[Federal Register Volume 85, Number 51 (Monday, March 16, 2020)]
[Rules and Regulations]
[Pages 14736-14756]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2020-05086]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[NRC-2017-0024]
RIN 3150-AJ93


Approval of American Society of Mechanical Engineers' Code Cases

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its 
regulations to incorporate by reference the latest revisions of three 
regulatory guides approving new, revised, and reaffirmed Code Cases 
published by the American Society of Mechanical Engineers. This action 
allows licensees and applicants to use the Code Cases listed in these 
regulatory guides as voluntary alternatives to engineering standards 
for the construction, inservice inspection, and inservice testing of 
nuclear power plant components. These engineering standards are set 
forth in the American Society of Mechanical Engineers' Boiler and 
Pressure Vessel Codes and American Society of Mechanical Engineers' 
Operation and Maintenance Codes, which are currently incorporated by 
reference into the NRC's regulations. Further, this final rule 
announces the availability of a related regulatory guide, not 
incorporated by reference into the NRC's regulations, that lists Code 
Cases that the NRC has not approved for use.

DATES: This final rule is effective on April 15, 2020. The 
incorporation by reference of certain publications listed in the 
regulation is approved by the Director of the Federal Register as of 
April 15, 2020.

ADDRESSES: Please refer to Docket ID NRC-2017-0024 when contacting the 
NRC about the availability of information for this action. You may 
obtain publicly-available information related to this action by any of 
the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0024. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions contact 
the individuals listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. For 
the convenience of the reader, instructions about obtaining materials 
referenced in this document are provided in the ``Availability of 
Documents'' section.

FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear 
Material Safety and Safeguards, telephone: 301-415-1519, email: 
[email protected]; and Bruce Lin, Office of Nuclear Regulatory 
Research, telephone: 301-415-2446; email: [email protected]. Both are 
staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.

SUPPLEMENTARY INFORMATION: 

Executive Summary

A. Need for the Regulatory Action

    The purpose of this regulatory action is to incorporate by 
reference into the NRC's regulations the latest revisions of three 
regulatory guides (RGs). The three RGs identify new, revised, and 
reaffirmed Code Cases published by the American Society of Mechanical 
Engineers (ASME), which the NRC has determined are acceptable for use 
as voluntary alternatives to compliance with certain provisions of the 
ASME Boiler and Pressure Vessel (BPV) Code and ASME Operation and 
Maintenance (OM) Code currently incorporated by reference into the 
NRC's regulations.

B. Major Provisions

    The three RGs that the NRC is incorporating by reference are RG 
1.84, ``Design, Fabrication, and Materials Code Case Acceptability, 
ASME Section III,'' Revision 38; RG 1.147, ``Inservice Inspection Code 
Case Acceptability, ASME Section XI, Division 1,'' Revision 19; and RG 
1.192, ``Operation and Maintenance Code Case Acceptability, ASME OM 
Code,'' Revision 3. This final rule allows nuclear power plant 
licensees and applicants for construction permits, operating licenses, 
combined licenses, standard design certifications, standard design 
approvals, and manufacturing licenses to voluntarily use the Code 
Cases, newly listed in these revised RGs, as alternatives to 
engineering standards for the design, construction, inservice 
inspection (ISI) and inservice testing (IST), and repair/replacement of 
nuclear power plant components. In this document, the NRC also notifies 
the public of the availability of RG 1.193, ``ASME Code Cases Not 
Approved for Use,'' Revision 6, which lists Code Cases that the NRC has 
not approved for generic use and will not be incorporated by reference 
into the NRC's regulations.
    The NRC prepared a regulatory analysis (ADAMS Accession No. 
ML19156A178) to identify the benefits and costs associated with this 
final rule. The regulatory analysis prepared for this final rule was 
used to determine if the rule is cost-effective, overall, and to help 
the NRC evaluate potentially costly conditions placed on specific 
provisions of the ASME Code Cases, which are the subject of this final 
rule. In addition, qualitative factors to be considered in the NRC's 
rulemaking decision are considered in the regulatory analysis. The 
analysis concluded that this rule would result in net savings to the 
industry and the NRC. Table 1 shows the estimated total net benefit 
relative to the regulatory baseline, the quantitative benefits outweigh 
the costs by a range from approximately $6.34 million (7 percent net 
present value (NPV)) to $7.20 million (3 percent NPV).

[[Page 14737]]



                                          Table 1--Cost Benefit Summary
----------------------------------------------------------------------------------------------------------------
                                                                            Total averted costs (costs)
                            Attribute                            -----------------------------------------------
                                                                   Undiscounted       7% NPV          3% NPV
----------------------------------------------------------------------------------------------------------------
Industry Implementation.........................................              $0              $0              $0
Industry Operation..............................................       5,620,000       4,470,000       5,080,000
                                                                 -----------------------------------------------
    Total Industry Costs........................................       5,620,000       4,470,000       5,080,000
NRC Implementation..............................................               0               0               0
NRC Operation...................................................       2,350,000       1,870,000       2,120,000
                                                                 -----------------------------------------------
    Total NRC Cost..............................................       2,350,000       1,870,000       2,120,000
                                                                 ===============================================
        Net.....................................................       7,970,000       6,340,000       7,200,000
----------------------------------------------------------------------------------------------------------------

    The regulatory analysis also considered the following qualitative 
considerations: (1) Flexibility and decreased uncertainty for licensees 
when making modifications or preparing to perform ISI or IST; (2) 
consistency with the provisions of the National Technology Transfer and 
Advancement Act of 1995 (NTTAA), which encourages Federal regulatory 
agencies to consider adopting voluntary consensus standards as an 
alternative to de novo agency development of standards affecting an 
industry; (3) consistency with the NRC's policy of evaluating the 
latest versions of consensus standards in terms of their suitability 
for endorsement by regulations and regulatory guides; and (4) 
consistency with the NRC's goal to harmonize with international 
standards to improve regulatory efficiency for both the NRC and 
international standards groups.
    The regulatory analysis concludes that this final rule should be 
adopted because it is justified when integrating the cost-beneficial 
quantitative results and the positive and supporting nonquantitative 
considerations in the decision.

Table of Contents

I. Background
II. Discussion
    A. ASME Code Cases Approved for Unconditional Use
    B. ASME Code Cases Approved for Use With Conditions
    1. ASME BPV Code, Section III Code Cases (RG 1.84)
    2. ASME BPV Code, Section XI Code Cases (RG 1.147)
    3. ASME OM Code Cases (RG 1.192)
    C. ASME Code Cases not Approved for Use (RG 1.193)
III. Opportunities for Public Participation
IV. Public Comment Analysis
V. Section-by-Section Analysis
VI. Regulatory Flexibility Certification
VII. Regulatory Analysis
VIII. Backfitting and Issue Finality
IX. Plain Writing
X. Environmental Assessment and Final Finding of No Significant 
Environmental Impact
XI. Paperwork Reduction Act Statement
XII. Congressional Review Act
XIII. Voluntary Consensus Standards
XIV. Incorporation by Reference--Reasonable Availability to 
Interested Parties
XV. Availability of Guidance
XVI. Availability of Documents

I. Background

    The ASME develops and publishes the ASME BPV Code, which contains 
requirements for the design, construction, and ISI examination of 
nuclear power plant components, and the ASME OM Code,\1\ which contains 
requirements for IST of nuclear power plant components. In response to 
BPV and OM Code user requests, the ASME develops Code Cases that 
provide voluntary alternatives to BPV and OM Code requirements under 
special circumstances.
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    \1\ The editions and addenda of the ASME Code for Operation and 
Maintenance of Nuclear Power Plants have had different titles from 
2005 to 2017, and are referred to collectively in this rule as the 
``OM Code.''
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    The NRC approves the ASME BPV and OM Codes in Sec.  50.55a of title 
10 of the Code of Federal Regulations (10 CFR), ``Codes and 
standards,'' through the process of incorporation by reference. As 
such, each provision of the ASME Codes incorporated by reference into, 
and mandated by, Sec.  50.55a constitutes a legally-binding NRC 
requirement imposed by rule. As noted previously, ASME Code Cases, for 
the most part, represent alternative approaches for complying with 
provisions of the ASME BPV and OM Codes. Accordingly, the NRC 
periodically amends Sec.  50.55a to incorporate by reference the NRC's 
RGs listing approved ASME Code Cases that may be used as voluntary 
alternatives to the BPV and OM Codes.\2\
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    \2\ See Federal Register notification (FRN), ``Incorporation by 
Reference of ASME BPV and OM Code Cases'' (68 FR 40469; July 8, 
2003).
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    This final rule is the latest in a series of rules that incorporate 
by reference new versions of several RGs identifying new, revised, and 
reaffirmed,\3\ and unconditionally or conditionally acceptable ASME 
Code Cases that the NRC approves for use. In developing these RGs, the 
NRC reviews ASME BPV and OM Code Cases, determines the acceptability of 
each Code Case, and publishes its findings in the RGs. The RGs are 
revised periodically as new Code Cases are published by ASME. The NRC 
incorporates by reference the RGs listing acceptable and conditionally 
acceptable ASME Code Cases into Sec.  50.55a. The NRC published a final 
rule dated January 17, 2018 (83 FR 2331) that incorporated by reference 
into Sec.  50.55a the previous versions of these RGs, which are: RG 
1.84, ``Design, Fabrication, and Materials Code Case Acceptability, 
ASME Section III,'' Revision 37; RG 1.147, ``Inservice Inspection Code 
Case Acceptability, ASME Section XI, Division 1,'' Revision 18; and RG 
1.192, ``Operation and Maintenance Code Case Acceptability, ASME OM 
Code,'' Revision 2.
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    \3\ Code Cases are categorized by ASME as one of three types: 
New, revised, or reaffirmed. A new Code Case provides for a new 
alternative to specific ASME Code provisions or addresses a new 
need. The ASME defines a revised Code Case to be 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 RGs that have been incorporated 
by reference into Sec.  50.55a. The ASME defines ``reaffirmed'' as 
an OM Code Case that does not have any change to technical content, 
but includes editorial changes.
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II. Discussion

    This final rule incorporates by reference the latest revisions of 
the NRC's RGs that list ASME BPV and OM Code Cases that the NRC finds 
to be acceptable, or acceptable with NRC-specified conditions 
(``conditionally acceptable''). Regulatory Guide 1.84, Revision 38, 
supersedes the incorporation by reference of Revision

[[Page 14738]]

37; RG 1.147, Revision 19, supersedes the incorporation by reference of 
Revision 18; and RG 1.192, Revision 3, supersedes the incorporation by 
reference of Revision 2.
    The ASME Code Cases that are the subject of this final rule are the 
new and revised Section III and Section XI Code Cases as listed in 
Supplement 11 to the 2010 BPV Code through Supplement 7 to the 2013 BPV 
Code, and the OM Code Cases published at the same time as the 2017 
Edition. Additional Section XI Code Cases published from the 2015 
Edition and the 2017 Edition of the BPV Code are also included at the 
request of the ASME.
    The latest editions and addenda of the ASME BPV and OM Codes that 
the NRC approved for use are referenced in Sec.  50.55a. The ASME also 
publishes Code Cases that provide alternatives to existing Code 
requirements that the ASME developed and approved. This final rule 
incorporates by reference RGs 1.84, 1.147, and 1.192 allowing nuclear 
power plant licensees, and applicants for combined licenses, standard 
design certifications, standard design approvals, and manufacturing 
licenses 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. The ASME publishes OM Code Cases at the same 
time as the specific editions of the ASME OM Code. However, the ASME OM 
Code Cases are published in a separate document from the ASME OM Code 
Editions. The ASME publishes BPV Code Cases in a separate document and 
at a different time from ASME BPV Code Editions. This final rule 
identifies Code Cases by the edition of the ASME BPV Code or ASME OM 
Code under which they were published by ASME. This final rule only 
accepts Code Cases for use in lieu of the specific editions and addenda 
of the ASME BPV and OM Codes incorporated by reference in Sec.  50.55a.
    The following general guidance applies to the use of the ASME Code 
Cases approved in the latest versions of the RGs that are incorporated 
by reference into Sec.  50.55a as part of this final rule. 
Specifically, the use of the Code Cases listed in RGs 1.84, 1.147, and 
1.192 are acceptable with the specified conditions when implementing 
the editions and addenda of the ASME BPV and OM Codes incorporated by 
reference in Sec.  50.55a.
    The approval of a Code Case in an NRC RG 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 is 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 may revise Code Cases for many reasons. For example, the 
ASME may revise a Code Case to incorporate operational examination and 
testing experience or 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. Therefore, when an applicant or 
a licensee initially implements a Code Case, Sec.  50.55a requires that 
the applicant or the licensee implement the most recent version of that 
Code Case, as listed in the RGs incorporated by reference. Code Cases 
superseded by revision are no longer acceptable for new applications 
unless otherwise indicated.
    Section III of the ASME BPV Code applies only to new construction 
(i.e., the edition and addenda to be used in the construction of a 
plant are selected based on the date of the construction permit and are 
not changed thereafter, except voluntarily by the applicant or the 
licensee). Hence, if a Section III Code Case is implemented by an 
applicant or a licensee and a later version of the Code Case is 
incorporated by reference into Sec.  50.55a and listed in the RG, the 
applicant or the licensee may use either version of the Code Case 
(subject, however, to whatever change requirements apply to its 
licensing basis (e.g., Sec.  50.59)) until the next mandatory ISI or 
IST update.
    A licensee's ISI and IST programs must be updated every 10 years to 
the latest edition and addenda of ASME BPV Code, Section XI, and the OM 
Code, respectively, that were incorporated by reference into Sec.  
50.55a and in effect 12 months prior to the start of the next 
inspection and testing interval. Licensees that 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., Licensees choosing to continue using a Code Case 
during the subsequent ISI or IST interval must implement the latest 
version incorporated by reference into Sec.  50.55a and listed in the 
RGs.
    The ASME may annul Code Cases that are no longer required, are 
determined to be inaccurate or inadequate, or have been incorporated 
into the BPV or OM Codes. 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 Sec.  50.55a(z). If 
a Code Case is incorporated by reference into Sec.  50.55a and later a 
revised version is issued by the ASME because experience has shown that 
the design analysis, construction method, examination method, or 
testing method is inadequate; the NRC will amend Sec.  50.55a and the 
relevant RG to remove the approval of the superseded Code Case. 
Applicants and licensees should not begin to implement such superseded 
Code Cases in advance of the rulemaking.

A. ASME Code Cases Approved for Unconditional Use

    The Code Cases discussed in Table I are new, revised, or reaffirmed 
Code Cases which the NRC approves for use without conditions. The table 
identifies the regulatory guide listing the applicable Code Case that 
the NRC approves for use.

