[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
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Total averted costs (costs)
Attribute -----------------------------------------------
Undiscounted 7% NPV 3% NPV
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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
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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
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Code Case No. Published with supplement Title
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Boiler and Pressure Vessel Code Section III
(addressed in RG 1.84, Table 1)
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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.
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Boiler and Pressure Vessel Code Section XI
(addressed in RG 1.147, Table 1)
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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)
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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.
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\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
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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.
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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.
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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.
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\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