[Federal Register Volume 79, Number 214 (Wednesday, November 5, 2014)]
[Rules and Regulations]
[Pages 65776-65814]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-25491]
[[Page 65775]]
Vol. 79
Wednesday,
No. 214
November 5, 2014
Part II
Nuclear Regulatory Commission
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10 CFR Part 50
Approval of American Society of Mechanical Engineers' Code Cases; Final
Rule
Federal Register / Vol. 79 , No. 214 / Wednesday, November 5, 2014 /
Rules and Regulations
[[Page 65776]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2009-0359; NRC-2013-0133]
RIN 3150-AI72
Approval of American Society of Mechanical Engineers' Code Cases
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its
regulations to incorporate by reference the latest revisions of three
NRC Regulatory Guides (RGs) approving new and revised Code Cases
published by the American Society of Mechanical Engineers. This action
allows nuclear power plant licensees, and applicants for construction
permits, operating licenses, combined licenses, standard design
certifications, standard design approvals, and manufacturing licenses,
to use the Code Cases listed in these RGs, as alternatives to
engineering standards for the construction, inservice inspection, and
inservice testing of nuclear power plant components. This final rule
changes NRC's regulations to address a petition for rulemaking (PRM),
PRM-50-89, submitted by Mr. Raymond West. The final rule also
restructures the NRC's requirements governing Codes and standards to
align with the Office of the Federal Register's guidelines for
incorporating documents by reference.
This final rule announces the availability of the final versions of
the three RGs that are being incorporated by reference, and a related
RG, not incorporated by reference into the NRC's regulations, that
lists Code Cases that the NRC has not approved for use. For additional
information on these RGs, see Section XVII, Availability of Regulatory
Guides, of this document.
DATES: This final rule is effective on December 5, 2014. The
incorporation by reference of RG 1.84, ``Design, Fabrication, and
Materials Code Case Acceptability, ASME Section III,'' Revision 36 (May
2014); RG 1.147, ``Inservice Inspection Code Case Acceptability, ASME
Section XI, Division 1,'' Revision 17 (May 2014); and RG 1.192,
``Operation and Maintenance Code Case Acceptability, ASME OM Code,''
Revision 1 (May 2014) is approved by the Director of the Office of the
Federal Register as of December 5, 2014.
ADDRESSES: Please refer to Docket ID NRC-2009-0359 when contacting the
NRC about the availability of information for this final rule and RGs
1.84, 1.147 and 1.192. Please refer to Docket ID NRC-2013-0133 when
contacting the NRC about the availability of information for RG 1.193.
You may obtain publicly-available information related to this final
rule by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2009-0359. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this final rule.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-Based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. For
the convenience of the reader, the ADAMS accession numbers are provided
in a table in the ``Availability of Documents'' section of this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Jenny Tobin, Office of Nuclear Reactor
Regulation; telephone: 301-415-2328, email: [email protected]; or
Wallace Norris, Office of Nuclear Regulatory Research, telephone: 301-
251-7650; email: [email protected]; both are staff of the U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001.
Executive Summary
The U.S. Nuclear Regulatory Commission (NRC) is amending its
regulations to incorporate by reference the latest revisions of three
NRC Regulatory Guides (RGs) approving new and revised Code Cases
published by the American Society of Mechanical Engineers (ASME). The
three RGs incorporated by reference are RG 1.84, Revision 36; RG 1.147,
Revision 17; and RG 1.192, Revision 1. This action allows nuclear power
plant licensees, and applicants for construction permits, operating
licenses, combined licenses, standard design certifications, standard
design approvals, and manufacturing licenses, to use the Code Cases
listed in these RGs as alternatives to engineering standards for the
construction, inservice inspection, and inservice testing of nuclear
power plant components.
The NRC is announcing the availability of the final versions of the
three RGs that are being incorporated by reference, and a final version
of RG 1.193, Revision 4, not incorporated by reference into the NRC's
regulations, that lists Code Cases that the NRC has not approved for
generic use.
This final rule also includes changes to the NRC's regulations that
address a petition for rulemaking (PRM), PRM-50-89, submitted by Mr.
Raymond West. Mr. West requested that the NRC amend its regulations to
allow consideration of alternatives to NRC-approved ASME Boiler and
Pressure Vessel and Operation and Maintenance of Nuclear Power Plants
Code Cases. This final rule resolves Mr. West's petition and represents
the NRC's final action on PRM-50-89.
Lastly, this final rule resequences the NRC's requirements in Sec.
50.55a of Title 10 of the Code of Federal Regulations (10 CFR),
governing Codes and standards to align with Office of the Federal
Register's guidelines for incorporating published standards by
reference.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Background
II. Opportunity for Public Participation
A. Overview of Public Comments
Table I--Comment Submissions Received on the Proposed Rule and
Draft Regulatory Guides
III. Public Comment Analysis
A. NRC Reponses to Public Comments on Proposed Rule
B. NRC Responses to Public Comments on Draft Regulatory Guides
IV. NRC Approval of New and Amended ASME Code Cases
A. ASME Code Cases Approved for Unconditional Use
Table II--Unconditionally Approved Code Cases
B. ASME Code Case Approved for Use With Conditions
Table III--Conditionally Approved Code Cases
C. ASME Code Cases Not Approved for Use
V. Petition for Rulemaking (PRM-50-89)
VI. Changes Addressing the Office of the Federal Register's
Guidelines on Incorporation by Reference
VII. Addition of Headings to Paragraphs
A. NRC's Convention for Headings and Subheadings
B. Readers Aids
VIII. Paragraph-by-Paragraph Discussion
IX. Regulatory Flexibility Certification
X. Regulatory Analysis
XI. Backfitting and Issue Finality
XII. Plain Writing
[[Page 65777]]
XIII. Finding of No Significant Environmental Impact: Environmental
Assessment
XIV. Paperwork Reduction Act Statement
XV. Congressional Review Act
XVI. Voluntary Consensus Standards
XVII. Availability of Regulatory Guides
XVIII. Availability of Documents
I. Background
The American Society of Mechanical Engineers (ASME) develops and
publishes the ASME Boiler and Pressure Vessel (BPV) Code, which
contains requirements for the design, construction, and inservice
inspection (ISI) and examination of nuclear power plant components, and
the ASME Code for Operation and Maintenance of Nuclear Power Plants
(OM) Code, which contains requirements for inservice testing (IST) of
nuclear power plant components. In response to BPV and OM Code user
requests, the ASME develops ASME Code Cases that provide alternatives
to BPV and OM Code requirements under special circumstances.