[[Page 14739]]



                                                     Table I
----------------------------------------------------------------------------------------------------------------
             Code Case No.                     Published with  supplement                      Title
----------------------------------------------------------------------------------------------------------------
                                   Boiler and Pressure Vessel Code Section III
                                         (addressed in RG 1.84, Table 1)
----------------------------------------------------------------------------------------------------------------
N-60-6................................  11 (2010 Edition).......................  Material for Core Support
                                                                                   Structures, Section III,
                                                                                   Division 1.
N-249-15..............................  7 (2013 Edition)........................  Additional Materials for
                                                                                   Subsection NF, Classes 1, 2,
                                                                                   3, and MC Supports Fabricated
                                                                                   Without Welding, Section III,
                                                                                   Division 1.
N-284-4...............................  11 (2010 Edition).......................  Metal Containment Shell
                                                                                   Buckling Design Methods,
                                                                                   Class MC, TC, and SC
                                                                                   Construction, Section III,
                                                                                   Divisions 1 and 3.
N-520-6...............................  1 (2013 Edition)........................  Alternative Rules for Renewal
                                                                                   of Active or Expired N-type
                                                                                   Certificates for Plants Not
                                                                                   in Active Construction,
                                                                                   Section III, Division 1.
N-801-1...............................  11 (2010 Edition).......................  Rules for Repair of N-Stamped
                                                                                   Class 1, 2, and 3 Components,
                                                                                   Section III, Division 1.
N-822-2...............................  7 (2013 Edition)........................  Application of the ASME
                                                                                   Certification Mark, Section
                                                                                   III, Divisions 1, 2, 3, and
                                                                                   5.
N-833.................................  1 (2013 Edition)........................  Minimum Non-prestressed
                                                                                   Reinforcement in the
                                                                                   Containment Base Mat or Slab
                                                                                   Required for Concrete Crack
                                                                                   Control, Section III,
                                                                                   Division 2.
N-834.................................  3 (2013 Edition)........................  ASTM A988/A988M-11 UNS S31603,
                                                                                   Subsection NB, Class 1
                                                                                   Components, Section III,
                                                                                   Division 1.
N-836.................................  3 (2013 Edition)........................  Heat Exchanger Tube Mechanical
                                                                                   Plugging, Class 1, Section
                                                                                   III, Division 1.
N-841.................................  4 (2013 Edition)........................  Exemptions to Mandatory Post
                                                                                   Weld Heat Treatment (PWHT) of
                                                                                   SA-738 Grade B for Class MC
                                                                                   Applications, Section III,
                                                                                   Division 1.
N-844.................................  5 (2013 Edition)........................  Alternatives to the
                                                                                   Requirements of NB-4250(c),
                                                                                   Section III, Division 1.
----------------------------------------------------------------------------------------------------------------
                                   Boiler and Pressure Vessel Code Section XI
                                        (addressed in RG 1.147, Table 1)
----------------------------------------------------------------------------------------------------------------
N-513-4...............................  6 (2013 Edition)........................  Evaluation of Criteria for
                                                                                   Temporary Acceptance of Flaws
                                                                                   in Moderate Energy Class 2 or
                                                                                   3 Piping, Section XI,
                                                                                   Division 1.
N-528-1...............................  5 (1998 Edition)........................  Purchase, Exchange, or
                                                                                   Transfer of Material Between
                                                                                   Nuclear Plant Sites, Section
                                                                                   XI, Division 1.
N-661-3...............................  6 (2015 Edition)........................  Alternative Requirements for
                                                                                   Wall Thickness Restoration of
                                                                                   Class 2 and 3 Carbon Steel
                                                                                   Piping for Raw Water Service,
                                                                                   Section XI, Division 1.
N-762-1...............................  3 (2013 Edition)........................  Temper Bead Procedure
                                                                                   Qualification Requirements
                                                                                   for Repair/Replacement
                                                                                   Activities without Postweld
                                                                                   Heat Treatment, Section XI,
                                                                                   Division 1.
N-789-2...............................  5 (2015 Edition)........................  Alternative Requirements for
                                                                                   Pad Reinforcement of Class 2
                                                                                   and 3 Moderate Energy Carbon
                                                                                   Steel Piping for Raw Water
                                                                                   Service, Section XI, Division
                                                                                   1.
N-823-1...............................  4 (2013 Edition)........................  Visual Examination, Section
                                                                                   XI, Division 1.
N-839.................................  7 (2013 Edition)........................  Similar and Dissimilar Metal
                                                                                   Welding Using Ambient
                                                                                   Temperature SMAW \1\ Temper
                                                                                   Bead Technique, Section XI,
                                                                                   Division 1.
N-842.................................  4 (2013 Edition)........................  Alternative Inspection Program
                                                                                   for Longer Fuel Cycles,
                                                                                   Section XI, Division 1.
N-853.................................  6 (2015 Edition)........................  PWR \2\ Class 1 Primary Piping
                                                                                   Alloy 600 Full Penetration
                                                                                   Branch Connection Weld Metal
                                                                                   Buildup for Material
                                                                                   Susceptible to Primary Water
                                                                                   Stress Corrosion Cracking,
                                                                                   Section XI, Division 1.
N-854.................................  1 (2015 Edition)........................  Alternative Pressure Testing
                                                                                   Requirements for Class 2 and
                                                                                   3 Components Connected to the
                                                                                   Class 1 Boundary, Section XI,
                                                                                   Division 1.
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                                                     OM Code
                                        (addressed in RG 1.192, Table 1)
----------------------------------------------------------------------------------------------------------------
OMN-16 Revision 2.....................  2017 Edition............................  Use of a Pump Curve for
                                                                                   Testing.
OMN-21................................  2017 Edition............................  Alternative Requirements for
                                                                                   Adjusting Hydraulic
                                                                                   Parameters to Specified
                                                                                   Reference Points.
----------------------------------------------------------------------------------------------------------------
\1\ Shielded metal arc welding.
\2\ Pressurized water reactor.

B. ASME Code Cases Approved for Use With Conditions

    The NRC determined that certain Code Cases, as issued by ASME, are 
generally acceptable for use, but that the alternative requirements 
specified in those Code Cases must be supplemented in order to provide 
an acceptable level of quality and safety. Accordingly, the NRC imposes 
conditions on the use of these Code Cases to modify, limit, or clarify 
their requirements. The conditions specify, for each applicable Code 
Case, 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, listed in 
Table II, are included in Table 2 of RG 1.84, RG 1.147, and RG 1.192. 
This section provides the NRC's evaluation of the Code Cases and the 
reasons for the NRC's conditions. Notations indicate the conditions 
duplicated from previous versions of the RG.
    It should also be noted that this section only addresses those Code 
Cases for which the NRC imposes condition(s), which are listed in the 
RG for the first time.

[[Page 14740]]



                                                    Table II
----------------------------------------------------------------------------------------------------------------
             Code Case No.                     Published with  supplement                      Title
----------------------------------------------------------------------------------------------------------------
                                   Boiler and Pressure Vessel Code Section III
                                         (addressed in RG 1.84, Table 2)
----------------------------------------------------------------------------------------------------------------
N-71-19...............................  0 (2013 Edition)........................  Additional Materials for
                                                                                   Subsection NF, Class 1, 2, 3,
                                                                                   and MC Supports Fabricated by
                                                                                   Welding, Section III,
                                                                                   Division 1.
----------------------------------------------------------------------------------------------------------------
                                   Boiler and Pressure Vessel Code Section XI
                                        (addressed in RG 1.147, Table 2)
----------------------------------------------------------------------------------------------------------------
N-516-4...............................  7 (2013 Edition)........................  Underwater Welding, Section
                                                                                   XI, Division 1.
N-597-3...............................  5 (2013 Edition)........................  Evaluation of Pipe Wall
                                                                                   Thinning, Section XI,
                                                                                   Division 1.
N-606-2...............................  2 (2013 Edition)........................  Similar and Dissimilar Metal
                                                                                   Welding Using Ambient
                                                                                   Temperature Machine GTAW \1\
                                                                                   Temper Bead Technique for BWR
                                                                                   \2\ CRD \3\ Housing/Stub Tube
                                                                                   Repairs, Section XI, Division
                                                                                   1.
N-638-7...............................  2 (2013 Edition)........................  Similar and Dissimilar Metal
                                                                                   Welding Using Ambient
                                                                                   Temperature Machine GTAW
                                                                                   Temper Bead Technique,
                                                                                   Section XI, Division 1.
N-648-2...............................  7 (2013 Edition)........................  Alternative Requirements for
                                                                                   Inner Radius Examinations of
                                                                                   Class 1 Reactor Vessel
                                                                                   Nozzles, Section XI, Division
                                                                                   1.
N-695-1...............................  0 (2015 Edition)........................  Qualification Requirements for
                                                                                   Dissimilar Metal Piping
                                                                                   Welds, Section XI, Division
                                                                                   1.
N-696-1...............................  6 (2013 Edition)........................  Qualification Requirements for
                                                                                   Mandatory Appendix VIII
                                                                                   Piping Examination Conducted
                                                                                   from the Inside Surface,
                                                                                   Section XI, Division 1.
N-702.................................  12 (2001 Edition).......................  Alternative Requirements for
                                                                                   Boiling Water Reactor (BWR)
                                                                                   Nozzle Inner Radius and
                                                                                   Nozzle-to-Shell Welds,
                                                                                   Section XI, Division 1.
N-705 (Errata)........................  11 (2010 Edition).......................  Evaluation Criteria for
                                                                                   Temporary Acceptance of
                                                                                   Degradation in Moderate
                                                                                   Energy Class 2 or 3 Vessels
                                                                                   and Tanks, Section XI,
                                                                                   Division 1.
N-711-1...............................  0 (2017 Edition)........................  Alternative Examination
                                                                                   Coverage Requirements for
                                                                                   Examination Category B-F, B-
                                                                                   J, C-F-1, C-F-2, and R[dash]A
                                                                                   Piping Welds, Section XI,
                                                                                   Division 1.
N-754-1...............................  1 (2013 Edition)........................  Optimized Structural
                                                                                   Dissimilar Metal Weld Overlay
                                                                                   for Mitigation of PWR Class 1
                                                                                   Items, Section XI, Division
                                                                                   1.
N-766-1...............................  1 (2013 Edition)........................  Nickel Alloy Reactor Coolant
                                                                                   Inlay and Onlay for
                                                                                   Mitigation of PWR Full
                                                                                   Penetration Circumferential
                                                                                   Nickel Alloy Dissimilar Metal
                                                                                   Welds in Class 1 Items,
                                                                                   Section XI, Division 1.
N-799.................................  4 (2010 Edition)........................  Dissimilar Metal Welds Joining
                                                                                   Vessel Nozzles to Components,
                                                                                   Section XI, Division 1.
N-824.................................  11 (2010 Edition).......................  Ultrasonic Examination of Cast
                                                                                   Austenitic Piping Welds From
                                                                                   the Outside Surface, Section
                                                                                   XI, Division 1.
N-829.................................  0 (2013 Edition)........................  Austenitic Stainless Steel
                                                                                   Cladding and Nickel Base
                                                                                   Cladding Using Ambient
                                                                                   Temperature Machine GTAW
                                                                                   Temper Bead Technique,
                                                                                   Section XI, Division 1.
N-830.................................  7 (2013 Edition)........................  Direct Use of Master Fracture
                                                                                   Toughness Curve for Pressure-
                                                                                   Retaining Materials of Class
                                                                                   1 Vessels, Section XI,
                                                                                   Division 1.
N-831.................................  0 (2017 Edition)........................  Ultrasonic Examination in Lieu
                                                                                   of Radiography for Welds in
                                                                                   Ferritic Pipe, Section XI,
                                                                                   Division 1.
N-838.................................  2 (2015 Edition)........................  Flaw Tolerance Evaluation of
                                                                                   Cast Austenitic Stainless
                                                                                   Steel Piping, Section XI,
                                                                                   Division 1.
N-843.................................  4 (2013 Edition)........................  Alternative Pressure Testing
                                                                                   Requirements Following
                                                                                   Repairs or Replacements for
                                                                                   Class 1 Piping between the
                                                                                   First and Second Injection
                                                                                   Isolation Valves, Section XI,
                                                                                   Division 1.
N-849.................................  7 (2013 Edition)........................  In situ VT-3 Examination of
                                                                                   Removable Core Support
                                                                                   Structures Without Removal,
                                                                                   Section XI, Division 1.
----------------------------------------------------------------------------------------------------------------
                                                     OM Code
                                        (addressed in RG 1.192, Table 2)
----------------------------------------------------------------------------------------------------------------
OMN-1 Revision 2......................  2017 Edition............................  Alternative Rules for
                                                                                   Preservice and Inservice
                                                                                   Testing of Active Electric
                                                                                   Motor.
OMN-3.................................  2017 Edition............................  Requirements for Safety
                                                                                   Significance Categorization
                                                                                   of Components Using Risk
                                                                                   Insights for Inservice
                                                                                   Testing of LWR \4\ Power
                                                                                   Plants.
OMN-4.................................  2017 Edition............................  Requirements for Risk Insights
                                                                                   for Inservice Testing of
                                                                                   Check Valves at LWR Power
                                                                                   Plants.
OMN-9.................................  2017 Edition............................  Use of a Pump Curve for
                                                                                   Testing.
OMN-12................................  2017 Edition............................  Alternative Requirements for
                                                                                   Inservice Testing Using Risk
                                                                                   Insights for Pneumatically
                                                                                   and Hydraulically Operated
                                                                                   Valve Assemblies in Light-
                                                                                   Water Reactor Power Plants
                                                                                   (OM-Code 1998, Subsection
                                                                                   ISTC).
OMN-13................................  2017 Edition............................  Performance-Based Requirements
                                                                                   for Extending Snubber
                                                                                   Inservice Visual Examination
                                                                                   Interval at [light water
                                                                                   reactor] LWR Power Plants.
OMN-18................................  2017 Edition............................  Alternate Testing Requirements
                                                                                   for Pumps Tested Quarterly
                                                                                   Within 20% of
                                                                                   Design Flow.
OMN-19................................  2017 Edition............................  Alternative Upper Limit for
                                                                                   the Comprehensive Pump Test.
OMN-20................................  2017 Edition............................  Inservice Test Frequency.
----------------------------------------------------------------------------------------------------------------
\1\ Gas tungsten arc welding.
\2\ Boiling water reactor.
\3\ Control rod drive.
\4\ Light water reactor.


