re-inspection 
requirements were not provided to verify structural integrity since the 
root cause may not be mitigated. The ASME made significant technical 
revisions to previous versions of this Code Case by applying the 
findings from a very similar application (i.e., Code Case N-661, 
``Alternative Requirements for Wall Thickness Restoration of Class 2 
and 3 Carbon Steel Piping for Raw Water Service'').
    A request to apply Code Case N-661 at the Edwin I. Hatch Nuclear 
Power Plant (Hatch Plant) was conditionally approved by the NRC in the 
Hatch Safety Evaluation Report (SER) (ADAMS Accession No. ML033280037). 
Code Case N-661 was subsequently approved with the same conditions in 
RG 1.147, Revision 15. The ASME used these same conditions in revising 
Code Case N-561-1 resulting in Code Case N-561-2. Based on the NRC 
staff's review of Code Cases N-561-2 and N-661, and on its experience 
applying Code Case N-661 at the Hatch Plant, the NRC proposes to 
approve Code Case N-561-2 with certain conditions. This is reflected in 
Table 2 of DG-1231. Five proposed conditions on this Code Case will be 
listed in Table 2 of DG-1231/RG 1.147. The proposed conditions are 
discussed in this section.
    The provisions of Code Cases N-561-2 and N-661-1 are similar in 
that the Code Cases apply to similar systems (i.e., Class 2 and High 
Energy Class 3 Carbon Steel Piping, Class 3 Moderate Energy Carbon 
Steel Piping, and Carbon Steel Piping for Raw Water Service). The 
provisions were developed by the ASME to perform an alternative repair 
of degraded components, which involves the application of weld metal 
overlay on the exterior of the piping system to restore the wall 
thickness of the component. Accordingly, the conditions identified in 
the SER regarding Code Case N-661 also apply to Code Case N-561-2.
    One of the conditions in the SER addressed the time period for 
which the repair would be considered acceptable. The definition 
established by the NRC was modified when added to Code Case N-561-2. In 
Code Case N-661, the repair is only acceptable until the ``next 
refueling outage.'' In contrast, Code Case N-561-2 states that the 
repair would be acceptable for ``one fuel cycle.'' The NRC believes 
that it is unclear in Code Case N-561-2 what one fuel cycle actually 
infers if a repair is performed at mid-cycle.
    It could be interpreted that the repair is acceptable for the 
remainder of the current fuel cycle plus the subsequent fuel cycle. 
This interpretation could double the time period. The NRC established 
this limitation on the acceptable life of the repair of the five 
because the Code Case does not require that the root cause of the 
degradation be determined. If the root cause of the degradation has not 
been determined, a suitable reinspection frequency cannot be 
established. In addition, the Code Case would allow repairs to be made 
by welding on surfaces that are wet or exposed to water. Performing 
through-

[[Page 37891]]

wall weld repairs on surfaces that are wet or exposed to water would 
greatly increase the chances of producing welds that include weld 
defects such as porosity, lack of fusion, and cracks. It is highly 
unlikely that a weld can be made on an open root joint with water 
present on the backside of the weld without having several weld 
defects. These types of weld defects can, and many times do, lead to 
premature failure of a weld joint.
    Accordingly, the NRC is proposing on Code Case N-561-2 two of the 
five conditions (identified as Conditions 1 and 3) in the DG-1231 to 
address these concerns. The first proposed condition addresses those 
situations where welds are fabricated with water present on the 
backside, defects are likely, and the service life time would be 
expected to be greatly reduced: ``Paragraph 5(b): [of Code Case N-561-
2] for repairs performed on a wet surface, the overlay is only 
acceptable until the next refueling outage.'' A second proposed 
condition is being added on Code Case N-561-2 that would not allow the 
exemption in Paragraph (6)(c)(1). Paragraph (6)(c)(1) states that 
``Class 3 weld overlays are exempt from volumetric examination when the 
Construction Code does not require that full-penetration butt welds in 
the same location be volumetrically examined.'' Many licensees are 
mitigating stress corrosion cracking through the addition of a weld 
overlay on the outside of the piping. The purpose of the overlay is to 
restore wall thickness. The NRC has approved this mitigation technique 
provided that the full thickness of the weld overlay as well as a 
certain portion of the base material can be volumetrically examined. 
The exemption in Paragraph (6)(c)(1) conflicts with the NRC position on 
this matter, and thus the third condition is proposed requiring the 
performance of a volumetric examination of the weld overlay.
    The third proposed condition on Code Case N-561-2 is: ``Paragraph 
7(c): if the cause of the degradation has not been determined, the 
repair is only acceptable until the next refueling outage.''
    The fourth condition on Code Case N-561-2 is proposed to address 
the NRC's concern that a preexisting flaw could grow through-wall after 
application of a weld overlay: ``The area where the weld overlay is to 
be applied must be examined using ultrasonic methods to demonstrate 
that no crack-like defects exist.'' The basis for this proposed 
condition is discussed in detail here. Weld overlays have been used as 
a mitigation method and as a repair method to address stress corrosion 
cracking in piping butt welds. The basis for applying a weld overlay is 
that it will result in compressive residual stresses on the inside 
surface of the pipe, thus preventing a flaw from growing. Analytical 
modeling has been used to predict post-weld repair residual stress 
distributions for common piping configurations. Many times, however, 
weld records are not available or are not complete with regard to weld 
repairs made during construction. The investigations using modeling to 
predict the residual stresses resulting from weld repairs have used 
various assumptions to address the lack of data from weld records.
    This raises a question whether a model can accurately predict 
residual stresses if the extent of repairs is unknown. Factors such as 
the number of weld passes, welding sequence, and heat input can greatly 
influence stress patterns. Thus, analytical modeling of typical piping 
weld configurations with a weld overlay has been used to determine 
whether application of a weld overlay would result in compressive 
residual stresses and impede the growth of a preexisting flaw. Because 
of the many assumptions that might be required, configurations have 
been analyzed with up to a 75 percent through thickness flaw.
    While the results of the analyses performed have shown that a weld 
overlay could produce compressive stresses on the inside diameter of 
the piping for repairs as great as 75 percent through-wall, the NRC 
continues to be concerned regarding the lack of repair information. For 
example, an investigation into a leak that occurred several years ago 
showed that at least four weld repairs had been performed. This case is 
not believed to be unique. Thus, the NRC does not believe that the 
analyses that have been conducted to date are bounding, nor that the 
analyses have demonstrated that a preexisting flaw would not continue 
to grow circumferentially and perhaps through-wall after application of 
a weld overlay. Accordingly, the NRC proposes that it must be shown, 
using ultrasonic methods that no flaws exist in the area where the weld 
overlay is to be applied.
    The fifth and last condition being proposed on Code Case N-561-2 is 
``Paragraph 4(b): All systems must be depressurized before welding.'' 
The need for this condition is the same as that for the first proposed 
condition, i.e., the Code Case would allow repairs to be made by 
welding on surfaces that is wet or exposed to water. As previously 
discussed, it is highly unlikely that a weld can be made on an open 
root joint with water present on the backside of the weld without 
having several weld defects, and these types of weld defects can lead 
to premature failure of a weld joint. Thus, depressurizing the system 
would decrease the chances of producing a suspect weld.
Code Case N-562-2
    Type: Revised
    Title: Alternative Requirements for Wall Thickness Restoration of 
Class 3 Moderate Energy Carbon Steel Piping, Section XI, Division 1
    Published: Supplement 1, 2007 Edition
    Code Case N-562-2 is nearly identical to Code Case N-561-2, which 
is discussed separately herein. The principal difference between the 
Code Cases is that N-562-2 addresses lower energy piping. However, the 
same concerns previously discussed regarding Code Case N-561-2 also 
apply to Code Case N-562-2. Accordingly, the same five conditions are 
being proposed for Code Case N-562-2.
Code Case N-661-2
    Type: Revised
    Title: Alternative Requirements for Wall Thickness Restoration of 
Classes 2 and 3 Carbon Steel Piping for Raw Water Service, Section XI, 
Division 1
    Published: Supplement 1, 2007 Edition
    As previously discussed with respect to Code Case N-561-2, Code 
Case N-661-2 is very similar to the other two Code Cases addressing 
restoration of wall thickness (namely N-561-1 and N-562-2), except that 
N-661-2 addresses raw water service systems.
    Conditions (1) and (3) in draft Revision 17 to RG 1.147 for Code 
Case N-661-2 were listed in Revision 16 to RG 1.147. Those conditions 
are: (1) Paragraph 4(b): for repairs performed on a wet surface, the 
overlay is only acceptable until the next refueling outage; and (3) 
paragraph 7(c): if the cause of the degradation has not been 
determined, the repair is only acceptable until the next refueling 
outage. As previously indicated in the discussion addressing Code Case 
N-561-2, the ASME made significant technical revisions to Code Cases N-
561-1, N-562-1, and N-661-1. Consistent with the technical 
justification addressing the proposed conditions for Code Case N-561-2, 
the NRC is proposing three new conditions for Code Case N-661-2. Those 
conditions are listed in draft Revision 17 to RG 1.147 as following: 
(2) Paragraph 6(c)(1): this exemption is not permitted; (4) The area 
where the weld overlay is to be applied must be examined using 
ultrasonic methods to

[[Page 37892]]

demonstrate that no crack-like defects exist; and (5) All systems must 
be depressurized before welding.
Code Case N-739-1 [Supplement 1]
    Type: Revised
    Title: Alternative Qualification Requirements for Personnel 
Performing Class CC Concrete and Post-Tensioning System Visual 
Examinations, Section XI, Division 1
    Published: Supplement 1, 2007 Edition
    The original version of this Code Case was not approved by the NRC 
for use. The NRC had concerns regarding the lack of detail provided on 
the instructional material to be covered in the qualification of 
personnel performing these inspections. The revised Code Case includes 
detailed instructional material regarding requirements for training. 
The NRC finds the added requirements to be acceptable. However, the 
reference in the Code Case to the American Concrete Institute (ACI) 
standard has been printed incorrectly. To ensure that the correct 
instructional material is used, the NRC is proposing to conditionally 
approve Code Case N-739-1 to indicate that the correct ACI reference is 
201.1.
OM Code Cases (DG-1232/RG 1.192)
    Code Case OMN-1
    Type: Revised
    Title: Alternative Rules for Preservice and Inservice Testing of 
Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor 
Power Plants
    Published: 2006 Addenda
    Proposed Revision 1 to RG 1.192 does not modify the conditions 
imposed on the implementation of Code Case OMN-1 that were listed in 
Revision 0 to RG 1.192, issued June 2003. The following discussion is 
included in the proposed rule to emphasize that caution is required 
when using risk insights to evaluate the performance of MOVs that have 
exercise intervals extended from quarterly to every refueling outage.
    In 1996, ASME issued Code Case OMN-1 that allows quarterly stroke-
time testing of motor-operated valves (MOVs) in the IST program to be 
replaced by a program of exercising on a refueling outage frequency and 
periodic diagnostic testing at intervals up to 10 years. In 1999, the 
NRC accepted the use of Code Case OMN-1 with conditions in 10 CFR 
50.55a(b)(3)(iii) as an alternative to the requirement in 10 CFR 
50.55a(b)(3)(ii) that licensees shall comply with the provisions for 
MOV stroke-time testing in the OM Code and shall establish a program to 
ensure that MOVs continue to be capable of performing their design-
basis safety functions.
    In June 2003, the NRC staff developed RG 1.192 and transferred the 
acceptance of Revision 0 to Code Case OMN-1 from 10 CFR 
50.55a(b)(3)(iii) to RG 1.192 with the following conditions. Those 
conditions are:
    (1) The adequacy of the diagnostic test interval for each MOV must 
be evaluated and adjusted as necessary, but not later than 5 years or 
three refueling outages (whichever is longer) from initial 
implementation of OMN-1.
    (2) When extending exercise test intervals for high risk MOVs 
beyond a quarterly frequency, licensees must ensure that the potential 
increase in Core Damage Frequency (CDF) and risk associated with the 
extension is small and consistent with the intent of the Commission's 
Safety Goal Policy Statement.
    (3) When applying risk insights as part of the implementation of 
OMN-1, licensees must categorize MOVs according to their safety 
significance using the methodology described in Code Case OMN-3, 
``Requirements for Safety Significance Categorization of Components 
Using Risk Insights for Inservice Testing of LWR Power Plants,'' with 
the conditions discussed in RG 1.192 or use other MOV risk-ranking 
methodologies accepted by the NRC on a plant-specific or industry-wide 
basis with the conditions in the applicable safety evaluations.
    Licensees may use Code Case OMN-1 in lieu of the provisions for 
stroke-time testing in Subsection ISTC of the 1995 Edition up to and 
including the 2000 Addenda of the ASME OM Code when applied in 
conjunction with the provisions for leakage rate testing in, as 
applicable, ISTC 4.3 (1995 Edition with the 1996 and 1997 Addenda) and 
ISTC-3600 (1998 Edition through the 1999 and 2000 Addenda). In 
addition, licensees who continue to implement Section XI of the ASME 
BPV Code as their Code of Record may use OMN-1 in lieu of the 
provisions for stroke-time testing specified in Paragraph 4.2.1 of 
ASME/ANSI OM Part 10 as required by 10 CFR 50.55a(b)(2)(vii) subject to 
the conditions in this regulatory guide. Licensees who choose to apply 
OMN-1 must apply all its provisions.
    It should be noted that ASME issued Code Case OMN-11, ``Risk-
Informed Testing for Motor-Operated Valves,'' in the 2001 Edition to 
provide more specific provisions for the application of risk insights 
as part of the MOV diagnostic testing alternative allowed in Code Case 
OMN-1. The NRC accepted the use of OMN-11 in Revision 0 of RG 1.192 
with conditions related to determination of acceptable MOV test 
intervals based on diagnostic data, evaluation of test results for 
grouped low-risk MOVs, and extension of the exercise interval for high-
risk MOVs similar to the condition in RG 1.192 for Code Case OMN-1.
    In the 2006 Addenda to the ASME OM Code, ASME issued an updated 
version of Code Case OMN-1 to clarify the guidance for users of the 
code case. In its updated version, Code Case OMN-1 incorporates the 
provisions of Code Case OMN-11 for applying risk insights as well as 
the conditions specified in the June 2003 version of RG 1.192 for the 
use of Code Case OMN-11.
    The NRC staff is not proposing to modify the conditions for the 
acceptability of Code Case OMN-1 based on the incorporation of 
provisions for applying risk insights from OMN-11. However, based on 
operating experience at nuclear power plants, the NRC emphasizes the 
importance of evaluating the performance of MOVs that have exercise 
intervals extended from quarterly to every refueling outage. As 
discussed in Federal Register Notice 51370 (dated September 22, 1999) 
on page 51386, and which the NRC finds is still applicable when using 
the 2006 version of Code Case OMN-1, the licensee should 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. This information may be 
obtained by grouping similar MOVs and staggering the exercising of the 
MOVs in the group equally over the refueling interval. Licensees are 
cautioned that, when implementing OMN-1, the benefits of performing a 
particular test should be balanced against the potential adverse 
effects placed on the valves or systems caused by this testing.
    Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants 
and Fuel Reprocessing Plants,'' to 10 CFR part 50 requires nuclear 
power plant licensees to evaluate deficiencies in the performance of 
safety-related MOVs. Where degradation in the performance of a high-
risk MOV is identified when exercised or tested at an extended 
interval, licensees should reapply the quarterly frequency for the 
exercise test interval for all high-risk MOVs and implement diagnostic 
testing of those MOVs at an interval that provides assurance of their 
design-basis capability throughout the test interval. Licensees should 
also incorporate the performance results for all MOVs into the 
probabilistic risk analysis to determine whether the risk ranking of

[[Page 37893]]

MOVs should be modified based on those results.
    For additional information on OMN-1, see the discussion on OMN-4 
and OMN-12 below.
Code Case OMN-3
    Type: Revised
    Title: Requirements for Safety Significance Categorization of 
Components Using Risk Insights for Inservice Testing of LWR Power 
Plants
    Published: 2004 Edition
    The NRC initially issued RG 1.192 in June 2003 accepting several 
ASME OM Code Cases, including Code Case OMN-3. Subsequently, on 
December 18, 2003, the Commission issued Staff Requirements Memorandum 
(SRM) COMNJD-03-0002, ``Stabilizing the PRA Quality Expectations and 
Requirements'' (ADAMS Accession No. ML033520457), which approved 
implementation of a phased approach to achieving an appropriate quality 
for probabilistic risk assessments (PRAs) for the NRC's risk-informed 
decisionmaking. In SECY-04-0118 dated July 13, 2004 (ADAMS Accession 
No. ML041470505), the NRC staff described its action plan to implement 
the SRM, which the Commission subsequently approved in an SRM dated 
October 6, 2004 (ADAMS Accession No. ML042800369).
    The central concept of the action plan specifies the development of 
consensus PRA standards and associated industry guidance documents, as 
discussed in RG 1.200 (March 2009), ``An Approach for Determining the 
Technical Adequacy of Probabilistic Risk Assessment Results for Risk-
Informed Activities.'' RG 1.200 clarifies that the staff anticipates 
that current good practice, (i.e., Capability Category II (CCII)) as 
explained in the appendices of RG 1.200, is the level of technical 
adequacy that is sufficient for the majority of applications. RG 1.200 
provides that licensees evaluate all deviations from CCII or higher and 
document why the PRA is sufficient for the proposed application.
    In a related action, the Commission published Section 69, ``Risk-
Informed Categorization and Treatment of Structures, Systems and 
Components for Nuclear Power Reactors'' in 10 CFR part 50 on November 
22, 2004. RG 1.201 (May 2006), ``Guidelines for Categorizing 
Structures, Systems, and Components in Nuclear Power Plants According 
to their Safety Significance,'' describes one acceptable method to 
categorize the safety significance of active components. Section 50.69 
specifies high level treatment requirements for low risk SSCs whereas 
SSC treatment is prescribed in more detail in several risk-informed 
ASME OM Code Cases.
    Based on a consideration of the information in Section 69 in 10 CFR 
part 50 and in the RG 1.201, the NRC proposes Conditions (5), (6) and 
(7) in RG 1.192 to its acceptance of Code Case OMN-3 included in the 
2004 Edition of the ASME OM Code. Licensees applying Code Case OMN-3, 
2004 Edition, will need to apply the conditions specified in the 
previous version of RG 1.192 issued in June 2003, and new Conditions 
(5), (6) and (7) discussed in this section. As stated in RG 1.192, if a 
licensee implements a Code Case and a later version of the Code Case is 
incorporated by reference into 10 CFR 50.55a and listed in Tables 1 and 
2 during the licensee's present 120-month IST program interval, that 
licensee may use either the later version or the previous version. An 
exception to this provision would be the inclusion of a limitation or 
condition on the use of Code Case that is necessary, for example, to 
enhance safety. The NRC staff has determined that a licensee currently 
using Code Case OMN-3 must use the later version of the Code Case 
listed in Table 2 of RG 1.192, Revision 1, after it is incorporated by 
reference into 10 CFR 50.55a.
    Condition (5) specifies that the implementation of Section 3.2, 
``Plant Specific PRA,'' in Code Case OMN-3 must 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. 
Condition (5) references RG 1.200 and 1.201 for guidance in satisfying 
this condition. For example, RG 1.200 includes descriptions of 
technical adequacy of PRA analyses beyond those modeling only internal 
initiating events, (e.g., for seismic and internal fire initiating 
events). RG 1.201 endorses the guidance described by the Nuclear Energy 
Institute (NEI) in Revision 0 to NEI 00-04, ``10 CFR 50.69 SSC 
Categorization Guideline,'' dated July 2005. This document describes 
how the importance of components relied on for seismic, fires, and 
other initiating events (and operating modes) should be included in the 
categorization process, including if no plant-specific PRA is available 
for the hazard.
    Condition (6) specifies that paragraph (b) in Section 4.2.4, 
``Reconciliation,'' in Code Case OMN-3 is not endorsed. Condition (6) 
states that the expert panel may not classify components that are 
ranked as a High Safety Significant Component (HSSC) by the results of 
a qualitative or quantitative PRA evaluation (excluding the sensitivity 
studies) or the defense-in-depth assessment to a Low Safety Significant 
Component (LSSC). RG 1.201 clarifies that a component, identified as 
high safety significant by any of the PRA (excluding the sensitivity 
studies) or defense-in-depth evaluations may not be re-categorized to 
low safety significant by the expert panel. The position in RG 1.201 
that an expert panel may not decide the PRA or defense-in-depth 
evaluations are in error and lower the safety significance assigned 
according to these evaluations is applicable to OMN-3 deliberations. 
Rather, the expert panel should provide information regarding its views 
to the PRA analysts so that the evaluations can be re-performed, if 
appropriate, to address the expert panel issue or document the 
appropriateness of the current analysis results.
    Condition (7) specifies that implementation of Section 3.3, 
``Living PRA,'' in Code Case OMN-3 must be consistent with the 
following: (1) to account for potential changes in the failure rates 
and other changes that could affect the PRA, changes to the plant must 
be reviewed, and, as appropriate, the PRA updated; (2) when the PRA is 
updated, the SSC categorization must be reviewed and changed if 
necessary to remain consistent with the categorization process; and (3) 
the review of plant changes must be performed in a timely manner and 
must be performed once every two refueling outages or as required by 
50.71(h)(2) for COL holders. Changes to the plant, including potential 
changes in failure rates, might affect the PRA evaluations, and changes 
to the PRA evaluations might affect the safety significance of the 
components developed from these evaluations. Therefore, the PRA must be 
periodically updated and the risk categorization reviewed when the PRA 
is updated. The period of two refueling outages as the maximum period 
between determinations of whether a PRA update is needed is consistent 
with the time span in 10 CFR 50.69.
    Code Case OMN-3 addresses safety significance categorization of 
components using risk insights as applied to inservice testing. Several 
new conditions are proposed with respect to Code Case OMN-3 (discussed 
earlier) that reflect current NRC regulatory positions on determining 
PRA technical adequacy when using risk insights in regulatory 
applications. Code Cases OMN-1, OMN-4, 2004 Edition, and OMN-12, 2004 
Edition, also address the use of risk insights for inservice testing. 
Accordingly, to ensure consistent

[[Page 37894]]

implementation among these Code Cases, a note has been added to Code 
Case OMN-4 and OMN-12. Paragraph 3.1 of Code Case OMN-12 states that 
``Valve assemblies shall be classified as either high safety 
significant or low safety significant in accordance with Code Case OMN-
3.'' However, given the interdependence of Code Cases OMN-1, OMN-3, 
OMN-4, and OMN-12, Note 3 has been added to Code Case OMN-12 as a 
reminder of the dependence on Code Case OMN-3 (i.e., paragraph 3.1). In 
addition, Note 2 has been added to Code Case OMN-4 as a reminder that 
the conditions with respect to allowable methodologies for OMN-3 risk 
ranking specified for the use of OMN-1 also apply to OMN-4.