The NRC approves and/or mandates the use of the ASME BPV and OM
Codes in Sec. 50.55a of Title 10 of the Code of Federal Regulations
(10 CFR) through the process of incorporation by reference (IBR). As
such, each provision of the ASME Codes incorporated by reference into,
and mandated by, Sec. 50.55a, ``Codes and standards,'' constitutes a
legally-binding NRC requirement imposed by rule. As noted previously,
ASME Code Cases, for the most part, represent alternative approaches
for complying with provisions of the ASME BPV and OM Codes.
Accordingly, the NRC periodically amends Sec. 50.55a to incorporate by
reference NRC Regulatory Guides (RGs) listing approved ASME Code Cases
that may be used as alternatives to the BPV and OM Codes. See Federal
Register notice (FRN), ``Incorporation by Reference of ASME BPV and OM
Code Cases'' (68 FR 40469; July 8, 2003).
This rulemaking is the latest in a series of rulemakings that
incorporate by reference new versions of several RGs identifying new
and revised \1\ unconditionally or conditionally acceptable ASME Code
Cases that are approved for use. In developing these RGs, the NRC staff
reviews ASME BPV and OM Code Cases, determines the acceptability of
each Code Case, and publishes its findings in the RGs. The RGs are
revised periodically as new Code Cases are published by the ASME. The
NRC incorporates by reference the RGs listing acceptable and
conditionally acceptable ASME Code Cases into Sec. 50.55a. Currently,
NRC RG 1.84, Revision 35, ``Design, Fabrication, and Materials Code
Case Acceptability, ASME Section III''; RG 1.147, Revision 16,
``Inservice Inspection Code Case Acceptability, ASME Section XI,
Division 1''; and RG 1.192, Revision 0, ``Operation and Maintenance
Code Case Acceptability, ASME OM Code,'' are incorporated into the
NRC's regulations at 10 CFR 50.55a, ``Codes and standards.''
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\1\ ASME Code Cases can be categorized as one of two types: New
or revised. A new Code Case provides for a new alternative to
specific ASME Code provisions or addresses a new need. A revised
Code Case is a revision (modification) to an existing Code Case to
address, for example, technological advancements in examination
techniques or to address NRC conditions imposed in one of the
regulatory guides that have been incorporated by reference into 10
CFR 50.55a.
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This final rule adds provisions that allow the NRC to authorize
alternatives to NRC-approved ASME BPV and OM Code Cases, as requested
in a petition for rulemaking (PRM) that was submitted to the NRC on
December 14, 2007, and revised on December 19, 2007, by Mr. Raymond
West (ADAMS Accession No. ML073600974). A detailed discussion of the
PRM is provided in Section V, ``Petition for Rulemaking (PRM-50-89),''
of this document.
II. Opportunity for Public Participation
On June 24, 2013 (78 FR 37886), the NRC published a proposed rule
in the Federal Register that would incorporate by reference RG 1.84,
Revision 36; RG 1.147, Revision 17; and RG 1.192, Revision 1. On the
same date, the NRC published a parallel FRN announcing the availability
of the three draft RGs and opportunity for public comment (78 FR 37721;
June 24, 2013). The NRC provided a 75-day public comment period for
both the proposed rule and the draft RGs, which ended on September 9,
2013.
A. Overview of Public Comments
The NRC received a total of 10 comment submissions. The submissions
were received from three private citizens, four utility organizations,
and three industry groups that provide engineering and inspection
services to the utilities. Table I lists the commenter's name and
affiliation, ADAMS accession number for the comment submission, and the
Code Case or subject of each comment.
Table I--Comment Submissions Received on the Proposed Rule and Draft Regulatory Guides
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Comment submission ADAMS Affected code cases/
Commenter name Affiliation Accession No. subject
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William Culp....................... Private Citizen....... ML13210A143 Proposed Rule.
Saige Stephens..................... Private Citizen....... ML13210A151 General.
Richard Swayne..................... ASME.................. ML13253A076 N-60-5.
ML13252A286 ** N-416-4.
N-561-2.
N-562-2.
N-597-2.
N-606-1.
N-619.
N-648-1.
N-661-2.
N-702.
N-739-1.
N-798.
N-800.
N-659-2.
Proposed Rule.
Mark Richter....................... Nuclear Energy ML13259A040 Proposed Rule.
Institute.
ML13254A080 **
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Edward Colie....................... South Carolina ML13254A082 Proposed Rule.
Electric and Gas.
Patricia Campbell.................. GE Hitachi Nuclear ML13259A038 1332-6.
Energy.
Devin Kelley....................... AREVA................. ML13259A039 N-71-18.
David Helker....................... Exelon Generation ML13269A371 N-60-5.
Company, LLC.
N-798.
N-800.
N-702.
Shawn Comstock..................... Private Citizen....... ML13182A081 OMN-1 (2006 Addenda).
OMN-11 (2006 Addenda).
OMN-12 (2004 Edition).
Roy Hall........................... Inservice Inspection ML13197A239 N-805.
Program Owners Group.
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** There are two ADAMS accession numbers for the submissions from ASME and the Nuclear Energy Institute because
each submission contained comments on the proposed rule and the drafts RGs. Both accession numbers are for the
same incoming submission, but one accession number is identified in ADAMS as a response to the Federal
Register notice soliciting comments on the proposed rule and the other is identified as a response for the
draft RGs.
III. Public Comment Analysis
The NRC has reviewed every comment submission and has identified 42
unique comments requiring NRC consideration and response. Comment
summaries and the NRC responses are presented in this section. Comment
responses have been organized in two categories: (A) NRC Responses to
Public Comments on Proposed Rule and (B) NRC Responses to Public
Comments on Draft RGs, further delineated by individual RG (i.e., RG
1.84, RG 1.147, and RG 1.192).
A. NRC Reponses to Public Comments on Proposed Rule
Proposed Rule
Comment: The commenter developed a proposed one-page revision to
the overall Codes and standards rule in Sec. 50.55a that reflects the
commenter's view of the current regulatory process and suggested
parsing the details of Sec. 50.55a to the appropriate RGs. The
commenter provided the background and bases for his proposed rule
structure, and stated that the purpose of his proposal is to simplify
the overall structure of Sec. 50.55a. (Culp-3)
NRC Response: The main purpose of this rulemaking is to amend Sec.
50.55a to incorporate by reference the latest revisions of three RGs
approving new and revised Code Cases published by ASME. This rulemaking
also proposes to: (1) Resolve a petition for rulemaking (PRM-50-89)
submitted by Mr. Raymond West, (2) resequence the NRC's requirements
governing Codes and standards in order to align with the latest
guidelines of the OFR for IBR, and (3) add headings (explanatory
titles) to paragraphs and lower-level subparagraphs of Sec. 50.55a.