[[Page 14741]]

1. ASME BPV Code, Section III Code Cases (RG 1.84)
Code Case N-71-19 [Supplement 0, 2013 Edition]
    Type: Revised.
    Title: Additional Materials for Subsection NF, Class 1, 2, 3, and 
MC Supports Fabricated by Welding, Section III, Division 1.
    The first condition on Code Case N-71-19 is identical to the first 
condition on Code Case N-71-18 that was first approved by the NRC in 
Revision 33 of RG 1.84 in August 2005. The condition stated that the 
maximum measured ultimate tensile strength of the component support 
material must not exceed 170 ksi in view of the susceptibility of high 
strength materials to brittleness and stress corrosion cracking. When 
ASME revised N-71, the Code Case was not modified in a way that would 
make it possible for the NRC to remove the first condition. Therefore, 
the first condition is retained in Revision 38 of RG 1.84.
    The second condition on Code Case N-71-18 is removed because it is 
related to materials of up to 190 ksi and the first condition has an 
ultimate tensile strength limit of 170 ksi on materials. The NRC is not 
aware of any materials listed in this Code Case to which this condition 
would apply, so the condition is removed and the subsequent conditions 
renumbered.
    The second condition on Code Case N-71-19 is an update to the third 
condition on Revision 18 of the Code Case. This condition has been 
modified so that it references the correct sentence and paragraph of 
the revised Code Case and now refers to paragraph 5.2 of the Code Case, 
instead of paragraph 5.5 to reference ``5.3.2.3, `Alternative 
Atmosphere Exposure Time Periods Established by Test,' of the AWS 
[American Welding Society] D1.1 Code for the evidence presented to and 
accepted by the Authorized Inspector concerning exposure of electrodes 
for a longer period of time.'' The basis for this change is that the 
paragraph of the Code Case identified by this condition has been 
renumbered and is now 5.2. When ASME revised N-71, the Code Case was 
not modified in a way that would make it possible for the NRC to remove 
the second condition. Therefore, the second condition is retained in 
Revision 38 of RG 1.84.
    The third condition on Code Case N-71-19 is substantively the same 
as the fourth condition on Code Case N-71-18 that was first approved by 
the NRC in Revision 33 of RG 1.84 in August 2005, except that it now 
references the renumbered paragraphs of the revised Code Case. The 
condition now states that paragraph 16.2.2 of Code Case N-71-19 is not 
acceptable as written and must be replaced with the following: ''When 
not exempted by 16.2.1 above, the post weld heat treatment must be 
performed in accordance with NF-4622 except that ASTM A-710 Grade A 
Material must be at least 1000 [deg]F (540 [deg]C) and must not exceed 
1150 [deg]F (620 [deg]C) for Class 1 and 2 material and 1175 [deg]F 
(640 [deg]C) for Class 3 material.'' When ASME revised N-71, the Code 
Case was not modified in a way that would make it possible for the NRC 
to remove the third condition. Therefore, the third condition is 
retained in Revision 38 of RG 1.84.
    The fourth condition on Code Case N-71-19 is identical to the fifth 
condition on Code Case N-71-18 that was first approved by the NRC in 
Revision 33 of RG 1.84 in August 2005. The condition stated that the 
new holding time-at-temperature for weld thickness (nominal) must be 30 
minutes for welds \1/2\ inch or less in thickness, 1 hour per inch of 
thickness for welds over \1/2\ inch to 5 inches, and for thicknesses 
over 5 inches, 5 hours plus 15 minutes for each additional inch over 5 
inches. When ASME revised N-71, the Code Case was not modified in a way 
that would make it possible for the NRC to remove the fourth condition. 
Therefore, the fourth condition is retained in Revision 38 of RG 1.84.
    The fifth condition on Code Case N-71-19 is identical to the sixth 
condition on Code Case N-71-18 that was first approved by the NRC in 
Revision 33 of RG 1.84 in August 2005. The condition stated that the 
fracture toughness requirements apply only to piping supports and not 
to Class 1, 2 and 3 component supports. When ASME revised N-71, the 
Code Case was not modified in a way that would make it possible for the 
NRC to remove the fifth condition. Therefore, the fifth condition is 
retained in Revision 38 of RG 1.84.
    The sixth condition is a new condition, which states that when 
welding P-Number materials listed in the Code Case, the corresponding 
S-Number welding requirements shall apply. Previous revisions of the 
Code Case assigned every material listed in the Code Case an S-Number 
designation. Welding requirements for materials in the Code Case are 
specified based on the S-Number. The current version of the Code Case 
was modified to assign corresponding P-Numbers to those Code Case 
materials, which are also listed in ASME Code Section IX and have a P-
Number designation. However, the Code Case was not modified to make 
clear that the Code Case requirements for welding S-Number materials 
are also applicable to the P-Number materials, all of which were 
previously listed with S-Numbers. Therefore, as written, if a user 
applies this Code Case and uses a P-Number material listed in the 
tables, it is not clear that the corresponding S-Number welding 
requirements apply. To clarify the application of S-Number welding 
requirements to P-Number materials, the NRC imposes the sixth condition 
as stated. This new condition does not impose any additional 
restrictions on the use of this Code Case from those placed on the 
previous revisions.
2. ASME BPV Code, Section XI Code Cases (RG 1.147)
Code Case N-516-4 [Supplement 7, 2013 Edition]
    Type: Revised.
    Title: Underwater Welding, Section XI, Division 1.
    The previously approved revision of this Code Case, N-516-3, was 
conditionally accepted in RG 1.147 to require that licensees obtain NRC 
approval in accordance with Sec.  50.55a(z) regarding the technique to 
be used in the weld repair or replacement of irradiated material 
underwater. The rationale for this condition was that it was known that 
materials subjected to high neutron fluence could not be welded without 
cracking (this is discussed in more detail in the next paragraph). 
However, the condition applied to Code Case N-516-3 did not provide any 
guidance on what level of neutron irradiation could be considered a 
threshold for weldability.
    The technical basis for imposing conditions on the welding of 
irradiated materials is that neutrons can generate helium atoms within 
the metal lattice through transmutation of various isotopes of boron 
and/or nickel. At high temperatures, such as those during welding, 
these helium atoms rapidly diffuse though the metal lattice, forming 
helium bubbles. In sufficient concentration, these helium atoms can 
cause grain boundary cracking that occurs in the fusion zones and heat 
affected zones during the heatup/cooldown cycle.
    In the final rule for the 2009-2013 Editions of the ASME Code, the 
NRC adopted conditions that should be applied to Section XI, Article 
IWA-4660 when performing underwater welding on irradiated materials. 
These conditions provide guidance on what level of neutron irradiation 
and/or helium content would require approval by the NRC because of the 
impact of neutron fluence on weldability. These

[[Page 14742]]

conditions provide separate criteria for three generic classes of 
material: Ferritic material, austenitic material other than P-No. 8 
(e.g., nickel based alloys), and austenitic P-No. 8 material (e.g., 
stainless steel alloys). These conditions are currently located in 
Sec.  50.55a(b)(2)(xii). Although these conditions apply to underwater 
welding performed in accordance with IWA-4660, they do not apply to 
underwater welding performed in accordance with Code Case N-516-4.
    Consequently, the NRC approves Code Case N-516-4 with the following 
conditions for underwater welding. The first condition captures the 
Sec.  50.55a(b)(2)(xii) requirement for underwater welding of ferritic 
materials, and states that licensees must obtain NRC approval in 
accordance with Sec.  50.55a(z) regarding the welding technique to be 
used prior to performing welding on ferritic material exposed to fast 
neutron fluence greater than 1 x 10\17\ n/cm\2\ (E > 1 MeV). The second 
condition captures the Sec.  50.55a(b)(2)(xii) requirement for 
underwater welding of austenitic material other than P-No. 8, and 
states that licensees must obtain NRC approval in accordance with Sec.  
50.55a(z) regarding the welding technique to be used prior to 
performing welding on austenitic material other than P-No. 8, exposed 
to thermal neutron fluence greater than 1 x 10\17\ n/cm\2\ (E < 0.5 
eV). The third condition captures the Sec.  50.55a(b)(2)(xii) 
requirement for underwater welding of austenitic P-No. 8 material, and 
states that licensees must obtain NRC approval in accordance with Sec.  
50.55a(z) regarding the welding technique to be used prior to 
performing welding on austenitic P-No. 8 material exposed to thermal 
neutron fluence greater than 1 x 10\17\ n/cm\2\ (E < 0.5 eV) and 
measured or calculated helium concentration of the material greater 
than 0.1 atomic parts per million.
Code Case N-597-3 [Supplement 5, 2013 Edition]
    Type: Revised.
    Title: Evaluation of Pipe Wall Thinning, Section XI, Division 1.
    The NRC revised the conditions to clarify their intent. The 
conditions on N-597-3 are all carryovers from the previous version of 
this Code Case N-597-2. The first condition on Code Case N-597-3 
addresses the NRC's concerns regarding how the corrosion rate and 
associated uncertainties will be determined when N-597-3 is applied to 
evaluate the wall thinning in pipes for degradation mechanisms other 
than flow accelerated corrosion. Therefore, the NRC imposes a condition 
that requires the corrosion rate be reviewed and approved by the NRC 
prior to the use of the Code Case.
    The second condition on Code Case N-597-3 has two parts that allow 
the use of this Code Case to mitigate flow accelerated corrosion, but 
only if both of the requirements of the condition are met. Due to the 
difficulty inherent in calculating wall thinning, the first part of 
Condition 2 requires that the use of N-597-3 on flow-accelerated 
corrosion piping must be supplemented by the provisions of Electric 
Power Research Institute (EPRI) Nuclear Safety Analysis Center Report 
202L- 2, ``Recommendations for an Effective Flow Accelerated Corrosion 
Program,'' April 1999, which contain rigorous provisions to minimize 
wall thinning.
    The first part of Condition 2 (i.e., (2)(a)) on Code Case N-597-3 
is identical to the first condition on Code Case N-597-2 that was first 
approved by the NRC in Revision 15 of RG 1.147 in October 2007. The 
condition stated that the Code Case must be supplemented by the 
provisions of EPRI Nuclear Safety Analysis Center Report (NSAC) 202L- 
2, ``Recommendations for an Effective Flow Accelerated Corrosion 
Program'' (Ref. 7), April 1999, for developing the inspection 
requirements, the method of predicting the rate of wall thickness loss, 
and the value of the predicted remaining wall thickness. As used in 
NSAC-202L-R2, the term ``should'' is to be applied as ''shall'' (i.e., 
a requirement). When ASME revised N-597, the Code Case was not modified 
in a way that would make it possible for the NRC to remove the first 
part of Condition 2. Therefore, the first part of Condition 2 is 
retained in Revision 19 of RG 1.147.
    The second part of Condition 2 (i.e., (2)(b)) on Code Case N-597-3 
is identical to the second condition on Code Case N-597-2 that was 
first approved by the NRC in Revision 15 of RG 1.147 in October 2007. 
The condition stated that components affected by flow-accelerated 
corrosion to which this Code Case are applied must be repaired or 
replaced in accordance with the construction code of record and owner's 
requirements or a later NRC approved edition of Section III, ''Rules 
for Construction of Nuclear Power Plant Components,'' of the ASME Code 
prior to the value of tp reaching the allowable minimum wall 
thickness, tmin, as specified in -3622.1(a)(1) of the Code 
Case. Alternatively, use of the Code Case is subject to NRC review and 
approval per Sec.  50.55a(z). When ASME revised N-597, the Code Case 
was not modified in a way that would make it possible for the NRC to 
remove the second part of Condition 2. Therefore, the second part of 
Condition 2 is retained in Revision 19 of RG 1.147.
    The third condition on Code Case N-597-3 is identical to the fourth 
condition on Code Case N-597-2 that was first approved by the NRC in 
Revision 15 of RG 1.147 in October 2007. The condition stated that for 
those components that do not require immediate repair or replacement, 
the rate of wall thickness loss is to be used to determine a suitable 
inspection frequency, so that repair or replacement occurs prior to 
reaching allowable minimum wall thickness. When ASME revised N-597, the 
Code Case was not modified in a way that would make it possible for the 
NRC to remove the third condition. Therefore, the third condition is 
retained in Revision 19 of RG 1.147.
    The fourth condition on Code Case N-597-3 is updated from the sixth 
condition on Code Case N-597-2 that was first approved by the NRC in 
Revision 17 of RG 1.147 in August 2014. This condition allows the use 
of Code Case N-597-3 to calculate wall thinning for moderate-energy 
Class 2 and 3 piping (using criteria in Code Case N-513-2) for 
temporary acceptance (until the next refueling outage). When ASME 
revised N-597, the Code Case was not modified in a way that would make 
it possible for the NRC to remove the fourth condition. Therefore, the 
fourth condition is retained in Revision 19 of RG 1.147.
    The fifth condition is also updated from the sixth condition on 
Code Case N-597-2 that was first approved by the NRC in Revision 17 of 
RG 1.147 in August 2014. This condition prohibits the use of this Code 
Case in evaluating through-wall leakage in high energy piping due to 
the consequences and safety implications associated with pipe failure. 
When ASME revised N-597, the Code Case was not modified in a way that 
would make it possible for the NRC to remove the fifth condition. 
Therefore, the fifth condition is retained in Revision 19 of RG 1.147.
Code Case N-606-2 [Supplement 2, 2013 Edition]
    Type: Revised.
    Title: Similar and Dissimilar Metal Welding Using Ambient 
Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stub 
Tube Repairs, Section XI, Division 1.
    The condition on Code Case N-606-2 is identical to the condition on 
Code Case N-606-1 that was first approved by the NRC in Revision 13 of 
RG 1.147 in January 2004. The condition stated that prior to welding, 
an examination or

[[Page 14743]]

verification must be performed to ensure proper preparation of the base 
metal, and that the surface is properly contoured so that an acceptable 
weld can be produced. This verification is required to be in the 
welding procedure. When ASME revised N-606, the Code Case was not 
modified in a way that would make it possible for the NRC to remove the 
condition. Therefore, the condition is retained in Revision 19 of RG 
1.147.
Code Case N-638-7 [Supplement 2, 2013 Edition]
    Type: Revised.
    Title: Similar and Dissimilar Metal Welding Using Ambient 
Temperature Machine GTAW Temper Bead Technique, Section XI, Division 1.
    The condition on Code Case N-638-7 is identical to the condition on 
Code Case N-638-6 that was first approved by the NRC in Revision 18 of 
RG 1.147 in the January 2018 final rule and states that demonstration 
for ultrasonic examination of the repaired volume is required using 
representative samples, which contain construction type flaws. When 
ASME revised N-638, the Code Case was not modified in a way that would 
make it possible for the NRC to remove the condition. Therefore, the 
condition is retained in Revision 19 of RG 1.147.
Code Case N-648-2 [Supplement 7, 2013 Edition]
    Type: Revised.
    Title: Alternative Requirements for Inner Radius Examinations of 
Class 1 Reactor Vessel Nozzles, Section XI, Division 1.
    The NRC imposes one condition for this Code Case related to 
preservice inspections. The condition on N-648-2 is that this Code Case 
shall not be used to eliminate the preservice or inservice volumetric 
examination of plants with a combined operating license pursuant to 10 
CFR part 52, or a plant that receives its operating license after 
October 22, 2015.
    The requirements for examinations of inner nozzle radii in several 
components were developed in the ASME BPV Code in reaction to the 
discovery of thermal fatigue cracks in the inner-radius section of 
boiling water reactor feedwater nozzles in the late 1970's and early 
1980's. Significant inspections and repairs were required in the late 
1970s and early 1980s to address these problems. The redesign of safe 
end/thermal sleeve configurations and feedwater spargers, coupled with 
changes in operating procedures, has been effective to date. No further 
occurrences of nozzle fatigue cracking have been reported for PWRs or 
BWRs. In addition to operating experience, fatigue analysis for a 
variety of plants shows that there is reasonable assurance that there 
will not be significant cracking at the nozzle inner radii before the 
end of the operating licenses of the nuclear power plants.
    The NRC's position regarding this Code Case is that the required 
preservice volumetric examinations should be performed on all vessel 
nozzles for comparison with volumetric examinations later, if 
indications of flaws are found. Eliminating the volumetric preservice 
or inservice examination is predicated on good operating experience for 
the existing fleet, which has not found any inner radius cracking in 
the nozzles within the scope of the Code Case. In addition to good 
operating experience, flaw tolerance evaluation and fatigue analysis of 
the nozzle inner radius were performed for each of the limiting sizes, 
geometries and operating conditions, including transients for the 
existing fleet that demonstrated large margins to failure and extremely 
low fatigue usage factors. At this time, the new reactor designs have 
no inspection history or operating experience available to support 
eliminating the periodic volumetric examination of the nozzles in 
question. Also, new reactors could have different geometries, sizes and 
operating conditions, including transients, that may not be bounded by 
the analysis performed for the existing fleet, and therefore would not 
have large margins to failure and extremely low fatigue usage factors 
that contributed in removing the requirement of volumetric examination 
of the nozzle inner radius. Use of Code Case N-648-2 would not 
eliminate preservice examinations for the existing fleet since all 
plants have already completed a preservice examination.
Code Case N-695-1 [Supplement 0, 2015 Edition]
    Type: Revised.
    Title: Qualification Requirements for Dissimilar Metal Piping 
Welds, Section XI, Division 1.
    The NRC approves Code Case N-695-1 with the following condition. 
Examiners qualified using the 0.25 root mean square (RMS) error for 
measuring the depths of flaws using N-695-1 are not qualified to depth-
size inner diameter (ID) surface breaking flaws greater than 50 percent 
through-wall in dissimilar metal welds 2.1 inches or greater in 
thickness. When an examiner qualified using N-695-1 measures a flaw as 
greater than 50 percent through-wall in a dissimilar metal weld from 
the ID, the flaw shall be considered to have an indeterminate depth.
    Code Case N-695-1 provides alternative rules for ultrasonic 
examinations of dissimilar metal welds from the inner and outer 
surfaces. Code Case N-695 was developed to allow for examinations from 
the inner surface in ASME Code Section XI editions prior to 2007. 
However, no examination vendor was able to meet the depth-sizing 
requirements of 0.125 inch RMS error of the original N-695. The NRC has 
granted relief to several licensees to allow the use of alternate 
depth-sizing requirements. The NRC reviewed the depth-sizing results at 
the Performance Demonstration Initiative (PDI) for procedures able to 
achieve an RMS error over 0.125 inches but less than 0.25 inches. The 
review found that the examiners tend to oversize small flaws and 
undersize deep flaws. The flaws sized by the examiners as 50 percent 
though-wall or less were accurately or conservatively measured. There 
were, however, some instances of very large flaws being measured as 
significantly smaller than the true state, but they were not measured 
as less than 50 percent through-wall.
    Code Case N-695-1 changes the depth sizing requirements for inner-
surface examinations of test blocks of 2.1 inches or greater thickness 
to 0.25 inches RMS error. This change is in line with the granted 
relief requests and with the NRC's review of the PDI test results.
    The depth-sizing capabilities of the examinations do not provide 
sufficient confidence in the ability of an inspector qualified using a 
0.25 inch RMS error to accurately measure the depth of deep flaws. The 
NRC imposes a condition on Code Case N-695-1 in that any surface-
connected flaw sized over 50 percent through-wall should be considered 
of indeterminate depth.
Code Case N-696-1 [Supplement 6, 2013 Edition]
    Type: Revised.
    Title: Qualification Requirements for Mandatory Appendix VIII 
Piping Examination Conducted from the Inside Surface, Section XI, 
Division 1.
    The NRC approves Code Case N-696-1 with the following condition. 
Examiners qualified using the 0.25 RMS error for measuring the depths 
of flaws using N-696-1 in dissimilar metal or austenitic welds are not 
qualified to depth-size ID surface breaking flaws greater than 50 
percent through-wall in dissimilar metal welds or austenitic weld metal 
welds 2.1 inches or greater in thickness. When a qualified examiner, 
uses N-696-1 and measures a flaw greater than 50 percent through-