C. NRC Proposals for Code Cases on Which NRC Received Public Comments 
in the 2009 Proposed ASME Code Case Rulemaking

    On June 2, 2009, the NRC published a proposed rule (74 FR 26303) 
and a parallel notice of availability of draft RGs (74 FR 26440) 
seeking public comments on incorporating by reference draft RG 1.84, 
Revision 35, and draft RG 1.147, Revision 16. The NRC received public 
comments on draft Revision 35 to RG 1.84 and draft Revision 16 to RG 
1.147 requesting that certain revised Code Cases that were not listed 
in those draft guides be approved in the final guides. These revised 
Code Cases that were the subject of comment in 2009 are N-520-2, N-655-
1, N-757-1, N-759-2, and N-782 for RG 1.84; and Code Cases N-508-4, N-
597-2, N-619, N-648-1, and N-702 for RG 1.147. In that earlier 
rulemaking, the NRC determined that the revised Code Cases represented 
changes significant enough to warrant broader public participation 
prior to the NRC making a final determination of them. Therefore, the 
final RG 1.84 and RG 1.147 associated with the 2010 final rule (75 FR 
61321; October 5, 2010) did not include these Code Cases.
    The NRC has reviewed these Code Cases, and now proposes to approve 
those Code Cases, in some cases with conditions. These Code Cases are 
discussed in this section, under the applicable draft regulatory guide.
Section III Code Cases (DG-1230/RG 1.84)
Code Case N-520-2
    Type: Revised
    Title: Alternative Rules for Renewal of Active or Expired N-type 
Certificates for Plants Not in Active Construction
    Published: Supplement 4, 2007 Edition
    Code Case N-520-1, the predecessor of Code Case N-520-2, was 
unconditionally approved in Revision 34 to RG 1.84. The objective of 
Code Case N-520-1 was to address situations where construction was 
halted on a nuclear power plant, interrupting ASME Code activities, but 
the Certificate Holder had maintained its certificate. Code Case N-520-
1 provided guidance on what a Certificate Holder had to do to document 
and stamp the completed construction work. On June 2, 2009, the NRC 
published a proposed rule (74 FR 26303) and a parallel notice of 
availability of draft RGs (74 FR 26440) seeking public comment on draft 
RG 1.84, Revision 35. The NRC received a public comment requesting that 
the NRC approve Code Case N-520-2 for inclusion in final Revision 35, 
noting that Code Case N-520-2 had been approved by the ASME on November 
1, 2007, and published in Supplement 4 to the 2007 Edition. Code Case 
N-520-2 was developed to allow an organization with an expired 
certificate to secure an ASME Temporary Certificate of Authorization. 
Because Code Case N-520-2 was not part of the June 2009 proposed rule 
and the changes reflected in N-520-2 were significant, the NRC did not 
adopt the public comment to list Code Case N-520-2 in final Revision 35 
to RG 1.84 (incorporated by reference in the final rule published on 
October 5, 2010 (75 FR 61321)).
    The NRC has now determined that the provisions of Code Case N-520-2 
are adequate for addressing a situation where a Certificate Holder has 
let its N-type certificates expire. The basis for this determination is 
that all completed in-process work must be clearly documented to ensure 
that remaining activities and Code responsibilities are readily 
identifiable. In addition, the ASME Temporary Certificate of 
Authorization is for the sole purpose of completing the required 
documentation and component stamping. Finally, this work must be 
completed under a contract with an Authorized Nuclear Inspection Agency 
(ANIA).
    The NRC is proposing to conditionally approve Code Case N-520-2 
because it believes that the wording of the Code Case may create 
confusion regarding the relationship between the ANIA and the 
Authorized Nuclear Inspector (ANI). The purpose of the condition in 
Table 2 of DG1230/RG 1.84, Revision 36, is to clearly indicate that the 
ANIA employs the ANI.
Code Cases N-655-1, N-757-1, N-759-2, N-782
    A comment responding to the June 2, 2009, proposed rule (74 FR 
26303) and a parallel notice of availability of draft RGs (74 FR 
26440), requested that the following four Code Cases used in the AP-
1000 design that were not included in draft Revision 35 of RG 1.84 be 
included in the final guide: Code Case N-655-1, ``Use of SA-738, Grade 
B, for Metal Containment Vessels, Class MC, Section III, Division 1;'' 
Code Case N-757-1, ``Alternative Rules for Acceptability for Class 2 
and 3 Valves, NPS 1 (DN25) and Smaller with Welded and Nonwelded End 
Connections other than Flanges, Section III, Division 1;'' Code Case N-
759-2, ``Alternative Rules for Determining Allowable External Pressure 
and Compressive Stresses for Cylinders, Cones, Spheres, and Formed 
Heads, Section III, Division 1;'' and Code Case N-782, ``Use of Code 
Editions, Addenda, and Cases Section III, Division 1.'' Draft Revision 
35 of RG 1.84 considered Code Cases published up to Supplement 0 to the 
2007 Edition. Code Cases N-655-1 and N-757-1 were published in 
Supplement 2 to the 2007 Edition. Code Case N-759-2 was published in 
Supplement 4 to the 2007 Edition. Code Case N-782 was published in 
Supplement 9 to the 2007 Edition. These four Code Cases were beyond the 
scope of the draft RG and thus had not been considered for inclusion in 
the draft RG.
    The NRC did not include these four Code Cases in final Revision 35 
of RG 1.84 because it would have been inappropriate to include them in 
the final RG without providing the public an opportunity for comment. 
In addition, these Code Cases were not referenced in the latest AP-1000 
Design Control Document.
    Code Cases N-655-1, N-759-2, and N-782 have been reviewed by the 
NRC and have been found to be acceptable. Accordingly, these Code Cases 
are listed in Table 1 of DG-1230/RG 1.84, Revision 36, and the NRC 
proposes to unconditionally approve them, as presented in Table I under 
``Code Cases Approved for ``Unconditional Use''.
    Code Case N-757-1 was reviewed and found to be conditionally 
acceptable. It is listed in Table 2 of the DG-1230/RG 1.84. The 
proposed condition for Code Case N-757-1 is discussed in the following 
discussion.
Code Case N-757-1 [Supplement 2]
    Type: Revised
    Title: Alternative Rules for Acceptability for Class 2 and 3 Valves 
NPS 1 (DN 25) and Smaller with Welded and Nonwelded End Connections 
Other than Flanges, Section III, Division 1
    Published: Supplement 2, 2007 Edition
    The NRC proposes to impose a condition on Code Case N-757-1 in 
Table 2 of RG 1.84 to prohibit the use

[[Page 37895]]

of the design provisions in ASME Section III, Division 1, Appendix 
XIII, for Class 3 valves. This would be accomplished by adding the 
condition to Table 2 of DG-1230/RG 1.84. The Code Case addresses the 
use of instrument, control, and sampling line valves, NPS 1 (DN 25) and 
smaller, with nonwelded end connections other than ASME B16.5 flanges 
for Section III, Division 1, Class 2 and Class 3 construction. The Code 
Case provides three options for the design of Class 2 and Class 3 
valves that do not meet the minimum thickness requirements in ASME 
B16.34. These options include the following: 1) the pressure design 
rules of Section III, paragraphs NC-3324 and ND-3324; 2) the 
experimental stress analysis rules in Section III, Appendix II; or 3) 
design based on the stress analysis rules in Section III, Appendix 
XIII.
    The NRC finds that the first option provides an acceptable 
alternative basis for the design of ASME Class 2 and Class 3 valves 
because it provides adequate design margin by using the vessel design 
rules accepted by the NRC in 10 CFR 50.55a. The second option is also 
acceptable for the design of ASME Class 2 and Class 3 valves because it 
allows the designer to use experimental stress analysis techniques to 
establish that the design provides acceptable ASME Code margins for 
parts in which theoretical stress analysis might not be possible or 
practical. The third option, however, is not acceptable to the NRC.
    Option 3 would allow a designer to use the criteria provided in 
Section III, Division 1, Appendix XIII. As defined by the scope of 
Appendix XIII, these Code rules are only applicable to the design of 
Class 2 vessels meeting the requirements of NC-3200. Further, Appendix 
XIII provides for design based on a stress analysis that uses criteria 
similar to that used for the design of ASME Class 1 components 
(including the ASME Class 1 stress intensity allowable limits). The 
stress intensity values in the acceptance criteria are greater than the 
allowable stress intensity values specified for the design of ASME 
Class 3 components. The NRC concludes that the criteria in Appendix 
XIII are not intended for the design of ASME Class 3 components, 
including the valves within the scope of N-757, and that a condition 
should be added to Table 2 of DG-1230/RG 1.84 that prohibits the use of 
these design provisions for Class 3 valves.
    It should be noted that the NRC staff approved this Code Case as it 
was considered by the cognizant ASME committees. However, upon further 
consideration as Code Cases were reviewed for inclusion in the subject 
RGs, the NRC determined that use of the Code Case was inappropriate for 
ASME Class 3 components. Therefore, the NRC proposes to impose a 
condition that would prohibit the use of the design provisions in ASME 
Section III, Division 1, Appendix XIII, for Class 3 valves.
Section XI Code Cases (DG-1231/RG 1.147)
Code Case N-508-4
    Type: New
    Title: Rotation of Snubbers and Pressure Retaining Items for the 
Purpose of Testing or Preventive Maintenance, Section XI, Division 1
    Published: Supplement 8, 2007 Edition
    Code Case N-508-3, the predecessor of Code Case N-508-4, was 
unconditionally approved in Revision 15 to RG 1.147. The objective of 
Code Case N-508-3 was to provide guidance on rotating snubbers and 
relief valves from stock for the purpose of testing or preventive 
maintenance. On June 2, 2009, the NRC published a proposed rule (74 FR 
26303) and a parallel notice of availability of draft RGs (74 FR 26440) 
seeking public comment on draft RG 1.147, Revision 16. The NRC received 
a public comment noting that Code Case N-508-4 had been approved by the 
ASME on January 26, 2009, and published in Supplement 8 to the 2007 
Edition, and requesting that the NRC approve Code Case N-508-4 in final 
Revision 16 rather than cease approval at Code Case N-508-3. Code Case 
N-508-4 significantly expands the list of components that may be 
rotated from stock for the purpose of testing or preventive maintenance 
(adds pumps, control rod drive mechanisms, and pump seal packages).
    Because Code Case N-508-4 was not part of the June 2009 proposed 
rule and the changes reflected in N-508-4 were significant, the NRC did 
not adopt the public comment to list Code Case N-508-4 in final 
Revision 16 to RG 1.147 (incorporated by reference in the final rule 
published on October 5, 2010 (75 FR 61321)). Instead, this Code Case is 
addressed in draft Revision 17 to RG 1.147.
    The NRC has not identified any technical reasons why additional 
components may be considered for the purpose of testing or preventive 
maintenance as described in the Code Case N-508-4. However, the NRC has 
identified an issue and proposes to condition Code Case N-508-4 to 
ensure that there is no conflict regarding the application of this Code 
Case. When Section XI is used to govern snubber examination and 
testing, Footnote 1 (which was later added to the Code Case) conflicts 
with Subsection IWF, Section XI, up to and including the 2004 Edition 
through the 2005 Addenda. Footnote 1 directs the user to implement the 
OM Code for snubber examination and testing. The OM Code was developed 
in order to have a separate Code for the development and maintenance of 
provisions for the IST of pumps and valves. In 1990, the ASME published 
the initial edition of the OM Code, thereby transferring responsibility 
for these provisions from Section XI to the OM Code Committee. While 
the use of the OM Code is an option under paragraph (b)(3)(v)(A), the 
examination and testing requirements for snubbers are also provided in 
the 2005 Addenda and earlier editions and addenda of Section XI. Thus, 
there is a conflict for editions and addenda up to the 2005 Addenda of 
Section XI, but there is no conflict for licensees who have adopted the 
2006 Addenda or later editions and addenda of Section XI.
    To resolve the conflict, the NRC is proposing to include in DG-
1231/RG 1.147, Revision 17, a condition to Code Case N-508-4 stating 
that Footnote 1 to the Code Case would not apply when the ISI Code of 
record is earlier than Section XI, 2006 Addenda, and Section XI 
requirements are used to govern the examination and testing of 
snubbers.

Code Case N-597-2

    Type: Revised
    Title: Requirements for Analytical Evaluation of Pipe Wall Thinning
    Listed: Revision 15 to RG 1.147
    Published: November 18, 2003
    Code Case N-597-2 was conditionally approved in Revision 15 to RG 
1.147. Two comments responding to the proposed rule published on June 
2, 2009 (74 FR 26303), and a parallel notice of availability of draft 
RGs (74 FR 26440) seeking public comment on draft RG 1.147, Revision 
16, suggested that the method in Code Case N-513-2, ``Evaluation 
Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 
or 3 Piping,'' used to evaluate local degradation, should be approved 
by the NRC for application to Code Case, N-597-2. The comments argued 
that the NRC has conditionally approved Code Case N-513-2 with an 
evaluation methodology to allow licensees to temporarily accept flaws 
in moderate energy Class 2 or 3 piping, whereas condition (2) on Code 
Case N-597-2 requires NRC approval for any amount of local degradation 
beyond that calculated by the hoop stress equation.

[[Page 37896]]

    Because Code Case N-513-2 was not part of the June 2, 2009, 
proposed rule and the changes reflected in N-513-2 are significant, the 
NRC did not adopt the public comments to allow the Code Case N-513-2 
evaluation to also be used with respect to Code Case N-597-2. While the 
NRC agrees that the flaw evaluation methodology for analyzing piping 
degradation contained in Code Case N-513-2 could under certain 
circumstances be applied for a Code Case N-597-2 evaluation (i.e., both 
Code Cases address the analytical evaluation of pipe wall thinning), 
the NRC disagrees with the comments that through-wall leakage should be 
included in the scope of such an evaluation. Code Case N-597-2 was not 
developed to address leakage; it is focused only on analytical 
evaluation of wall thinning. The comments discuss local degradation up 
to and including through-wall leakage and believe it would be 
appropriate to allow such leakage for all ASME Code class components. 
This implies that such leakage from high temperature, high pressure 
systems is no different from leakage from low temperature, low pressure 
systems. Permitting degradation up to and including through-wall 
leakage in certain systems would violate 10 CFR part 50, appendix A, 
Criterion General Design Criteria (GDC) 14, ``Reactor coolant pressure 
boundary,'' and/or similar provisions in the licensing basis for these 
facilities, which require that the reactor coolant pressure boundary be 
tested to ensure an extremely low probability of abnormal leakage, of 
propagating failure, and of gross rupture. In addition, there have been 
pipe breaks and leakage in high temperature, high pressure lines 
throughout the world and some have been sudden and catastrophic. Code 
Case N-597-2 is applicable to all ASME Code class piping, including 
high energy piping; whereas, Code Case N-513-2 is limited to Class 2 
and 3 moderate energy piping. The NRC has only approved temporary 
acceptance of flaws for moderate energy Class 2 or 3 piping (maximum 
operating temperature does not exceed 200[emsp14][deg]F (93 [deg]C) and 
maximum operating pressure does not exceed 275 psig (1.9 MPa)). The 
comments' requested change would redefine the defense-in-depth concept. 
Rather than performing inspections to detect flaws before structural 
integrity is compromised, degradation would be managed in effect after 
leakage is discovered.
    The NRC agrees, however, that it should be permissible under 
certain circumstances for licensees to evaluate local thinning using 
the acceptance criteria of the Code Case without NRC review and 
acceptance. Thus, a sixth condition is being proposed for Code Case N-
597-2 in DG-1231/RG 1.147, Revision 17. The condition would propose 
that, on moderate-energy Class 2 and 3 piping, wall thinning acceptance 
criteria may be used on a temporary basis based on the provisions of 
Code Case N-513-2, and that Code Case N-597-2 cannot be used to 
evaluate through-wall leakage conditions.
Code Cases N-619
    Type: Conditionally approved
    Title: Alternative Requirements for Nozzle Inner Radius Inspections 
for Class 1 Pressurizer and Steam Generator Nozzles Published
    Published: April 8, 2002
Code Case N-648-1
    Type: Conditionally approved for the first time
    Title: Alternative Requirements for Inner Radius Examination of 
Class 1 Reactor Pressure Vessel Nozzles
    Published: September 18, 2001
    A comment on the proposed rule published on June 2, 2009 (74 FR 
26303), and a parallel notice of availability of draft RGs (74 FR 
26440) seeking public comment on draft RG 1.147, Revision 16, requested 
that the NRC reconsider the conditions placed on Code Case N-619, 
``Alternative Requirements for Nozzle Inner Radius Inspections for 
Class 1 Pressurizer and Steam Generator Nozzles,'' and Code Case N-648-
1, ``Alternative Requirements for Inner Radius Examination of Class 1 
Reactor Pressure Vessel Nozzles.'' The comment states that the 
conditions on the two Code Cases requiring a wire standard to 
demonstrate the resolution capability of remote visual examination 
systems should be changed to the ASME 0.044-inch characters because 
those characters have been recognized to be a better resolution 
standard than the wire standard.
    Because Code Case N-619 and Code Case N-648-1 were not part of the 
June 2, 2009, proposed rule and the changes reflected in N-619 and N-
648-1 are significant; the NRC did not adopt the public comment to use 
characters rather than the wire standard.
    The NRC is addressing the comment as part of this rulemaking. The 
NRC agrees with the 2009 comment that characters have been demonstrated 
to be a better resolution standard than the wire standard. However, the 
NRC believes that the shift to characters should be part of broader 
changes to the visual testing standards. Visual examinations are used 
in certain situations as alternatives to volumetric and/or surface 
examination tests where it is not possible to conduct volumetric 
examination (e.g., where there are limitations due to access or 
geometry) or to reduce occupational exposure in high radiation fields. 
Visual testing experts had believed that if the camera and lighting 
were sufficient to see a 12 [mu]m (0.0005 in.) diameter wire, then the 
camera system had a resolution sufficiently high for the inspection. 
Subsequent investigation of the effectiveness and reliability of visual 
examinations has shown that the wire resolution standard is not 
sufficient to determine the visual acuity of a remote system (i.e., 
there are important differences between visually detecting a wire and a 
crack). Research conducted at the Pacific Northwest National Laboratory 
showed that other calibration standards should be adapted for visual 
testing such as reading charts and resolution targets. Results 
supporting this recommendation were published in NUREG/CR-6943, ``A 
Study of Remote Visual Methods to Detect Cracking in Reactor 
Components'' (ADAMS Accession No. ML073110060). As also discussed in 
the report, other parameters such as crack size, lighting conditions, 
camera resolution, and surface conditions were assessed. The NRC 
concluded from the investigation that a significant fraction of the 
cracks that have been reported in nuclear power plant components are at 
the lower end of the capabilities of the visual testing equipment 
currently being used. Code Case N-619 addresses the examination of the 
nozzle inner radius of Class 1 pressurizers and steam generators. Code 
Case N-648-1 provides an alternative for examining the inner radius of 
Class 1 reactor vessel nozzles. The NRC investigation of crack opening 
dimensions of service-induced cracks in nuclear components included 
thermal fatigue, mechanical fatigue, and stress corrosion cracks. The 
NRC concluded that current visual testing systems may not reliably 
detect a significant number of these cracks, and the research results 
showed that detection of these cracks under field conditions is 
strongly dependent on camera magnification, lighting, inspector 
training, and inspector vigilance. While this research supports the use 
of characters in lieu of a wire standard, the research also showed that 
other changes should be considered to visual testing as related to 
these two Code Cases. The NRC and the Electric Power Research Institute 
(EPRI) are currently conducting a collaborative

[[Page 37897]]

research project investigating these parameters. The results of the 
collaborative research will be assessed by the NRC and the industry to 
determine what changes should be made to visual testing requirements in 
the future.
    The comment also indicated that it is unclear how allowable flaw 
lengths would be determined from Table IWB-3512-1. The condition on the 
two Code Cases states that ``licensees may perform a visual examination 
with enhanced magnification that has a resolution sensitivity to detect 
a 1-mil width wire or crack, utilizing the allowable flaw length 
criteria of Table IWB-3512-1 with limiting assumptions on the flaw 
aspect ratio.'' Table IWB-3512-1 does not specifically provide 
allowable flaw length criteria. The commenter recommended that the 
acceptance criteria be modified as following: ``Crack-like surface 
flaws exceeding the acceptance criteria of Table IWB-3510-3 are 
unacceptable for continued service unless the vessel meets the 
requirements of IWB-2142.2, IWB-3142.3, or IWB-3142.4. The component 
thickness, t, to be applied in calculating the allowable surface flaw, 
I, in Table IWB-3510-3 shall be selected as specified in Table IWB-
3512-2.''
    The NRC does not agree with the suggestion. Table IWB-3512-1 was 
selected because it is the only table that considers the inside corner 
region. In determining an acceptable flaw size, the limiting aspect 
ratio is assumed, which is 0.5. The surface flaw allowable size divided 
by the limiting aspect ratio yields the limiting surface flaw size in 
terms of the l/t. In the case of wall thickness sizes provided in Table 
IWB-3512-1, the acceptance criteria are the same as those in Table IWB-
3510-3. The NRC does not intend to make any changes to the table 
referred to for acceptance criteria, because Table IWB-3512-1 is the 
only table to refer to the inside corner region.
    Finally, the commenter believes that the condition on Code Case N-
648-1 describing the surfaces to be examined is unnecessary because the 
Code Case describes the same examination surfaces. The NRC agrees and 
proposes to eliminate this condition in Table 2 of DG-1231/RG 1.147, 
Revision 17.
Code Case N-702
    Type: New
    Title: Alternative Requirements for Boiling Water Reactor (BWR) 
Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division 1
    Published: Supplement 12, 2001 Edition
    Two comments on the proposed rule published on June 2, 2009 (74 FR 
26303), and a parallel notice of availability of draft RGs (74 FR 
26440) seeking public comment on draft RG 1.147, Revision 16, requested 
that Code Case N-702, ``Alternative Requirements for Boiling Water 
Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section 
XI, Division 1,'' be conditionally approved in the final guide. Code 
Case N-702 had been listed in draft RG 1.193, Revision 3, ``ASME Code 
Cases Not Approved for Use,'' because at the time that draft Revision 
16 to RG 1.147 was published (October 2007), the NRC staff was 
considering the industry response to the NRC staff's request for 
additional information relative to the acceptability of ``BWRVIP-108: 
BWR Vessel and Internals Project (VIP), Technical Basis for the 
Reduction of Inspection Requirements for the Boiling Water Reactor 
Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,'' EPRI Technical 
Report 1003557, October 2002 (ADAMS Accession No. ML023330203). BWRVIP-
108 provides the technical basis supporting Code Case N-702. 
Subsequently, the NRC conditionally approved a licensee's request to 
use the Code Case on the basis of the NRC's Safety Evaluation (ADAMS 
Accession No. ML073600374; December 18, 2007).
    The Safety Evaluation discussed the NRC's review of BWRVIP-108 and 
the conditions under which it could be used. The commenters believed 
that the conditions in the Safety Evaluation provided a basis for the 
NRC to conditionally approve Code Case N-702 in final RG 1.147, 
Revision 16. The NRC did not adopt the public comment to approve the 
Code Case in final Revision 16 to RG 1.147. Code Case N-702 is an 
alternative to provisions in the ASME Code to reduce the inspection 
requirements of BWR reactor vessel nozzle-to-shell welds and nozzle 
blend radii. BWRVIP-108 discusses the probabilistic fracture mechanics 
evaluation that was performed to demonstrate that the probability of 
failure considering these inspection changes meets NRC requirements. 
While the NRC believes that the Safety Evaluation and BWRVIP report 
provide a basis for conditionally approving the Code Case on a generic 
basis, the NRC did not believe that it would have been appropriate to 
move the Code Case from RG 1.193 to RG 1.147 without first having 
sought public comment. Thus, the NRC is proposing to conditionally 
approve Code Case N-702 in DG-1231/RG 1.147, Revision 17, based on the 
conditions that were discussed in the Safety Evaluation. The 
applicability of Code Case N-702 must be shown by demonstrating that 
the criteria in Section 5.0 of the NRC Safety Evaluation regarding 
BWRVIP-108 dated December 18, 2007, are met. The evaluation 
demonstrating the applicability of the Code Case must be reviewed and 
approved by the NRC prior to the application of the Code Case.
Code Case N-747
    Type: New
    Title: Reactor Vessel Head-to Flange Weld Examinations, Section XI, 
Division 1
    Published: Supplement 9, 2004 Edition
    A comment on the proposed rule published on June 2, 2009 (74 FR 
26303), and a parallel notice of availability of draft RGs (74 FR 
26440) seeking public comment on draft RG 1.147, Revision 16, suggested 
that the basis for listing Code Case N-747, ``Reactor Vessel Head-to 
Flange Weld Examinations, Section XI, Division 1,'' in draft RG 1.193 
was flawed, and that the Code Case should be unconditionally accepted 
in final Revision 16. Additional technical information to support 
approval of the Code Case was provided in the comment letter (ADAMS 
Accession No. ML092190138). The NRC did not adopt the public comment to 
list Code Case N-747 in final Revision 16 to RG 1.147 (incorporated by 
reference in the final rule published on October 5, 2010, (75 FR 
61321)), because the NRC determined that the public should have an 
opportunity to comment on the additional information that was submitted 
by the commenter.
    The NRC has reviewed the information provided in the comment, which 
deals with the expected fluence levels of reactor vessel head-to-flange 
welds. Based on this information, the NRC believes that an adequate 
technical basis has been provided to support a conclusion that the 
fracture toughness will remain high. The key points discussed in the 
additional information are that calculations show that the fluence in 
the upper head region will be low, even after 60 years of service. 
Therefore, there will be no irradiation induced change in 
RTNDT. In addition, the industry has calculated 
RTNDT for the upper head region for early Westinghouse plant 
designs using the Standard Review Plan (NUREG-0800) and determined that 
the fracture toughness is high. Therefore, the NRC proposes to 
unconditionally approve

[[Page 37898]]

Code Case N-747 in Table 1 of DG-1231/RG 1.147, Revision 17.