The NRC is not proposing a major restructuring or simplification of
the requirements in Sec. 50.55a. As explained in the statement of
considerations in the proposed rule, the proposed editorial, non-
substantive changes were made to align with the IBR guidance for
multiple standards that is included in Chapter 6 of the OFR's,
``Federal Register Document Drafting Handbook,'' January 2011 Revision.
These changes will structure NRC's regulations consistent with other
Federal regulations that incorporate by reference multiple standards.
Although NRC welcomes public comments on the revised structure of Sec.
50.55a, the NRC is limited in the types of changes it can make in
response to public comments on the revised structure and must align
with the OFR's guidance.
Adding headings at the paragraph and subparagraph levels of Sec.
50.55a will enhance the reader's ability to identify the subject matter
of each paragraph and subparagraph. These headings are a first step
toward addressing longstanding complaints about the readability and
complex structure of Sec. 50.55a. The NRC is not making significant
structural changes to the rule at this time, but may, in the future,
consider doing so in a separate rulemaking. The NRC would consider the
commenter's suggestions and proposed rule language if and when NRC
conducts that rulemaking. At this time, however, the NRC considers the
commenter's suggestion to be outside the scope of this proposed
rulemaking.
No change was made to the final rule as a result of this comment.
Comment: The purpose and scope of the rule has changed over time,
and no longer reflects the actual regulatory process for review of
consensus industry Codes and standards that have been found acceptable
to the NRC staff on a generic basis or as part of a plant-specific
review process that covers more than the Codes and standards mentioned.
It does not seem appropriate for Sec. 50.55a to reference Codes and
standards that have been withdrawn (e.g., IEEE 279). The content of
Sec. 50.55a represents an archive of once-upon-a-time requirements,
not contemporary Codes and standards. It is not necessary to
recapitulate what Codes and standards were approved on individual
applications; applicants retain design and safety responsibility
(including identification of unreviewed safety questions) that might
arise from new regulatory guides, Codes and standards, and operating
experience. The following Codes, standards, and Code Cases in the
proposed regulation are not the latest and conditions are imposed on
the use of superseded documents which would preferably not be used for
new design or ISI activities (the conditions are most likely fully
documented in the licenses, safety analyses, and ISI programs for
individual nuclear power plants as approved by the NRC): (Culp-3.1,
3.3, 3.9)
a. ASME III and Code Case N-729-1 (N-729-4 Is Approved by ASME)
b. ASME XI
c. IEEE 279
NRC Response: The NRC disagrees with the assertion that the
proposed rule does not reflect the actual regulatory process for review
of consensus industry Codes and standards that have been found
acceptable to the NRC staff. Section II, ``Discussion,'' of the
proposed rule described the three-step process that the NRC follows to
determine the acceptability of new and revised Code Cases and the need
for regulatory positions on the uses of these
[[Page 65779]]
Code Cases. The fundamental process has not changed over time. Also,
the Code of Record for design and construction does not change over
time unless there is a voluntary update by the licensee. As such, these
codes and standards must be referenced in Sec. 50.55a as long as they
are in use.
Any Code or standard still in use must continue to be listed in the
regulation, or licensees would have to discontinue their use when the
rule becomes effective and immediately implement the latest version.
These Codes and Code Cases are still in use and, therefore, may not be
removed from Sec. 50.55a without unacceptably changing their legal
status from mandatory requirements or approved for use, to guidance.
No change was made to the final rule as a result of this comment.
Comment: The current language and structure of Sec. 50.55a blurs
the lines between the requirements for a quality program and for
safety. (Culp-3.2)
NRC Response: The NRC believes this is an out of scope comment
because it addresses the clarity of the requirements in Sec. 50.55a in
this rulemaking. The scope of this rulemaking is to: (1) Incorporate by
reference the three Regulatory Guides identifying NRC-approved ASME
Code Cases; and (2) to reorganize the section to address Office of the
Federal Register requirements for incorporation by reference.
However, the NRC provides the following response to the out of
scope comment. The NRC notes that the commenter did not provide any
rationale why the rulemaking blurs the distinction between quality
assurance and safety. In addition, the NRC notes that the
reorganization of Sec. 50.55a fundamentally addressed the paragraph
identifying the ASME and IEEE codes that are incorporated by reference.
The reorganization did not change any of the NRC requirements with
respect to quality assurance or safety.
No change was made to the final rule as a result of this comment.
Comment: The proposed reorganization of Sec. 50.55a uses the
unconventional numbering hierarchy (a), (1), (i), (A). This is
difficult to follow in the existing rule which is very long. It is even
more difficult to follow in the proposed regulation with or without
added introductory statements. (Culp-3.4)
NRC Response: The NRC has added headings to the paragraph and
subparagraph levels of Sec. 50.55a to aid the reader of this
regulation. The hierarchy used in Sec. 50.55a is that which is used
throughout the Code of Federal Regulations and is dictated by the OFR.
The NRC is also considering developing additional user aides.
No change was made to the final rule as a result of this comment.
Comment: The proposed regulation states that the regulation is
consistent with a policy to review and accept industry standards
instead of writing regulations; this is not achieved in practice due to
delays in endorsing new Code editions and addenda. In at least some
cases, the unendorsed newer Code revisions have been specifically made
to incorporate the conditions, exceptions, and limitations in Sec.
50.55a. (Culp-3.5)
NRC Response: The NRC appreciates the ASME's efforts to consider
the NRC's concerns as addressed in conditions to Sec. 50.55a. The NRC
agrees that delays in approving new ASME Code editions and Code Cases
can be counterproductive with respect to implementation of improvements
in ASME Code requirements. The NRC continues to assess ways to improve
the rulemaking process to find schedule efficiencies.
No change was made to the final rule as a result of this comment.
Comment: There is too much detail in the proposed regulation; NRC
concerns should be more appropriately organized and put into consensus
Code and Code Case work and topical regulatory guides. The proposed
regulation is excessively detailed and covers an extraordinary range of
subjects; the diverse NRC conditions ranging from grease caps to relief
valve testing facility capabilities could be better organized and
documented in regulatory guides on the specific topic (e.g., RG 1.90).
(Culp-3.6)
NRC Response: The NRC agrees that there are many conditions in
Sec. 50.55a. It should be noted, that certain conditions are necessary
because applicants and licensees continue to use many different Code
editions and addenda. Accordingly, it is necessary to continue to list
conditions that may have been addressed by a later Code edition because
the earlier Code edition is still in use. The NRC determined that other
conditions, such as those addressing grease caps, are necessary to
ensure that safety-related concerns are adequately addressed.
With respect to the suggestion to use RGs, the NRC notes that RGs
normally provide guidance and describe approaches that would be
acceptable to the NRC for implementing a rule. Under the approach
suggested in the comment, the RG would have to be incorporated by
reference into Sec. 50.55a in order for the provisions in the
regulatory guides to continue to be legally-binding. In enclosure 5 to
the comments submitted by the ASME, the ASME encouraged the NRC to
consider alternative methods for endorsing ASME Codes and standards,
such as moving many of the requirements currently specified in Sec.