[[Page 14744]]

wall in a dissimilar metal weld or austenitic weld metal from the ID, 
the flaw shall be considered to have an indeterminate depth. Code Case 
N-696-1 provides alternative rules for ultrasonic examinations of 
Supplement 2, 3 and 10 welds from the inner and outer surfaces. Code 
Case N-696 was developed to allow for examinations for welds from the 
inner surface in ASME Code Section XI editions prior to 2007. However, 
no examination vendor was able to meet the depth-sizing requirements of 
0.125 inch RMS error required by the original N-696. The NRC granted 
relief to several licensees to allow the use of alternate depth-sizing 
requirements. The NRC reviewed the depth-sizing results at the PDI for 
procedures able to achieve an RMS error over 0.125 inches but less than 
0.25 inches. The review found that the examiners tend to oversize small 
flaws and undersize deep flaws. The flaws sized by the examiners as 50 
percent though-wall or less were accurately or conservatively measured. 
There were, however, some instances of very large flaws being measured 
as significantly smaller than the true state, but they were not 
measured as less than 50 percent through-wall.
    Code Case N-696-1 changes the depth sizing requirements for inner-
surface examinations of test blocks of 2.1 inches or greater thickness 
to 0.25 inch RMS error. This change is consistent with the granted 
relief requests and with the NRC review of the PDI test results. The 
depth-sizing capabilities of the examinations does not provide 
sufficient confidence in the ability of an examiner qualified using a 
0.25 inch RMS error to accurately measure the depth of deep flaws. 
Therefore, the NRC imposes a condition on Code Case N-696-1 that any 
surface-connected flaw sized over 50 percent through-wall should be 
considered of indeterminate depth.
Code Case N-702 [Supplement 12, 2001 Edition]
    Type: Revised.
    Title: Alternative Requirements for Boiling Water Reactor (BWR) 
Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division 1.
    The NRC previously accepted with conditions Code Case N-702 in RG 
1.147, Revision 18. For Revision 19 of RG 1.147 the NRC has revised the 
conditions on Code Case N-702. The original conditions in RG 1.147, 
Revision 17, were consistent with the established review procedure for 
applications for use of Code Case N-702 before August 2014 for the 
original 40 years of operation. The previous conditions on Code Case N-
702 required licensees to prepare and submit for NRC review and 
approval an evaluation demonstrating the applicability of Code Case N-
702 prior to the application of Code Case N-702. Subsequent reviews by 
the NRC of requests to utilize the provisions of Code Case N-702 show 
that all licensees have adequately evaluated the applicability of Code 
Case N-702 during the original 40 years of operation. Therefore, future 
review by the NRC is not needed. For the period of extended operation, 
the application of Code Case N-702 is not approved. Licensees that wish 
to use Code Case N-702 in the period of extended operation may submit 
relief requests based on BWRVIP-241, Appendix A, ``BWR Nozzle Radii and 
Nozzle-to-Vessel Welds Demonstration of Compliance with the Technical 
Information Requirements of the License Renewal Rule (10 CFR 54.21),'' 
approved on April 26, 2017, or plant-specific probabilistic fracture 
mechanics analyses. Therefore, the NRC has revised the RG 1.147, 
Revision 17, condition to reflect these changes.
    Consistent with the safety evaluations for all prior ASME Code Case 
N-702 requests, a condition on visual examination is being added to 
clarify that the NRC is not relaxing the licensees' practice on VT-1 on 
nozzle inner radii.
    The revised conditions on Code Case N-702 states that the 
applicability of Code Case N-702 for the first 40 years of operation 
must be demonstrated by satisfying the criteria in Section 5.0 of NRC 
Safety Evaluation regarding BWRVIP-108 dated December 18, 2007, (ADAMS 
Accession No. ML073600374) or Section 5.0 of NRC Safety Evaluation 
regarding BWRVIP-241 dated April 19, 2013 (ADAMS Accession No. 
ML13071A240).
    The use of Code Case N-702 in the period of extended operation is 
not approved. If VT-1 is used, it shall utilize ASME Code Case N-648-2, 
``Alternative Requirements for Inner Radius Examination of Class 1 
Reactor Vessel Nozzles, Section XI Division 1,'' with the associated 
required conditions specified in Regulatory Guide 1.147.
Code Case N-705 (Errata) [Supplement 11, 2010 Edition]
    Type: Revised.
    Title: Evaluation Criteria for Temporary Acceptance of Degradation 
in Moderate Energy Class 2 or 3 Vessels and Tanks, Section XI, Division 
1.
    The NRC has already accepted Code Case N-705 in Regulatory Guide 
1.147, Revision 16, without conditions. The revised Code Case in 
Supplement 11 contains only editorial changes. However, the NRC has 
identified an area of concern. The Code Case is applicable to the 
temporary acceptance of degradation, which could be a through wall 
leak, and would permit a vessel or tank to leak coolant for 26 months 
without repair or replacement. Paragraph 1(d) of Code Case N-705 states 
that the evaluation period is the operational time for which the 
temporary acceptance criteria are satisfied (i.e., evaluation period <= 
tallow) but not greater than 26 months from the initial 
discovery of the condition. As discussed later in the comment 
resolution section the NRC finds that flaws, which are not through-
wall, that have been evaluated in accordance with the Code Case should 
be allowed to remain in service for the entire length of the period 
evaluated by the Code Case (i.e. up to 26 months). The evaluation 
methods of the Code Case reasonably assure that the structural 
integrity of the component will not be impacted during the period of 
the evaluation. However, the NRC finds that through-wall flaws accepted 
in accordance with the Code Case should be subject to repair/
replacement at the next refueling outage. Therefore, the NRC imposes 
the following condition on Code Case N-705: The ASME Code repair or 
replacement activity temporarily deferred under the provisions of this 
Code Case shall be performed no later than the next scheduled refueling 
outage for through-wall flaws. This is consistent with the current 
regulations for the use of ASME Code, Section XI, Non-Mandatory 
Appendix U which is where the ASME Code has incorporated this case into 
ASME Section XI.
Code Case N-711-1 [Supplement 0, 2017 Edition]
    Type: Revised.
    Title: Alternative Examination Coverage Requirements for 
Examination Category B-F, B-J, C-F-1, C-F-2, and R-A Piping Welds, 
Section XI, Division 1.
    Code Case N-711 was first listed as unacceptable for use by the NRC 
in Revision 3 of RG 1.193 in October 2010. Code Case N-711-1 was 
created to incorporate several NRC conditions for the use of Code Case 
N-711. This Code Case provides requirements for determining an 
alternative required examination volume, which is defined as the volume 
of primary interest based on the postulated degradation mechanism in a 
particular piping weld.
    The NRC finds Code Case N-711-1 acceptable with one condition. The 
Code Case shall not be used to redefine the required examination volume 
for

[[Page 14745]]

preservice examinations or when the postulated degradation mechanism 
for piping welds is primary water stress corrosion cracking (PWSCC) or 
crevice corrosion. For PWSCC, the NRC finds that the examination volume 
must meet the requirements of ASME Code Case N-770-1 as conditioned by 
Sec.  50.55a(g)(6)(ii)(F). For crevice corrosion, the Code Case does 
not define a volume of primary interest and therefore it cannot be used 
for this degradation mechanism. The Code Case requires selection of an 
alternative inspection location within the same risk region or category 
if it will improve the examination coverage of the volume of primary 
interest. Use of the Code Case must be identified in the licensee's 90-
day post outage report of activities identifying the examination 
category, weld number, weld description, percent coverage and a 
description of limitation. The NRC determined that the Code Case 
provides a suitable process for identifying the appropriate volume of 
primary interest based on the degradation mechanism postulated by the 
degradation mechanism analysis, except as noted in the condition.
    The NRC determined that the case should not be used to reduce the 
required examination volume for preservice examinations because for 
newer reactors 50.55a regulations require new plants be designed for 
accessibility for inservice inspection. For preservice examinations 
related to repair/replacements activities ASME Section XI, IWA-4000 
makes it clear that preservice exams are required and IWA-1400 says the 
owner's responsibility includes design and arrangement of system 
components to include adequate access and clearances for conduct of 
examination and tests.
Code Case N-754-1 [Supplement 1, 2013 Edition]
    Type: Revised.
    Title: Optimized Structural Dissimilar Metal Weld Overlay for 
Mitigation of PWR Class 1 Items, Section XI, Division 1.
    The first condition on Code Case N-754-1 is the same as the first 
condition on N-754 that was first approved by the NRC in Revision 18 of 
RG 1.147 in January 2018. The condition stated that the conditions 
imposed on the optimized weld overlay design in the NRC safety 
evaluation for MRP-169, Revision 1-A (ADAMS Accession Nos. ML101620010 
and ML101660468) must be satisfied. When ASME revised N-754, the Code 
Case was not modified in a way that would make it possible for the NRC 
to remove the first condition. Therefore, the first condition is 
retained in Revision 19 of RG 1.147.
    The second condition on Code Case N-754-1 is the same as the second 
condition on N-754 that was first approved by the NRC in Revision 18 of 
RG 1.147 in January 2018. The condition stated that the preservice and 
inservice inspections of the overlaid weld must satisfy 10 CFR 
50.55a(g)(6)(ii)(F). When ASME revised N-754, the Code Case was not 
modified in a way that would make it possible for the NRC to remove the 
second condition. Therefore, the second condition is retained in 
Revision 19 of RG 1.147.
    The proposed rule included a third condition. The NRC has decided 
not to include that condition in the final rule. The basis for removing 
the proposed third condition is discussed in the Public Comment 
Analysis section.
Code Case N-766-1 [Supplement 1, 2013 Edition]
    Type: Revised.
    Title: Nickel Alloy Reactor Coolant Inlay and Onlay for Mitigation 
of PWR Full Penetration Circumferential Nickel Alloy Dissimilar Metal 
Welds in Class 1 Items, Section XI, Division 1.
    Code Case N-766-1 contains provisions for repairing nickel-based 
Alloy 82/182 dissimilar metal butt welds in Class 1 piping using weld 
inlay and onlay. The NRC notes that the Code Case provides adequate 
requirements on the design, installation, pressure testing, and 
examinations of the inlay and onlay. The NRC finds that the weld inlay 
and onlay using the Code Case provides reasonable assurance that 
structural integrity of the repaired pipe will be maintained. However, 
certain provisions of the Code Case are inadequate and therefore the 
NRC imposes five new conditions. The NRC notes that the preservice and 
inservice inspection requirements of inlay and onlay are specified in 
Code Case N-770-1, as stated in Section 3(e) of Code Case N-766-1.
    The first condition on Code Case N-766-1 prohibits the reduction of 
preservice and inservice inspection requirements specified by this Code 
Case for inlays or onlays applied to Alloy 82/182 dissimilar metal 
welds, which contain an axial indication that has a depth of more than 
25 percent of the pipe wall thickness and a length of more than half 
axial width of the dissimilar metal weld, or a circumferential 
indication that has a depth of more than 25 percent of the pipe wall 
thickness and a length of more than 20 percent of the circumference of 
the pipe. Paragraph 1(c)(1) of the Code Case states that:

    . . . Indications detected in the examination of 3(b)(1) that 
exceed the acceptance standards of IWB-3514 shall be corrected in 
accordance with the defect removal requirements of IWA-4000. 
Alternatively, indications that do not meet the acceptance standards 
of IWB-3514 may be accepted by analytical evaluation in accordance 
with IWB-3600 . . .

    This alternative would allow a flaw with a maximum depth of 75 
percent through wall to remain in service in accordance with the ASME 
Code, Section XI, IWB-3643. Even if the inlay or onlay will isolate the 
dissimilar metal weld from the reactor coolant to minimize the 
potential for stress corrosion cracking, the NRC finds that having a 75 
percent flaw in the Alloy 82/182 weld does not provide reasonable 
assurance of structural integrity of the affected pipe. The NRC finds 
that the indication in the Alloy 82/182 weld needs to be limited in 
size to ensure structural integrity of the weld.
    The second condition on Code Case N-766-1 modifies the Code Case to 
require that pipe with any thickness of inlay or onlay must be 
evaluated for weld shrinkage, pipe system flexibility, and additional 
weight of the inlay or onlay. Paragraph 2(e) of the Code Case states 
that:

    . . . If the inlay or onlay deposited in accordance with this 
Case is thicker than 1/8t, where t is the original nominal DMW 
[Dissimilar Metal Weld] thickness, the effects of any change in 
applied loads, as a result of weld shrinkage from the entire inlay 
or onlay, on other items in the piping system (e.g., support loads 
and clearances, nozzle loads, and changes in system flexibility and 
weight due to the inlay or onlay) shall be evaluated. Existing flaws 
previously accepted by analytical evaluation shall be evaluated in 
accordance with IWB-3640 . . .