D. ASME Code Cases Not Approved for Use

    The ASME Code Cases that are currently issued by the ASME but not 
approved for generic use by the NRC are listed in RG 1.193, ``ASME Code 
Cases Not Approved for Use.'' In addition to ASME Code Cases that the 
NRC has found to be technically or programmatically unacceptable, RG 
1.193 includes Code Cases on reactor designs for high-temperature gas-
cooled reactors and liquid metal reactors, reactor designs not 
currently licensed by the NRC, and certain requirements in Section III, 
Division 2, for submerged spent fuel waste casks, that are not endorsed 
by the NRC. Regulatory Guide 1.193 complements RGs 1.84, 1.147, and 
1.192. It should be noted that RG 1.193 is not part of this rulemaking 
because the NRC is not proposing to adopt any of the Code Cases listed 
in that RG. Comments have been submitted in the past, however, on 
certain Code Cases listed in RG 1.193 where the commenter believed that 
additional technical information was available that might not have been 
considered by the NRC in its determination not to approve the use of 
these Code Cases. While the NRC will consider those comments, any 
changes in the NRC's non-approval of such Code Cases will be the 
subject of an additional opportunity for public comment.

IV. Petition for Rulemaking (PRM-50-89)

    On December 14, 2007, Mr. Raymond West (the petitioner) submitted a 
PRM requesting the NRC to amend 10 CFR 50.55a to allow consideration of 
alternatives to the ASME BPV and OM Code Cases. The petitioner 
submitted an amended petition on December 19, 2007 (ADAMS Accession No. 
ML073600974). The petition was docketed by the NRC as PRM-50-89. The 
petitioner requested that the regulations be amended to provide 
applicants and licensees a process for requesting NRC approval of 
changes or modifications to ASME Code Cases that are listed in the 
relevant NRC-approved RGs cited in the current regulations. The 
petitioner stated that the current requirements do not allow changes or 
modifications to be proposed as alternatives to NRC-approved ASME Code 
Cases, and asserted that such changes or modifications should be 
allowed as alternatives to NRC Code Cases. Overall, the petitioner 
requested that the regulations be amended to allow applicants and 
licensees to request authorization of NRC-approved Code Cases with 
proposed modifications directly through Sec.  50.55a(a)(3).
    The NRC believes that Code Cases often provide alternatives that 
have technical merit and, in many instances, are incorporated into 
future ASME Code editions. The ASME Code Case process itself 
constitutes a method of how an applicant or licensee can seek to obtain 
ASME approval for a variation of a previously-approved Code provision. 
Section 50.55a(a)(3) currently provides specific approaches for 
obtaining NRC authorization of alternatives to ASME Code provisions. 
Inasmuch as ASME Code Cases are analogous to ASME Code provisions, it 
is not unreasonable to provide an analogous regulatory approach for 
obtaining NRC authorization of alternatives to ASME Code Cases. For 
these reasons, the NRC determined that the issues raised in this PRM 
should be considered in the NRC's rulemaking process, and the NRC 
published a FRN with this determination on April 22, 2009 (74 FR 
18303). Accordingly, the NRC is addressing PRM-50-89 in this proposed 
rule.
    On the basis of the previous discussion, the NRC is proposing to 
include language in proposed 10 CFR 50.55a(z) (existing 50.55a(a)(3)) 
that would allow applicants and licensees to request authorization of 
alternatives for changes to conditions on NRC-approved ASME Code Cases 
in current paragraphs (b)(4), (b)(5), and (b)(6) of Sec.  50.55a. In 
addition, the NRC proposes extending the scope of the petitioner's 
request for allowing alternatives to NRC-approved Code Case conditions 
to allow applicants and licensees to request authorization of 
alternatives for changes to conditions on Section III and XI of the 
ASME BPV Code and OM Code in current paragraphs (b)(1), (b)(2), and 
(b)(3).

V. Changes Addressing Office of the Federal Register Guidelines on 
Incorporation by Reference

    This proposed rule includes changes to 10 CFR 50.54, 50.55, and 
50.55a. These changes were made in accordance with the guidance for 
incorporation by reference of multiple standards that is included in 
Chapter 6 of the OFR's ``Federal Register Document Drafting Handbook,'' 
January 2011 Revision. This latest revision of the OFR's guidance 
provides several options for incorporating by reference multiple 
standards into regulations.
    The NRC proposes to incorporate by reference, in a single 
paragraph, the multiple standards mentioned in 10 CFR 50.55a. For the 
least disruption to the existing structure of the section, the NRC 
proposes to incorporate by reference the multiple standards into 10 CFR 
50.55a(a), the first paragraph of the section. Each national consensus 
standard that is being incorporated by reference in 10 CFR 50.55a has 
been listed separately. Accordingly, the regulatory language of 10 CFR 
50.54, 50.55, and 50.55a has been reorganized by moving existing 
paragraphs, creating new paragraphs, and revising introductory and 
regulatory texts.
    The NRC has made conforming changes to references throughout 10 CFR 
50.55a to reflect this reorganization. A detailed discussion of the 
affected paragraphs, other than the aforementioned reference changes, 
is provided in Section VII, ``Paragraph-by-Paragraph Discussion,'' of 
this document. The regulatory text of 10 CFR 50.55a has been set out in 
its entirety for the convenience of the reader. The staff has also 
developed reader aids to help users understand these changes (see 
Section VI of this document.)

VI. Addition of Headings to Paragraph

    The NRC is proposing to add headings (explanatory titles) to 
paragraphs and all lower-level subparagraphs of 10 CFR 50.55a. These 
headings are intended to enhance the readers' ability to identify the 
paragraphs (e.g., paragraphs (a), (b), (c)) and subparagraphs with the 
same subject matter. The NRC's proposal addresses longstanding 
complaints by external and internal stakeholders on the readability and 
complex structure of the requirements in 10 CFR 50.55a. To address this 
concern, the NRC evaluated a range of solutions, including the creation 
of new regulations and relocation of existing requirements from 10 CFR 
50.55a to the new regulations.
    Some alternatives the NRC considered were a new regulation adjacent 
to 10 CFR 50.55a (e.g., Sec. Sec.  50.55b, 50.55c, 50.55d), a new 
subpart containing a new series of regulations at the end of 10 CFR 
part 50 (e.g., subpart B beginning at Sec.  50.200, and continuing with 
Sec. Sec.  50.201, 50.202, 50.203), or a new part (designated for codes 
and standards) containing a new series of regulations addressing codes 
and standards approved for incorporation by reference by the OFR. The 
relocation of each existing requirement to a new regulation (or set of 
regulations) would follow a set of organizing principles established by 
the NRC after consideration of stakeholder's views.
    Upon consideration of these alternatives, the NRC decided that 
these alternatives should not be adopted--at least not at this time 
without further stakeholder input--and instead that the

[[Page 37899]]

NRC should develop and adopt headings for paragraphs and subparagraphs. 
The primary reason for the NRC's decision is external stakeholders' 
objections to a previous attempt by the NRC to re-designate paragraphs 
in Sec.  50.55a (75 FR 24324; May 4, 2010). As the NRC understands it, 
many nuclear power plant licensees' procedures reference specific 
paragraphs and subparagraphs of Sec.  50.55a. It would require 
substantial rewriting of these procedures and documents to correct the 
references to the old (superseded) section, paragraphs and 
subparagraphs. In addition, currently-approved design certification 
rules may require conforming amendments to be made to correct 
references to ASME Code provisions on design (and possibly ISI and 
IST).
    The NRC requests public comment on whether the NRC should adopt one 
of these approaches, either as a follow-on activity to the addition of 
headings, or as a substitute for the addition of headings. The most 
helpful comments would identify a specific approach, and set forth the 
reasons why the proposed approach should be adopted, taking into 
account the factors considered by the NRC in selecting the headings 
approach.

NRC's Proposal: Convention for Headings and Subheadings

    The NRC is proposing to add headings to all first, second, third, 
fourth, and some fifth-level paragraphs for certain sections of 10 CFR 
50.55a to add clarity and a user-friendly method for following sublevel 
contents within a regulation. The proposed heading for a fourth-level 
would follow the same convention, but may designate the provision 
number only. Fifth-level paragraphs are proposed for only newly 
incorporated Code Cases. Each first-level paragraph (designated using 
letters, (e.g., (a), (b), (c))) would have a heading that concisely 
describes the general subject matter addressed in that paragraph. Each 
second-level paragraph (designated using numbers, e.g., (1), (2), (3)) 
would have a heading comprised of a summary of the first-level 
paragraph's heading and a semicolon (``;''), followed by a concise 
description of the subject matter addressed in the second paragraph. 
The proposed heading for a third-level paragraph would follow the same 
convention (i.e., a heading comprised of a summary level of the higher-
level paragraph's title and a semicolon, followed by a concise 
description of the subject matter addressed in that subparagraph). The 
proposed heading for a fourth-level paragraph would follow the same 
convention, but may designate the provision number only. The proposed 
fifth-level paragraph is applied to only paragraph (a) for 
incorporation by reference of approved editions and addenda to the ASME 
BPV and OM Codes.

Reader Aids

    The staff has developed a table showing the proposed structure of 
10 CFR 50.55a. This table, ``Proposed Reorganization of Paragraphs and 
Subparagraphs in 10 CFR 50.55a, Codes and standards'' (ADAMS Accession 
No. ML12289A121) is available in a separate document and outlines the 
section showing all paragraph designations, including the new paragraph 
headings. The staff has also developed Cross-Reference tables showing 
the current designations for 10 CFR 50.54, 50.55, and 50.55a 
regulations and the proposed designations for these sections. These 
tables contain the new headings and a description of each change and 
are available in a separate document (ADAMS Accession No. ML12289A114).

VII. Paragraph-by-Paragraph Discussion

Section 50.54

    In Sec.  50.54, the introductory statement would be revised to 
include a reference to Sec.  50.55a. This revision would clarify that 
nuclear power plant licensees, as described in the introductory 
paragraph of Sec.  50.54, also are subject to the applicable 
requirements delineated in Sec.  50.55a. In addition, the NRC proposes 
to revise the introductory text of this section, add and reserve 
paragraph (ii), and add paragraph (jj) to include a condition of every 
license. This requirement is currently contained in Sec.  50.55a(a)(1), 
and no change to the requirement is intended by the transfer of this 
requirement from Sec.  50.55a(a)(1) to Sec.  50.54(jj).

Section 50.55

    In Sec.  50.55, the introductory text would be revised to include 
references to existing Sec.  50.55a, and paragraphs (g) and (h) would 
be added and reserved for future use. Further, existing Sec.  
50.55a(a)(1) would be moved to a newly created Sec.  50.55(i).

Section 50.55a

    In Sec.  50.55a, the current introductory statement would be 
relocated to Sec.  50.54(jj), 50.55(i), and 50.55a.

Paragraph (a)

    A new paragraph (a) would be created in Sec.  50.55a to incorporate 
by reference the multiple standards currently identified in existing 
Sec.  50.55a. The heading would be revised to read ``Documents approved 
for incorporation by reference.''
    Paragraph (a)(1): This paragraph ``American Society of Mechanical 
Engineers (ASME)'' would be added to group all ASME Sections.
    Paragraph (a)(1)(i): This paragraph, ``ASME Boiler and Pressure 
Vessel Code, Section III,'' would be added to discuss the availability 
of standards referenced in current paragraph (b)(1). This change would 
bring the NRC's requirements into compliance with the OFR's revised 
guidelines for incorporating by reference consensus standards in 
regulations.
    Paragraph (a)(1)(i)(A): This paragraph, ``Rules for Construction of 
Nuclear Vessels,'' would be added to group all the individual standards 
referenced regarding the subject matter included in current paragraph 
(b)(1). This change would bring the NRC's requirements into compliance 
with the OFR's revised guidelines for incorporating by reference 
consensus standards in regulations.
    Paragraph (a)(1)(i)(B): This paragraph, ``Rules for Construction of 
Nuclear Power Plant Components,'' would be added to group all the 
individual standards referenced regarding the subject matter included 
in current paragraph (b)(1). This change would bring the NRC's 
requirements into compliance with the OFR's revised guidelines for 
incorporating by reference consensus standards in regulations.
    Paragraph (a)(1)(i)(C): This paragraph, ``Division I Rules for 
Construction of Nuclear Power Plant Components,'' would be added to 
group all the individual standards referenced regarding the subject 
matter included in current paragraph (b)(1). This change would bring 
the NRC's requirements into compliance with the OFR's revised 
guidelines for incorporating by reference consensus standards in 
regulations.
    Paragraph (a)(1)(i)(D): This paragraph, ``Rules for Construction of 
Nuclear Power Plant Components--Division 1,'' would be added to group 
all the individual standards referenced regarding the subject matter 
included in current paragraph (b)(1). This change would bring the NRC's 
requirements into compliance with the OFR's revised guidelines for 
incorporating by reference consensus standards in regulations.
    Paragraph (a)(1)(i)(E): This paragraph, ``Rules for Construction of 
Nuclear Facility Components--Division 1,'' would be added to group all 
the individual standards referenced regarding the subject matter 
included in

[[Page 37900]]

current paragraph (b)(1). This change would bring the NRC's 
requirements into compliance with the OFR's revised guidelines for 
incorporating by reference consensus standards in regulations.
    Paragraph (a)(1)(ii)(A): This paragraph, ``Rules for Inservice 
Inspection of Nuclear Reactor Coolant Systems,'' would be added to 
discuss the availability of individual standards referenced regarding 
the subject matter included in current paragraph (b)(2). This change 
would bring the NRC's requirements into compliance with the OFR's 
revised guidelines for incorporating by reference consensus standards 
in regulations.
    Paragraph (a)(1)(ii)(B): This paragraph, ``Rules for Inservice 
Inspection of Nuclear Power Plant Components,'' would be added to 
discuss the availability of individual standards referenced regarding 
the subject matter included in current paragraph (b)(2). This change 
would bring the NRC's requirements into compliance with the OFR's 
revised guidelines for incorporating by reference consensus standards 
in regulations.
    Paragraph (a)(1)(ii)(C): This paragraph, ``Rules for Inservice 
Inspection of Nuclear Power Plant Components--Division 1,'' would be 
added to discuss the availability of individual standards referenced 
regarding the subject matter included in current paragraph (b)(2). This 
change would bring the NRC's requirements into compliance with the 
OFR's revised guidelines for incorporating by reference consensus 
standards in regulations.
    Paragraph (a)(1)(iii): This paragraph, ``ASME Code Cases: Nuclear 
Components,'' would be added to discuss the newly approved Code Cases 
referenced regarding the subject matter in current paragraph (b). This 
change would bring the NRC's requirements into compliance with the 
OFR's revised guidelines for incorporating by reference consensus 
standards in regulations.
    Paragraph (a)(1)(iii)(A): This paragraph, ``ASME Code Case N-722-
1,'' would be added to discuss the newly approved Code Case referenced 
regarding the subject matter in current paragraph (b). This change 
would bring the NRC's requirements into compliance with the OFR's 
revised guidelines for incorporating by reference consensus standards 
in regulations.
    Paragraph (a)(1)(iii)(B): This paragraph, ``ASME Code Case N-729-
1,'' would be added to discuss the newly approved Code Case referenced 
regarding the subject matter in current paragraph (b). This change 
would bring the NRC's requirements into compliance with the OFR's 
revised guidelines for incorporating by reference consensus standards 
in regulations.
    Paragraph (a)(1)(iii)(C): This paragraph, ``ASME Code Case N-770-
1,'' would be added to discuss the newly approved Code Case referenced 
regarding the subject matter in current paragraph (b). This change 
would bring the NRC's requirements into compliance with the OFR's 
revised guidelines for incorporating by reference consensus standards 
in regulations.
    Paragraph (a)(1)(iv): This paragraph, ``ASME Operation and 
Maintenance Code,'' would be added to group all the individual 
standards referenced in current paragraph (b). This change would bring 
the NRC's requirements into compliance with the OFR's revised 
guidelines for incorporating by reference consensus standards in 
regulations.
    Paragraph (a)(1)(iv)(A): This paragraph, ``Code for Operation and 
Maintenance of Nuclear Power Plants,'' would be added to group all the 
individual standards referenced in current paragraph (b).
    Paragraph (a)(1)(iv)(B): This paragraph would be added and reserved 
for future use.
    Paragraph (a)(2): This paragraph, ``Institute of Electrical and 
Electronics Engineers (IEEE) Service Center,'' would be added to list 
all IEEE sections.
    Paragraph (a)(2)(i): This paragraph, ``IEEE Standard 279-1971,'' 
would be added to discuss the availability of standards referenced in 
current paragraph (h)(2). This would be done in compliance with OFR 
revised guidelines for incorporation by reference standards in 
regulations.
    Paragraph (a)(2)(ii): This paragraph, ``IEEE Standard 603-1991,'' 
would be added to discuss the availability of the standard referenced 
in current paragraph (h)(2) and (h)(3). This would be done in 
compliance with OFR revised guidelines for incorporation by reference 
standards in regulations.
    Paragraph (a)(2)(iii): This paragraph, ``IEEE Standard 603-1991 
correction sheet,'' would be added to discuss the availability of the 
standard referenced in current paragraphs (h)(2) and (h)(3). This would 
be done in compliance with OFR revised guidelines for incorporation by 
reference standards in regulations.
    Paragraph (a)(3): This paragraph, ``U.S. Nuclear Regulatory 
Commission (NRC) Reproduction and Distribution Services Section,'' 
lists all regulatory guides being incorporated by reference. This would 
be done in compliance with OFR revised guidelines for incorporation by 
reference standards in regulations.
    Paragraph (a)(3)(i): This paragraph, ``NRC Regulatory Guide 1.84, 
Revision 36,'' would be added to discuss the availability of the 
standard. This would be done in compliance with OFR revised guidelines 
for incorporation by reference standards in regulations.
    Paragraph (a)(3)(ii): This paragraph, ``NRC Regulatory Guide 1.147, 
Revision 17,'' would be added to discuss the availability of the 
standard. This would be done in compliance with OFR revised guidelines 
for incorporation by reference standards in regulations.
    Paragraph (a)(3)(iii): This paragraph, ``NRC Regulatory Guide 
1.192, Revision 1,'' would be added to discuss the availability of the 
standard. This would be done in compliance with OFR revised guidelines 
for incorporation by reference standards in regulations.
    Paragraph (b): The paragraph heading would be revised to ``Use and 
conditions on the use of standards.'' The contents would be moved, in 
part, to 50.55a(a) for compliance with OFR revised guidelines for 
incorporation by reference standards in regulations.
    Paragraph (c): Introductory text would be added to the existing 
paragraph (c). Explanatory headings would be added for subparagraphs.
    Paragraph (d): The new paragraph would add introductory text to 
``Quality Group B components,'' as part of the NRC initiative of adding 
headings and providing clarity. Explanatory headings would be added for 
subparagraphs.
    Paragraph (e): The new paragraph would add introductory text to 
``Quality Group C components,'' as part of the NRC initiative of adding 
headings and providing clarity. Explanatory headings would be added for 
subparagraphs.
    Paragraph (f): Introductory text would be revised and expanded in 
``Inservice testing requirements,'' as part of the NRC initiative of 
adding headings and providing clarity. Explanatory headings would be 
added for subparagraphs.
    Paragraph (g): Introductory text would be revised and expanded in 
``Inservice inspection requirements,'' as part of the NRC initiative of 
adding headings and providing clarity. Explanatory headings would be 
added for subparagraphs.
    Paragraphs (b)(5), (f)(2), (f)(3)(iii)(A), (f)(3)(iv)(A), 
(f)(4)(ii), (g)(2), (g)(3)(i), (g)(3)(ii), (g)(4)(i), and (g)(4)(ii): 
References to the revision number for RG 1.147 would be changed from 
``Revision 16'' to ``Revision 17.''
    Paragraph (h)(1): This paragraph would be designated as reserved 
because the informational content from

[[Page 37901]]

current (h)(1) would be moved to proposed paragraph (a)(2).
    Paragraphs (i)-(y): The paragraphs would be added and reserved for 
future use.
    Paragraph (z): This paragraph would be added to contain information 
that would be relocated from the introductory text of current paragraph 
(a)(3) and current subparagraphs (a)(3)(i)-(ii) as a result of the 
NRC's compliance with the OFR's revised guidelines for incorporating by 
reference. Paragraph (z) would also be revised to allow applicants and 
licensees to request alternatives to the requirements in paragraph (b) 
of this section.
Overall Considerations on the Use of ASME Code Cases
    This rulemaking would amend 10 CFR 50.55a to incorporate by 
reference RG 1.84, Revision 36, which would supersede Revision 35; RG 
1.147, Revision 17, which would supersede Revision 16; and RG 1.192, 
Revision 1, which would supersede Revision 0. The following general 
guidance applies to the use of the ASME Code Cases approved in the 
latest versions of the RGs that are incorporated by reference into 10 
CFR 50.55a as part of this rulemaking.
    The approval of a Code Case in the NRC RGs constitutes acceptance 
of its technical position for applications that are not precluded by 
regulatory or other requirements or by the recommendations in these or 
other RGs. The applicant and/or licensee are responsible for ensuring 
that use of the Code Case does not conflict with regulatory 
requirements or licensee commitments. The Code Cases listed in the RGs 
are acceptable for use within the limits specified in the Code Cases. 
If the RG states an NRC condition on the use of a Code Case, then the 
NRC condition supplements and does not supersede any condition(s) 
specified in the Code Case, unless otherwise stated in the NRC 
condition.
    The ASME Code Cases may be revised for many reasons, (e.g., to 
incorporate operational examination and testing experience and to 
update material requirements based on research results). On occasion, 
an inaccuracy in an equation is discovered or an examination, as 
practiced, is found not to be adequate to detect a newly discovered 
degradation mechanism. Hence, when an applicant or a licensee initially 
implements a Code Case, 10 CFR 50.55a requires that the applicant or 
the licensee implement the most recent version of that Code Case as 
listed in the RGs incorporated by reference. Code Cases superseded by 
revision are no longer acceptable for new applications unless otherwise 
indicated.
    Section III of the ASME BPV Code applies only to new construction 
(i.e., the edition and addenda to be used in the construction of a 
plant are selected based on the date of the construction permit and are 
not changed thereafter, except voluntarily by the applicant or the 
licensee). Hence, if a Section III Code Case is implemented by an 
applicant or a licensee and a later version of the Code Case is 
incorporated by reference into 10 CFR 50.55a and listed in the RGs, the 
applicant or the licensee may use either version of the Code Case 
(subject, however, to whatever change requirements apply to its 
licensing basis, (e.g., 10 CFR 50.59)).
    A licensee's ISI and IST programs must be updated every 10 years to 
the latest edition and addenda of Section XI and the OM Code, 
respectively, that were incorporated by reference into 10 CFR 50.55a 
and in effect 12 months prior to the start of the next inspection and 
testing interval. Licensees who were using a Code Case prior to the 
effective date of its revision may continue to use the previous version 
for the remainder of the 120-month ISI or IST interval. This relieves 
licensees of the burden of having to update their ISI or IST program 
each time a Code Case is revised by the ASME and approved for use by 
the NRC. Code Cases apply to specific editions and addenda, and Code 
Cases may be revised if they are no longer accurate or adequate, so 
licensees choosing to continue using a Code Case during the subsequent 
ISI or IST interval must implement the latest version incorporated by 
reference into 10 CFR 50.55a and listed in the RGs.
    The ASME may annul Code Cases that are no longer required, are 
determined to be inaccurate or inadequate, or have been incorporated 
into the BPV or OM Codes. If an applicant or a licensee applied a Code 
Case before it was listed as annulled, the applicant or the licensee 
may continue to use the Code Case until the applicant or the licensee 
updates its construction Code of Record (in the case of an applicant, 
updates its application) or until the licensee's 120-month ISI or IST 
update interval expires, after which the continued use of the Code Case 
is prohibited unless NRC authorization is given under the current 10 
CFR 50.55a(a)(3). If a Code Case is incorporated by reference into 10 
CFR 50.55a and later annulled by the ASME because experience has shown 
that the design analysis, construction method, examination method, or 
testing method is inadequate; the NRC will amend 10 CFR 50.55a and the 
relevant RG to remove the approval of the annulled Code Case. 
Applicants and licensees should not begin to implement such annulled 
Code Cases in advance of the rulemaking.
    A Code Case may be revised, for example, to incorporate user 
experience. The older or superseded version of the Code Case cannot be 
applied by the licensee or applicant for the first time.
    If an applicant or a licensee applied a Code Case before it was 
listed as superseded, the applicant or the licensee may continue to use 
the Code Case until the applicant or the licensee updates its 
construction Code of Record (in the case of an applicant, updates its 
application) or until the licensee's 120-month ISI or IST update 
interval expires, after which the continued use of the Code Case is 
prohibited unless NRC authorization is given under proposed 10 CFR 
50.55a(z). If a Code Case is incorporated by reference into 10 CFR 
50.55a and later a revised version is issued by the ASME because 
experience has shown that the design analysis, construction method, 
examination method, or testing method is inadequate; the NRC will amend 
10 CFR 50.55a and the relevant RG to remove the approval of the 
superseded Code Case. Applicants and licensees should not begin to 
implement such superseded Code Cases in advance of the rulemaking.