50.55a into a suitable regulatory guide that can be referenced within
the regulation. The NRC agrees that the format and organization of
Sec. 50.55a could be improved, and the NRC may, in the future, conduct
a rulemaking to restructure and simplify Sec. 50.55a. The public would
be given opportunity to comment before implementation.
No change was made to the final rule as a result of this comment.
Comment: There are multiple reviews and opportunities for staff
review and public comment without necessarily also requiring comment on
the proposed regulations to ``incorporate by reference'' what started
as a simple reference to ASME III. The process of a comment in Code
committee, comment on proposed regulatory guides, and comment on Code
Cases seems adequate. Yet, comments from NRC representatives in Code
meetings do not, according to their own words, ``carry the weight of
the NRC staff endorsement,'' and some conditions have arisen after Code
committees have finished reviews and published revisions. (Culp-3.7)
NRC Response: The NRC staff representatives on ASME Code committees
have the opportunity to participate during the consideration of the
Code cases during the ASME standards process. These individuals can
provide input to the cases both before and after ASME endorsement.
However, this participation is not a substitute for the technical,
legal, and management reviews that must be conducted with respect to a
complete rulemaking prior to issuance.
The second issue in this comment concerns public involvement in the
rulemaking process involved in incorporating by reference those Code
cases that the NRC has reviewed and approved. In accordance with the
Administrative Procedures Act, the public is afforded an opportunity
for review and comment, unless there is reasonable likelihood that
there will be no ``significant adverse comment'' on a proposed rule.
Past NRC experience suggests that the NRC will receive at least one
``significant adverse comment'' on each Sec. 50.55a proposed rule.
No change was made to the final rule as a result of this comment.
Comment: The proposed revision to Sec. 50.55a is very complicated
and seems to be contrary to multiple claims in the discussion points in
the proposed rule regarding: (Culp-3.8)
a. Paperwork reduction
[[Page 65780]]
b. Regulatory flexibility
c. Plain writing
d. Backfitting and issue finality
NRC Response: The NRC does not agree with the comment. The comment
did not explain why the proposed Paperwork Reduction Act statement,
Regulatory Flexibility Certification, Plain Writing discussion, or
Backfitting and Issue Finality discussion is contrary to the proposed
regulation. Complexity by itself does not mean that the NRC's proposed
discussions on the four areas are inadequate or in error. Furthermore,
the bulk of the changes in this rulemaking involve the reorganization
of the rule. Therefore, the comment incorrectly implies that this
rulemaking is the reason for the ``complexity'' of Sec. 50.55a.
No change was made to the final rule as a result of this comment.
Comment: Should Mechanical Engineers become the new regulated
embodiment of manufacturing arms? Change administration using
international standards. (Stephens-4.1)
NRC Response: The NRC is unable to respond to this comment because
of its ambiguous nature.
No change was made to the final rule as a result of this comment.
Comment: The NRC should amend its regulations to allow
consideration of alternatives to the ASME BPV and OM Code Cases, as
requested in a petition for rulemaking submitted by Mr. Raymond West
(PRM-50-89) (ADAMS Accession No. ML073600974). The possibility of
implementing an alternative to a Code Case approved by the Director of
the Office of Nuclear Reactor Regulation will reduce the administrative
burden on licensees and significantly reduce the lengthy process of
proposing and gaining acceptance for a change or modification to a Code
Case. The ASME supports the proposed changes in Sec. 50.55a(z) to
address PRM-50-89. (NEI-6.2, ASME-5.5.1)
NRC Response: The NRC agrees. Authorizing an alternative to an NRC-
approved ASME Code Case reduces the administrative burden on the NRC
and licensees. A complete discussion of the bases is set forth in
Section V, ``Petition for Rulemaking (PRM-50-89).''
The final rule includes a provision in 50.55a(z) allowing the NRC
to authorize alternatives to NRC-approved ASME Code Cases.
Comment: The ASME believes changes for Federal Register guidelines
have been crafted to minimize administrative burden. (ASME-5.5.2)
NRC Response: No response is necessary.
Comment: Paragraph headings will improve readability. (ASME-5.5.3)
NRC Response: No response is necessary.
Comment: In general, the proposed RGs and related documents are
written in a clear and effective manner, consistent with the Plain
Writing Act and the Presidential Memorandum, ``Plain Language in
Government Writing.'' Well-written regulatory guidance documents
support their correct interpretation and implementation (NEI-6.2).
NRC Response: No response necessary.
Comment: The proposed changes to 10 CFR 50.55a would place a large
burden on licensees. As discussed in Section VI, these changes would
``require substantial rewriting of these procedures and documents to
correct the references to the old (superseded) sections, paragraphs and
subparagraphs.'' For licensees, these revisions would include licensing
documentation. None of the proposed organizational changes to 10 CFR
50.55a pertain to any of the provisions of 10 CFR 50.109(a)(4), since
no information is changing and is merely reorganized. This means that
in order to reorganize 10 CFR 50.55a, backfit analysis would have to be
performed in accordance with 10 CFR 50.109. There is no need to change
the location of the content in 10 CFR 50.55a (South Carolina Electric
and Gas-7.1).
NRC Response: As indicated in Section V, ``Changes Addressing
Office of the Federal Register's Guidelines on Incorporation by
Reference,'' of the proposed rule, the reorganization of content was
made in accordance with the revised guidance for incorporation by
reference of multiple standards that is included in Chapter 6 of the
OFR's, ``Federal Register Document Drafting Handbook,'' January 2011
Revision. All Federal agencies were directed to align with the
guidelines. The OFR's guidance provided several options for
incorporating by reference multiple standards into regulations. The NRC
found moving the incorporation by reference of multiple standards into
the first paragraph of Sec. 50.55a(a) to be the least disruptive
option. These changes, which are required by the OFR, are not within
the purview of the backfit rule, and no further consideration of
backfitting is needed to address the OFR-mandated reorganization.
No change was made to the final rule as a result of this comment.
Comment: The NRC should consider adding hyperlinks and indentation
to Sec. 50.55a because it would aid readers in navigating the rule.
(South Carolina Electric and Gas-7.2)
NRC Response: The NRC appreciates these practical suggestions and
agrees that adding hyperlinks or indentation would aid the readers in
navigating Sec. 50.55a. However, the NRC is unable to add hyperlinks
or indentation to a rule published in the Code of Federal Regulations.