    The NRC finds that a pipe with any thickness of inlay or onlay must 
be evaluated for weld shrinkage, pipe system flexibility, and 
additional weight of the inlay or onlay.
    The third condition on Code Case N-766-1 sets re-examination 
requirements for inlay or onlay when applied to an Alloy 82/182 
dissimilar metal weld with any indication that the weld exceeds the 
acceptance standards of IWB-3514 and is accepted for continued service 
in accordance with IWB-3132.3 or IWB-3142.4. This condition states that 
the subject weld must be inspected in three successive examinations 
after the installation of the inlay or onlay. The NRC notes that the 
Code Case permits indications exceeding IWB-3514 to remain in service 
after inlay or onlay installation, based on analytical

[[Page 14746]]

evaluation of IWB-3600. The IWB-2420 requires three successive 
examinations for indications that are permitted to remain in service 
per IWB-3600. The Code Case does not discuss the three successive 
examinations. The NRC finds that if an inlay or onlay is applied to an 
Alloy 82/182 dissimilar metal weld that contains an indication that 
exceeds the acceptance standards of IWB-3514 and is accepted for 
continued service in accordance with IWB-3132.3 or IWB-3142.4, the 
subject weld must be inspected in three successive examinations after 
inlay or onlay installation. The NRC imposes this condition to ensure 
that the three successive examinations will be performed such that 
structural integrity of the affected pipe is maintained.
    The fourth condition on Code Case N-766-1 prohibits an inlay or 
onlay with detectable subsurface indication discovered by eddy current 
testing in the acceptance examinations from remaining in service. 
Operational experience has shown that subsurface flaws on Alloy 52 
welds for upper heads may be very near the surface. However, these 
flaws are undetectable by liquid dye penetrant, as there are no surface 
breaking aspects during initial construction. Nevertheless, in multiple 
cases, after a plant goes through one or two cycles of operation, these 
defects become exposed to the primary coolant. The exposure of these 
subsurface defects to primary coolant challenges the effectiveness of 
the Alloy 52 weld mitigation of only 3 mm in total thickness. In the 
repair of reactor vessel upper head nozzle penetrations, these welds 
are inspected each outage after the repair. In order to allow the 
extension of the inspection frequency to that defined by Sec.  
50.55a(g)(6)(ii)(F), the NRC found that all detectable subsurface 
indications by eddy current examination should be removed from the 
Alloy 52 weld layer.
    The fifth condition on Code Case N-766-1 requires that the flaw 
analysis of paragraph 2(d) of the Code Case shall also consider primary 
water stress corrosion cracking growth in the circumferential and axial 
directions, in accordance with IWB-3640. The postulated flaw evaluation 
in the Code Case only requires a fatigue analysis. Conservative generic 
analysis by the NRC has raised the concern that a PWSCC flaw could 
potentially grow through the inner Alloy 52 weld layer and into the 
highly susceptible Alloy 82/182 weld material, to a depth of 75 percent 
through-wall, within the period of reexamination frequency required by 
Sec.  50.55a(g)(6)(ii)(F). Therefore, users of this Code Case will 
verify, for each weld, that a primary water stress corrosion crack will 
not reach a depth of 75 percent through-wall within the required re-
inspection interval.
Code Case N-799 [Supplement 4, 2010 Edition]
    Type: Revised.
    Title: Dissimilar Metal Welds Joining Vessel Nozzles to Components, 
Section XI, Division 1.
    The January 2018 final rule included a response to a public comment 
about Code Case N-799 (83 FR 2348). In the public comment response, the 
NRC described how the conditions on Code Case N-799 were being changed 
to four conditions. However the change to the conditions were not 
reflected in Revision 18 to RG 1.147. As an administrative correction, 
the conditions on N-799 are corrected in Revision 19 to RG 1.147, Table 
2, as described in the January 2018 final rule.
Code Case N-824 [Supplement 11, 2010 Edition]
    Type: New.
    Title: Ultrasonic Examination of Cast Austenitic Piping Welds From 
the Outside Surface, Section XI, Division 1.
    Code Case N-824 is a new Code Case for the examination of cast 
austenitic piping welds from the outside surface. The NRC, using NUREG/
CR-6933 and NUREG/CR-7122, determined that inspections of cast 
austenitic stainless steel (CASS) materials are very challenging, and 
sufficient technical basis exists to condition the Code Case to bring 
the Code Case into agreement with the NUREG/CR reports. The NUREG/CR 
reports also show that CASS materials produce high levels of coherent 
noise. The noise signals can be confusing and mask flaw indications.
    The optimum inspection frequencies for examining CASS components of 
various thicknesses are described in NUREG/CR-6933 and NUREG/CR-7122. 
For this reason, the NRC added a condition to require that ultrasonic 
examinations performed to implement ASME BPV Code Case N-824 on piping 
greater than 1.6 inches thick shall use a phased array search unit with 
a center frequency of 500 kHz with a tolerance of +/- 20 percent.
    The NUREG/CR-6933 shows that the grain structure of CASS can reduce 
the effectiveness of some inspection angles, namely angles including, 
but not limited to, 30 to 55 degrees with a maximum increment of 5 
degrees. For this reason, the NRC imposes a condition to require that 
ultrasonic examinations performed to implement ASME BPV Code Case N-824 
shall use angles including, but not limited to, 30 to 55 degrees with a 
maximum increment of 5 degrees. Therefore, the NRC finds Code Case N-
824 acceptable with the following conditions: (1) Instead of paragraph 
1(c)(1)(-c)(-2), licensees shall use a search unit with a center 
frequency of 500 kHz with a tolerance of  20 percent, and 
(2) instead of Paragraph 1(c)(1)(-d), the search unit must produce 
angles including, but not limited to, 30 to 55 degrees with a maximum 
increment of 5 degrees.
    Existing regulations in Sec.  50.55a(a)(1)(iii)(E) and 
(b)(2)(xxxvii) discuss N-824 and the associated conditions. The NRC 
previously incorporated Code Case N-824 by reference directly in Sec.  
50.55a and provided conditions for its use in a final rule dated July 
18, 2017 (82 FR 32934), to allow licensees to use recent advances in 
inspection technology and perform effective inservice inspection of 
CASS components. Because N-824 will now be incorporated in RG 1.147, 
the existing requirements are redundant. These paragraphs are removed.
Code Case N-829 [Supplement 0, 2013 Edition]
    Type: New.
    Title: Austenitic Stainless Steel Cladding and Nickel Base Cladding 
Using Ambient Temperature Machine GTAW Temper Bead Technique, Section 
XI, Division 1.
    Code Case N-829 is a new Code Case for the use of automatic or 
machine GTAW temper bead technique for the repair of stainless steel 
cladding and nickel-base cladding without the specified preheat or 
postweld heat treatment in Section XI, Paragraph IWA-4411.
    The NRC finds the Code Case acceptable on the condition that the 
provisions of Code Case N-829, paragraph 3(e)(2) or 3(e)(3) may only be 
used when it is impractical to use the interpass temperature 
measurement methods described in 3(e)(1), such as in situations where 
the weldment area is inaccessible (e.g., internal bore welding) or when 
there are extenuating radiological conditions. The NRC determined that 
interpass temperature measurement is critical to obtaining acceptable 
corrosion resistance and/or notch toughness in a weld. Only in areas 
which are totally inaccessible to temperature measurement devices or 
when there are extenuating radiological conditions shall alternate 
methods be allowed such as the calculation method from section 3(e)(2) 
in ASME Code Case N-829 or the weld coupon test method shown in section 
3(e)(3) in ASME Code Case N-829.

[[Page 14747]]

Code Case N-830 [Supplement 7, 2013 Edition]
    Type: New.
    Title: Direct Use of Master Fracture Toughness Curve for Pressure-
Retaining Materials of Class 1 Vessels, Section XI, Division 1.
    Code Case N-830 is a new Code Case introduced in the 2013 Edition 
of the ASME Code. This Code Case outlines the use of a material 
specific master curve as an alternative fracture toughness curve for 
crack initiation, KIC, in Section XI, Division 1, Appendices 
A and G, for Class 1 pressure retaining materials, other than bolting.
    The NRC finds the Code Case acceptable with one condition to 
prohibit the use of the provision in Paragraph (f) of the Code Case 
that allows for the use of an alternative to limiting the lower shelf 
of the 95 percent lower tolerance bound Master Curve toughness, 
KJC-lower 95, to a value consistent with 
the current KIC curve. Code Case N-830 contains provisions 
for using the KJC-lower 95 curve and the 
master curve-based reference temperature To as an 
alternative to the KIC curve and the nil-ductility 
transition reference temperature RTNDT in Appendices A and G 
of the ASME Code, Section XI. To is determined in accordance 
with ASTM International Standard E 1921, ``Standard Test Method for the 
Determination of Reference Temperature, To, for Ferritic 
Steels in the Transition Range,'' from direct fracture toughness 
testing data. The RTNDT is determined in accordance with 
ASME Code, Section III, NB-2330, ``Test Requirements and Acceptance 
Standards,'' from indirect Charpy V-notch testing data, and RG 1.99, 
Revision 2, ``Radiation Embrittlement of Reactor Vessel Materials.'' 
Considering the entire test data at a wide range of T-RTNDT 
(-400 [deg]F to 100 [deg]F), the NRC found that the current 
KIC curve also represents approximately a 95 percent lower 
tolerance bound for the data. Thus, using the KJC-lower 
95 curve based on the Master Curve is acceptable. 
However, since Paragraph (f) provides a significant deviation from the 
KJC-lower 95 curve for (T-To) 
below -115 [deg]F in a non-conservative manner without justification, 
the NRC determined that Paragraph (f) of N-830 must not be applied when 
using N-830.
Code Case N-831 [Supplement 0, 2017 Edition]
    Type: New.
    Title: Ultrasonic Examination in Lieu of Radiography for Welds in 
Ferritic Pipe, Section XI, Division 1.
    Code Case N-831 is a new Code Case, which provides an alternative 
to radiographic testing when it is required by the construction code 
for Section Xl repair/replacement activities. This Code Case describes 
the requirements for inspecting ferritic welds for fabrication flaws 
using Ultrasonic Testing as an alternative to the current requirements 
to use radiography. The Code Case describes the scanning methods, 
recordkeeping and performance demonstration qualification requirements 
for the ultrasonic procedures, equipment, and personnel.
    The NRC finds the Code Case acceptable with the condition that it 
is prohibited for use in new reactor construction. History has shown 
that the combined use of radiographic testing for weld fabrication 
examinations followed by the use of Ultrasonic Testing for pre-service 
inspections and ISI ensures that workmanship is maintained (with 
radiographic testing) while potentially critical planar fabrication 
flaws are not put into service (with Ultrasonic Testing). Until studies 
are completed that demonstrate the ability of Ultrasonic Testing to 
replace radiographic testing (repair/replacement activity), the NRC 
will not generically allow the substitute of Ultrasonic Testing in lieu 
of radiographic testing for weld fabrication examinations. In addition, 
ultrasonic examinations are not equivalent to radiographic examinations 
as they use different physical mechanisms to detect and characterize 
discontinuities. These differences in physical mechanisms result in 
several key differences in sensitivity and discrimination capability. 
As a result of these differences, as well as in consideration of the 
inherent strengths of each of the methods, the two methods are not 
considered to be interchangeable, but are considered complementary. In 
addition, using ultrasonic examinations instead of radiographic testing 
has a particular advantage for operating plants that is not present 
during new reactor construction. Operating plants must take into 
account the additional dose from irradiated plant equipment, which may 
present challenges to keeping radiological dose (man-rem) as low as 
reasonably achievable. In contrast, there is no irradiated plant 
equipment present during new reactor construction. Thus, the additional 
dose that may be received during radiographic testing in operating 
plants may present a hardship or unusually difficulty without an equal 
compensating increase in the level of quality or safety for operating 
plants, but does not justify the reduction in quality assurance for new 
construction. In addition, performing ultrasonic examination under a 
repair or replacement activity for operating plants allows the 
ultrasonic examination results to be available for comparison in future 
inservice inspections that use ultrasonic examination. Therefore, the 
NRC has determined that this Code Case is not acceptable for use on new 
reactor construction.
Code Case N-838 [Supplement 2, 2015 Edition]
    Type: New.
    Title: Flaw Tolerance Evaluation of Cast Austenitic Stainless Steel 
Piping, Section XI, Division 1.
    The NRC approves Code Case N-838 with the following condition: Code 
Case N-838 shall not be used to evaluate flaws in cast austenitic 
stainless steel piping where the delta ferrite content exceeds 25 
percent.
    Code Case N-838 contains provisions for performing a postulated 
flaw tolerance evaluation of ASME Class 1 and 2 CASS piping with delta 
ferrite exceeding 20 percent. The Code Case provides a recommended 
target flaw size for the qualification of nondestructive examination 
methods, along with an approach that may be used to justify a larger 
target flaw size, if needed. The Code Case is intended for the flaw 
tolerance evaluation of postulated flaws in CASS base metal adjacent to 
welds, in conjunction with license renewal commitments. The NRC notes 
that the Code Case is limited in application and provides restrictions 
so that the Code Case will not be misused. For example, the Code Case 
is applicable to portions of Class 1 and 2 piping comprised of SA-351 
statically- or centrifugally-cast Grades CF3, CF3A, CF3M, CF8, CF8A and 
CF8M base metal with delta ferrite exceeding 20 percent and niobium or 
columbium content not greater than 0.2 weight percent. This Code Case 
is limited to be applied to thermally aged CASS material types as 
listed with normal operating temperatures between 500 [deg]F and 662 
[deg]F. The Code Case is not applicable for evaluation of detected 
flaws. Section 3 of the Code Case provides specific analytical 
evaluation procedures for the pipe mean-radius-to-thickness ratio 
greater than 10 and for those with a ratio less than 10. Tables 1 
through 4 provide the maximum tolerable flaw depth-to-thickness ratio 
for circumference and axial flaws.
    However, the NRC finds paragraph 3(c) of the Code Case to be 
inadequate. Paragraph 3(c) specifies that for delta ferrite exceeding 
25 percent, or pipe mean-radius-to-thickness ratio exceeding 10, the 
flaw tolerance evaluation shall be performed, except

[[Page 14748]]