VIII. Plain Writing

    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). The NRC 
requests comment on the proposed rule with respect to the clarity and 
effectiveness of the language used.

IX. Availability of Documents

    The NRC is making the documents identified in Table II 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.

[[Page 37902]]



                                                    Table II
----------------------------------------------------------------------------------------------------------------
                    Document                         PDR         WEB                   NRC Library.
----------------------------------------------------------------------------------------------------------------
Proposed Rule--Regulatory Analysis.............          X           X   ML103060189.
Proposed Rule Federal Register Notice..........          X           X   ML103060003.
Proposed Reorganization of Paragraphs and                X           X   ML12289A121.
 Subparagraphs.
Cross-Reference Tables.........................          X           X   ML12289A114.
RG 1.84, Revision 36 (DG-1230).................          X           X   ML102590003.
RG 1.147, Revision 17 (DG-1231)................          X           X   ML102590004.
RG 1.192, Revision 1 (DG-1232).................          X           X   ML102600001.
RG 1.200, Revision 2, An Approach for                    X           X   ML090410014.
 Determining the Technical Adequacy of
 Probabilistic Risk Assessment Results for Risk-
 informed Activities.
RG 1.201, Revision 1, Guidelines for                     X           X   ML061090627.
 Categorizing Structures, Systems, and
 Components in Nuclear Power Plants According
 to Their Safety Significance.
2007/12/19--Petition for Rulemaking PRM-50-89            X           X   ML073600974.
 submitted by Ray West regarding, ``To Amend
 CFR 5-.55a--Codes and Standards--Revision 1''.
Hatch Plant Report--Hatch, Units 1 & 2, Farley,          X           X   ML033280037.
 Units 1 & 2, Vogtle, Units 1 & 2, Safety
 Evaluation Re. Request to Use ASME Code Case N-
 661.
EPRI Technical Report--Project No. 704--BWRVIP-          X           X   ML023330203.
 108: BWR Vessel & Internals Project, Technical
 Basis for Reduction of Inspection Requirements
 for Boiling Water Reactor Nozzle-to-Vessel
 Shell Welds & Nozzle Blend Radii.
Safety Evaluation of Proprietary EPRI Report--           X           X   ML073600374.
 BWR Vessel and Internals Project, Technical
 Basis for the Reduction of Inspection
 Requirements for the Boiling Water Reactor
 Nozzle-to-Vessel Shell Welds and Nozzle Inner
 Radius (BWRVIP-108).
Comment Letter--Comment (4) of Bryan A. Erler            X           X   ML092190138.
 on Behalf of ASME Supporting Draft Regulatory
 Guides DG-1191, DG-1192, DG-1193, and the
 Proposed Rule Incorporating the Final
 Revisions of these Regulatory Guides into 10
 CFR 50.55a.
SRM-COMNJD-03-0002--Stabilizing the PRA Quality          X           X   ML033520457.
 Expectations and Requirements.
SECY-04-0118--Plan for the Implementation of             X           X   ML041470505.
 the Commission's Phased Approach to
 Probabilistic Risk Assessment Quality.
SRM-SECY-04-0118--Plan for the Implementation            X           X   ML042800369.
 of the Commission's Phased Approach to
 Probabilistic Risk Assessment Quality.
NUREG-0800--Chapter 4, Section 4.5.1, Revision           X           X   ML070230007.
 3, Control Rod Drive Structural Materials,
 dated March 2007.
NUREG-0800--Chapter 5, Section 5.2.3, Revision           X           X   ML063190006.
 3, Reactor Coolant Pressure Boundary
 Materials, dated March 2007.
NUREG/CR-6943--A Study of Remote Visual Methods          X           X   ML073110060.
 to Detect Cracking in Reactor Components.
----------------------------------------------------------------------------------------------------------------

X. Voluntary Consensus Standards

    Section 12(d)(3) of the National Technology Transfer and 
Advancement Act (NTTAA) of 1995, Public Law 104-113, and implementing 
guidance in U.S. Office of Management and Budget (OMB) Circular A-119 
(February 10, 1998), require each Federal government agency (should it 
decide that regulation is necessary) to use a voluntary consensus 
standard instead of developing a government-unique standard. An 
exception to using a voluntary consensus standard is allowed where the 
use of such a standard is inconsistent with applicable law or is 
otherwise impractical. The NTTAA requires Federal agencies to use 
industry consensus standards to the extent practical; it does not 
require Federal agencies to endorse a standard in its entirety. Neither 
the NTTAA nor OMB Circular A-119 prohibit an agency from adopting a 
voluntary consensus standard while taking exception to specific 
portions of the standard, if those provisions are deemed to be 
``inconsistent with applicable law or otherwise impractical.'' 
Furthermore, taking specific exceptions furthers the Congressional 
intent of Federal reliance on voluntary consensus standards because it 
allows the adoption of substantial portions of consensus standards 
without the need to reject the standards in their entirety because of 
limited provisions that are not acceptable to the agency.
    In this rulemaking, the NRC is continuing its existing practice of 
approving the use of ASME BPV and OM Code Cases, which are ASME-
approved alternatives to compliance with various provisions of the ASME 
BPV and OM Code. The NRC's approval of the ASME Code Cases is 
accomplished by amending the NRC's regulations to incorporate by 
reference the latest revisions of the following, which are the subject 
of this rulemaking, into 10 CFR 50.55a: RG 1.84, ``Design, Fabrication, 
and Materials Code Case Acceptability, ASME Section III,'' Revision 36; 
RG 1.147, ``Inservice Inspection Code Case Acceptability, ASME Section 
XI, Division 1,'' Revision 17; and RG 1.192, ``Operation and 
Maintenance Code Case Acceptability, ASME Code,'' Revision 1. These RGs 
list the ASME Code Cases that the NRC has approved for use. The ASME 
Code Cases are national consensus standards as defined in the NTTAA and 
OMB Circular A-119. The ASME Code Cases constitute voluntary consensus 
standards, in which all interested parties (including the NRC and 
licensees of nuclear power plants) participate. Therefore, the NRC's 
approval of the use of the ASME Code Cases identified in RGs 1.84, 
Revision 36; RG 1.147, Revision 17; and RG 1.192, Revision 1, which are 
the subject of this rulemaking, is consistent with the overall 
objectives of the NTTAA and OMB Circular A-119.
    The NRC reviews each Section III, Section XI, and OM Code Case 
published by the ASME to ascertain whether it is consistent with the 
safe operation of nuclear power plants. The Code Cases found to be 
generically acceptable are listed in the RGs that are incorporated by 
reference in 10 CFR 50.55a. The Code Cases found to be unacceptable are 
listed in RG 1.193, but licensees may still seek the NRC's approval to 
apply these Code Cases through the processes in 10 CFR 50.55a for 
requesting the approval of alternatives or for relief. Code Cases that 
the NRC finds to be conditionally acceptable are also listed in RGs 
1.84,

[[Page 37903]]

1.147, and 1.192, which are the subject of this rulemaking, together 
with the conditions that must be used if the Code Case is applied. The 
NRC believes that this rule complies with the NTTAA and OMB Circular A-
119 despite these conditions. If the NRC did not conditionally accept 
ASME Code Cases, it would disapprove these Code Cases entirely. The 
effect would be that licensees and applicants would submit a larger 
number of requests for use of alternatives under the current 10 CFR 
50.55a(a)(3), requests for relief under 10 CFR 50.55a(f) and (g), or 
requests for exemptions under 10 CFR 50.12 and/or 10 CFR 52.7. For 
these reasons, the treatment of ASME Code Cases and any conditions 
proposed to be placed on them in this proposed rule do not conflict 
with any policy on agency use of consensus standards specified in OMB 
Circular A-119.
    The NRC did not identify any other voluntary consensus standards 
developed by the United States voluntary consensus standards bodies for 
use within the United States that the NRC could approve instead of the 
ASME Code Cases.
    The NRC also did not identify any voluntary consensus standards 
developed by multinational voluntary consensus standards bodies for use 
on a multinational basis that the NRC could incorporate by reference 
instead of the ASME Code Cases. This is because no other multinational 
voluntary consensus body would develop alternatives to a voluntary 
consensus standard (i.e., either the ASME BPV Code or the ASME OM Code) 
for which they did not develop and do not maintain.
    In summary, this proposed rule satisfies the requirements of 
Section 12(d)(3) of the NTTAA and OMB Circular A-119.

XI. Finding of No Significant Environmental Impact: Environmental 
Assessment

    This proposed action stems from the Commission's practice of 
incorporating by reference the RGs listing the most recent set of NRC-
approved ASME Code Cases. The purpose of this proposed action is to 
allow licensees to use the Code Cases listed in the RGs as alternatives 
to requirements in the ASME BPV and OM Codes for the construction, ISI, 
and IST of nuclear power plant components. This proposed action is 
intended to advance the NRC's strategic goal of ensuring adequate 
protection of public health and safety and the environment. It also 
demonstrates the agency's commitment to participate in the national 
consensus standards process under the National Technology Transfer and 
Advancement Act of 1995, Pub. L.104-113.
    The National Environmental Policy Act (NEPA), as amended, requires 
Federal government agencies to study the impacts of their ``major 
Federal actions significantly affecting the quality of the human 
environment'' and prepare detailed statements on the environmental 
impacts of the action and alternatives to the action (United States 
Code (U.S.C), Volume 42, Section 4332(C) [42 U.S.C. Sec. 4332(C)]; NEPA 
Sec. 102(C)).
    The Commission has determined under NEPA, as amended, and the 
Commission's regulations in subpart A of 10 CFR part 51, that this 
proposed rule would not be a major Federal action significantly 
affecting the quality of the human environment. Therefore, an 
environmental impact statement is not required.
    As alternatives to the ASME Code, NRC-approved Code Cases provide 
an equivalent level of safety. Therefore, the probability or 
consequences of accidents is not changed. There are also no 
significant, non-radiological impacts associated with this action 
because no changes would be made affecting non-radiological plant 
effluents and because no changes would be made in activities that would 
adversely affect the environment. The determination of this 
environmental assessment is that there will be no significant offsite 
impact to the public from this proposed action.

XII. Paperwork Reduction Act Statement

    This proposed rule contains new or amended information collection 
requirements that are subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501 et seq). This rule has been submitted to the Office of 
Management and Budget (OMB) for review and approval of the information 
collection requirements.
    Type of submission, new or revision: Revision.
    The title of the information collection: Domestic Licensing of 
Production and Utilization Facilities: Updates to Incorporation by 
Reference and Regulatory Guides.
    The form number if applicable: Not applicable.
    How often the collection is required: On occasion.
    Who will be required or asked to report: Power reactor licensees 
and applicants for power reactors under construction.
    An estimate of the number of annual responses: -185.
    The estimated number of annual respondents: 109.
    An estimate of the total number of hours needed annually to 
complete the requirement or request: A reduction of 14,800 reporting 
hours.
    Abstract: This proposed rule is the latest in a series of 
rulemakings that incorporate by reference the latest versions of 
several Regulatory Guides identifying new and revised unconditionally 
or conditionally acceptable ASME Code Cases that are approved for use. 
The incorporation by reference of these Code Cases will reduce the 
number of alternative requests submitted by licensees under proposed 10 
CFR 50.55a(z) by an estimated 185 requests annually.
    The U.S. Nuclear Regulatory Commission is seeking public comment on 
the potential impact of the information collections contained in this 
proposed rule (or proposed policy statement) and on the following 
issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of burden accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?
    The public may examine and have copied for a fee publicly available 
documents, including the draft supporting statement at the NRC's PDR, 
One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville, 
Maryland 20852. The OMB clearance requests are available at the NRC's 
Web site: http://www.nrc.gov/public-involve/doc-comment/omb/. The 
document will be available on the NRC home page site for 60 days after 
the signature date of this notice.
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
above issues, by July 24, 2013 to the Information Services Branch (T-5 
F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or 
by email to INFOCOLLECTS.RESOURCE@NRC.GOV and to the Desk Officer, Chad 
Whiteman, Office of Information and Regulatory Affairs, NEOB-10202, 
(3150-0011), Office of Management and Budget, Washington, DC 20503. 
Comments on the proposed information collections may also be submitted 
via the Federal eRulemaking Portal http://www.regulations.gov, docket 
 NRC-2009-0359. Comments received after this date will be 
considered if it is

[[Page 37904]]

practical to do so, but assurance of consideration cannot be given to 
comments received after this date. Comments can also be emailed to 
Chad_S_Whiteman@omb.eop.gov or submitted by telephone at 202-395-
4718.
Public Protection Notification
    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
unless the requesting document displays a currently valid OMB control 
number.

XIII. Regulatory Analysis

    The ASME Code Cases listed in the RGs to be incorporated by 
reference provide voluntary alternatives to the provisions in the ASME 
BPV and OM Codes for design, construction, ISI, and IST of specific 
structures, systems, and components used in nuclear power plants. 
Implementation of these Code Cases is not required. Licensees and 
applicants use NRC-approved ASME Code Cases to reduce unnecessary 
regulatory burden or gain additional operational flexibility. It would 
be difficult for the NRC to provide these advantages independently of 
the ASME Code Case publication process without expending considerable 
additional resources. The NRC has prepared a regulatory analysis 
addressing the qualitative benefits of the alternatives considered in 
this proposed rulemaking and comparing the costs associated with each 
alternative (ADAMS Accession No. ML103060189). The NRC invites public 
comment on this draft regulatory analysis. Copies of the regulatory 
analysis are available to the public as indicated in Section IX, 
``Availability of Documents,'' of this document.
    In addition to the general opportunity to submit comments on the 
proposed rule, the NRC also requests comments on the NRC's cost and 
benefit estimates as shown in the proposed rule regulatory analysis.

XIV. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the 
Commission certifies that this proposed rule would not impose a 
significant economical impact on a substantial number of small 
entities. This proposed rule would affect only the licensing and 
operation of nuclear power plants. The companies that own these plants 
are not ``small entities'' as defined in the Regulatory Flexibility Act 
or the size standards established by the NRC (10 CFR 2.810).

XV. Backfitting and Issue Finality

    The provisions in this proposed rulemaking would allow licensees 
and applicants to voluntarily apply NRC-approved Code Cases, sometimes 
with NRC-specified conditions. The approved Code Cases are listed in 
three regulatory guides that are incorporated by references into 10 CFR 
50.55a.
    An applicant's and/or licensees voluntary application of an 
approved Code Cases does not constitute backfitting, inasmuch as there 
is no imposition of a new requirement or new position. Similarly, 
voluntary application of an approved Code Case by a part 52 applicant 
or licensee does not represent NRC imposition of a requirement or 
action which is inconsistent with any issue finality provision in part 
52. For these reasons the NRC finds that this proposed rule does not 
involve any provisions requiring the preparation of a backfit analysis 
or documentation demonstrating that one or more of the issue finality 
criteria in part 52 are met.

List of Subjects in 10 CFR Part 50

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

    For the reasons set forth in the preamble, and under the authority 
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553, the NRC is proposing to 
adopt 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 secs. 102, 103, 104, 105, 147, 
149, 161, 181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133, 
2134, 2135, 2167, 2169, 2201, 2231, 2232, 2233, 2236, 2239, 2273, 
2282); Energy Reorganization Act secs. 201, 202, 206 (42 U.S.C. 
5841, 5842, 5846); Nuclear Waste Policy Act sec. 306 (42 U.S.C. 
10226); Government Paperwork Elimination Act sec. 1704 (44 U.S.C. 
3504 note); Energy Policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 
194 (2005). Section 50.7 also issued under Pub. L. 95-601, sec. 10, 
as amended by Pub. L. 102-486, sec. 2902 (42 U.S.C. 5851). Section 
50.10 also issued under Atomic Energy Act secs. 101, 185 (42 U.S.C. 
2131, 2235); National Environmental Policy Act sec. 102 (42 U.S.C. 
4332). Sections 50.13, 50.54(dd), and 50.103 also issued under 
Atomic Energy Act sec. 108 (42 U.S.C. 2138).
    Sections 50.23, 50.35, 50.55, and 50.56 also issued under Atomic 
Energy Act sec. 185 (42 U.S.C. 2235). Appendix Q also issued under 
National Environmental Policy Act sec. 102 (42 U.S.C. 4332). 
Sections 50.34 and 50.54 also issued under sec. 204 (42 U.S.C. 
5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 
97-415 (42 U.S.C. 2239). Section 50.78 also issued under Atomic 
Energy Act sec. 122 (42 U.S.C. 2152). Sections 50.80--50.81 also 
issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).

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


Sec.  50.54  Conditions of licenses.

    The following paragraphs of this section, with the exception of 
paragraphs (r) and (gg), and the applicable requirements of 10 CFR 
50.55a, are conditions in every nuclear power reactor operating license 
issued under this part. The following paragraphs with the exception of 
paragraph (r), (s), and (u) of this section are conditions in every 
combined license issued under part 52 of this chapter, provided, 
however, that paragraphs (i), (i-1), (j), (k), (l), (m), (n), (w), (x), 
(y), and (z) of this section are only applicable after the Commission 
makes the finding under Sec.  52.103(g) of this chapter.
* * * * *
    (ii) [Reserved]
    (jj) Structures, systems, and components must be designed, 
fabricated, erected, constructed, tested, and inspected to quality 
standards commensurate with the importance of the safety function to be 
performed.
0
3. In Sec.  50.55, revise the introductory text of the section, add and 
reserve paragraphs (g) and (h), and add paragraph (i) to read as 
follows:


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

    Each construction permit is subject to the following terms and 
conditions and the applicable requirements of 10 CFR 50.55a; each early 
site permit is subject to the terms and conditions in paragraph (f) of 
this section; each manufacturing license is subject to the terms and 
conditions in paragraphs (e) and (f) of this section and the applicable 
requirements of 10 CFR 50.55a; and each combined license is subject to 
the terms and conditions in paragraphs (e) and (f) of this section and 
the applicable requirements of 10 CFR 50.55a until the date that the 
Commission makes the

[[Page 37905]]

finding under Sec.  52.103(g) of this chapter:
* * * * *
    (g) [Reserved]
    (h) [Reserved]
    (i) Structures, systems, and components must be designed, 
fabricated, erected, constructed, tested, and inspected to quality 
standards commensurate with the importance of the safety function to be 
performed.
0
4. Revise Sec.  50.55a to read as follows:


Sec.  50.55a  Codes and standards.

    (a) Documents approved for incorporation by reference. The 
standards listed in this paragraph have been approved for incorporation 
by reference by the Director of the Federal Register pursuant to 5 
U.S.C. 552(a) and 1 CFR Part 51. The standards are available for 
inspection at the NRC Technical Library, 11545 Rockville Pike, 
Rockville, Maryland 20852; or at the National Archives and Records 
Administration (NARA). For information on the availability of this 
material at NARA, call 202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.
    (1) American Society of Mechanical Engineers (ASME), Three Park 
Avenue, New York, NY 10016 (telephone 800-843-2763), http://www.asme.org/Codes/.
    (i) ASME Boiler and Pressure Vessel Code, Section III. The editions 
and addenda for Section III of the ASME Boiler and Pressure Vessel Code 
are listed below, but limited to those provisions identified in 
paragraph (b)(1) of this section.
    (A) ``Rules for Construction of Nuclear Vessels:''

(1) 1963 Edition,
(2) Summer 1964 Addenda,
(3) Winter 1964 Addenda,
(4) 1965 Edition
(5) 1965 Summer Addenda,
(6) 1965 Winter Addenda,
(7) 1966 Summer Addenda,
(8) 1966 Winter Addenda,
(9) 1967 Summer Addenda,
(10) 1967 Winter Addenda,
(11) 1968 Edition,
(12) 1968 Summer Addenda,
(13)1968 Winter Addenda,
(14) 1969 Summer Addenda,
(15) 1969 Winter Addenda,
(16) 1970 Summer Addenda, and
(17) 1970 Winter Addenda.

    (B) ``Rules for Construction of Nuclear Power Plant Components:''

(1) 1971 Edition,
(2) 1971 Summer Addenda,
(3) 1971 Winter Addenda,
(4) 1972 Summer Addenda,
(5) 1972 Winter Addenda,
(6) 1973 Summer Addenda, and
(7) 1973 Winter Addenda.

    (C) ``Division 1 Rules for Construction of Nuclear Power Plant 
Components:''

(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda,
(4) 1975 Summer Addenda,
(5) 1975 Winter Addenda,
(6) 1976 Summer Addenda, and
(7) 1976 Winter Addenda;

    (D) ``Rules for Construction of Nuclear Power Plant Components--
Division 1;''

(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Summer Addenda,
(10) 1980 Winter Addenda,
(11) 1981 Summer Addenda,
(12) 1981 Winter Addenda,
(13) 1982 Summer Addenda,
(14) 1982 Winter Addenda,
(15) 1983 Edition,
(16) 1983 Summer Addenda,
(17) 1983 Winter Addenda,
(18) 1984 Summer Addenda,
(19) 1984 Winter Addenda,
(20) 1985 Summer Addenda,
(21) 1985 Winter Addenda,
(22) 1986 Edition,
(23) 1986 Addenda,
(24) 1987 Addenda,
(25) 1988 Addenda,
(26) 1989 Edition,
(27) 1989 Addenda,
(28) 1990 Addenda,
(29) 1991 Addenda,
(30) 1992 Edition,
(31) 1992 Addenda,
(32) 1993 Addenda,
(33) 1994 Addenda,
(34) 1995 Edition,
(35)1995 Addenda,
(36)1996 Addenda, and
(37) 1997 Addenda.