Format requirements for the Code of Federal Regulations are established
and enforced by the OFR, and do not permit inclusion of hyperlinks or a
different indentation scheme. Please note that the NRC has prepared two
documents to aid the reader in navigating Sec. 50.55a: ``Final
Reorganization of Paragraphs and Subparagraphs in 10 CFR 50.55a, `Codes
and standards' '' (ADAMS Accession No. ML14015A191) and ``Cross-
Reference Tables'' (ADAMS Accession No. ML14211A050--package with two
tables). The NRC is currently considering developing several
alternatives to improve the format and organization of Sec. 50.55a in
a potential future rulemaking. The NRC plans to seek public interaction
as part of the rulemaking process.
No change was made to the final rule as a result of this comment.
B. NRC Responses to Public Comments on Draft Regulatory Guides
Regulatory Guide 1.84, Revision 36 (DG-1230)
Code Case N-60-5
Comment: Text in the proposed condition should be corrected to
change ``stain-hardened'' to ``strain-hardened.'' (ASME-5.1.1, Exelon-
10.1)
NRC Response: The NRC agrees with the comment.
RG 1.84, Revision 36 has been corrected in accordance with the
comment.
Code Case 1332-6
Comment: Appendix C of DG-1230 states that Code Case 1332-6 is
contained in Table 5. However, Code Case 1332-6 does not appear in
Table 5. (GE Hitachi Nuclear Energy-8.1)
NRC Response: The NRC agrees with this comment. Code Case 1332-6
has been added to Table 5 in RG 1.84, Revision 36, which lists those
Section III Code Cases that have been superseded by revised Code Cases.
Code Case N-71-18
Comment: The American Welding Society (AWS) Code D1.1 was
reformatted, and the provisions in paragraph 4.5.2.2 were relocated to
paragraph 5.3.2.3 in the AWS Code. The paragraph references for AWS
D1.1 in condition No. 3 to Code Case N-71-18
[[Page 65781]]
should be revised accordingly. (AREVA-9.1)
NRC Response: The NRC agrees with this comment. The reference in
condition 3 to Code Case N-71-18 has been corrected in RG 1.84,
Revision 36 by referring to paragraph ``5.3.2.3.''
Regulatory Guide 1.147, Revision 17 (DG-1231)
Code Case N-416-4
Comment: The NRC condition on this Code Case requiring
nondestructive examination of welded or brazed repairs, and fabricated
and installed joints, in accordance with the construction code of
record, imposes an unnecessary burden on licensees and is not necessary
to ensure safe operation. The BPV Code has long relied on a specified
relationship between NDE and allowable stresses, i.e., vintage codes,
such as American National Standards Institute (ANSI) B31.1 or Section
III, have lower allowable stresses, due to the fact that NDE is
generally not required, whereas nuclear codes (ASME Section III and
B31.7) have higher allowable stress intensities for Class 1 components
relative to Class 2 and 3 components (due mostly to the additional
examinations required for Class 1 components).
The NRC stated that ``A system pressure test or hydrostatic
pressure test does not verify the structural integrity of the repaired
piping components.'' The ASME has never established any relationship
between the test pressure to which a component is subjected and any
other material or design characteristic. The primary technical
consideration in development of the required test pressure is to ensure
that it is low enough to prevent yielding of the material. Hydrostatic
testing does not prove structural integrity; it proves only leak
tightness. Similarly, NDE alone does not ensure structural integrity.
The ASME Code ensures structural integrity through a combination of
many factors, including material testing, design formulas, design
factors, and qualification of personnel. Adding more NDE than required
by the Construction Code (be it ASME Section III or B31.1) is not
required to ensure structural integrity. (ASME-5.2.1)
NRC Response: The NRC disagrees with the comment that the
additional NDE requirements imposed when using Code Case N-416-4 are
unnecessary and imply that existing components are unsuitable. The NRC
does agree that hydrostatic pressure testing or NDE alone does not
ensure structural integrity. The original Construction Codes ensured
structural integrity through a combination of many factors including
material testing, design formulas, design factors, qualification of
procedures, qualification of personnel, NDE, and hydrostatic testing.
Code Case N-416-4 would allow a system leakage test to be performed in
lieu of (1) a hydrostatic pressure test prior to return to service of
Class 1, 2, and 3 welded or brazed repairs; (2) fabrication welds or
brazed joints for replacement parts and piping subassemblies; or (3)
installation of replacement items by welding or brazing.
The NRC believes that the rigorous NDE requirements of Section III
should be performed when the hydrostatic pressure test is not
performed. The reason for this condition is that some earlier
Construction Codes have less stringent NDE requirements than Section
III; however, they require a greater pressure for the Code Case N-416-4
required hydrostatic test. Section III NDE requirements for Class 1, 2,
and 3 components generally require either surface or volumetric
examinations or possibly both. The NRC believes that these NDE
requirements along with a system leakage test provide the same level of
quality and safety as the higher pressure hydrostatic test and reduced
NDE requirements of earlier Construction Codes.
No changes were made to RG 1.147, Revision 17, as a result of this
comment.
Code Case N-561-2
Comment: Proposed Conditions (1) and (3) should be eliminated.
Proposed Conditions (1) and (3) limit the life of the repair ``until
the next refueling outage'' for repairs performed on a wet surface or
if the cause of the degradation has not been determined. The Code Case
already limits the life of the repair to ``one fuel cycle'' for these
same situations. The ASME Code committee considered both phrases when
revising this Code Case to add these restrictions, and intentionally
chose ``one fuel cycle'' instead of ``next refueling outage'' so as not
to imply that such weld overlays could not be performed while a plant
is shut down for a refueling outage. In such a case, literal
application of ``next refueling outage'' could mean the current
refueling outage, which could be an extreme hardship, depending on the
timing of the discovery of the need for a weld overlay. Use of the term
``one fuel cycle'' clearly requires that the overlay be removed during
the subsequent fuel cycle no later than the same point in the cycle at
which the overlay was applied. In the vast majority of cases, this will
happen during the next refueling outage; otherwise, a special outage or
a special limiting condition of operation would be required mid-cycle
in order to effect its removal. (ASME-5.2.2.a)
NRC Response: The NRC disagrees with the comment on the ``next
refueling outage.'' The NRC finds that the suggested phrase, ``next
fuel cycle,'' is not as conservative as ``the next refueling outage''
phrase because the ``next fuel cycle'' condition would permit longer
service time to the repair that is performed on a wet surface, or the
cause of the degradation has not been determined.
To clarify the difference between the ``next refueling outage'' vs.
``one fuel cycle,'' the NRC staff uses the following example. Assume
fuel cycle No. 1 is followed by refueling outage No. 1, fuel cycle No.