that representative data shall be used to determine the maximum 
tolerable flaw depths applicable to the CASS base metal and mean-
radius-to-thickness ratio, in lieu of Tables 1 through 4 of the Code 
Case.
    The NRC notes that there are insufficient fracture toughness data 
for cast austenitic stainless steel that is greater than 25 percent in 
the open source literature. As such, the NRC needs to review flaw 
tolerance evaluations to ensure that they are performed with adequate 
conservatism. Therefore, the NRC imposes a condition to prohibit the 
use of this Code Case where delta ferrite in cast austenitic stainless 
steel piping exceeds 25 percent.
Code Case N-843 [Supplement 4, 2013 Edition]
    Type: New.
    Title: Alternative Pressure Testing Requirements Following Repairs 
or Replacements for Class 1 Piping between the First and Second 
Inspection Isolation Valves, Section XI, Division 1.
    Code Case N-843 is consistent with alternatives that have been 
granted by the NRC. The NRC is concerned about return lines being 
included that could allow significantly lower pressures to be used on 
Class 1 portions of return lines. Therefore, the NRC imposes a 
condition to ensure that the injection lines are tested at the highest 
pressure of the line's intended safety function. If the portions of the 
system requiring pressure testing are associated with more than one 
safety function, the pressure test and visual examination VT-2 shall be 
performed during a test conducted at the higher of the operating 
pressures for the respective system safety functions.
Code Case N-849 [Supplement 7, 2013 Edition]
    Type: New.
    Title: In Situ VT-3 Examination of Removable Core Support 
Structures Without Removal, Section XI, Division 1.
    Code Case N-849 is a new Code Case introduced in the 2013 Edition 
of ASME Code. This Code Case is meant to provide guidelines for 
allowing the VT-3 inspection requirements of Table IWB-2500-1 for 
preservice or inservice inspections of the core support structures to 
be performed without the removal of the core support structure. The NRC 
finds the Code Case acceptable with two new conditions.
    The first condition on Code Case N-849 limits the use of the Code 
Case to plants that are designed with accessible core support 
structures to allow for in situ inspection. Code Case N-849 allows the 
performance of VT-3 preservice or inservice visual examinations of 
removable core support structures in situ using a remote examination 
system. A provision of the Code Case is that all surfaces accessible 
for examination when the structure is removed shall be accessible when 
the structure is in situ, except for load bearing and contact surfaces, 
which would only be inspected when the core barrel is removed. Designs 
for new reactors, such as certain small modular reactors, may include 
accessibility of the annulus between the core barrel and the reactor 
vessel. Unlike some new reactor designs, currently operating plants 
were not designed to allow in situ VT-3 examinations. There are no 
industry survey results of the current fleet to provide an evaluation 
of operating plant inspection findings. Therefore, applicability to the 
designs of currently operating plants has not been satisfactorily 
addressed.
    The second condition on Code Case N-849 requires that prior to 
initial plant startup, the VT-3 preservice examination shall be 
performed with the core support structure removed, as required by ASME 
Section XI, IWB-2500-1, and shall include all surfaces that are 
accessible when the core support structure is removed, including all 
load bearing and contact surfaces. The NRC has concerns that a 
preservice examination would not be performed on the load bearing and 
contact surfaces even though the surfaces would be accessible prior to 
installing the core support structure. There is also no evidence that 
the in situ examination will achieve the same coverage as the 
examination with the core support structure removed.
3. ASME Operation and Maintenance Code Cases (RG 1.192)
Code Case OMN-1 Revision 2 [2017 Edition]
    Type: Revised.
    Title: Alternative Rules for Preservice and Inservice Testing of 
Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor 
Power Plants.
    The conditions on Code Case OMN-1, Revision 2 [2017 Edition] are 
identical to the conditions on OMN-1 Revision 1 [2012 Edition] that 
were approved by the NRC in Revision 2 of RG 1.192 in January 2018. 
When ASME revised OMN-1, the Code Case was not modified in a way that 
would make it possible for the NRC to remove the conditions. Therefore 
the conditions are retained in Revision 3 of RG 1.192.
Code Case OMN-3 [2017 Edition]
    Type: Reaffirmed.
    Title: Requirements for Safety Significance Categorization of 
Components Using Risk Insights for Inservice Testing of LWR Power 
Plants.
    The conditions on Code Case OMN-3 [2017 Edition] are identical to 
the conditions on OMN-3 [2012 Edition] that were approved by the NRC in 
Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-3, the 
Code Case was not modified in a way that would make it possible for the 
NRC to remove the conditions. Therefore the conditions are retained in 
Revision 3 of RG 1.192.
Code Case OMN-4 [2017 Edition]
    Type: Reaffirmed.
    Title: Requirements for Risk Insights for Inservice Testing of 
Check Valves at LWR Power Plants.
    The conditions on Code Case OMN-4 [2017 Edition] are identical to 
the conditions on OMN-4 [2012 Edition] that were approved by the NRC in 
Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-4, the 
Code Case was not modified in a way that would make it possible for the 
NRC to remove the conditions. Therefore, the conditions are retained in 
Revision 3 of RG 1.192.
Code Case OMN-9 [2017 Edition]
    Type: Reaffirmed.
    Title: Use of a Pump Curve for Testing.
    The conditions on Code Case OMN-9 [2017 Edition] are identical to 
the conditions on OMN-9 [2012 Edition] that were approved by the NRC in 
Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-9, the 
Code Case was not modified in a way that would make it possible for the 
NRC to remove the conditions. Therefore, the conditions are retained in 
Revision 3 of RG 1.192.
Code Case OMN-12 [2017 Edition]
    Type: Reaffirmed.
    Title: Alternative Requirements for Inservice Testing Using Risk 
Insights for Pneumatically and Hydraulically Operated Valve Assemblies 
in Light-Water Reactor Power Plants (OM-Code 1998, Subsection ISTC).
    The conditions on Code Case OMN-12 [2017 Edition] are identical to 
the conditions on OMN-12 [2012 Edition] that were approved by the NRC 
in Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-12, 
the Code Case was not modified in a way that would make it possible for 
the NRC to remove the conditions. Therefore, the

[[Page 14749]]

conditions are retained in Revision 3 of RG 1.192.
Code Case OMN-13 Revision 2 [2017 Edition]
    Type: Reaffirmed.
    Title: Performance-Based Requirements for Extending Snubber 
Inservice Visual Examination Interval at LWR Power Plants.
    The NRC has moved Code Case OMN-13, Revision 2 (2017 Edition) to 
Table 2 in RG 1.192 to clarify its acceptance for use with all editions 
and addenda of the OM Code listed in Sec.  50.55a(a)(1)(iv).
Code Case OMN-18 [2017 Edition]
    Type: Reaffirmed.
    Title: Alternate Testing Requirements for Pumps Tested Quarterly 
Within 20 Percent of Design Flow.
    The conditions on Code Case OMN-18 [2017 Edition] are identical to 
the conditions on OMN-18 [2012 Edition] that were approved by the NRC 
in Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-18, 
the Code Case was not modified in a way that would make it possible for 
the NRC to remove the conditions. Therefore, the conditions are 
retained in Revision 3 of RG 1.192.
Code Case OMN-19 [2017 Edition]
    Type: Reaffirmed.
    Title: Alternative Upper Limit for the Comprehensive Pump Test.
    The conditions on Code Case OMN-19 [2017 Edition] are identical to 
the conditions on OMN-19 [2012 Edition] that were approved by the NRC 
in Revision 2 of RG 1.192 in January 2018. When ASME revised OMN-19, 
the Code Case was not modified in a way that would make it possible for 
the NRC to remove the conditions. Therefore, the conditions are 
retained in Revision 3 of RG 1.192.
Code Case OMN-20 [2017 Edition]
    Type: Reaffirmed.
    Title: Inservice Test Frequency.
    This Code Case is applicable to the editions and addenda of the OM 
Code listed in Sec.  50.55a(a)(1)(iv).
    With the acceptance of Code Case OMN-20 in RG 1.192, Revision 3, 
paragraphs (a)(1)(iii)(G) and (b)(3)(x) in Sec.  50.55a accepting Code 
Case OMN-20 are unnecessary. The paragraphs in Sec.  50.55a are removed 
with this final rule.

C. ASME Code Cases not Approved for Use (RG 1.193)

    The ASME Code Cases that are currently issued by ASME but not 
approved for generic use by the NRC are listed in RG 1.193, ``ASME Code 
Cases not Approved for Use.'' 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. The NRC is not adopting any of the Code Cases listed in RG 
1.193.

III. Opportunities for Public Participation

    The proposed rule and draft RGs were published in the Federal 
Register on August 16, 2018 (83 FR 40685), for a 75-day comment period. 
The public comment period closed on October 30, 2018. The NRC did not 
seek public comments on the draft revision to RG 1.193. Any 
reconsideration for approval by the NRC of such Code Cases will include 
an opportunity for public comment.

IV. Public Comment Analysis

    The NRC received a total of five comment submissions on the 
proposed rule and draft RGs, for a total of 20 comments. The NRC 
reviewed every comment submission and identified 12 unique comments 
requiring the NRC's consideration and response. Comment summaries and 
the NRC's responses are presented in this section. At the beginning of 
each summary, the individual comments represented by the summary are 
identified in the form [XX-YY] where XX represents the Submission ID in 
Table III and YY represents the sequential comment within the 
submission. Multiple comments expressed general support for the 
rulemaking. Those comments are listed at the bottom of Table III, but 
no specific changes were made to the final rule in response to those 
comments.

                                                    Table III
----------------------------------------------------------------------------------------------------------------
                                      Sequential                                                       ADAMS
           Submission ID              comment No.         Commenter              Code case         Accession No.
----------------------------------------------------------------------------------------------------------------
                                    Public Comments To Modify the Rule or RGs
----------------------------------------------------------------------------------------------------------------
NRC-2017-0024-0006................             6-1  Jungbao Zhang........  N-841................     ML18282A102
NRC-2017-0024-0007................             7-1  Glen Palmer..........  OMN-13...............     ML18298A186
NRC-2017-0024-0008................             8-1  Christian Sanna of     n/a..................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................            8-10  Christian Sanna of     N-831................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................            8-11  Christian Sanna of     N-795................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-4  Christian Sanna of     N-702................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-5  Christian Sanna of     N-705................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-7  Christian Sanna of     N-711-1..............     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-8  Christian Sanna of     N-711-1..............     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-9  Christian Sanna of     N-831................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0009................             9-1  Douglas Kull & Carl    N-695-1..............     ML18303A377
                                                     Latiolias of EPRI.
NRC-2017-0024-0009................             9-2  Douglas Kull & Carl    N-711-1..............     ML18303A377
                                                     Latiolias of EPRI.
NRC-2017-0024-0009................             9-3  Douglas Kull & Carl    N-711-1..............     ML18303A377
                                                     Latiolias of EPRI.
NRC-2017-0024-0009................             9-4  Douglas Kull & Carl    N-754-1..............     ML18303A377
                                                     Latiolias of EPRI.
NRC-2017-0024-0009................             9-5  Douglas Kull & Carl    N-831................     ML18303A377
                                                     Latiolias of EPRI.

[[Page 14750]]

 
NRC-2017-0024-0010................            10-1  Justin Wheat of SNO--  N-702................     ML18304A266
                                                     Southern Nuclear
                                                     Operating Company.
----------------------------------------------------------------------------------------------------------------
                                       Public Comments Supporting the Rule
----------------------------------------------------------------------------------------------------------------
NRC-2017-0024-0008................            8-12  Christian Sanna of     n/a..................     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-2  Christian Sanna of     N-661-3, N-789-2, N-      ML18303A362
                                                     ASME Board on          853, and N-854.
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-3  Christian Sanna of     N-516-4, N-695-1, N-      ML18303A362
                                                     ASME Board on          696-1.
                                                     Nuclear Codes and
                                                     Standards.
NRC-2017-0024-0008................             8-6  Christian Sanna of     N-711-1..............     ML18303A362
                                                     ASME Board on
                                                     Nuclear Codes and
                                                     Standards.
----------------------------------------------------------------------------------------------------------------

Regulatory Guide 1.84, Revision 38 (Draft Regulatory Guide (DG) 1345)

Code Case N-841 Exemptions to Mandatory Post Weld Heat Treatment (PWHT) 
of SA-738 Grade B for Class MC Applications Section III, Division 1
    Comment [6-1]: The comment raises issues with the use of shielded 
metal arc welding (SMAW) electrodes identified with a diffusible 
hydrogen content of H-8 or lower and states that, ``Currently, for 
pressure vessels, diffusible hydrogen designator is H4 or lower.'' The 
comment also raises issues with the minimum heat input of 66,000 
Joules/inch (26,000 Joules/Centimeter) and states, ``For ensuring HAZ 
[heat affected zone] properties, the heat input shall be as low as 
possible, normally, 14,000-30,000 Joules/centimeter.'' The comment 
recommends moving N-841 to Table 2 and adding a condition which states, 
``when using the SMAW process the welding electrodes are identified 
with a diffusible hydrogen designator of H4 or lower and the heat input 
shall be specified according to the PQR.''
    NRC Response: The NRC disagrees with this comment. Concerning the 
use of electrodes identified with diffusible hydrogen content of H4 or 
lower, ASME Code, Section III, Subsection NE (Class MC components), 
does not require the use of H4 or lower designated SMAW electrodes. 
Subsection NB (Class 1 components) does require the use of H4 or lower 
designated SMAW electrodes when employing the temper bead welding 
technique at ambient temperature. Code Case N-841 is for Class MC, does 
not entail the use of the temper bead welding technique, nor does it 
permit welding at ambient temperature. For SMAW welding, the Code Case 
requires a minimum preheat of 250 [deg]F.
    Concerning minimum heat input comment, during the development of 
the Code Case, Y-groove testing was performed using the SMAW process. 
The testing performed showed that weld heat input below 66,000 Joules/
inch with a preheat below 250 [deg]F can increase the probability of 
HAZ cracking.
    No change was made to this final rule as a result of this comment.

Regulatory Guide 1.147, Revision 19 (DG-1342)

Generic Comment Clarification of the Term ``Superseded''
    Comment [8-1]: One comment asked whether the word ``superseded'' 
used in RG 1.147, applies to those Code Cases that are superseded by 
ASME or those Code Cases that are listed as superseded in Table 5 of 
Regulatory Guide 1.147. The comment recommended revising the second 
sentence of this paragraph to clarify that the older or superseded 
version of the Code Case, if listed in Table 5, cannot be applied by 
the licensee or applicant for the first time.
    NRC Response: The NRC agrees with this comment. The proposed 
additional text will clarify the information presented in Table 5. The 
introductory paragraph to Table 5 in RG 1.147 has been revised to 
include the statement, ``The versions of the Code Cases listed in Table 
5 cannot be applied by the licensee or applicant for the first time 
after the effective date of this RG.'' at the end of the explanatory 
text above Table 5.
Code Case N-696-1 Qualification Requirements for Mandatory Appendix 
VIII Piping Examinations Conducted From the Inside Surface, Section XI, 
Div. 1
    Condition: Inspectors qualified using the 0.25 RMS error for 
measuring the depths of flaws using N-695-1 are not qualified to depth-
size inner diameter (ID) surface breaking flaws greater than 50 percent 
through-wall in dissimilar metal welds 2.1 inches or greater in 
thickness. When an inspector qualified using N-695-1 measures a flaw as 
greater than 50 percent through-wall in a dissimilar metal weld from 
the ID, the flaw shall be considered to have an indeterminate depth.
    Comment [9-1]: The discussion of the condition as found in the 
Federal Register Vol. 83, No. 159, focused mainly on dissimilar metal 
welds (DMW) whereas the condition defined in DG-1342 applies to the 
coordinated implementation of Supplements 2, 3, & 10 from the ID 
surface. Section 3.3 of the Code Case require users to follow 
Supplement 10 (Alt. CC N-695-1) for DMW and Supplement 3 for ferritic 
welds. As conditioned, Code Case N-695-1, includes depth sizing 
acceptance criteria of 0.25 RMS and Supplement 3 depth sizing 
acceptance criteria remains unchanged at 0.125. As written the proposed 
condition on Code Case N-696-1 would require examiners qualified to 
depth size flaws in ferritic and austenitic welds, from the ID surface, 
to report flaws greater than 50 percent through wall as having an 
indeterminate depth, which is inconsistent with discussion included in 
the Federal Register Vol. 83, No. 159, and in the regulatory analysis 
for the proposed rule.
    NRC Response: The NRC agrees with the comment. The FRN for the 
proposed rule only mentioned dissimilar metal welds when ASME Code Case 
N-696-1 applies to ferritic, dissimilar metal welds, and austenitic 
welds. The condition is intended for procedures, equipment, and 
personnel qualified to examine dissimilar and austenitic welds greater 
than 2.1 inches. In response to this comment, the condition on N-696-1 
in RG 1.147 has been revised to clarify the weld types to which the 
condition applies.