    (E) ``Rules for Construction of Nuclear Facility Components--
Division 1:''

(1) 1998 Edition,
(2) 1998 Addenda,
(3) 1999 Addenda,
(4) 2000 Addenda,
(5) 2001 Edition,
(6) 2001 Addenda,
(7) 2002 Addenda,
(8) 2003 Addenda,
(9) 2004 Edition,
(10) 2005 Addenda,
(11) 2006 Addenda,
(12) 2007 Edition, and
(13) 2008 Addenda.

    (ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions 
and addenda for Section XI of the ASME Boiler and Pressure Vessel Code 
are listed below, but limited to those provisions identified in 
paragraph (b)(2) of this section.
    (A) ``Rules for Inservice Inspection of Nuclear Reactor Coolant 
Systems:''

(1) 1970 Edition,
(2) 1971 Edition,
(3) 1971 Summer Addenda,
(4) 1971 Winter Addenda,
(5) 1972 Summer Addenda,
(6) 1972 Winter Addenda,
(7) 1973 Summer Addenda, and
(8) 1973 Winter Addenda.

    (B) ``Rules for Inservice Inspection of Nuclear Power Plant 
Components:''

(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda, and
(4) 1975 Summer Addenda.

    (C) ``Rules for Inservice Inspection of Nuclear Power Plant 
Components--Division 1:''

(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Winter Addenda,
(10) 1981 Summer Addenda,
(11) 1981 Winter Addenda,
(12) 1982 Summer Addenda,
(13) 1982 Winter Addenda,
(14) 1983 Edition,
(15) 1983 Summer Addenda,
(16) 1983 Winter Addenda,
(17) 1984 Summer Addenda,
(18) 1984 Winter Addenda,
(19) 1985 Summer Addenda,
(20) 1985 Winter Addenda,
(21) 1986 Edition,
(22) 1986 Addenda,
(23) 1987 Addenda,
(24) 1988 Addenda,
(25) 1989 Edition,
(26) 1989 Addenda,
(27) 1990 Addenda,
(28) 1991 Addenda,
(28) 1992 Edition,
(30) 1992 Addenda,
(31) 1993 Addenda,
(32) 1994 Addenda,
(33) 1995 Edition,
(34) 1995 Addenda,
(35) 1996 Addenda,
(36) 1997 Addenda,
(37) 1998 Edition,
(38) 1998 Addenda,
(39) 1999 Addenda,
(40) 2000 Addenda,
(41) 2001 Edition,
(42) 2001 Addenda,
(43) 2002 Addenda,
(44) 2003 Addenda,

[[Page 37906]]

(45) 2004 Edition,
(46) 2005 Addenda,
(47) 2006 Addenda,
(48) 2007 Edition, and
(49) 2008 Addenda.

    (iii) ASME Code Cases: Nuclear Components
    (A) ASME Code Case N-722-1. ASME Code Case N-722-1, ``Additional 
Examinations for PWR Pressure Retaining Welds in Class 1 Components 
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' 
(Approval Date: January 26, 2009), with the conditions in paragraph 
(g)(6)(ii)(E) of this section.
    (B) ASME Code Case N-729-1. ASME Code Case N-729-1, ``Alternative 
Examination Requirements for PWR Reactor Vessel Upper Heads With 
Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section 
XI, Division 1'' (Approval Date: March 28, 2006), with the conditions 
in paragraph (g)(6)(ii)(D) of this section.
    (C) ASME Code Case N-770-1. ASME Code Case N-770-1, ``Additional 
Examinations for PWR Pressure Retaining Welds in Class 1 Components 
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1'' 
(Approval Date: December 25, 2009), with the conditions in paragraph 
(g)(6)(ii)(F) of this section.
    (iv) ASME Operation and Maintenance Code. The editions and addenda 
for the ASME Code for Operation and Maintenance of Nuclear Power Plants 
are listed below, but limited to those provisions identified in 
paragraph (b)(3) of this section.
    (A) ``Code for Operation and Maintenance of Nuclear Power Plants:''
(1) 1995 Edition,
(2) 1996 Addenda,
(3) 1997 Addenda,
(4) 1998 Edition,
(5) 1999 Addenda,
(6) 2000 Addenda,
(7) 2001 Edition,
(8) 2002 Addenda,
(9) 2003 Addenda,
(10) 2004 Edition,
(11) 2005 Addenda, and
(12) 2006 Addenda.

    (B) [Reserved]
    (2) Institute of Electrical and Electronics Engineers (IEEE) 
Service Center, 445 Hoes Lane, Piscataway, NJ 08855.
    (i) IEEE standard 279-1971. (IEEE Std 279-1971), ``Criteria for 
Protection Systems for Nuclear Power Generating Stations'' (Approval 
Date: June 3, 1971), referenced in paragraphs (h)(2) of this section.
    (ii) IEEE Standard 603-1991. (IEEE Std 603-1991), ``Standard 
Criteria for Safety Systems for Nuclear Power Generating Stations'' 
(Approval Date: June 27, 1971), referenced in paragraphs (h)(2) and 
(h)(3) of this section. All other standards that are referenced in IEEE 
Std 603-1991 are not approved incorporation by reference.
    (iii) IEEE standard 603-1991, correction sheet. (IEEE Std 603-1991 
correction sheet), ``Standard Criteria for Safety Systems for Nuclear 
Power Generating Stations, Correction Sheet, Issued January 30, 1995, 
'' referenced in paragraphs (h)(2) and (h)(3) of this section. (Copies 
of this correction sheet may be purchased from Thomson Reuters, 3916 
Ranchero Dr., Ann Arbor, MI 48108, http://www.techstreet.com.)
    (3) U.S. Nuclear Regulatory Commission (NRC) Reproduction and 
Distribution Services Section, Washington, DC 20555- 0001; fax: 301-
415-2289; email: Distribution.Resource@nrc.gov.
    (i) NRC Regulatory Guide 1.84, Revision 36. NRC Regulatory Guide 
1.84, Revision 36, ``Design, Fabrication, and Materials Code Case 
Acceptability, ASME Section III,'' [INSERT DATE OF FINAL RULE 
PUBLICATION IN THE Federal Register], with the requirements in 
paragraph (b)(4) of this section.
    (ii) NRC Regulatory Guide 1.147, Revision 17. NRC Regulatory Guide 
1.147, Revision 17, ``Inservice Inspection Code Case Acceptability, 
ASME Section XI, Division 1,'' [INSERT DATE OF FINAL RULE PUBLICATION 
IN THE Federal Register], which lists ASME Code Cases that the NRC has 
approved in accordance with the requirements in paragraph (b)(5) of 
this section.
    (iii) NRC Regulatory Guide 1.192, Revision 1. NRC Regulatory Guide 
1.192, Revision 1, ``Operation and Maintenance Code Case Acceptability, 
ASME OM Code,'' [INSERT DATE OF FINAL RULE PUBLICATION IN THE Federal 
Register], which lists ASME Code Cases that the NRC has approved in 
accordance with the requirements in paragraph (b)(6) of this section.
    (b) Use and conditions on the use of standards. Systems and 
components of boiling and pressurized water-cooled nuclear power 
reactors must meet the requirements of the ASME Boiler and Pressure 
Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance 
of Nuclear Power Plants (OM Code) as specified in this paragraph. Each 
combined license for a utilization facility is subject to the following 
conditions.
    (1) Conditions on ASME BPV Code Section III. Each manufacturing 
license, standard design approval, and design certification under Part 
52 of this chapter is subject to the following conditions. As used in 
this section, references to Section III refer to Section III of the 
ASME Boiler and Pressure Vessel Code and include the 1963 Edition 
through 1973 Winter Addenda and the 1974 Edition (Division 1) through 
the 2008 Addenda (Division 1), subject to the following conditions:
    (i) Section III condition: Section III materials. When applying the 
1992 Edition of Section III, applicants or licensees must apply the 
1992 Edition with the 1992 Addenda of Section II of the ASME Boiler and 
Pressure Vessel Code.
    (ii) Section III condition: Weld leg dimensions. When applying the 
1989 Addenda through the latest edition and addenda, applicants or 
licensees may not apply subparagraphs NB-3683.4(c)(1) and NB-
3683.4(c)(2) or Footnote 11 from the 1989 Addenda through the 2003 
Addenda, or Footnote 13 from the 2004 Edition through the 2008 Addenda 
to Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg size 
less than 1.09 tn.
    (iii) Section III condition: Seismic design of piping. Applicants 
or licensees may use Subarticles NB-3200, NB-3600, NC-3600, and ND-3600 
for seismic design of piping, up to and including the 1993 Addenda, 
subject to the condition specified in paragraph (b)(1)(ii) of this 
section. Applicants or licensees may not use these subarticles for 
seismic design of piping in the 1994 Addenda through the 2005 Addenda 
incorporated by reference in paragraph (a)(1) of this section, except 
that Subarticle NB-3200 in the 2004 Edition through the 2008 Addenda 
may be used by applicants and licensees, subject to the condition in 
paragraph (b)(1)(iii)(A) of this section. Applicants or licensees may 
use Subarticles NB-3600, NC-3600, and ND-3600 for the seismic design of 
piping in the 2006 Addenda through the 2008 Addenda, subject to the 
conditions of this paragraph corresponding to those subarticles.
    (A) Seismic design of piping: first provision. When applying Note 
(1) of Figure NB-3222-1 for Level B service limits, the calculation of 
Pb stresses must include reversing dynamic loads (including 
inertia earthquake effects) if evaluation of these loads is required by 
NB-3223(b).
    (B) Seismic design of piping: second provision. For Class 1 piping, 
the material and Do/t requirements of NB-3656(b) must be met 
for all Service Limits when the Service Limits include reversing 
dynamic loads, and the alternative rules for reversing dynamic loads 
are used.

[[Page 37907]]

    (iv) Section III condition: Quality assurance. When applying 
editions and addenda later than the 1989 Edition of Section III, the 
requirements of NQA-1, ``Quality Assurance Requirements for Nuclear 
Facilities,'' 1986 Edition through the 1994 Edition, are acceptable for 
use, provided that the edition and addenda of NQA-1 specified in NCA-
4000 is used in conjunction with the administrative, quality, and 
technical provisions contained in the edition and addenda of Section 
III being used.
    (v) Section III condition: Independence of inspection. Applicants 
or licensees may not apply NCA-4134.10(a) of Section III, 1995 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1) of this section.
    (vi) Section III condition: Subsection NH. The provisions in 
Subsection NH, ``Class 1 Components in Elevated Temperature Service,'' 
1995 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1) of this section, may only be used for the 
design and construction of Type 316 stainless steel pressurizer heater 
sleeves where service conditions do not cause the components to reach 
temperatures exceeding 900[emsp14][deg]F.
    (vii) Section III condition: Capacity certification and 
demonstration of function of incompressible-fluid pressure-relief 
valves. When applying the 2006 Addenda through the 2007 Edition up to 
and including the 2008 Addenda, applicants and licensees may use 
paragraph NB-7742, except that paragraph NB-7742(a)(2) may not be used. 
For a valve design of a single size to be certified over a range of set 
pressures, the demonstration of function tests under paragraph NB-7742 
must be conducted as prescribed in NB-7732.2 on two valves covering the 
minimum set pressure for the design and the maximum set pressure that 
can be accommodated at the demonstration facility selected for the 
test.
    (2) Conditions on ASME BPV Code Section XI. As used in this 
section, references to Section XI refer to Section XI, Division 1, of 
the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition 
through the 1976 Winter Addenda and the 1977 Edition through the 2007 
Edition with the 2008 Addenda, subject to the following conditions:
    (i) [Reserved]
    (ii) Section XI condition: Pressure-retaining welds in ASME Code 
Class 1 piping (applies to Table IWB-2500 and IWB-2500-1 and Category 
B-J). If the facility's application for a construction permit was 
docketed prior to July 1, 1978, the extent of examination for Code 
Class 1 pipe welds may be determined by the requirements of Table IWB-
2500 and Table IWB-2600 Category B-J of Section XI of the ASME BPV Code 
in the 1974 Edition and Addenda through the Summer 1975 Addenda or 
other requirements the NRC may adopt.
    (iii) [Reserved]
    (iv) [Reserved]
    (v) [Reserved]
    (vi) Section XI condition: Effective edition and addenda of 
Subsection IWE and Subsection IWL. Applicants or licensees may use 
either the 1992 Edition with the 1992 Addenda or the 1995 Edition with 
the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned 
by the requirements in paragraphs (b)(2)(viii) and (b)(2)(ix) of this 
section, when implementing the initial 120-month inspection interval 
for the containment inservice inspection requirements of this section. 
Successive 120-month interval updates must be implemented in accordance 
with paragraph (g)(4)(ii) of this section.
    (vii) Section XI condition: Section XI references to OM Part 4, OM 
Part 6, and OM Part 10 (Table IWA-1600-1). When using Table IWA-1600-1, 
``Referenced Standards and Specifications,'' in the Section XI, 
Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified 
``Revision Date or Indicator'' for ASME/ANSI OM part 4, ASME/ANSI part 
6, and ASME/ANSI part 10 must be the OMa--1988 Addenda to the OM-1987 
Edition. These requirements have been incorporated into the OM Code, 
which is incorporated by reference in paragraph (a)(1)(iv) of this 
section.
    (viii) Section XI condition: Concrete containment examinations. 
Applicants or licensees applying Subsection IWL, 1992 Edition with the 
1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through 
(b)(2)(viii)(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 (b)(2)(viii)(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 (b)(2)(viii)(G) of this section. Applicants or licensees 
applying Subsection IWL, 2007 Edition through the latest edition and 
addenda incorporated by reference in paragraph (a)(1)(ii) of this 
section, must apply paragraph (b)(2)(viii)(E) of this section.
    (A) Concrete containment examinations: first provision. Grease caps 
that are accessible must be visually examined to detect grease leakage 
or grease cap deformations. Grease caps must be removed for this 
examination when there is evidence of grease cap deformation that 
indicates deterioration of anchorage hardware.
    (B) Concrete containment examinations: second provision. When 
evaluation of consecutive surveillances of prestressing forces for the 
same tendon or tendons in a group indicates a trend of prestress loss 
such that the tendon force(s) would be less than the minimum design 
prestress requirements before the next inspection interval, an 
evaluation must be performed and reported in the Engineering Evaluation 
Report as prescribed in IWL-3300.
    (C) Concrete containment examinations: third provision. When the 
elongation corresponding to a specific load (adjusted for effective 
wires or strands) during retensioning of tendons differs by more than 
10 percent from that recorded during the last measurement, an 
evaluation must be performed to determine whether the difference is 
related to wire failures or slip of wires in anchorage. A difference of 
more than 10 percent must be identified in the ISI Summary Report 
required by IWA-6000.
    (D) Concrete containment examinations: fourth provision. The 
applicant or licensee must report the following conditions, if they 
occur, in the ISI Summary Report required by IWA-6000:
    (1) The sampled sheathing filler grease contains chemically 
combined water exceeding 10 percent by weight or the presence of free 
water;
    (2) The absolute difference between the amount removed and the 
amount replaced exceeds 10 percent of the tendon net duct volume; and
    (3) Grease leakage is detected during general visual examination of 
the containment surface.
    (E) Concrete containment examinations: fifth provision. For Class 
CC applications, the applicant or licensee must evaluate the 
acceptability of inaccessible areas when conditions exist in accessible 
areas that could indicate the presence of or the result in degradation 
to such inaccessible areas. For each inaccessible area identified, the 
applicant or licensee must provide the following in the ISI Summary 
Report required by IWA-6000:
    (1) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (2) An evaluation of each area, and the result of the evaluation; 
and

[[Page 37908]]

    (3) A description of necessary corrective actions.
    (F) Concrete containment examinations: sixth provision. Personnel 
that examine containment concrete surfaces and tendon hardware, wires, 
or strands must meet the qualification provisions in IWA-2300. The 
``owner-defined'' personnel qualification provisions in IWL-2310(d) are 
not approved for use.
    (G) Concrete containment examinations: seventh provision. Corrosion 
protection material must be restored following concrete containment 
post-tensioning system repair and replacement activities in accordance 
with the quality assurance program requirements specified in IWA-1400.
    (ix) Section XI condition: Metal containment examinations. 
Applicants or licensees applying Subsection IWE, 1992 Edition with the 
1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy 
the requirements of paragraphs (b)(2)(ix)(A) through (b)(2)(ix)(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), (b)(2)(ix)(B), and 
(b)(2)(ix)(F) through (b)(2)(ix)(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), (b)(2)(ix)(B), and (b)(2)(ix)(F) through (b)(2)(ix)(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 addenda incorporated by reference in paragraph (a)(1)(ii) of 
this section, must satisfy the requirements of paragraphs 
(b)(2)(ix)(A)(2), (b)(2)(ix)(B), and (b)(2)(ix)(J) of this section.
    (A) Metal containment examinations: first provision. For Class MC 
applications, the following apply to inaccessible areas.
    (1) The applicant or licensee must evaluate the acceptability of 
inaccessible areas when conditions exist in accessible areas that could 
indicate the presence of or could result in degradation to such 
inaccessible areas.
    (2) For each inaccessible area identified for evaluation, the 
applicant or licensee must provide the following in the ISI Summary 
Report as required by IWA-6000:
    (i) A description of the type and estimated extent of degradation, 
and the conditions that led to the degradation;
    (ii) An evaluation of each area, and the result of the evaluation; 
and
    (iii) A description of necessary corrective actions.
    (B) Metal containment examinations: second provision. When 
performing remotely the visual examinations required by Subsection IWE, 
the maximum direct examination distance specified in Table IWA-2210-1 
may be extended and the minimum illumination requirements specified in 
Table IWA-2210-1 may be decreased provided that the conditions or 
indications for which the visual examination is performed can be 
detected at the chosen distance and illumination.
    (C) Metal containment examinations: third provision. The 
examinations specified in Examination Category E-B, Pressure Retaining 
Welds, and Examination Category E-F, Pressure Retaining Dissimilar 
Metal Welds, are optional.
    (D) Metal containment examinations: fourth provision. This 
paragraph (b)(2)(ix)(D) may be used as an alternative to the 
requirements of IWE-2430.
    (1) If the examinations reveal flaws or areas of degradation 
exceeding the acceptance standards of Table IWE-3410-1, an evaluation 
must be performed to determine whether additional component 
examinations are required. For each flaw or area of degradation 
identified that exceeds acceptance standards, the applicant or licensee 
must provide the following in the ISI Summary Report required by IWA-
6000:
    (i) A description of each flaw or area, including the extent of 
degradation, and the conditions that led to the degradation;
    (ii) The acceptability of each flaw or area and the need for 
additional examinations to verify that similar degradation does not 
exist in similar components; and
    (iii) A description of necessary corrective actions.
    (2) The number and type of additional examinations to ensure 
detection of similar degradation in similar components.
    (E) Metal containment examinations: fifth provision. A general 
visual examination as required by Subsection IWE must be performed once 
each period.
    (F) Metal containment examinations: sixth provision. VT-1 and VT-3 
examinations must be conducted in accordance with IWA-2200. Personnel 
conducting examinations in accordance with the VT-1 or VT-3 examination 
method must be qualified in accordance with IWA-2300. The ``owner-
defined'' personnel qualification provisions in IWE-2330(a) for 
personnel that conduct VT-1 and VT-3 examinations are not approved for 
use.
    (G) Metal containment examinations: seventh provision. The VT-3 
examination method must be used to conduct the examinations in Items 
E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method 
must be used to conduct the examination in Item E4.11 of Table IWE-
2500-1. An examination of the pressure-retaining bolted connections in 
Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must 
be conducted once each interval. The ``owner-defined'' visual 
examination provisions in IWE-2310(a) are not approved for use for VT-1 
and VT-3 examinations.
    (H) Metal containment examinations: eighth provision. Containment 
bolted connections that are disassembled during the scheduled 
performance of the examinations in Item E1.11 of Table IWE-2500-1 must 
be examined using the VT-3 examination method. Flaws or degradation 
identified during the performance of a VT-3 examination must be 
examined in accordance with the VT-1 examination method. The criteria 
in the material specification or IWB-3517.1 must be used to evaluate 
containment bolting flaws or degradation. As an alternative to 
performing VT-3 examinations of containment bolted connections that are 
disassembled during the scheduled performance of Item E1.11, VT-3 
examinations of containment bolted connections may be conducted 
whenever containment bolted connections are disassembled for any 
reason.
    (I) Metal containment examinations: ninth provision. The ultrasonic 
examination acceptance standard specified in IWE-3511.3 for Class MC 
pressure-retaining components must also be applied to metallic liners 
of Class CC pressure-retaining components.
    (J) Metal containment examinations: tenth provision. In general, a 
repair/replacement activity such as replacing a large containment 
penetration, cutting a large construction opening in the containment 
pressure boundary to replace steam generators, reactor vessel heads, 
pressurizers, or other major equipment; or other similar modification 
is considered a major containment modification. When applying IWE-5000 
to Class MC pressure-retaining components, any major containment 
modification or repair/replacement must be followed by a Type A test to 
provide assurance of