2, and refueling outage No. 2. Under the ``next refueling outage''
condition, if a repair is performed during fuel cycle No. 1, regardless
whether on the first day or last day of fuel cycle No. 1, the ``next
refueling outage'' would be refueling outage No. 1 during which time
the repair needs to be removed. If the repair is performed during
refueling outage No. 1, the next refueling outage would be refueling
outage No. 2 during which time the repair needs to be removed. Under
the ``next fuel cycle'' condition, if a repair is performed in the
middle of fuel cycle No. 1, the next fuel cycle would mean fuel cycle
No. 2 during which time the repair needs to be removed. However, this
condition does not specify exactly when in the next fuel cycle (fuel
cycle No. 2) the repair must be removed. A licensee could interpret the
next fuel cycle as the entire fuel cycle No. 2 and remove the repair
after fuel cycle No. 2 is completed. This means that the licensee could
remove the repair during refueling outage No. 2. Some licensees may
choose to remove the overlay during refueling outage No. 1 as the
comment stated, but based on the interpretation described earlier, the
repair does not need to be removed during refueling outage No. 1.
No changes were made to RG 1.147, Revision 17, as a result of this
comment.
Code Case N-561-2
Comment: Proposed Condition (2) on Code Case N-561-2 should be
eliminated. Proposed Condition (2) prohibits the use of the exemption
listed in paragraph 6(c)(1) of this case. The provisions in paragraph
6(c)(1) are identical to existing, approved provisions of IWA 4520,
Examination, in the 2001 Edition of ASME Section XI.
Weld overlays are base metal repairs, and are therefore already
exempt by Section XI, IWA-4520 (2001 and later editions and addenda).
This exemption
[[Page 65782]]
was only included in revision 2 of Code Cases N-561 and N-562; and also
in Revision 1 of Code Case N-661-2 which was approved by Regulatory
Guide 1.147, Rev. 16, without this condition, to enable plants not yet
implementing the 2001 or later edition and addenda to apply the
exemption which had been accepted by the NRC in Sec. 50.55a.
Paragraph 6(a) of the case requires a surface examination of the
completed weld overlay to provide additional assurance of the quality
of the repair weld. ASME believes that this requirement is sufficient
for Class 3 applications in locations where the Construction Code would
not require volumetric examination of full penetration butt welds in
that location. Further, with the added condition of ultrasonically
examining the base metal to verify absence of cracking, the benefit of/
need for volumetric examination is significantly reduced. (ASME-
5.2.2.b)
NRC Response: The NRC agrees that proposed condition (2) can be
eliminated. Paragraph 6(c)(1) of the Code Case states that ``Class 3
weld overlays are exempt from volumetric examination when the
Construction Code does not require the full penetration butt welds in
the same location be volumetrically examined.'' Section XI, paragraph
IWA-4520(a)(1), 2001 Edition and later, states that ``Base metal
repairs on Class 3 items are not required to be volumetrically examined
when the Construction Code does not require that full-penetration butt
welds in the same location be volumetrically examined.'' As indicated
in the comment, the exemptions are identical. The NRC unconditionally
approved paragraph IWA-4520(a)(1) in the 2001 Edition through 2008
Addenda. Therefore, it would be inconsistent to retain the condition on
the Code Case.
The NRC has removed proposed Condition (2) on Code Case N-561-2
from the final RG 1.147, Revision 17.
Code Case N-561-2 and N-661.2
Comment: Proposed Condition (5) on Code Case N-561-2 is unwarranted
and should be removed or modified.
The rationale for this condition is to reduce the chances of
producing a suspect weld (i.e., one made on a wet surface).
Additionally, proposed Conditions (1), (2), (3), and (5) are
unwarranted for reasons listed in comments provided on Code Case N 561-
2.
Footnote 6 in Code Cases N-561-2 and N-661-2 (and footnote 5 in N-
562-2) states: ``Testing has shown that piping with areas of wall
thickness less than the diameter of the electrode may burn-through
during application of a water-backed weld overlay.'' Testing performed
by the Electric Power Research Institute (EPRI) and described in EPRI
Report TR-108131, ``Weld Repair of Class 2 and 3 Ferritic Piping,''
demonstrated that this criteria applies to application of weld overlays
under both pressurized (up to 500 psi during the testing) and non-
pressurized conditions (during this testing, specimens that burned-
through were successfully welded-up using the shielded metal arc
welding process with water leaking from the pipe; and those specimens
passed the subsequent burst testing at pressures beyond the minimum
burst pressure of new pipe). The results were the same in both
situations--if the electrode diameter exceeded the thickness being
welded, burn-through was likely--irrespective of internal pressure. If
the thickness of the base metal equaled the thickness of the electrode,
burn through would not occur, regardless of internal pressure. To
require depressurization in such cases--in order to reduce the chances
of producing a suspect weld--would cause extreme hardships, with no
technical justification.
Code Cases N-561-1, N-562-1, and N-661-1 each contained the
statement: ``4(b) Piping with wall thickness less than the diameter of
the electrode shall be depressurized before welding.'' This was changed
to a footnote for editorial purposes in revision 2 of each Code Case.
If the NRC believes that Condition (5) must be retained in Table 2 of
RG 1.147, the ASME recommends that this condition be revised to read
``Piping with wall thickness less than the diameter of the electrode
shall be depressurized before welding.'' This wording is consistent
with that specified in paragraph 4(b) of Code Case N-661-1, which is
currently listed in Table 2 of RG 1.147. (ASME-5.2.2.c and ASME-5.2.7)
NRC Response: The NRC agrees with the comment.
The NRC staff has reviewed the EPRI report and finds that the ASME
recommendation has merit because it is supported by experimental data.
The results of the research shows that if the thickness of the base
metal equals the thickness of the electrode then burn through will not
occur regardless of internal pressure. There were five conditions in
the draft regulatory guide issued for public comment. The NRC agreed in
a response to a separate comment (follows below) to remove condition
(2) regarding the exemption from volumetric examination of Class 3 weld
overlays. Condition (5) in the draft regulatory guide has therefore
been renumbered as condition (4) in the final regulatory guide, and the
NRC has revised it consistent with the ASME recommendation.
Comment: Proposed Conditions (1), (2), (3), and (5) are unwarranted
for reasons listed in comments provided on Code Case N-561-2. However,
if the NRC believes that Condition (5) must be retained in Table 2 of
RG 1.147, this condition be revised to read ``Piping with wall
thickness less than the diameter of the electrode shall be
depressurized before welding.'' This wording is consistent with that
specified in paragraph 4(b) of Code Case N-661-1, which is currently
listed in Table 2 of RG 1.147. (ASME-5.2.3)
NRC Response: Code Case N-562-2 is similar to Code Case N-561-2.