[[Page 14751]]

Code Case N-702 Alternative Requirements for Boiling Water Reactor 
(BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, 
Division 1
    Condition: The applicability of Code Case N-702 for the first 40 
years of operation must be demonstrated by satisfying the criteria in 
Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated 
December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation 
regarding BWRVIP-241 dated April 19, 2013 (ML13071A240). The use of 
Code Case N-702 in the period of extended operation is prohibited.
    Comment (8-4, 10-1): The proposed conditions on Code Case N-702 
state, in part, that ``The use of Code Case N-702 in the period of 
extended operation is prohibited.'' Two comment submissions suggest 
that the proposed condition be revised to provide better guidance to 
licensees on how this case may be used during the period of extended 
operation, rather than to simply prohibit its use. Specifically, one 
comment suggests that the above condition be replaced with the 
following to better describe the explanation provided in the Federal 
Register document for the proposed rule:
    ``The use of Code Case N-702 after the first 40 years of operation 
is not approved. Licensees that wish to use Code Case N-702 after the 
first 40 years of operation may submit relief requests based on BWRVIP-
241, Appendix A, `BWR Nozzle Radii and Nozzle-to-Vessel Welds 
Demonstration of Compliance with the Technical Information Requirements 
of the License Renewal Rule (10 CFR 54.21).' ''
    NRC Response: The NRC disagrees with the comment. Because all 
licensees may propose an alternative to the code requirements under 
Sec.  50.55a(z) ``Alternatives to codes and standards requirements,'' 
there is no need to repeat that option here. The language proposed in 
the comment could be viewed as limiting the potential alternatives that 
could be proposed by licensees.
    No change was made to this final rule as a result of this comment.
Code Case N-705 Evaluation Criteria for Temporary Acceptance of 
Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks Section 
XI, Division 1
    Condition: The ASME Code repair or replacement activity temporarily 
deferred under the provisions of this Code Case shall be performed 
during the next scheduled refueling outage. If a flaw is detected 
during a scheduled shutdown, an ASME Code repair is required before 
plant restart.
    Comment [8-5]: In the proposed rule, the NRC has indicated a 
concern with use of this case to permit a component with through-wall 
leakage to operate for up to 26 months before repairs are made. 
However, the proposed condition applies to all applications of this 
case, including those where through-wall leakage has not occurred. One 
comment suggests that the proposed condition could be revised to read 
as follows to address this concern:

    ``The ASME Code repair or replacement activity temporarily 
deferred under the provisions of this Code Case shall be performed 
during the next scheduled refueling outage for any through-wall 
flaws. If a through-wall flaw is detected during a scheduled 
shutdown, an ASME code repair is required before plant restart.''

    NRC Response: The NRC agrees with the comment. Flaws that are not 
through-wall and have been evaluated in accordance with the Code Case 
should be allowed to remain in service the entire length of the period 
evaluated by the Code Case (i.e., up to 26 months). The evaluation 
methods of the Code Case reasonably assure the structural integrity of 
the component will not be impacted during the period of the evaluation. 
The NRC believes through wall flaws accepted in accordance with the 
Code Case should be subject to repair/replacement at the next refueling 
outage. The NRC also removed the second sentence in the proposed 
condition, which would have required an ASME code repair of the tank 
before plant restart if a through-wall flaw is detected during a 
scheduled shutdown. The NRC finds that the second sentence of the 
proposed condition is not necessary because the time period evaluated 
under the Code Case is greater than the period between refueling 
outages and the evaluation methods of the Code Case reasonably assure 
that the structural integrity of the component will not be impacted 
during that period. In the RG 1.147, the condition on N-705 has been 
revised in response to this comment.
Code Case N-711-1 Alternative Examination Coverage Requirements for 
Examination Category B-F, B-J, C-F-1, C-F-2, and R-A Piping Welds 
Section XI, Division 1
    Condition: Code Case N-711-1 shall not be used to redefine the 
required examination volume for preservice examinations or when the 
postulated degradation mechanism for piping welds is PWSCC, 
Intergranular Stress Corrosion Cracking (IGSCC) or crevice corrosion 
(CC) degradation mechanisms.
    Comment [8-7, 9-2]: Two comment submissions stated that the 
proposed RG 1.147, Table 2, condition should not prohibit the use of 
Code Case N-711-1 for preservice examinations for piping welds where 
use of this case is not prohibited for inservice examination. The 
preservice examination volume serves as a baseline for subsequent 
inservice examinations which should interrogate the same volume.
    NRC Response: The NRC disagrees with this comment in that the Code 
Case should not be applied to new reactors since regulations require 
new plants be designed for accessibility for inservice inspection. For 
preservice examinations related to repair/replacements activities, IWA-
4000 makes it clear that preservice exams are required. IWA-1400 also 
says the owner's responsibility includes design and arrangement of 
system components to include adequate access and clearances for conduct 
of examination and tests.
    No change was made to this final rule as a result of this comment.
    Comment [8-8, 9-3]: Two comment submissions stated that the 
proposed condition, prohibiting the use of this case to redefine the 
required examination volume when the postulated degradation mechanism 
for piping welds is Intergranular Stress Corrosion Cracking (IGSCC), is 
unnecessary for the following reasons:
    1. For boiling water reactor (BWR) plants, this case does not 
provide alternative examination volumes.
    2. For pressurized water reactor (PWR) plants, Table 2 of the case 
requires compliance with the examination requirements of B-F, B-J, C-F-
1, C-F-2, or R-A, as applicable, so this case specifies an appropriate 
volume of primary interest for IGSCC.
    NRC Response: The NRC agrees with this comment. The Code Case 
appropriately requires the correct volume to be examined for IGSCC in 
PWR plants. The condition to Code Case N-711-1 in RG 1.147 has been 
revised in response to these comments.
Code Case N-754-1 Optimized Structural Dissimilar Metal Weld Overlay 
for Mitigation of PWR Class 1 Items, Section XI, Division 1
    Condition: (3) The optimized weld overlay in this Code Case can 
only be installed on an Alloy 82/182 weld where the outer 25 percent of 
weld wall thickness does not contain indications that are greater than 
1/16 inch in length or depth.
    Comment [9-4]: The use of optimized weld overlays is most 
beneficial in applications with large bore components where the outer 
25 percent

[[Page 14752]]

can represent a significant volume of weld metal. One comment stated 
that it is not unreasonable to expect that fabrication flaws that meet 
the original pre-service acceptance standards defined in IWB-3514 to be 
present within the volume of a weld.
    Currently Code Case N-754-1 references Code Case N-770 for the 
acceptance standards for optimized weld overlays. Code Case N-770 
states that the preservice examination acceptance standards of IWB-3514 
shall be met for flaws in the weld overlay material and the outer 25 
percent of the original weld/base material, which is consistent with 
the original ASME Section XI acceptance standards of the original 
structural butt weld.
    Additionally, the current condition refers to ``indications'' that 
are greater than 1/16 inch in length or depth it is important to note 
that indications are not always synonymous with flaws. Indications can 
be attributed to geometric features, metallurgical responses or other 
non-flaw attributes. One comment suggested replacing the word 
indications with the word flaws.
    Another comment stated that the condition limiting the use of this 
Code Case to welds with no indications greater than 1/16 inch in depth 
or length exceeds the original ASME section XI, acceptance standards of 
the weld when it was initially put in service. This condition would 
lead to increase examination time and unnecessary radiation exposure 
due to numerous repairs to remove benign, previously acceptable 
fabrication flaws or other non-relevant indications. These repairs 
could also result in undesirable residual stress profiles in the post 
overlaid weldment that can reduce the functional properties 
(compressive stresses) of the installed overlay. For these reasons, the 
comment submission recommends the elimination of this condition.
    NRC Response: The NRC agrees with these comments. The technical 
basis of the optimized weld overlay in Code Case N-754-1 is that the 
structural integrity of the optimized weld overlay is supported by the 
combination of the outer 25 percent of the original weld and the 
deposited weld overlay on the pipe so that the thickness of the weld 
overlay could be less than the thickness of a full structural weld 
overlay. The Reply Section in Code Case N-754-1 states that it is for 
mitigation of flaws that do not exceed more than 50 percent in depth 
from the inside surface.
    The NRC notes that the ASME Code, Section III, NB-5331(b), 
Ultrasonic Acceptance Standards, requires that indications 
characterized as cracks, lack of fusion, or incomplete penetration are 
unacceptable regardless of length. The NRC understands that the 
hardship of satisfying limiting flaw size in the proposed condition 
would lead to radiation exposure due to repairs to remove fabrication 
flaws prior to weld overlay installation. The NRC also notes that there 
is measurement uncertainty associated with ultrasonic examinations. 
Based on these considerations, the NRC removed the proposed condition 
number 3 from Code Case N-754-1 in RG 1.147.
Code Case N-795 Alternative Requirements for BWR Class 1 System Leakage 
Test Pressure Following Repair/Replacement Activities, Section XI, 
Division 1
    Condition: (1) The use of nuclear heat to conduct the BWR Class 1 
system leakage test is prohibited (i.e., the reactor must be in a non-
critical state), except during refueling outages in which the ASME 
Section XI Category B-P pressure test has already been performed, or at 
the end of mid-cycle maintenance outages fourteen (14) days or less in 
duration. (2) The test condition holding time, after pressurization to 
test conditions, and before the visual examinations commence, shall be 
1 hour for non-insulated components.
    Comment [8-11]: Use of Code Case N-795 is limited to BWR Class 1 
pressure tests following repair/replacement activities and does not 
apply to Class 1 system leakage tests performed in accordance with IWB-
2500, Table IWB-2500-1, Examination Category B-P. Requirements for 
pressure tests following repair/replacement activities on Class 1 
components are specified in IWA--4540. Requirements for pressure test 
holding time for tests following repair/replacement activities are 
specified in IWA-5213. IWA--5213(b) requires that for system pressure 
tests required by IWA-4540, a 10 minutes holding time for noninsulated 
components, or 4 hour holding time for insulated components, is 
required after attaining test pressure. ASME often develops technical 
bases for Code Cases. The technical basis for the increased hold time 
of 15 minutes in Code Case N-795 is as follows:

    Indication of leakage identified through visual VT-2 
examinations during a test at either the 100 [percent] power 
pressure or at 87 [percent] of that value will not be significantly 
different between the two tests. Higher pressure under the otherwise 
same conditions will produce a higher flow rate but the difference 
is not significant. A pressure test at 87 [percent] of the 100 
[percent] rated power pressure would produce a flow rate 
approximately 7 [percent] below the full test pressure. This 
alternate differential pressure (>/=900 psi) is still adequate to 
provide evidence of leakage should a through-wall flaw exist. Since 
the reduced pressure would generate an approximate 7 [percent] 
reduction in flow rate, then, a 7 [percent] increase in the required 
hold time should allow for the equivalent amount of total leakage 
from any existing leak location. This Code Case requires a 50 
[percent] increase in the hold time, which will allow for more 
leakage than is currently generated and therefore a better 
indication of the leak.

    For reasons identified above, the comment asserts that the 1 hour 
hold time imposed by Table 2 of Regulatory Guide 1.147, Rev. 18 is 
unnecessary, and the comment recommends that this condition be removed.
    NRC Response: The NRC disagrees with this comment. The ASME's 
technical basis for the 15 minute hold time in Code Case N-795 relies 
on an argument that the time for leakage to manifest increases linearly 
with the decrease in flow rate corresponding to the reduction in leak 
test pressure. However, the relationship of the time for leakage to 
manifest to the flow rate may not be linear, given tight cracks, which 
result in a torturous path. The NRC does not consider a one hour hold 
time to be an excessive burden.
    No change was made to this final rule as a result of this comment.
Code Case N-831 Ultrasonic Examination in Lieu of Radiography for Welds 
in Ferritic Pipe, Section Xl, Division 1
    Condition: Code Case N-831 is prohibited for use in new reactor 
construction.
    Comment [8-9]: Table 2 in draft revision 19 of Regulatory Guide 
1.147 includes a proposed condition that prohibits Code Case N-831 for 
use in new reactor construction. A comment submission stated that the 
proposed condition is unnecessary and should be removed, for the 
following reasons:
    1. Use of any Section XI Code Case is not permissible until initial 
construction of a component is complete, when the rules of Section XI 
become mandatory. As such, if the Construction Code requires 
radiography as part of the initial construction of a component, then 
radiography is mandatory and ultrasonic examination cannot be 
substituted for radiography.
    2. Application of Code Case N-831 is limited to Section XI repair/
replacement activities where compliance with the Construction Code 
nondestructive examination requirements would require the performance 
of radiography. Ultrasonic examination is preferred when performing a 
repair/replacement

[[Page 14753]]

activity because the ultrasonic examination results will be available 
to compare against future inservice examination ultrasonic examination 
results.
    Comment [9-5]: Paragraph (a) of this Code Case specifies it is 
limited to Section XI repair/replacement activities which excludes its 
use in new construction applications, which is performed under Section 
III. One comment recommends the elimination of this condition since it 
is already included in the Code Case.
    NRC Response: The NRC disagrees with these comments. The subject 
Code Case states that it is limited to Section XI repair/replacement 
activities. However, the preface in Section XI of the ASME Code also 
states that Section XI is allowed for repairs and replacement 
activities once the system has certification marks applied and 
therefore the requirements of the construction code is met. Therefore, 
Section XI would allow the use of ultrasonic examination in lieu of 
radiography for a repair and/or replacement of a new reactor system 
prior to initial fuel load. The condition is to prevent this type of 
use of the Code Case.
    No change was made to this final rule as a result of these 
comments.
    Comment [8-10]: Section 50.55a(b)(2)(xix) includes a Section XI 
condition about substitution of alternative methods. One comment 
recommends that the condition be revised, to specifically allow for 
substitution of examination methods, a combination of methods, or 
techniques other than those specified by the Construction Code, when 
permitted by Code Cases that are acceptable for use in Regulatory Guide 
1.147. Without this clarification, there could be a conflict between 10 
CFR 50.55a(b)(2)(xix) and use of Code Case N-831 in accordance with 
Table 2 of draft Regulatory Guide 1.147.
    NRC Response: The NRC disagrees with the comment. There is no 
conflict as ASME Code Case N-831 is an alternative to Section XI, IWA-
4000 ``Welding, Brazing, Metal Removal, and Installation,'' including 
paragraph IWA-4520(c). Additionally, the condition described in Sec.  
50.55a(b)(2)(xix) does not address ASME Code Case N-831 and is 
therefore not in the scope of this final rule.
    No change was made to this final rule as a result of this comment.

Regulatory Guide 1.192, Revision 3 (DG-1343)

Code Case OMN-13 Performance-Based Requirements for Extending Snubber 
Inservice Visual Examination Interval at LWR3 Power Plants
    Comment [7-1]: The proposed rule referenced DG-1343 as supplemental 
information. DG-1343 identifies Code Case OMN-13, Revision 2 (2017 
Edition), in Table 1 as an acceptable OM Code Case without condition. 
The 2017 Edition of the OM Code, page C-1, OM Code Cases (for Division 
1), identifies applicability of Code Case OMN-13, Revision 2, as 1995 
up to and including 2017. However, Code Case OMN-13, Revision 2, 
itself, includes an applicability statement that identifies ASME OM 
Code-1995 Edition through 2011 Addenda. One comment requested 
clarification of the OM Code edition/addenda applicability for Code 
Case OMN-13, Revision 2, that the NRC is approving for use.
    NRC Response: The NRC agrees with this comment. The NRC has moved 
Code Case OMN-13, Revision 2 (2017 Edition), to Table 2, 
``Conditionally Acceptable OM Code Cases,'' in RG 1.192 to clarify its 
acceptance for use with all editions and addenda of the OM Code listed 
in Sec.  50.55a(a)(1)(iv). Similarly, the NRC noted that Code Case OMN-
20 has an applicability statement that is more restrictive than 
necessary. Therefore, Table 2 in RG 1.192 has been revised in response 
to this comment.

Regulatory Guide 1.193, Revision 6 (DG-1344)

    The NRC received no public comment submittals regarding DG-1344.

V. Section-by-Section Analysis

    The following paragraphs in Sec.  50.55a are revised as follows:

Paragraph (a)(1)(iii)(E)

    This final rule removes and reserves paragraph (a)(1)(iii)(E).

Paragraph (a)(1)(iii)(G)

    This final rule removes and reserves paragraph (a)(1)(iii)(G).

Paragraph (a)(3)

    This final rule adds a condition in paragraph (a)(3) stating that 
the Code Cases listed in RGs 1.84, 1.147, and 1.192 may be applied with 
the specified conditions when implementing the editions and addenda of 
the ASME BPV and OM Codes incorporated by reference in Sec.  50.55a.

Paragraph (a)(3)(i)

    This final rule revises the reference to ``NRC Regulatory Guide 
1.84, Revision 37,'' by removing ``Revision 37'' and adding in its 
place ``Revision 38.''

Paragraph (a)(3)(ii)

    This final rule revises the reference to ``NRC Regulatory Guide 
1.147, Revision 18,'' by removing ``Revision 18'' and adding in its 
place ``Revision 19.''