[[Page 37909]]

both containment structural integrity and leaktight integrity prior to 
returning to service, in accordance with 10 CFR Part 50, Appendix J, 
Option A or Option B on which the applicant's or licensee's Containment 
Leak-Rate Testing Program is based. When applying IWE-5000, if a Type 
A, B, or C Test is performed, the test pressure and acceptance standard 
for the test must be in accordance with 10 CFR Part 50, Appendix J.
    (x) Section XI condition: Quality assurance. When applying Section 
XI editions and addenda later than the 1989 Edition, the requirements 
of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,'' 
1979 Addenda through the 1989 Edition, are acceptable as permitted by 
IWA-1400 of Section XI, if the licensee uses its 10 CFR Part 50, 
Appendix B, quality assurance program, in conjunction with Section XI 
requirements. Commitments contained in the licensee's quality assurance 
program description that are more stringent than those contained in 
NQA-1 must govern Section XI activities. Further, where NQA-1 and 
Section XI do not address the commitments contained in the licensee's 
Appendix B quality assurance program description, the commitments must 
be applied to Section XI activities.
    (xi) [Reserved]
    (xii) Section XI condition: Underwater welding. The provisions in 
IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through 
the latest edition and addenda incorporated by reference in paragraph 
(a)(1)(ii) of this section, are not approved for use on irradiated 
material.
    (xiii) [Reserved]
    (xiv) Section XI condition: Appendix VIII personnel qualification. 
All personnel qualified for performing ultrasonic examinations in 
accordance with Appendix VIII must receive 8 hours of annual hands-on 
training on specimens that contain cracks. Licensees applying the 1999 
Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section may use the annual 
practice requirements in VII-4240 of Appendix VII of Section XI in 
place of the 8 hours of annual hands-on training provided that the 
supplemental practice is performed on material or welds that contain 
cracks, or by analyzing prerecorded data from material or welds that 
contain cracks. In either case, training must be completed no earlier 
than 6 months prior to performing ultrasonic examinations at a 
licensee's facility.
    (xv) Section XI condition: Appendix VIII specimen set and 
qualification requirements. Licensees using Appendix VIII in the 1995 
Edition through the 2001 Edition of the ASME Boiler and Pressure Vessel 
Code may elect to comply with all of the provisions in paragraphs 
(b)(2)(xv)(A) through (b)(2)(xv)(M) of this section, except for 
paragraph (b)(2)(xv)(F) of this section, which may be used at the 
licensee's option. Licensees using editions and addenda after 2001 
Edition through the 2006 Addenda must use the 2001 Edition of Appendix 
VIII and may elect to comply with all of the provisions in paragraphs 
(b)(2)(xv)(A) through (b)(2)(xv)(M) of this section, except for 
paragraph (b)(2)(xv)(F) of this section, which may be used at the 
licensee's option.
    (A) Specimen set and qualification: first provision. When applying 
Supplements 2, 3, and 10 to Appendix VIII, the following examination 
coverage criteria requirements must be used:
    (1) Piping must be examined in two axial directions, and when 
examination in the circumferential direction is required, the 
circumferential examination must be performed in two directions, 
provided access is available. Dissimilar metal welds must be examined 
axially and circumferentially.
    (2) Where examination from both sides is not possible, full 
coverage credit may be claimed from a single side for ferritic welds. 
Where examination from both sides is not possible on austenitic welds 
or dissimilar metal welds, full coverage credit from a single side may 
be claimed only after completing a successful single-sided Appendix 
VIII demonstration using flaws on the opposite side of the weld. 
Dissimilar metal weld qualifications must be demonstrated from the 
austenitic side of the weld, and the qualification may be expanded for 
austenitic welds with no austenitic sides using a separate add-on 
performance demonstration. Dissimilar metal welds may be examined from 
either side of the weld.
    (B) Specimen set and qualification: second provision. The following 
conditions must be used in addition to the requirements of Supplement 4 
to Appendix VIII:
    (1) Paragraph 3.1, Detection acceptance criteria--Personnel are 
qualified for detection if the results of the performance demonstration 
satisfy the detection requirements of ASME Section XI, Appendix VIII, 
Table VIII-S4-1, and no flaw greater than 0.25 inch through-wall 
dimension is missed.
    (2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the 
50 percent of allowable flaw size, as defined in IWB-3500, need not be 
included as detection flaws. For procedures applied from the inside 
surface, use the minimum thickness specified in the scope of the 
procedure to calculate a/t. For procedures applied from the outside 
surface, the actual thickness of the test specimen is to be used to 
calculate a/t.
    (C) Specimen set and qualification: third provision. When applying 
Supplement 4 to Appendix VIII, the following conditions must be used:
    (1) A depth sizing requirement of 0.15 inch RMS must be used in 
lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a 
length sizing requirement of 0.75 inch RMS must be used in lieu of the 
requirement in Subparagraph 3.2(b).
    (2) In lieu of the location acceptance criteria requirements of 
Subparagraph 2.1(b), a flaw will be considered detected when reported 
within 1.0 inch or 10 percent of the metal path to the flaw, whichever 
is greater, of its true location in the X and Y directions.
    (3) In lieu of the flaw type requirements of Subparagraph 
1.1(e)(1), a minimum of 70 percent of the flaws in the detection and 
sizing tests must be cracks. Notches, if used, must be limited by the 
following:
    (i) Notches must be limited to the case where examinations are 
performed from the clad surface.
    (ii) Notches must be semielliptical with a tip width of less than 
or equal to 0.010 inches.
    (iii) Notches must be perpendicular to the surface within 2 degrees.
    (4) In lieu of the detection test matrix requirements in paragraphs 
1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain 
a representative distribution of flaw orientations, sizes, and 
locations.
    (D) Specimen set and qualification: fourth provision. The following 
conditions must be used in addition to the requirements of Supplement 6 
to Appendix VIII:
    (1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are 
qualified for detection if:
    (i) No surface connected flaw greater than 0.25 inch through-wall 
has been missed.
    (ii) No embedded flaw greater than 0.50 inch through-wall has been 
missed.
    (2) Paragraph 3.1, Detection Acceptance Criteria--For procedure 
qualification, all flaws within the scope of the procedure are 
detected.
    (3) Paragraph 1.1(b) for detection and sizing test flaws and 
locations--Flaws smaller than the 50 percent of allowable flaw size, as 
defined in IWB-3500, need not be included as detection flaws. Flaws 
that are less than the allowable flaw size, as defined in IWB-3500, may 
be used as detection and sizing flaws.

[[Page 37910]]

    (4) Notches are not permitted.
    (E) Specimen set and qualification: fifth provision. When applying 
Supplement 6 to Appendix VIII, the following conditions must be used:
    (1) A depth sizing requirement of 0.25 inch RMS must be used in 
lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and 
3.2(c)(3).
    (2) In lieu of the location acceptance criteria requirements in 
Subparagraph 2.1(b), a flaw will be considered detected when reported 
within 1.0 inch or 10 percent of the metal path to the flaw, whichever 
is greater, of its true location in the X and Y directions.
    (3) In lieu of the length sizing criteria requirements of 
Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch 
RMS must be used.
    (4) In lieu of the detection specimen requirements in Subparagraph 
1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The 
remaining flaws may be cracks or fabrication type flaws, such as slag 
and lack of fusion. The use of notches is not allowed.
    (5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test 
matrix, personnel demonstration test sets must contain a representative 
distribution of flaw orientations, sizes, and locations.
    (F) Specimen set and qualification: sixth provision. The following 
conditions may be used for personnel qualification for combined 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII 
qualification. Licensees choosing to apply this combined qualification 
must apply all of the provisions of Supplements 4 and 6 including the 
following conditions:
    (1) For detection and sizing, the total number of flaws must be at 
least 10. A minimum of 5 flaws must be from Supplement 4, and a minimum 
of 50 percent of the flaws must be from Supplement 6. At least 50 
percent of the flaws in any sizing must be cracks. Notches are not 
acceptable for Supplement 6.
    (2) Examination personnel are qualified for detection and length 
sizing when the results of any combined performance demonstration 
satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
    (3) Examination personnel are qualified for depth sizing when 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws 
are sized within the respective acceptance criteria of those 
supplements.
    (G) Specimen set and qualification: seventh provision. When 
applying Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, 
or combined Supplement 4 and Supplement 6 qualification, the following 
additional conditions must be used, and examination coverage must 
include:
    (1) The clad-to-base-metal-interface, including a minimum of 15 
percent T (measured from the clad-to-base-metal-interface), must be 
examined from four orthogonal directions using procedures and personnel 
qualified in accordance with Supplement 4 to Appendix VIII.
    (2) If the clad-to-base-metal-interface procedure demonstrates 
detectability of flaws with a tilt angle relative to the weld 
centerline of at least 45 degrees, the remainder of the examination 
volume is considered fully examined if coverage is obtained in one 
parallel and one perpendicular direction. This must be accomplished 
using a procedure and personnel qualified for single-side examination 
in accordance with Supplement 6. Subsequent examinations of this volume 
may be performed using examination techniques qualified for a tilt 
angle of at least 10 degrees.
    (3) The examination volume not addressed by paragraph 
(b)(2)(xv)(G)(1) of this section is considered fully examined if 
coverage is obtained in one parallel and one perpendicular direction, 
using a procedure and personnel qualified for single sided examination 
when the conditions in paragraph (b)(2)(xv)(G)(2) are met.
    (H) Specimen set and qualification: eighth provision. When applying 
Supplement 5 to Appendix VIII, at least 50 percent of the flaws in the 
demonstration test set must be cracks and the maximum misorientation 
must be demonstrated with cracks. Flaws in nozzles with bore diameters 
equal to or less than 4 inches may be notches.
    (I) Specimen set and qualification: ninth provision. When applying 
Supplement 5, Paragraph (a), to Appendix VIII, the number of false 
calls allowed must be D/10, with a maximum of 3, where D is the 
diameter of the nozzle.
    (J) [Reserved]
    (K) Specimen set and qualification: eleventh provision. When 
performing nozzle-to-vessel weld examinations, the following conditions 
must be used when the requirements contained in Supplement 7 to 
Appendix VIII are applied for nozzle-to-vessel welds in conjunction 
with Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or 
combined Supplement 4 and Supplement 6 qualification.
    (1) For examination of nozzle-to-vessel welds conducted from the 
bore, the following conditions are required to qualify the procedures, 
equipment, and personnel:
    (i) For detection, a minimum of four flaws in one or more full-
scale nozzle mock-ups must be added to the test set. The specimens must 
comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for 
flaw locations specified in Table VIII S6-1. Flaws may be notches, 
fabrication flaws, or cracks. Seventy-five (75) percent of the flaws 
must be cracks or fabrication flaws. Flaw locations and orientations 
must be selected from the choices shown in paragraph (b)(2)(xv)(K)(4) 
of this section, Table VIII-S7-1--Modified, with the exception that 
flaws in the outer eighty-five (85) percent of the weld need not be 
perpendicular to the weld. There may be no more than two flaws from 
each category, and at least one subsurface flaw must be included.
    (ii) For length sizing, a minimum of four flaws as in paragraph 
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set. 
The length sizing results must be added to the results of combined 
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The 
combined results must meet the acceptance standards contained in 
paragraph (b)(2)(xv)(E)(3) of this section.
    (iii) For depth sizing, a minimum of four flaws as in paragraph 
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set. 
Their depths must be distributed over the ranges of Supplement 4, 
Paragraph 1.1, to Appendix VIII, for the inner 15 percent of the wall 
thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the 
remainder of the wall thickness. The depth sizing results must be 
combined with the sizing results from Supplement 4 to Appendix VIII for 
the inner 15 percent and to Supplement 6 to Appendix VIII for the 
remainder of the wall thickness. The combined results must meet the 
depth sizing acceptance criteria contained in paragraphs 
(b)(2)(xv)(C)(1), (b)(2)(xv)(E)(1), and (b)(2)(xv)(F)(3) of this 
section.
    (2) For examination of reactor pressure vessel nozzle-to-vessel 
welds conducted from the inside of the vessel, the following conditions 
are required:
    (i) The clad-to-base-metal-interface and the adjacent examination 
volume to a minimum depth of 15 percent T (measured from the clad-to-
base-metal-interface) must be examined from four orthogonal directions 
using a procedure and personnel qualified in accordance with Supplement 
4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and 
(b)(2)(xv)(C) of this section.
    (ii) When the examination volume defined in paragraph 
(b)(2)(xv)(K)(2)(i)

[[Page 37911]]

of this section cannot be effectively examined in all four directions, 
the examination must be augmented by examination from the nozzle bore 
using a procedure and personnel qualified in accordance with paragraph 
(b)(2)(xv)(K)(1) of this section.
    (iii) The remainder of the examination volume not covered by 
paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of 
paragraphs (b)(2)(xv)(K)(2)(i) and (b)(2)(xv)(K)(2)(ii) of this 
section, must be examined from the nozzle bore using a procedure and 
personnel qualified in accordance with paragraph (b)(2)(xv)(K)(1) of 
this section, or from the vessel shell using a procedure and personnel 
qualified for single sided examination in accordance with Supplement 6 
to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through 
(b)(2)(xv)(G) of this section.
    (3) For examination of reactor pressure vessel nozzle-to-shell 
welds conducted from the outside of the vessel, the following 
conditions are required:
    (i) The clad-to-base-metal-interface and the adjacent metal to a 
depth of 15 percent T (measured from the clad-to-base-metal-interface) 
must be examined from one radial and two opposing circumferential 
directions using a procedure and personnel qualified in accordance with 
Supplement 4 to Appendix VIII, as conditioned by paragraphs 
(b)(2)(xv)(B) and (b)(2)(xv)(C) of this section, for examinations 
performed in the radial direction, and Supplement 5 to Appendix VIII, 
as conditioned by paragraph (b)(2)(xv)(J) of this section, for 
examinations performed in the circumferential direction.
    (ii) The examination volume not addressed by paragraph 
(b)(2)(xv)(K)(3)(i) of this section must be examined in a minimum of 
one radial direction using a procedure and personnel qualified for 
single sided examination in accordance with Supplement 6 to Appendix 
VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (b)(2)(xv)(G) 
of this section.
    (4) Table VIII-S7-1, ``Flaw Locations and Orientations,'' 
Supplement 7 to Appendix VIII, is conditioned as follows:

                        Table VIII-S7-1--Modified
------------------------------------------------------------------------
                     Flaw locations and orientations
-------------------------------------------------------------------------
                                             Parallel   Perpendicular to
                                              to weld         weld
------------------------------------------------------------------------
Inner 15 percent..........................          X                 X
Outside Diameter Surface..................          X   ................
Subsurface................................          X   ................
------------------------------------------------------------------------

    (L) Specimen set and qualification: twelfth provision. As a 
condition to the requirements of Supplement 8, Subparagraph 1.1(c), to 
Appendix VIII, notches may be located within one diameter of each end 
of the bolt or stud.
    (M) Specimen set and qualification: thirteenth provision. When 
implementing Supplement 12 to Appendix VIII, only the provisions 
related to the coordinated implementation of Supplement 3 to Supplement 
2 performance demonstrations are to be applied.
    (xvi) Section XI condition: Appendix VIII single side ferritic 
vessel and piping and stainless steel piping examinations. When 
applying editions and addenda prior to the 2007 Edition of Section XI, 
the following conditions apply.
    (A) Ferritic and stainless steel piping examinations: first 
provision. Examinations performed from one side of a ferritic vessel 
weld must be conducted with equipment, procedures, and personnel that 
have demonstrated proficiency with single side examinations. To 
demonstrate equivalency to two sided examinations, the demonstration 
must be performed to the requirements of Appendix VIII, as conditioned 
by this paragraph and paragraphs (b)(2)(xv)(B) through (b)(2)(xv)(G) of 
this section, on specimens containing flaws with non-optimum sound 
energy reflecting characteristics or flaws similar to those in the 
vessel being examined.
    (B) Ferritic and stainless steel piping examinations: second 
provision. Examinations performed from one side of a ferritic or 
stainless steel pipe weld must be conducted with equipment, procedures, 
and personnel that have demonstrated proficiency with single side 
examinations. To demonstrate equivalency to two sided examinations, the 
demonstration must be performed to the requirements of Appendix VIII, 
as conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this 
section.
    (xvii) Section XI condition: Reconciliation of quality 
requirements. When purchasing replacement items, in addition to the 
reconciliation provisions of IWA-4200, 1995 Addenda through 1998 
Edition, the replacement items must be purchased, to the extent 
necessary, in accordance with the licensee's quality assurance program 
description required by 10 CFR 50.34(b)(6)(ii).
    (xviii) Section XI condition: NDE personnel certification.
    (A) NDE personnel certification: first provision. Level I and II 
nondestructive examination personnel must be recertified on a 3-year 
interval in lieu of the 5-year interval specified in the 1997 Addenda 
and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 
1999 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section.
    (B) NDE personnel certification: second provision. When applying 
editions and addenda prior to the 2007 Edition of Section XI, paragraph 
IWA-2316 may only be used to qualify personnel that observe leakage 
during system leakage and hydrostatic tests conducted in accordance 
with IWA 5211(a) and (b).
    (C) NDE personnel certification: third provision. When applying 
editions and addenda prior to the 2005 Addenda of Section XI, 
licensee's qualifying visual examination personnel for VT-3 visual 
examination under paragraph IWA-2317 of Section XI must demonstrate the 
proficiency of the training by administering an initial qualification 
examination and administering subsequent examinations on a 3-year 
interval.
    (xix) Section XI condition: Substitution of alternative methods. 
The provisions for substituting alternative examination methods, a 
combination of methods, or newly developed techniques in the 1997 
Addenda of IWA-2240 must be applied when using the 1998 Edition through 
the 2004 Edition of Section XI of the ASME BPV Code. The provisions in 
IWA-4520(c), 1997 Addenda through the 2004 Edition, allowing the 
substitution of alternative methods, a combination of methods, or newly 
developed techniques for the methods specified in the Construction 
Code, are not approved for use. The provisions in IWA-4520(b)(2) and 
IWA-4521 of the 2008 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, 
allowing the substitution of ultrasonic examination for radiographic 
examination specified in the Construction Code, are not approved for 
use.
    (xx) Section XI condition: System leakage tests.
    (A) System leakage tests: first provision. When performing system 
leakage tests in accordance with IWA-5213(a), 1997 through 2002 
Addenda,

[[Page 37912]]

the licensee must maintain a 10-minute hold time after test pressure 
has been reached for Class 2 and Class 3 components that are not in use 
during normal operating conditions. No hold time is required for the 
remaining Class 2 and Class 3 components provided that the system has 
been in operation for at least 4 hours for insulated components or 10 
minutes for uninsulated components.
    (B) System leakage tests: second provision. The NDE provision in 
IWA-4540(a)(2) of the 2002 Addenda of Section XI must be applied when 
performing system leakage tests after repair and replacement activities 
performed by welding or brazing on a pressure retaining boundary using 
the 2003 Addenda through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section.
    (xxi) Section XI condition: Table IWB-2500-1 examination 
requirements.
    (A) Table IWB-2500-1 examination requirements: first provision. The 
provisions of Table IWB-2500-1, Examination Category B-D, Full 
Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 
(Inspection Program A) and Items B3.120 and B3.140 (Inspection Program 
B) of the 1998 Edition must be applied when using the 1999 Addenda 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1)(ii) of this section. A visual examination with 
magnification that has a resolution sensitivity to detect a 1-mil width 
wire or crack, utilizing the allowable flaw length criteria in Table 
IWB-3512-1, 1997 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, with 
a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may 
be performed instead of an ultrasonic examination.
    (B) [Reserved]
    (xxii) Section XI condition: Surface examination. The use of the 
provision in IWA-2220, ``Surface Examination,'' of Section XI, 2001 
Edition through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section, that allows use of 
an ultrasonic examination method is prohibited.
    (xxiii) Section XI condition: Evaluation of thermally cut surfaces. 
The use of the provisions for eliminating mechanical processing of 
thermally cut surfaces in IWA-4461.4.2 of Section XI, 2001 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1)(ii) of this section, is prohibited.
    (xxiv) Section XI condition: Incorporation of the performance 
demonstration initiative and addition of ultrasonic examination 
criteria. The use of Appendix VIII and the supplements to Appendix VIII 
and Article I-3000 of Section XI of the ASME BPV Code, 2002 Addenda 
through the 2006 Addenda, is prohibited.
    (xxv) Section XI condition: Mitigation of defects by modification. 
The use of the provisions in IWA-4340, ``Mitigation of Defects by 
Modification,'' Section XI, 2001 Edition through the latest edition and 
addenda incorporated by reference in paragraph (a)(1)(ii) of this 
section are prohibited.
    (xxvi) Section XI condition: Pressure testing Class 1, 2 and 3 
mechanical joints. The repair and replacement activity provisions in 
IWA-4540(c) of the 1998 Edition of Section XI for pressure testing 
Class 1, 2, and 3 mechanical joints must be applied when using the 2001 
Edition through the latest edition and addenda incorporated by 
reference in paragraph (a)(1)(ii) of this section.
    (xxvii) Section XI condition: Removal of insulation. When 
performing visual examination in accordance with IWA-5242 of Section XI 
of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-
5241 of the 2007 Edition through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, 
insulation must be removed from 17-4 PH or 410 stainless steel studs or 
bolts aged at a temperature below 1100[emsp14][deg]F or having a 
Rockwell Method C hardness value above 30, and from A-286 stainless 
steel studs or bolts preloaded to 100,000 pounds per square inch or 
higher.
    (xxviii) Section XI condition: Analysis of flaws. Licensees using 
ASME BPV Code, Section XI, Appendix A, must use the following 
conditions when implementing Equation (2) in A-4300(b)(1):
    For R < 0, [Delta]KI depends on the crack depth (a), and 
the flow stress ([sigma]f). The flow stress is defined by 
[sigma]f = 1/2([sigma]ys+ [sigma]ult), 
where [sigma]ys is the yield strength and 
[sigma]ult is the ultimate tensile strength in units ksi 
(MPa) and (a) is in units in. (mm). For -2 <= R <= 0 and 
Kmax- Kmin <= 0.8 x 1.12 
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI = 
Kmax. For R < -2 and Kmax- Kmin <= 0.8 
x 1.12 [sigma]f[radic]([pi]a), S = 1 and 
[Delta]KI= (1 - R) Kmax/3. For R < 0 and 
Kmax - Kmin > 0.8 x 1.12 
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI= 
Kmax- Kmin.
    (xxix) Section XI condition: Nonmandatory Appendix R. Nonmandatory 
Appendix R, ``Risk-Informed Inspection Requirements for Piping,'' of 
Section XI, 2005 Addenda through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(ii) of this section, may 
not be implemented without prior NRC authorization of the proposed 
alternative in accordance with paragraph (z) of this section.
    (3) Conditions on ASME OM Code. As used in this section, references 
to the OM Code refer to the ASME Code for Operation and Maintenance of 
Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory 
Appendices I and II, and Nonmandatory Appendices A through H and J, 
including the 1995 Edition through the 2006 Addenda, subject to the 
following conditions:
    (i) OM condition: Quality assurance. When applying editions and 
addenda of the OM Code, the requirements of NQA-1, ``Quality Assurance 
Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as 
permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-
1500 of the 1998 Edition through the latest edition and addenda 
incorporated by reference in paragraph (a)(1)(iv) of this section, 
provided the licensee uses its 10 CFR Part 50, Appendix B, quality 
assurance program in conjunction with the OM Code requirements. 
Commitments contained in the licensee's quality assurance program 
description that are more stringent than those contained in NQA-1 
govern OM Code activities. If NQA-1 and the OM Code do not address the 
commitments contained in the licensee's Appendix B quality assurance 
program description, the commitments must be applied to OM Code 
activities.
    (ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees 
must comply with the provisions for MOV testing in OM Code ISTC 4.2, 
1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition 
through the latest edition and addenda incorporated by reference in 
paragraph (a)(1)(iv) of this section, and must establish a program to 
ensure that motor-operated valves continue to be capable of performing 
their design basis safety functions.
    (iii) [Reserved]
    (iv) OM condition: Check valves (Appendix II). Licensees applying 
Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM 
Code, 1995 Edition with the 1996 and 1997 Addenda, must satisfy the 
requirements of paragraphs (b)(3)(iv)(A), (b)(3)(iv)(B), and 
(b)(3)(iv)(C) of this section. Licensees applying Appendix II, 1998 
Edition through the 2002 Addenda, must satisfy the requirements of 
paragraphs (b)(3)(iv)(A), (b)(3)(iv)(B), and (b)(3)(iv)(D) of this 
section.
    (A) Check valves: first provision. Valve opening and closing 
functions