Therefore, the NRC's position on conditions in Code Case N-561-2 are
also applicable to Code Case N-562-2. Therefore, the NRC has determined
to retain Conditions (1) and (3) as proposed. Proposed Condition (2)
has been removed; paragraph 6(c)(1) of the Code Case states that
``Class 3 weld overlays are exempt from volumetric examination when the
Construction Code does not require the full penetration butt welds in
the same location be volumetrically examined.'' Section XI, paragraph
IWA-4520(a)(1), 2001 Edition and later, states that ``Base metal
repairs on Class 3 items are not required to be volumetrically examined
when the Construction Code does not require that full-penetration butt
welds in the same location be volumetrically examined.'' As indicated
in the comment, the exemptions are identical. The NRC unconditionally
approved paragraph IWA-4520(a)(1) in the 2001 Edition through 2008
Addenda. Therefore, it would be inconsistent to retain the condition on
the Code Case.
Due to the removal of Condition (2), proposed Conditions (3), (4),
and (5) have been renumbered as Conditions (2), (3), and (4). Proposed
Condition (5) has been revised as recommended in the comment.
Code Case N-597-2
Comment: It is unclear whether proposed Condition (6) prohibits the
use of the Code Case for moderate-energy Class 2 and 3 piping. If the
intent of this condition is to allow the use of this case only until
the next refueling outage for moderate-energy Class 2 and 3 piping,
this condition should be clarified. In addition, the reference to Code
Case N-513-2 should be removed from the proposed condition since Code
Case N-513-3 is listed in Table 2 of RG 1.147. Because the condition
imposed on the use of Code Case N-513-3 already restricts the use of N-
513-3 until a
[[Page 65783]]
repair/replacement activity can be performed during the next refueling
outage, the proposed condition is not needed for Code Case N-597-2.
Proposed Condition (6) should, therefore, be removed or revised to
clarify the intent. (ASME-5.2.4)
NRC Response: The NRC disagrees with this comment. As discussed in
the statement of considerations for the proposed rule (78 FR 37886;
June 24, 2013), the NRC had received a comment in a previous rulemaking
(74 FR 26303; June 2, 2009), suggesting that the method described in
Code Case N-513-2 for the temporary acceptance of flaws in moderate
energy piping be added to Code Case N-597-2. The NRC agreed that it
should be permissible under certain circumstances for licensees to
evaluate local pipe wall thinning under Code Case N-597-2 without the
NRC review and acceptance. The intent of Condition (6) was to reference
the method in Code Case N-513-2 so that all of the provisions,
formulas, graphs, and figures would not have to be duplicated in
conditions to Code Case N-597-2.
As also discussed in the statement of considerations for the
proposed rule, the circumstances under which such an evaluation is
conducted must be limited, because Code Case N-597-2 is applicable to
all the ASME Code class piping (including high energy piping), whereas
Code Case N-513-2 is limited to Class 2 and 3 moderate energy piping.
The NRC has only approved temporary acceptance of flaws for moderate
energy Class 2 or 3 piping (maximum operating temperature does not
exceed 200[emsp14][deg]F (93 [deg]C) and maximum operating pressure
does not exceed 275 psig (1.9 MPa)). In addition, it is not appropriate
to apply the method under Code Case N-597-2 to evaluate through-wall
leakage conditions.
Condition (6) in the proposed rule stated, ``For moderate-energy
Class 2 and 3 piping, wall thinning acceptance criteria may be
determined on a temporary basis (until the next refueling outage) based
on the provisions of Code Case N-513-2. Moderate-energy piping is
defined as Class 2 and 3 piping whose maximum operating temperature
does not exceed 200[emsp14][deg]F (93 [deg]C) and whose maximum
operating pressure does not exceed 275 psig (1.9 MPa). Code Case N-597-
2 shall not be used to evaluate through-wall leakage conditions.''
This condition has been revised in RG 1.147, Revision 17, to read
as follows: ``The evaluation criteria in Code Case N-513-2 may be
applied to Code Case N-597-2 for the temporary acceptance of wall
thinning (until the next refueling outage) for moderate-energy Class 2
and 3 piping. Moderate-energy piping is defined as Class 2 and 3 piping
whose maximum operating temperature does not exceed 200[emsp14][deg]F
(93 [deg]C) and whose maximum operating pressure does not exceed 275
psig (1.9 MPa). Code Case N-597-2 shall not be used to evaluate
through-wall leakage conditions.''
Code Case N-606-1
Comment: The proposed condition to Code Case N-606-1 is already
inherently required.
The surface preparation and cleaning prior to welding are
considered to be standard requirements by Welding Programs complying
with Sec. 50.55a specified Codes and 10 CFR part 50, appendix B
Quality Assurance Programs. Furthermore, these requirements are already
required/implied by the reference to the ASME Section IX and paragraph
3(e) of the Case. Many other instances where welding is performed, even
temper bead welding, can be found in Code Cases and in Code that do not
explicitly specify this level of detail since such details are included
in the Owner's or the Owner's Repair Organization's Welding Procedure
Specification/Welding Program. Therefore, this condition should be
removed from the regulatory guide. (ASME-5.2.5)
NRC Response: The NRC agrees that, the second sentence of the
proposed condition is redundant with requirements in Section III NB-
4412. The NRC removed the second sentence of the condition.
The NRC disagrees with the comment's suggestion to remove the first
and third sentences of the condition. The original version of Code Case
N-606, and other temper bead Code Cases (such as N-638-5), require that
prior to welding base metal, a surface examination shall be performed
on the area to be welded, so there is precedent for this level of
detail in temper bead Code Cases. This verification is not required by
Section IX of the ASME Code. The NRC has determined that this
verification is necessary to assure the necessary quality level for
temper bead welding. Therefore, the condition is necessary. No change
was made to the first and third sentences of the condition in response
to this comment.
Code Case N-619 and N-648-1
Comment: The NRC should not include the condition to Code Case N-
619 and N-648-1 which requires the 1-mil wire standard for
qualification of visual examinations for components within the scope of
these code cases. Research has shown that characters on a printed chart
are a better resolution standard than the use of 1-mil wire.
The use of printed characters for qualification will improve the
resolution of visual examinations, thus improving the capability of the
technique in detecting indications for which the examinations are
performed. (ASME-5.2.6.a, ASME-5.2.6.b)
NRC Response: Visual resolution sensitivity techniques are used to
ensure the capabilities of the examiner, and that a camera, when used,
is operating properly. The NRC conducted a preliminary assessment of
remote visual testing at Pacific Northwest National Laboratory. The
results were published in NUREG/CR-6860, ``An Assessment of Visual
Testing,'' which is available on the NRC's public Web site at http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/. The 1-mil wire
standard had been implemented in response to the requirement in the
condition for a resolution sensitivity of 1-mil. The preliminary
assessment identified issues with respect to the accuracy of using a
wire as a performance demonstration standard. Other issues were also
identified. This led to the development of a cooperative research
program between the NRC and the EPRI. This is the research effort
referenced in ASME's comment. While issues had been identified with the
use of a wire standard, the NRC decided to not consider changes in the
condition to Code Case N-619 until the cooperative research had
progressed, and it could be determined if there were other issues that
should be considered regarding visual examination.