Paragraph (a)(3)(iii)

    This final rule revises the reference to ``NRC Regulatory Guide 
1.192, Revision 2,'' by removing ``Revision 2'' and adding in its place 
``Revision 3.''

Paragraph (b)(2)(xxxvii)

    This final rule removes paragraph (b)(2)(xxxvii).

Paragraph (b)(3)(x)

    This final rule removes and reserves paragraph (b)(3)(x).

VI. Regulatory Flexibility Certification

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

VII. Regulatory Analysis

    The NRC has prepared a regulatory analysis on this regulation. The 
analysis examines the costs and benefits of the alternatives considered 
by the NRC. The NRC did not receive public comments on the regulatory 
analysis. The regulatory analysis is available as indicated in the 
``Availability of Documents'' section of this document.

VIII. Backfitting and Issue Finality

    The provisions in this final rule allow licensees and applicants to 
voluntarily apply NRC-approved Code Cases, sometimes with NRC-specified 
conditions. The approved Code Cases are listed in three RGs that are 
incorporated by reference into Sec.  50.55a. An applicant's or a 
licensee's voluntary application of an approved Code Case does not 
constitute backfitting, inasmuch as there is no imposition of a new 
requirement or new position. Similarly, voluntary application of an 
approved Code Case by a 10 CFR part 52 applicant or licensee does not 
represent NRC imposition of a requirement or action, and therefore is 
not inconsistent with any issue finality provision in 10 CFR part 52. 
For these

[[Page 14754]]

reasons, the NRC finds that this final rule does not involve any 
provisions requiring the preparation of a backfit analysis or 
documentation demonstrating that one or more of the issue finality 
criteria in 10 CFR part 52 are met.

IX. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, and well-organized 
manner. The NRC has written this document to be consistent with the 
Plain Writing Act as well as the Presidential Memorandum, ``Plain 
Language in Government Writing,'' published June 10, 1998 (63 FR 
31883).

X. Environmental Assessment and Final Finding of No Significant 
Environmental Impact

    The Commission has determined under the National Environmental 
Policy Act (NEPA) of 1969, as amended, and the Commission's regulations 
in subpart A of 10 CFR part 51, that this rule, if adopted, would not 
be a major Federal action significantly affecting the quality of the 
human environment; therefore, an environmental impact statement is not 
required.
    The determination of this environmental assessment is that there 
will be no significant effect on the quality of the human environment 
from this action. The NRC did not receive public comments regarding any 
aspect of this environmental assessment.
    As voluntary alternatives to the ASME Code, NRC-approved Code Cases 
provide an equivalent level of safety. Therefore, the probability or 
consequences of accidents is not changed. There are also no 
significant, non-radiological impacts associated with this action 
because no changes would be made affecting non-radiological plant 
effluents and because no changes would be made in activities that would 
adversely affect the environment. The determination of this 
environmental assessment is that there will be no significant offsite 
impact to the public from this action.

XI. Paperwork Reduction Act Statement

    This final rule amends collections of information subject to the 
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). The 
collections of information were approved by the Office of Management 
and Budget, approval number 3150-0011.
    Because the rule will reduce the burden for existing information 
collections, the public burden for the information collections is 
expected to be decreased by 380 hours per response. This reduction 
includes the time for reviewing instructions, searching existing data 
sources, gathering and maintaining the data needed, and completing and 
reviewing the information collection.
    The information collection is being conducted to document the plans 
for and the results of inservice inspection and inservice testing 
programs. The records are generally historical in nature and provide 
data on which future activities can be based. Information will be used 
by the NRC to determine if ASME BPV and OM Code provisions for 
construction, inservice inspection, repairs, and inservice testing are 
being properly implemented in accordance with Sec.  [thinsp]50.55a of 
the NRC regulations, or whether specific enforcement actions are 
necessary. Responses to this collection of information are generally 
mandatory under Sec.  [thinsp]50.55a.
    You may submit comments on any aspect of the information 
collections, including suggestions for reducing the burden, by the 
following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0024.
     Mail comments to: Information Services Branch, Office of 
the Chief Information Officer, Mail Stop: T6-A10M, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001 or to the OMB reviewer 
at: OMB Office of Information and Regulatory Affairs (3150-0011), Attn: 
Desk Officer for the Nuclear Regulatory Commission, 725 17th Street NW, 
Washington, DC 20503; email: [email protected].

Public Protection Notification

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

XII. Congressional Review Act

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

XIII. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires that Federal agencies use technical 
standards that are developed or adopted by voluntary consensus 
standards bodies unless using such a standard is inconsistent with 
applicable law or is otherwise impractical. In this rule, the NRC is 
continuing to use ASME BPV and OM Code Cases, which are ASME-approved 
voluntary alternatives to compliance with various provisions of the 
ASME BPV and OM Codes. The NRC's approval of the ASME Code Cases is 
accomplished by amending the NRC's regulations to incorporate by 
reference the latest revisions of the following, which are the subject 
of this rulemaking, into Sec.  50.55a: RG 1.84, Revision 38; RG 1.147, 
Revision 19; and RG 1.192, Revision 3. 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 National Technology 
Transfer and Advancement Act of 1995 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.

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

    The NRC is incorporating by reference three NRC RGs that list new 
and revised ASME Code Cases that the NRC has approved as voluntary 
alternatives to certain provisions of NRC-required Editions and Addenda 
of the ASME BPV Code and the ASME OM Code. These regulatory guides are: 
RG 1.84, Revision 38; RG 1.147, Revision 19; and RG 1.192, Revision 3.
    The NRC is required by law to obtain approval for incorporation by 
reference from the Office of the Federal Register (OFR). The OFR's 
requirements for incorporation by reference are set forth in 1 CFR part 
51. On November 7, 2014, the OFR adopted changes to its regulations 
governing incorporation by reference (79 FR 66267). The discussion in 
this section complies with the requirement for final rules as set forth 
in 1 CFR 51.5(a)(1).
    The NRC considers ``interested parties'' to include all potential 
NRC stakeholders, not only the individuals and entities regulated or 
otherwise subject to the NRC's regulatory oversight. These NRC 
stakeholders are not a homogenous group, so the considerations for 
determining ``reasonable availability'' vary by class of interested 
parties. The NRC identifies six classes of interested parties with 
regard to the material to be incorporated by reference in an NRC rule:
     Individuals and small entities regulated or otherwise 
subject to the NRC's regulatory oversight. This class includes 
applicants and potential applicants for licenses and other NRC

[[Page 14755]]

regulatory approvals, and who are subject to the material to be 
incorporated by reference. In this context, ``small entities'' has the 
same meaning as set out in 10 CFR 2.810.
     Large entities otherwise subject to the NRC's regulatory 
oversight. This class includes applicants and potential applicants for 
licenses and other NRC regulatory approvals, and who are subject to the 
material to be incorporated by reference. In this context, a ``large 
entity'' is one that does not qualify as a ``small entity'' under 10 
CFR 2.810.
     Non-governmental organizations with institutional 
interests in the matters regulated by the NRC.
     Other Federal agencies, states, local governmental bodies 
(within the meaning of 10 CFR 2.315(c)).
     Federally-recognized and State-recognized \4\ Indian 
tribes.
---------------------------------------------------------------------------

    \4\ State-recognized Indian tribes are not within the scope of 
10 CFR 2.315(c). However, for purposes of the NRC's compliance with 
1 CFR 51.5, ``interested parties'' includes a broad set of 
stakeholders including State-recognized Indian tribes.
---------------------------------------------------------------------------

     Members of the general public (i.e., individual, 
unaffiliated members of the public who are not regulated or otherwise 
subject to the NRC's regulatory oversight) and who need access to the 
materials that the NRC proposes to incorporate by reference in order to 
participate in the rulemaking.
    The three RGs that the NRC is incorporating by reference in this 
final rule are available without cost and can be read online, 
downloaded, or viewed, by appointment, at the NRC Technical Library, 
which is located at Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland 20852; telephone: 301-415-7000; email: 
[email protected].
    Because access to the three regulatory guides, are available in 
various forms at no cost, the NRC determines that the three regulatory 
guides 1.84, Revision 38; RG 1.147, Revision 19; and RG 1.192, Revision 
3, as approved by the OFR for incorporation by reference, are 
reasonably available to all interested parties.

 Table IV--Regulatory Guides Incorporated by Reference in 10 CFR 50.55a
------------------------------------------------------------------------
                                          ADAMS Accession No. Federal
           Document title                      Register citation
------------------------------------------------------------------------
RG 1.84, ``Design, Fabrication, and   ML19128A276
 Materials Code Case Acceptability,
 ASME Section III,'' Revision 38.
RG 1.147, ``Inservice Inspection      ML19128A244
 Code Case Acceptability, ASME
 Section XI, Division 1,'' Revision
 19.
RG 1.192, ``Operation and             ML19128A261
 Maintenance Code Case
 Acceptability, ASME OM Code,''
 Revision 3.
------------------------------------------------------------------------

XV. Availability of Guidance

    The NRC is issuing revised guidance, RG 1.193, ``ASME Code Cases 
Not Approved for Use,'' Revision 6, for the implementation of the 
requirements in this final rule. The guidance is available in ADAMS 
under Accession No. ML19128A269. You may access information and comment 
submissions related to the guidance by searching on https://www.regulations.gov under Docket ID NRC-2017-0024.
    The regulatory guide lists Code Cases that the NRC has not approved 
for generic use and will not be incorporated by reference into the 
NRC's regulations. Regulatory Guide 1.193 complements RGs 1.84, 1.147, 
and 1.192.

XVI. Availability of Documents

    The documents identified in the following tables are available to 
interested persons through one or more of the following methods, as 
indicated. Throughout the development of this rule, the NRC has posted 
documents related to this rule, including public comments, on the 
Federal rulemaking website at: https://www.regulations.gov under Docket 
ID NRC-2017-0024. The Federal rulemaking website allows you to receive 
alerts when changes or additions occur in a docket folder. To 
subscribe: (1) Navigate to the docket folder (NRC-2017-0024); (2) click 
the ``Sign up for Email Alerts'' link; and (3) enter your email address 
and select how frequently you would like to receive emails (daily, 
weekly, or monthly).

                  Table V--Rulemaking Related Documents
------------------------------------------------------------------------
                                         ADAMS Accession No./ Federal
           Document title                      Register citation
------------------------------------------------------------------------
ASME-OM-2017, ``Operation and         Available for purchase.
 Maintenance of Nuclear Power
 Plants,'' May 31, 2017..
Final Rule--``Incorporation by        68 FR 40469.
 Reference of ASME BPV and OM Code
 Cases,'' July 8, 2003..
Final Rule--``Fracture Toughness      60 FR 65456.
 Requirements for Light Water
 Reactor Pressure Vessels,''
 December 19, 1995..
Assessment of Crack Detection in      ML071020409.
 Heavy-Walled Cast Stainless Steel
 Piping Welds Using Advanced Low-
 Frequency Ultrasonic Methods (NUREG/
 CR-6933), March 2007..
An Evaluation of Ultrasonic Phased    ML12087A004.
 Array Testing for Cast Austenitic
 Stainless Steel Pressurizer Surge
 Line Piping Welds (NUREG/CR-7122),
 March 2012..
Final Safety Evaluation for Nuclear   ML101620010.
 Energy Institute ``Topical Report    ML101660468.
 Materials Reliability Program
 (MRP): Technical Basis for
 Preemptive Weld Overlays for Alloy
 82/182 Butt Welds in Pressurized
 Water Reactors (MRP-169) Revision 1-
 A,'' August 9, 2010..
EPRI Nuclear Safety Analysis Center   Available for purchase.
 Report 202L[dash]2,
 ``Recommendations for an Effective
 Flow Accelerated Corrosion
 Program,'' April 1999..
ASTM International Standard E 1921,   Available for purchase.
 ``Standard Test Method for the
 Determination of Reference
 Temperature, To, for Ferritic
 Steels in the Transition Range.''.
ASME Code, Section III, NB-2330,      Available for purchase.
 ``Test Requirements and Acceptance
 Standards.''.
Regulatory Guide 1.99, Revision 2,    ML102310298.
 ``Radiation Embrittlement of
 Reactor Vessel Materials.''.
Final Rule--``Approval of American    83 FR 2331.
 Society of Mechanical Engineers'
 Code Cases'' dated January 17,
 2018..

[[Page 14756]]

 
Draft Guide 1345, ``Design,           ML18114A228.
 Fabrication, and Materials Code
 Case Acceptability, ASME Section
 III,'' (draft RG 1.84, Revision
 38)..
Draft Guide 1342, ``Inservice         ML18114A225.
 Inspection Code Case Acceptability,
 ASME Section XI, Division 1,''
 (draft RG 1.147, Revision 19)..
Draft Guide 1343, ``Operation and     ML18114A226.
 Maintenance Code Case
 Acceptability, ASME OM Code,''
 (draft RG 1.192, Revision 3)..
Draft Guide 1344, ``ASME Code Cases   ML18114A227.
 Not Approved for Use,'' (draft RG
 1.193, Revision 6)..
RG 1.84, ``Design, Fabrication, and   ML19128A276.
 Materials Code Case Acceptability,
 ASME Section III,'' Revision 38..
RG 1.147, ``Inservice Inspection      ML19128A244.
 Code Case Acceptability, ASME
 Section XI, Division 1,'' Revision
 19..
RG 1.192, ``Operation and             ML19128A261.
 Maintenance Code Case
 Acceptability, ASME OM Code,''
 Revision 3..
RG 1.193, ``ASME Code Cases Not       ML19128A269.
 Approved for Use,'' Revision 6..
Draft Regulatory Analysis...........  ML18099A054.
Final Regulatory Analysis...........  ML19156A178.
------------------------------------------------------------------------

List of Subjects in 10 CFR Part 50

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

    For the reasons set forth in the preamble, and under the authority 
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting 
the following amendments to 10 CFR part 50:

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
1. The authority citation for part 50 continues to read as follows:

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


0
2. In Sec.  50.55a:
0
a. Remove and reserve paragraphs (a)(1)(iii)(E) and (G);
0
b. Revise paragraph (a)(3) introductory text;
0
 c. In paragraph (a)(3)(i), wherever it appears remove the phrase 
``Revision 37'' and add in its place the phrase ``Revision 38'';
0
d. In paragraph (a)(3)(ii), wherever it appears remove the phrase 
``Revision 18'' and add in its place the phrase ``Revision 19'';
0
e. In paragraph (a)(3)(iii), wherever it appears remove the phrase 
``Revision 2'' and add in its place the phrase ``Revision 3''; and
0
f. Remove paragraph (b)(2)(xxxvii) and remove and reserve paragraph 
(b)(3)(x).
    The revision reads as follows:


Sec.  50.55a  Codes and standards.

    (a) * * *
    (3) U.S. Nuclear Regulatory Commission (NRC) Public Document Room, 
11555 Rockville Pike, Rockville, Maryland 20852; telephone: 1-800-397-
4209; email: [email protected]; https://www.nrc.gov/reading-rm/doc-collections/reg-guides/. The use of Code Cases listed in the NRC 
regulatory guides in paragraphs (a)(1)(i) through (iii) of this section 
is acceptable with the specified conditions in those guides when 
implementing the editions and addenda of the ASME BPV Code and ASME OM 
Code incorporated by reference in paragraph (a)(1) of this section.
* * * * *

    Dated at Rockville, Maryland, this 2nd day of March, 2020.

    For the Nuclear Regulatory Commission.
Ho K. Nieh, Director,
Office of Nuclear Reactor Regulation.
[FR Doc. 2020-05086 Filed 3-13-20; 8:45 am]
 BILLING CODE 7590-01-P