[[Page 37913]]

must be demonstrated when flow testing or examination methods 
(nonintrusive, or disassembly and inspection) are used;
    (B) Check valves: second provision. The initial interval for tests 
and associated examinations may not exceed two fuel cycles or 3 years, 
whichever is longer; any extension of this interval may not exceed one 
fuel cycle per extension with the maximum interval not to exceed 10 
years. Trending and evaluation of existing data must be used to reduce 
or extend the time interval between tests.
    (C) Check valves: third provision. If the Appendix II condition 
monitoring program is discontinued, then the requirements of ISTC 4.5.1 
through 4.5.4 must be implemented.
    (D) Check valves: fourth provision. The applicable provisions of 
subsection ISTC must be implemented if the Appendix II condition 
monitoring program is discontinued.
    (v) OM condition: Snubbers ISTD. Article IWF-5000, ``Inservice 
Inspection Requirements for Snubbers,'' of the ASME BPV Code, Section 
XI, must be used when performing inservice inspection examinations and 
tests of snubbers at nuclear power plants, except as conditioned in 
paragraphs (b)(3)(v)(A) and (b)(3)(v)(B) of this section.
    (A) Snubbers: first provision. Licensees may use Subsection ISTD, 
``Preservice and Inservice Examination and Testing of Dynamic 
Restraints (Snubbers) in Light-Water Reactor Power Plants,'' ASME OM 
Code, 1995 Edition through the latest edition and addenda incorporated 
by reference in paragraph (a)(1)(iv) of this section, in place of the 
requirements for snubbers in the editions and addenda up to the 2005 
Addenda of the ASME BPV Code, Section XI, IWF-5200(a) and (b) and IWF-
5300(a) and (b), by making appropriate changes to their technical 
specifications or licensee-controlled documents. Preservice and 
inservice examinations must be performed using the VT-3 visual 
examination method described in IWA-2213.
    (B) Snubbers: second provision. Licensees must comply with the 
provisions for examining and testing snubbers in Subsection ISTD of the 
ASME OM Code and make appropriate changes to their technical 
specifications or licensee-controlled documents when using the 2006 
Addenda and later editions and addenda of Section XI of the ASME BPV 
Code.
    (vi) OM condition: Exercise interval for manual valves. Manual 
valves must be exercised on a 2-year interval rather than the 5-year 
interval specified in paragraph ISTC-3540 of the 1999 through the 2005 
Addenda of the ASME OM Code, provided that adverse conditions do not 
require more frequent testing.
    (4) Conditions on Design, Fabrication, and Materials Code Cases. 
Each manufacturing license, standard design approval, and design 
certification application under Part 52 of this chapter is subject to 
the following conditions. Licensees may apply the ASME BPV Code Cases 
listed in NRC Regulatory Guide 1.84, Revision 36, without prior NRC 
approval, subject to the following conditions:
    (i) Design, Fabrication, and Materials Code Case condition: 
Applying Code Cases. When an applicant or licensee initially applies a 
listed Code Case, the applicant or licensee must apply the most recent 
version of that Code Case incorporated by reference in paragraph (a) of 
this section.
    (ii) Design, Fabrication, and Materials Code Case condition: 
Applying different revisions of Code Cases. If an applicant or licensee 
has previously applied a Code Case and a later version of the Code Case 
is incorporated by reference in paragraph (a) of this section, the 
applicant or licensee may continue to apply the previous version of the 
Code Case as authorized or may apply the later version of the Code 
Case, including any NRC-specified conditions placed on its use, until 
it updates its Code of Record for the component being constructed.
    (iii) Design, Fabrication, and Materials Code Case condition: 
Applying annulled Code Cases. Application of an annulled Code Case is 
prohibited unless an applicant or licensee applied the listed Code Case 
prior to it being listed as annulled in Regulatory Guide 1.84. If an 
applicant or licensee has applied a listed Code Case that is later 
listed as annulled in Regulatory Guide 1.84, the applicant or licensee 
may continue to apply the Code Case until it updates its Code of Record 
for the component being constructed.
    (5) Conditions on inservice inspection Code Cases. Licensees may 
apply the ASME BPV Code Cases listed in Regulatory Guide 1.147, 
Revision 17, without prior NRC approval, subject to the following:
    (i) ISI Code Case condition: Applying Code Cases. When a licensee 
initially applies a listed Code Case, the licensee must apply the most 
recent version of that Code Case incorporated by reference in paragraph 
(a) of this section.
    (ii) ISI Code Case condition: Applying different revisions of Code 
Cases. If a licensee has previously applied a Code Case and a later 
version of the Code Case is incorporated by reference in paragraph (a) 
of this section, the licensee may continue to apply, to the end of the 
current 120-month interval, the previous version of the Code Case, as 
authorized, or may apply the later version of the Code Case, including 
any NRC-specified conditions placed on its use. Licensees who choose to 
continue use of the Code Case during subsequent 120-month ISI program 
intervals will be required to implement the latest version incorporated 
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of 
Regulatory Guide 1.147, Revision 17.
    (iii) ISI Code Case condition: Applying annulled Code Cases. 
Application of an annulled Code Case is prohibited unless a licensee 
previously applied the listed Code Case prior to it being listed as 
annulled in Regulatory Guide 1.147. If a licensee has applied a listed 
Code Case that is later listed as annulled in Regulatory Guide 1.147, 
the licensee may continue to apply the Code Case to the end of the 
current 120-month interval.
    (6) Conditions on Operation and Maintenance of Nuclear Power Plants 
Code Cases. Licensees may apply the ASME Operation and Maintenance Code 
Cases listed in Regulatory Guide 1.192, Revision 1, without prior NRC 
approval, subject to the following:
    (i) OM Code Case condition: Applying Code Cases. When a licensee 
initially applies a listed Code Case, the licensee must apply the most 
recent version of that Code Case incorporated by reference in paragraph 
(a) of this section.
    (ii) OM Code Case condition: Applying different revisions of Code 
Cases. If a licensee has previously applied a Code Case and a later 
version of the Code Case is incorporated by reference in paragraph (a) 
of this section, the licensee may continue to apply, to the end of the 
current 120-month interval, the previous version of the Code Case, as 
authorized, or may apply the later version of the Code Case, including 
any NRC-specified conditions placed on its use. Licensees who choose to 
continue use of the Code Case during subsequent 120-month ISI program 
intervals will be required to implement the latest version incorporated 
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of 
Regulatory Guide 1.192, Revision 1.
    (iii) OM Code Case condition: Applying annulled Code Cases. 
Application of an annulled Code Case is prohibited unless a licensee 
previously applied the listed Code Case prior to it being listed as 
annulled in Regulatory Guide 1.192. If a licensee has applied a listed 
Code Case that is later listed as

[[Page 37914]]

annulled in Regulatory Guide 1.192, the licensee may continue to apply 
the Code Case to the end of the current 120-month interval.
    (c) Reactor coolant pressure boundary. Systems and components of 
boiling and pressurized water-cooled nuclear power reactors must meet 
the requirements of the ASME BPV Code as specified in this paragraph. 
Each manufacturing license, standard design approval, and design 
certification application under Part 52 of this chapter and each 
combined license for a utilization facility is subject to the following 
conditions:
    (1) Standards requirement for reactor coolant pressure boundary 
components. Components that are part of the reactor coolant pressure 
boundary must meet the requirements for Class 1 components in Section 
III\4,5\ of the ASME BPV Code, except as provided in paragraphs (c)(2), 
(c)(3), and (c)(4) of this section.
    (2) Exceptions to reactor coolant pressure boundary standards 
requirement. Components that are connected to the reactor coolant 
system and are part of the reactor coolant pressure boundary as defined 
in Sec.  50.2 need not meet the requirements of paragraph (c)(1) of 
this section, provided that:
    (i) Exceptions: Shutdown and cooling capability. In the event of 
postulated failure of the component during normal reactor operation, 
the reactor can be shut down and cooled down in an orderly manner, 
assuming makeup is provided by the reactor coolant makeup system; or
    (ii) Exceptions: Isolation capability. The component is or can be 
isolated from the reactor coolant system by two valves in series (both 
closed, both open, or one closed and the other open). Each open valve 
must be capable of automatic actuation and, assuming the other valve is 
open, its closure time must be such that, in the event of postulated 
failure of the component during normal reactor operation, each valve 
remains operable and the reactor can be shut down and cooled down in an 
orderly manner, assuming makeup is provided by the reactor coolant 
makeup system only.
    (3) Applicable Code and Code Cases and conditions on their use. The 
Code edition, addenda, and optional ASME Code Cases to be applied to 
components of the reactor coolant pressure boundary must be determined 
by the provisions of paragraph NCA-1140, Subsection NCA of Section III 
of the ASME BPV Code, subject to the following conditions:
    (i) Reactor coolant pressure boundary condition: Code edition and 
addenda. The edition and addenda applied to a component must be those 
that are incorporated by reference in paragraph (a)(1)(i) of this 
section;
    (ii) Reactor coolant pressure boundary condition: Earliest edition 
and addenda for pressure vessel. The ASME Code provisions applied to 
the pressure vessel may be dated no earlier than the summer 1972 
Addenda of the 1971 Edition;
    (iii) Reactor coolant pressure boundary condition: Earliest edition 
and addenda for piping, pumps, and valves. The ASME Code provisions 
applied to piping, pumps, and valves may be dated no earlier than the 
Winter 1972 Addenda of the 1971 Edition; and
    (iv) Reactor coolant pressure boundary condition: Use of Code 
Cases. The optional Code Cases applied to a component must be those 
listed in NRC Regulatory Guide 1.84 that is incorporated by reference 
in paragraph (a)(3)(i) of this section.
    (4) Standards requirement for components in older plants. For a 
nuclear power plant whose construction permit was issued prior to May 
14, 1984, the applicable Code edition and addenda for a component of 
the reactor coolant pressure boundary continue to be that Code edition 
and addenda that were required by Commission regulations for such a 
component at the time of issuance of the construction permit.
    (d) Quality Group B components. Systems and components of boiling 
and pressurized water-cooled nuclear power reactors must meet the 
requirements of the ASME BPV Code as specified in this paragraph. Each 
manufacturing license, standard design approval, and design 
certification application under Part 52 of this chapter, and each 
combined license for a utilization facility is subject to the following 
conditions:
    (1) Standards requirement for Quality Group B components. For a 
nuclear power plant whose application for a construction permit under 
this part, or a combined license or manufacturing license under Part 52 
of this chapter, docketed after May 14, 1984, or for an application for 
a standard design approval or a standard design certification docketed 
after May 14, 1984, components classified Quality Group B\9\ must meet 
the requirements for Class 2 Components in Section III of the ASME BPV 
Code.
    (2) Quality Group B: Applicable Code and Code Cases and conditions 
on their use. The Code edition, addenda, and optional ASME Code Cases 
to be applied to the systems and components identified in paragraph 
(d)(1) of this section must be determined by the rules of paragraph 
NCA-1140, Subsection NCA of Section III of the ASME BPV Code, subject 
to the following conditions:
    (i) Quality Group B condition: Code edition and addenda. The 
edition and addenda must be those that are incorporated by reference in 
paragraph (a)(1)(i) of this section;
    (ii) Quality Group B condition: Earliest edition and addenda for 
components. The ASME Code provisions applied to the systems and 
components may be dated no earlier than the 1980 Edition; and
    (iii) Quality Group B condition: Use of Code Cases. The optional 
Code Cases must be those listed in NRC Regulatory Guide 1.84 that is 
incorporated by reference in paragraph (a)(3)(i) of this section.
    (e) Quality Group C components. Systems and components of boiling 
and pressurized water-cooled nuclear power reactors must meet the 
requirements of the ASME BPV Code as specified in this paragraph. Each 
manufacturing license, standard design approval, and design 
certification application under Part 52 of this chapter and each 
combined license for a utilization facility is subject to the following 
conditions.
    (1) Standards requirement for Quality Group C components. For a 
nuclear power plant whose application for a construction permit under 
this part, or a combined license or manufacturing license under Part 52 
of this chapter, docketed after May 14, 1984, or for an application for 
a standard design approval or a standard design certification docketed 
after May 14, 1984, components classified Quality Group C\9\ must meet 
the requirements for Class 3 components in Section III of the ASME BPV 
Code.
    (2) Quality Group C applicable Code and Code Cases and conditions 
on their use. The Code edition, addenda, and optional ASME Code Cases 
to be applied to the systems and components identified in paragraph 
(e)(1) of this section must be determined by the rules of paragraph 
NCA-1140, subsection NCA of Section III of the ASME BPV Code, subject 
to the following conditions:
    (i) Quality Group C condition: Code edition and addenda. The 
edition and addenda must be those incorporated by reference in 
paragraph (a)(1)(i) of this section;
    (ii) Quality Group C condition: Earliest edition and addenda for 
components. The ASME Code provisions applied to the systems and 
components may be dated no earlier than the 1980 Edition; and
    (iii) Quality Group C condition: Use of Code Cases. The optional 
Code Cases

[[Page 37915]]

must be those listed in NRC Regulatory Guide 1.84 that is incorporated 
by reference in paragraph (a)(3)(i) of this section.
    (f) Inservice testing requirements. Systems and components of 
boiling and pressurized water-cooled nuclear power reactors must meet 
the requirements of the ASME BPV Code and ASME Code for Operation and 
Maintenance of Nuclear Power Plants as specified in this paragraph. 
Each operating license for a boiling or pressurized water-cooled 
nuclear facility is subject to the following conditions. Each combined 
license for a boiling or pressurized water-cooled nuclear facility is 
subject to the following conditions, but the conditions in paragraphs 
(f)(4), (f)(5), and (f)(6) of this section must be met only after the 
Commission makes the finding under Sec.  52.103(g) of this chapter. 
Requirements for inservice inspection of Class 1, Class 2, Class 3, 
Class MC, and Class CC components (including their supports) are 
located in Sec.  50.55a(g).
    (1) Inservice testing requirements for older plants (pre-1971 CPs). 
For a boiling or pressurized water-cooled nuclear power facility whose 
construction permit was issued prior to January 1, 1971, pumps and 
valves must meet the test requirements of paragraphs (f)(4) and (f)(5) 
of this section to the extent practical. Pumps and valves that are part 
of the reactor coolant pressure boundary must meet the requirements 
applicable to components that are classified as ASME Code Class 1. 
Other pumps and valves that perform a function to shut down the reactor 
or maintain the reactor in a safe shutdown condition, mitigate the 
consequences of an accident, or provide overpressure protection for 
safety-related systems (in meeting the requirements of the 1986 
Edition, or later, of the BPV or OM Code) must meet the test 
requirements applicable to components that are classified as ASME Code 
Class 2 or Class 3.
    (2) Design and accessibility requirements for performing inservice 
testing in plants with CPs issued between 1971 and 1974. For a boiling 
or pressurized water-cooled nuclear power facility whose construction 
permit was issued on or after January 1, 1971, but before July 1, 1974, 
pumps and valves that are classified as ASME Code Class 1 and Class 2 
must be designed and provided with access to enable the performance of 
inservice tests for operational readiness set forth in editions and 
addenda of Section XI of the ASME BPV incorporated by reference in 
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases 
listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 
1.192, Revision 1, that are incorporated by reference in paragraphs 
(a)(3)(ii) and (a)(3)(iii) of this section, respectively) in effect 6 
months before the date of issuance of the construction permit. The 
pumps and valves may meet the inservice test requirements set forth in 
subsequent editions of this Code and addenda that are incorporated by 
reference in paragraph (a)(1)(ii) of this section (or the optional ASME 
Code Cases listed in NRC Regulatory Guide 1.147, Revision 17; or 
Regulatory Guide 1.192, Revision 1, that are incorporated by reference 
in paragraphs (a)(3)(ii) and (a)(3)(iii) of this section, 
respectively), subject to the applicable conditions listed therein.
    (3) Design and accessibility requirements for performing inservice 
testing in plants with CPs issued after 1974. For a boiling or 
pressurized water-cooled nuclear power facility whose construction 
permit under this part or design approval, design certification, 
combined license, or manufacturing license under Part 52 of this 
chapter was issued on or after July 1, 1974:
    (i)-(ii) [Reserved]
    (iii) IST design and accessibility requirements: Class 1 pumps and 
valves.
    (A) Class 1 pumps and valves: first provision. In facilities whose 
construction permit was issued before November 22, 1999, pumps and 
valves that are classified as ASME Code Class 1 must be designed and 
provided with access to enable the performance of inservice testing of 
the pumps and valves for assessing operational readiness set forth in 
the editions and addenda of Section XI of the ASME BPV Code 
incorporated by reference in paragraph (a)(1)(ii) of this section (or 
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, 
Revision 17, or Regulatory Guide 1.192, Revision 1, that are 
incorporated by reference in paragraphs (a)(3)(ii) and (a)(3)(iii) of 
this section, respectively) applied to the construction of the 
particular pump or valve or the summer 1973 Addenda, whichever is 
later.
    (B) Class1 pumps and valves: second provision. In facilities whose 
construction permit under this part, or design certification, design 
approval, combined license, or manufacturing license under Part 52 of 
this chapter, issued on or after November 22, 1999, pumps and valves 
that are classified as ASME Code Class 1 must be designed and provided 
with access to enable the performance of inservice testing of the pumps 
and valves for assessing operational readiness set forth in editions 
and addenda of the ASME OM Code (or the optional ASME Code Cases listed 
in NRC Regulatory Guide 1.192, Revision 1, that are incorporated by 
reference in paragraph (a)(3)(iii) of this section), incorporated by 
reference in paragraph (a)(1)(iv) of this section at the time the 
construction permit, combined license, manufacturing license, design 
certification, or design approval is issued.
    (iv) IST design and accessibility requirements: Class 2 and 3 pumps 
and valves.
    (A) Class 2 and 3 pumps and valves: first provision. In facilities 
whose construction permit was issued before November 22, 1999, pumps 
and valves that are classified as ASME Code Class 2 and Class 3 must be 
designed and be provided with access to enable the performance of 
inservice testing of the pumps and valves for assessing operational 
readiness set forth in the editions and addenda of Section XI of the 
ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this 
section (or the optional ASME Code Cases listed in NRC Regulatory Guide 
1.147, Revision 17, that are incorporated by reference in paragraph 
(a)(3)(ii) of this section) applied to the construction of the 
particular pump or valve or the Summer 1973 Addenda, whichever is 
later.
    (B) Class 2 and 3 pumps and valves: second provision. In facilities 
whose construction permit under this part, or design certification, 
design approval, combined license, or manufacturing license under Part 
52 of this chapter, issued on or after November 22, 1999, pumps and 
valves that are classified as ASME Code Class 2 and 3 must be designed 
and provided with access to enable the performance of inservice testing 
of the pumps and valves for assessing operational readiness set forth 
in editions and addenda of the ASME OM Code (or the optional ASME OM 
Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that are 
incorporated by reference in paragraph (a)(3)(iii) of this section), 
incorporated by reference in paragraph (a)(1)(iv) of this section at 
the time the construction permit, combined license, or design 
certification is issued.
    (v) IST design and accessibility requirements: Meeting later IST 
requirements. All pumps and valves may meet the test requirements set 
forth in subsequent editions of codes and addenda or portions thereof 
that are incorporated by reference in paragraph (a) of this section, 
subject to the conditions listed in paragraph (b) of this section.

[[Page 37916]]

    (4) Inservice testing standards requirement for operating plants. 
Throughout the service life of a boiling or pressurized water-cooled 
nuclear power facility, pumps and valves that are classified as ASME 
Code Class 1, Class 2, and Class 3 must meet the inservice test 
requirements (except design and access provisions) set forth in the 
ASME OM Code and addenda that become effective subsequent to editions 
and addenda specified in paragraphs (f)(2) and (f)(3) of this section 
and that are incorporated by reference in paragraph (a)(1)(iv) of this 
section, to the extent practical within the limitations of design, 
geometry, and materials of construction of the components.
    (i) Applicable IST Code: Initial 120-month interval. Inservice 
tests to verify operational readiness of pumps and valves, whose 
function is required for safety, conducted during the initial 120-month 
interval must comply with the requirements in the latest edition and 
addenda of the OM Code incorporated by reference in paragraph 
(a)(1)(iv) of this section on the date 12 months before the date of 
issuance of the operating license under this part, or 12 months before 
the date scheduled for initial loading of fuel under a combined license 
under Part 52 of this chapter (or the optional ASME Code Cases listed 
in NRC Regulatory Guide 1.192, Revision 1, that is incorporated by 
reference in paragraph (a)(3)(iii) of this section, subject to the 
conditions listed in paragraph (b) of this section.
    (ii) Applicable IST Code: Successive 120-month intervals. Inservice 
tests to verify operational readiness of pumps and valves, whose 
function is required for safety, conducted during successive 120-month 
intervals must comply with the requirements of the latest edition and 
addenda of the OM Code incorporated by reference in paragraph 
(a)(1)(iv) of this section 12 months before the start of the 120-month 
interval (or the optional ASME Code Cases listed in NRC Regulatory 
Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that 
are incorporated by reference in paragraphs (a)(3)(ii) and (a)(3)(iii) 
of this section, respectively), subject to the conditions listed in 
paragraph (b) of this section.
    (iii) [Reserved]
    (iv) Applicable IST Code: Use of later Code editions and addenda. 
Inservice tests of pumps and valves may meet the requirements set forth 
in subsequent editions and addenda that are incorporated by reference 
in paragraph (a)(1)(iv) of this section, subject to the conditions 
listed in paragraph (b) of this section, and subject to NRC approval. 
Portions of editions or addenda may be used, provided that all related 
requirements of the respective editions or addenda are met.
    (5) Requirements for updating IST programs.
    (i) IST program update: Applicable IST Code editions and addenda. 
The inservice test program for a boiling or pressurized water-cooled 
nuclear power facility must be revised by the licensee, as necessary, 
to meet the requirements of paragraph (f)(4) of this section.
    (ii) IST program update: Conflicting IST Code requirements with 
technical specifications. If a revised inservice test program for a 
facility conflicts with the technical specifications for the facility, 
the licensee must apply to the Commission for amendment of the 
technical specifications to conform the technical specifications to the 
revised program. The licensee must submit this application, as 
specified in Sec.  50.4, at least 6 months before the start of the 
period during which the provisions become applicable, as determined by 
paragraph (f)(4) of this section.
    (iii) IST program update: Notification of impractical IST Code 
requirements. If the licensee has determined that conformance with 
certain Code requirements is impractical for its facility, the licensee 
must notify the Commission and submit, as specified in Sec.  50.4, 
information to support the determination.
    (iv) IST program update: Schedule for completing impracticality 
determinations. Where a pump or valve test requirement by the Code or 
addenda is determined to be impractical by the licensee and is not 
included in the revised inservice test program (as permitted by 
paragraph (f)(4) of this section), the basis for this determination 
must be submitted for NRC review and approval not later than 12 months 
after the expiration of the initial 120-month interval of operation 
from the start of facility commercial operation and each subsequent 
120-month interval of operation during which the test is determined to 
be impractical.
    (6) Actions by the Commission for evaluating impractical and 
augmented IST Code requirements.
    (i) Impractical IST requirements: Granting of relief. The 
Commission will evaluate determinations under paragraph (f)(5) of this 
section that code requirements are impractical. The Commission may 
grant relief and may impose such alternative requirements as it 
determines are authorized by law, will not endanger life or property or 
the common defense and security, and are otherwise in the public 
interest, giving due consideration to the burden upon the licensee that 
could result if the requirements were imposed on the facility.
    (ii) Augmented IST requirements. The Commission may require the 
licensee to follow an augmented inservice test program for pumps and 
valves for which the Commission deems that added assurance of 
operational readiness is necessary.
    (g) Inservice inspection requirements. Systems and components of 
boiling and pressurized water-cooled nuclear power reactors must meet 
the requirements of the ASME BPV Code as specified in this paragraph. 
Each operating license for a boiling or pressurized water-cooled 
nuclear facility is subject to the following conditions. Each combined 
license for a boiling or pressurized water-cooled nuclear facility is 
subject to the following conditions, but the conditions in paragraphs 
(g)(4), (g)(5), and (g)(6) of this section must be met only after the 
Commission makes the finding under Sec.  52.103(g) of this chapter. 
Requirements for inservice testing of Class 1, Class 2, and Class 3 
pumps and valves are located in Sec.  50.55a(f).
    (1) Inservice inspection requirements for older plants (pre-1971 
CPs). For a boiling or pressurized water-cooled nuclear power facility 
whose construction permit was issued before January 1, 1971, components 
(including supports) must meet the requirements of paragraphs (g)(4) 
and (g)(5) of this section to the extent practical. Components that are 
part of the reactor coolant pressure boundary and their supports must 
meet the requirements applicable to components that are classified as 
ASME Code Class 1. Other safety-related pressure vessels, piping, pumps 
and valves, and their supports must meet the requirements applicable to 
components that are classified as ASME Code Class 2 or Class 3.
    (2) Design and accessibility requirements for performing inservice 
inspection in plants with CPs issued between 1971 and 1974. For a 
boiling or pressurized water-cooled nuclear power facility whose 
construction permit was issued on or after January 1, 1971, but before 
July 1, 1974, components (including supports) that are classified as 
ASME Code Class 1 and Class 2 must be designed and be provided with 
access to enable the performance of inservice examination of such 
components (including supports) and must meet the preservice 
examination requirements set forth in editions and addenda of Section 
III or Section XI of the ASME BPV Code incorporated by reference in 
paragraph (a)(1) of this section (or the optional ASME Code

[[Page 37917]]

Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are 
incorporated by reference in paragraph (a)(3)(ii) of this section) in 
effect 6 months before the date of issuance of the construction permit. 
The components (including supports) may meet the requirements set forth 
in subsequent editions and addenda of this Code that are incorporated 
by reference in paragraph (a) of this section (or the optional ASME 
Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are 
incorporated by reference in paragraph (a)(3)(ii) of this section), 
subject to the applicable limitations and modifications.
    (3) Design and accessibility requirements for performing inservice 
inspection in plants with CPs issued after 1974. For a boiling or 
pressurized water-cooled nuclear power facility, whose construction 
permit under this part, or design certification, design approval, 
combined license, or manufacturing license under Part 52 of this 
chapter, was issued on or after July 1, 1974, the following are 
required:
    (i) ISI design and accessibility requirements: Class 1 components 
and supports. Components (including supports) that are classified as 
ASME Code Class 1 must be designed and be provided with access to 
enable the performance of inservice examination of these components and 
must meet the preservice examination requirements set forth in the 
editions and addenda of Section III or Section XI of the ASME BPV Code 
incorporated by reference in paragraph (a)(1) of this section (or the 
optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 
17, that are incorporated by reference in paragraph (a)(3)(ii) of this 
section) applied to the construction of the particular component.
    (ii) ISI design and accessibility requirements: Class 2 and 3 
components and supports. Components that are classified as ASME Code 
Class 2 and Class 3 and supports for components that are classified as 
ASME Code Class 1, Class 2, and Class 3 must be designed and provided 
with access to enable the performance of inservice examination of these 
components and must meet the preservice examination requirements set 
forth in the editions and addenda of Section XI of the ASME BPV Code 
incorporated by reference in paragraph (a)(1)(ii) of this section (or 
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, 
Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) 
of this section) applied to the construction of the particular 
component.
    (iii)-(iv) [Reserved]
    (v) ISI design and accessibility requirements: Meeting later ISI 
requirements. All components (including supports) may meet the 
requirements set forth in subsequent editions of codes and addenda or 
portions thereof that are incorporated by reference in paragraph (a) of 
this section, subject to the conditions listed therein.
    (4) Inservice inspection standards requirement for operating 
plants. Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) that are 
classified as ASME Code Class 1, Class 2, and Class 3 must meet the 
requirements, except design and access provisions and preservice 
examination requirements, set forth in Section XI of editions and 
addenda of the ASME BPV Code (or ASME OM Code for snubber examination 
and testing) that become effective subsequent to editions specified in 
paragraphs (g)(2) and (g)(3) of this section and that are incorporated 
by reference in paragraph (a)(1)(ii) or (a)(1)(iv) for snubber 
examination and testing of this section, to the extent practical within 
the limitations of design, geometry, and materials of construction of 
the components. Components that are classified as Class MC pressure 
retaining components and their integral attachments, and components 
that are classified as Class CC pressure retaining components and their 
integral attachments, must meet the requirements, except design and 
access provisions and preservice examination requirements, set forth in 
Section XI of the ASME BPV Code and addenda that are incorporated by 
reference in paragraph (a)(1)(ii) of this section, subject to the 
condition listed in paragraph (b)(2)(vi) of this section and the 
conditions listed in paragraphs (b)(2)(viii) and (b)(2)(ix) of this 
section, to the extent practical within the limitation of design, 
geometry, and materials of construction of the components.
    (i) Applicable ISI Code: Initial 120-month interval. Inservice 
examination of components and system pressure tests conducted during 
the initial 120-month inspection interval must comply with the 
requirements in the latest edition and addenda of the Code incorporated 
by reference in paragraph (a) of this section on the date 12 months 
before the date of issuance of the operating license under this part, 
or 12 months before the date scheduled for initial loading of fuel 
under a combined license under Part 52 of this chapter (or the optional 
ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when 
using Section XI, or Regulatory Guide 1.192, Revision 1, when using the 
OM Code, that are incorporated by reference in paragraphs (a)(3)(ii) 
and (a)(3)(iii) of this section, respectively), subject to the 
conditions listed in paragraph (b) of this section.
    (ii) Applicable ISI Code: Successive 120-month intervals. Inservice 
examination of components and system pressure tests conducted during 
successive 120-month inspection intervals must comply with the 
requirements of the latest edition and addenda of the Code incorporated 
by reference in paragraph (a) of this section 12 months before the 
start of the 120-month inspection interval (or the optional ASME Code 
Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using 
Section XI, or Regulatory Guide 1.192, Revision 1, when using the OM 
Code, that are incorporated by reference in paragraphs (a)(3)(ii) and 
(a)(3)(iii) of this section), subject to the conditions listed in 
paragraph (b) of this section. However, a licensee whose inservice 
inspection interval commences during the 12 through 18-month period 
after July 21, 2011, may delay the update of their Appendix VIII 
program by up to 18 months after July 21, 2011.
    (iii) Applicable ISI Code: Optional surface examination 
requirement. When applying editions and addenda prior to the 2003 
Addenda of Section XI of the ASME BPV Code, licensees may, but are not 
required to, perform the surface examinations of high-pressure safety 
injection systems specified in Table IWB-2500-1, Examination Category 
B-J, Item Numbers B9.20, B9.21, and B9.22.
    (iv) Applicable ISI Code: Use of subsequent Code editions and 
addenda. Inservice examination of components and system pressure tests 
may meet the requirements set forth in subsequent editions and addenda 
that are incorporated by reference in paragraph (a) of this section, 
subject to the conditions listed in paragraph (b) of this section, and 
subject to Commission approval. Portions of editions or addenda may be 
used, provided that all related requirements of the respective editions 
or addenda are met.
    (v) Applicable ISI Code: Metal and concrete containments. For a 
boiling or pressurized water-cooled nuclear power facility whose 
construction permit under this part or combined license under Part 52 
of this chapter was issued after January 1, 1956, the following are 
required:
    (A) Metal and concrete containments: first provision. Metal 
containment pressure retaining components and their

[[Page 37918]]

integral attachments must meet the inservice inspection, repair, and 
replacement requirements applicable to components that are classified 
as ASME Code Class MC;
    (B) Metal and concrete containments: second provision. Metallic 
shell and penetration liners that are pressure retaining components and 
their integral attachments in concrete containments must meet the 
inservice inspection, repair, and replacement requirements applicable 
to components that are classified as ASME Code Class MC; and
    (C) Metal and concrete containments: third provision. Concrete 
containment pressure retaining components and their integral 
attachments, and the post-tensioning systems of concrete containments, 
must meet the inservice inspections, repair, and replacement 
requirements applicable to components that are classified as ASME Code 
Class CC.
    (5) Requirements for updating ISI programs.
    (i) ISI program update: Applicable ISI Code editions and addenda. 
The inservice inspection program for a boiling or pressurized water-
cooled nuclear power facility must be revised by the licensee, as 
necessary, to meet the requirements of paragraph (g)(4) of this 
section.
    (ii) ISI program update: Conflicting ISI Code requirements with 
technical specifications. If a revised inservice inspection program for 
a facility conflicts with the technical specifications for the 
facility, the licensee must apply to the Commission for amendment of 
the technical specifications to conform the technical specifications to 
the revised program. The licensee must submit this application, as 
specified in Sec.  50.4, at least six months before the start of the 
period during which the provisions become applicable, as determined by 
paragraph (g)(4) of this section.
    (iii) ISI program update: Notification of impractical ISI Code 
requirements. If the licensee has determined that conformance with a 
Code requirement is impractical for its facility the licensee must 
notify the NRC and submit, as specified in Sec.  50.4, information to 
support the determinations. Determinations of impracticality in 
accordance with this section must be based on the demonstrated 
limitations experienced when attempting to comply with the Code 
requirements during the inservice inspection interval for which the 
request is being submitted. Requests for relief made in accordance with 
this section must be submitted to the NRC no later than 12 months after 
the expiration of the initial or subsequent 120-month inspection 
interval for which relief is sought.
    (iv) ISI program update: Schedule for completing impracticality 
determinations. Where the licensee determines that an examination 
required by Code edition or addenda is impractical, the basis for this 
determination must be submitted for NRC review and approval not later 
than 12 months after the expiration of the initial or subsequent 120-
month inspection interval for which relief is sought.
    (6) Actions by the Commission for evaluating impractical and 
augmented ISI Code requirements.
    (i) Impractical ISI requirements: Granting of relief. The 
Commission will evaluate determinations under paragraph (g)(5) of this 
section that code requirements are impractical. The Commission may 
grant such relief and may impose such alternative requirements as it 
determines are authorized by law, will not endanger life or property or 
the common defense and security, and are otherwise in the public 
interest giving due consideration to the burden upon the licensee that 
could result if the requirements were imposed on the facility.
    (ii) Augmented ISI program. The Commission may require the licensee 
to follow an augmented inservice inspection program for systems and 
components for which the Commission deems that added assurance of 
structural reliability is necessary.
    (A) [Reserved]
    (B) Augmented ISI requirements: Submitting containment ISI 
programs. Licensees do not have to submit to the NRC for approval of 
their containment inservice inspection programs that were developed to 
satisfy the requirements of Subsection IWE and Subsection IWL with 
specified conditions. The program elements and the required 
documentation must be maintained on site for audit.
    (C) Augmented ISI requirements: Implementation of Appendix VIII to 
Section XI.
    (1) Appendix VIII and the supplements to Appendix VIII to Section 
XI, Division 1, 1995 Edition with the 1996 Addenda of the ASME BPV Code 
must be implemented in accordance with the following schedule: Appendix 
VIII and Supplements 1, 2, 3, and 8--May 22, 2000; Supplements 4 and 
6--November 22, 2000; Supplement 11--November 22, 2001; and Supplements 
5, 7, and 10--November 22, 2002.
    (2) Licensees implementing the 1989 Edition and earlier editions 
and addenda of IWA-2232 of Section XI, Division 1, of the ASME BPV Code 
must implement the 1995 Edition with the 1996 Addenda of Appendix VIII 
and the supplements to Appendix VIII of Section XI, Division 1, of the 
ASME BPV Code.
    (D) Augmented ISI requirements: Reactor vessel head inspections.
    (1) All licensees of pressurized water reactors must augment their 
inservice inspection program with ASME Code Case N-729-1, subject to 
the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of 
this section. Licensees of existing operating reactors as of September 
10, 2008, must implement their augmented inservice inspection program 
by December 31, 2008. Once a licensee implements this requirement, the 
First Revised NRC Order EA-03-009 no longer applies to that licensee 
and must be deemed to be withdrawn.
    (2) Note 9 of ASME Code Case N-729-1 must not be implemented.
    (3) Instead of the specified ``examination method'' requirements 
for volumetric and surface examinations in Note 6 of Table 1 of Code 
Case N-729-1, the licensee must perform volumetric and/or surface 
examination of essentially 100 percent of the required volume or 
equivalent surfaces of the nozzle tube, as identified by Figure 2 of 
ASME Code Case N-729-1. A demonstrated volumetric or surface leak path 
assessment through all J-groove welds must be performed. If a surface 
examination is being substituted for a volumetric examination on a 
portion of a penetration nozzle that is below the toe of the J-groove 
weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface 
examination must be of the inside and outside wetted surface of the 
penetration nozzle not examined volumetrically.
    (4) By September 1, 2009, ultrasonic examinations must be performed 
using personnel, procedures, and equipment that have been qualified by 
blind demonstration on representative mockups using a methodology that 
meets the conditions specified in paragraphs (g)(6)(ii)(D)(4)(i) 
through (g)(6)(ii)(D)(4)(iv), instead of the qualification requirements 
of Paragraph -2500 of ASME Code Case N-729-1. References herein to 
Section XI, Appendix VIII, must be to the 2004 Edition with no addenda 
of the ASME BPV Code.
    (i) The specimen set must have an applicable thickness 
qualification range of +25 percent to -40 percent for nominal depth 
through-wall thickness. The specimen set must include geometric and 
material conditions that normally require discrimination from

[[Page 37919]]

primary water stress corrosion cracking (PWSCC) flaws.
    (ii) The specimen set must have a minimum of ten (10) flaws that 
provide an acoustic response similar to PWSCC indications. All flaws 
must be greater than 10 percent of the nominal pipe wall thickness. A 
minimum of 20 percent of the total flaws must initiate from the inside 
surface and 20 percent from the outside surface. At least 20 percent of 
the flaws must be in the depth ranges of 10-30 percent through-wall 
thickness and at least 20 percent within a depth range of 31-50 percent 
through-wall thickness. At least 20 percent and no more than 60 percent 
of the flaws must be oriented axially.
    (iii) Procedures must identify the equipment and essential 
variables and settings used for the qualification, in accordance with 
Subarticle VIII-2100 of Section XI, Appendix VIII. The procedure must 
be requalified when an essential variable is changed outside the 
demonstration range as defined by Subarticle VIII-3130 of Section XI, 
Appendix VIII, and as allowed by Articles VIII-4100, VIII-4200, and 
VIII-4300 of Section XI, Appendix VIII. Procedure qualification must 
include the equivalent of at least three personnel performance 
demonstration test sets. Procedure qualification requires at least one 
successful personnel performance demonstration.
    (iv) Personnel performance demonstration test acceptance criteria 
must meet the personnel performance demonstration detection test 
acceptance criteria of Table VIII--S10-1 of Section XI, Appendix VIII, 
Supplement 10. Examination procedures, equipment, and personnel are 
qualified for depth sizing and length sizing when the RMS error, as 
defined by Subarticle VIII-3120 of Section XI, Appendix VIII, of the 
flaw depth measurements, as compared to the true flaw depths, do not 
exceed \1/8\ inch (3 mm) and the root mean square (RMS) error of the 
flaw length measurements, as compared to the true flaw lengths, do not 
exceed \3/8\ inch (10 mm), respectively.
    (5) If flaws attributed to PWSCC have been identified, whether 
acceptable or not for continued service under Paragraphs -3130 or -3140 
of ASME Code Case N-729-1, the re-inspection interval must be each 
refueling outage instead of the re-inspection intervals required by 
Table 1, Note (8), of ASME Code Case N-729-1.
    (6) Appendix I of ASME Code Case N-729-1 must not be implemented 
without prior NRC approval.
    (E) Augmented ISI requirements: Reactor coolant pressure boundary 
visual inspections.
    (1) All licensees of pressurized water reactors must augment their 
inservice inspection program by implementing ASME Code Case N-722-1, 
subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) 
through (g)(6)(ii)(E)(4) of this section. The inspection requirements 
of ASME Code Case N-722-1 do not apply to components with pressure 
retaining welds fabricated with Alloy 600/82/182 materials that have 
been mitigated by weld overlay or stress improvement.
    (2) If a visual examination determines that leakage is occurring 
from a specific item listed in Table 1 of ASME Code Case N-722-1 that 
is not exempted by the ASME Code, Section XI, IWB-1220(b)(1), 
additional actions must be performed to characterize the location, 
orientation, and length of a crack or cracks in Alloy 600 nozzle 
wrought material and location, orientation, and length of a crack or 
cracks in Alloy 82/182 butt welds. Alternatively, licensees may replace 
the Alloy 600/82/182 materials in all the components under the item 
number of the leaking component.
    (3) If the actions in paragraph (g)(6)(ii)(E)(2) of this section 
determine that a flaw is circumferentially oriented and potentially a 
result of primary water stress corrosion cracking, licensees must 
perform non-visual NDE inspections of components that fall under that 
ASME Code Case N-722-1 item number. The number of components inspected 
must equal or exceed the number of components found to be leaking under 
that item number. If circumferential cracking is identified in the 
sample, non-visual NDE must be performed in the remaining components 
under that item number.
    (4) If ultrasonic examinations of butt welds are used to meet the 
NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) of 
this section, they must be performed using the appropriate supplement 
of Section XI, Appendix VIII, of the ASME BPV Code.
    (F) Augmented ISI requirements: Examination requirements for Class 
1 piping and nozzle dissimilar-metal butt welds.
    (1) Licensees of existing, operating pressurized-water reactors as 
of July 21, 2011, must implement the requirements of ASME Code Case N-
770-1, subject to the conditions specified in paragraphs 
(g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the 
first refueling outage after August 22, 2011.
    (2) Full structural weld overlays authorized by the NRC staff may 
be categorized as Inspection Items C or F, as appropriate. Welds that 
have been mitigated by the Mechanical Stress Improvement Process 
(MSIPTM) may be categorized as Inspection Items D or E, as 
appropriate, provided the criteria in Appendix I of the Code Case have 
been met. For ISI frequencies, all other butt welds that rely on Alloy 
82/182 for structural integrity must be categorized as Inspection Items 
A-1, A-2 or B until the NRC staff has reviewed the mitigation and 
authorized an alternative Code Case Inspection Item for the mitigated 
weld, or until an alternative Code Case Inspection Item is used based 
on conformance with an ASME mitigation Code Case endorsed in Regulatory 
Guide 1.147 with conditions, if applicable, and incorporated by 
reference in this section.
    (3) Baseline examinations for welds in Table 1, Inspection Items A-
1, A-2, and B, must be completed by the end of the next refueling 
outage after January 20, 2012. Previous examinations of these welds can 
be credited for baseline examinations if they were performed within the 
re-inspection period for the weld item in Table 1 using Section XI, 
Appendix VIII, requirements and met the Code required examination 
volume of essentially 100 percent. Other previous examinations that do 
not meet these requirements can be used to meet the baseline 
examination requirement, provided NRC approval of alternative 
inspection requirements in accordance with paragraphs (z)(1) or (z)(2) 
of this section is granted prior to the end of the next refueling 
outage after January 20, 2012.
    (4) The axial examination coverage requirements of Paragraph--
2500(c) may not be considered to be satisfied unless essentially 100 
percent coverage is achieved.
    (5) All hot-leg operating temperature welds in Inspection Items G, 
H, J, and K must be inspected each inspection interval. A 25 percent 
sample of Inspection Items G, H, J, and K cold-leg operating 
temperature welds must be inspected whenever the core barrel is removed 
(unless it has already been inspected within the past 10 years) or 20 
years, whichever is less.
    (6) For any mitigated weld whose volumetric examination detects 
growth of existing flaws in the required examination volume that exceed 
the previous IWB-3600 flaw evaluations or new flaws, a report 
summarizing the evaluation, along with inputs, methodologies, 
assumptions, and causes of the new flaw or flaw growth is to be 
provided to the NRC prior to the weld being placed in service other 
than modes 5 or 6.
    (7) For Inspection Items G, H, J, and K, when applying the 
acceptance

[[Page 37920]]

standards of ASME BPV Code, Section XI, IWB-3514, for planar flaws 
contained within the inlay or onlay, the thickness ``t'' in IWB-3514 is 
the thickness of the inlay or onlay. For planar flaws in the balance of 
the dissimilar metal weld examination volume, the thickness ``t'' in 
IWB-3514 is the combined thickness of the inlay or onlay and the 
dissimilar metal weld.
    (8) Welds mitigated by optimized weld overlays in Inspection Items 
D and E are not permitted to be placed into a population to be examined 
on a sample basis and must be examined once each inspection interval.
    (9) Replace the first two sentences of Extent and Frequency of 
Examination for Inspection Item D in Table 1 of Code Case N-770-1 with, 
``Examine all welds no sooner than the third refueling outage and no 
later than 10 years following stress improvement application.'' Replace 
the first two sentences of Note (11)(b)(2) in Code Case N-770-1 with, 
``The first examination following weld inlay, onlay, weld overlay, or 
stress improvement for Inspection Items D through K must be performed 
as specified.''
    (10) General Note (b) to Figure 5(a) of Code Case N-770-1 
pertaining to alternative examination volume for optimized weld 
overlays may not be applied unless NRC approval is authorized under 
paragraphs (z)(1) or (z)(2) of this section.
    (h) Protection and safety systems. Protection systems of nuclear 
power reactors of all types must meet the requirements specified in 
this paragraph. Each combined license for a utilization facility is 
subject to the following conditions.
    (1) [Reserved]
    (2) Protection systems. For nuclear power plants with construction 
permits issued after January 1, 1971, but before May 13, 1999, 
protection systems must meet the requirements stated in either IEEE 
Std. 279, ``Criteria for Protection Systems for Nuclear Power 
Generating Stations,'' or in IEEE Std. 603-1991, ``Criteria for Safety 
Systems for Nuclear Power Generating Stations,'' and the correction 
sheet dated January 30, 1995. For nuclear power plants with 
construction permits issued before January 1, 1971, protection systems 
must be consistent with their licensing basis or may meet the 
requirements of IEEE Std. 603-1991 and the correction sheet dated 
January 30, 1995.
    (3) Safety systems. Applications filed on or after May 13, 1999, 
for construction permits and operating licenses under this part, and 
for design approvals, design certifications, and combined licenses 
under Part 52 of this chapter, must meet the requirements for safety 
systems in IEEE Std. 603-1991 and the correction sheet dated January 
30, 1995.
    (i) through (y) [Reserved]
    (z) Alternatives to codes and standards requirements. Alternatives 
to the requirements of paragraphs (b), (c), (d), (e), (f), (g), and (h) 
of this section or portions thereof may be used when authorized by the 
Director, Office of Nuclear Reactor Regulation, or Director, Office of 
New Reactors, as appropriate. A proposed alternative must be submitted 
and authorized prior to implementation. The applicant or licensee must 
demonstrate that:
    (1) Acceptable level of quality and safety. The proposed 
alternative would provide an acceptable level of quality and safety; or
    (2) Hardship without a compensating increase in quality and safety. 
Compliance with the specified requirements of this section would result 
in hardship or unusual difficulty without a compensating increase in 
the level of quality and safety.
    Footnotes to Sec.  50.55a:

    \1\ For inspections to be conducted once per interval, the 
inspections must be performed in accordance with the schedule in 
Section XI, paragraph IWB-2400, except for plants with inservice 
inspection programs based on a Section XI edition or addenda prior 
to the 1994 Addenda. For plants with inservice inspection programs 
based on a Section XI edition or addenda prior to the 1994 Addenda, 
the inspection must be performed in accordance with the schedule in 
Section XI, paragraph IWB-2400, of the 1994 Addenda.
    2-3 [Reserved]
    \4\ USAS and ASME Code addenda issued prior to the winter 1977 
Addenda are considered to be ``in effect'' or ``effective'' 6 months 
after their date of issuance and after they are incorporated by 
reference in paragraph (a) of this section. Addenda to the ASME Code 
issued after the summer 1977 Addenda are considered to be ``in 
effect'' or ``effective'' after the date of publication of the 
addenda and after they are incorporated by reference in paragraph 
(a) of this section.
    \5\ For ASME Code editions and addenda issued prior to the 
winter 1977 Addenda, the Code edition and addenda applicable to the 
component is governed by the order or contract date for the 
component, not the contract date for the nuclear energy system. For 
the winter 1977 Addenda and subsequent editions and addenda the 
method for determining the applicable Code editions and addenda is 
contained in Paragraph NCA 1140 of Section III of the ASME Code.
    6-8 [Reserved]
    \9\ Guidance for quality group classifications of components 
that are to be included in the safety analysis reports pursuant to 
Sec.  50.34(a) and Sec.  50.34(b) may be found in Regulatory Guide 
1.26, ``Quality Group Classifications and Standards for Water-, 
Steam-, and Radiological-Waste-Containing Components of Nuclear 
Power Plants,'' and in Section 3.2.2 of NUREG-0800, ``Standard 
Review Plan for Review of Safety Analysis Reports for Nuclear Power 
Plants.''

    Dated at Rockville, Maryland, this 7th day of June 2013.

    For the Nuclear Regulatory Commission.
Jennifer L. Uhle,
Deputy Director, Reactor Safety Programs, Office of Nuclear Reactor 
Regulation.
[FR Doc. 2013-15022 Filed 6-21-13; 8:45 am]
BILLING CODE 7590-01-P