The research has not identified any issues calling into question
the use of characters as a resolution standard. In addition as
described in NUREG/CR-6860, the research demonstrated that the
character resolution standard was superior to the wire standard. The
NRC finds the ASME's suggestion to remove the requirement for a 1-mil
wire for VT-1 procedure demonstration acceptable.
The condition has been revised to remove the 1-mil wire standard
and to allow the use of printed characters.
Code Case N-702
Comment: The proposed condition for Code Case N-702 should be
modified to reference BWRVIP-241: BWR Vessel and Internals Project,
``Probabilistic Fracture Mechanics Evaluation for the Boiling Water
Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,'' EPRI
Technical Report 1021005, October 2010 (ADAMS Accession No.
ML11119A041). The proposed condition should be revised to read as
follows: (ASME-5.2.8)
[[Page 65784]]
The technical basis supporting the implementation of this Code
Case is addressed by BWRVIP-108, and BWRVIP-241. The applicability
of Code Case N-702 must be shown by demonstrating that the criteria
in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated
December 18, 2007 (ADAMS Accession No. ML073600374), or Section 5.0
of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013
(ADAMS Accession No. ML13071A240), are met. The evaluation
demonstrating the applicability of the Code Case shall be reviewed
and approved by the NRC prior to the application of the Code Case.
NRC Response: The NRC agrees with the suggestion to reference
BWRVIP-241 in the condition. By letter dated April 19, 2013 (ADAMS
Accession No. ML13071A233), to the Chairman of the BWR Vessel and
Internals Project, the NRC stated that BWRVIP-241 was acceptable for
referencing subject to the limitations specified in the technical
report and in the NRC Safety Evaluation. The BWRVIP-241 was not
referenced in the proposed condition to ASME Code Case N-702 because
the draft RG was already in the review process when the NRC Safety
Evaluation for BWRVIP-241 was released. The basis for including BWRVIP-
241 in the reference is as follows.
The BWRVIP-108 provides the technical basis document for ASME Code
Case N-702 regarding reduction of the inspection of reactor pressure
vessel (RPV) nozzle-to-vessel shell welds and nozzle inner radius areas
from 100 percent to 25 percent for each nozzle type every 10 years. The
BWRVIP-241 provides additional probabilistic fracture mechanics (PFM)
analyses to support its proposed changes to the NRC staff's criteria
specified in the Safety Evaluation on BWRVIP-108. Based on the
additional PFM results supporting the revised criteria, along with BWR
RPV inspection results which show no indications of inservice
degradation, the NRC staff determined that the inspection of 25 percent
of each RPV nozzle type each 10-year interval is justified.
Licensees who plan to request relief from the ASME Code, Section XI
requirements for RPV nozzle-to-vessel shell welds and nozzle inner
radius sections may reference the BWRVIP-241 report as the technical
basis for the use of ASME Code Case N-702 as an alternative. However,
licensees should demonstrate the plant-specific applicability of the
BWRVIP-241 report to their units in the relief request by addressing
the conditions and limitations specified in Section 5.0 of the NRC
Safety Evaluation for BWRVIP-241. The suggested condition is identical
to the proposed condition in the draft RG other than adding the
reference to BWRVIP-241 in two places. Therefore, the NRC finds the
comment's proposal to be acceptable.
The condition on ASME Code Case N-702 has been revised to reference
BWRVIP-241.
Code Case N-739-1
Comment: The American Concrete Institute (ACI) report referenced in
the condition to Code Case N-739-1 should be clarified to reference ACI
201.1R. Note that the ASME has taken action to issue an erratum to
correct this error in the Code Case and Section XI. The reference to
ACI 201.1 R is correctly shown in Table IWA-1600-1. (ASME-5.2.9)
NRC Response: The NRC agrees with the comment. The letter ``R'' was
missing in the reference in Code Case N-739-1. The ACI uses the letter
``R'' to distinguish reports from standards. With the ASME approval of
an erratum to the Code Case restoring the letter ``R,'' the NRC can
remove the condition in final RG 1.147, Revision 17.
The NRC has unconditionally approved Code Case N-739-1 in RG 1.147,
Revision 17.
Code Cases N-798 and N-800
Comment: Although Code Cases N-798 and N-800 have not been included
in DG-1231, the NRC should include both of these cases in the next
draft revision to RG 1.147. Until such time that N-798 and N-800 are
included in RG 1.147, owners will continue to seek relief pursuant to
Sec. 50.55a(a)(3) [Sec. 50.55a(z) in the draft rule] to use
provisions of these cases or similar alternatives. (ASME-5.2.10)
NRC Response: The NRC agrees with the comment and plans to address
these code cases in Supplement 11 to the 2007 Edition through
Supplement 10 to the 2010 Edition in draft Revision 18 to RG 1.147.
Code Cases N-798 and N-800 were not included in the draft regulatory
guide because they were issued in Supplement 4 to the 2010 Edition,
which was not considered for this regulatory guide.
No change was made to this final rule as a result of this comment.
Regulatory Guide 1.192, Revision 1 (DG-1232)
Code Case OMN-1
Comment: DG-1232 incorrectly identifies ASME Code Case OMN-1 (2006
Addenda) as ``Revision 0.'' The version of OMN-1 published with the
2006 Addenda does not include the identifier, ``Revision 0.''
(Comstock-2.1)
NRC Response: The NRC agrees with this comment. The ASME OMN-1 Code
Case published with the 2006 Addenda did not include the identifier
``Revision 0.'' Accordingly, RG 1.192, Revision 1, has been revised to
remove the words ``Revision 0'' from the first sentence of the first
paragraph in Table 2, under OMN-1 conditions.
Comment: The descriptions in the first and second sentence say OMN-
1 may be used in lieu of the provisions for stroke time testing.
However, OMN-1 says it may be used in place of all provisions with the
exception of leak testing. The conditions placed on the use of OMN-1
restrict its use in place of existing other ISTC requirements, such as
position indication verification and periodic (quarterly, cold
shutdown, refueling outage) exercising. All provisions of ISTC are
implemented in OMN-1 with the exception of leak testing. The leak
testing requirement of ISTC is referenced as a necessary requirement by
the Code Case. Strike out the words ``stroke-time'' in the first and
second sentences of Table 2 in DG-1232 to resolve this problem.
(Comstock-2.2)
NRC Response: The NRC disagrees with this comment. The general
discrepancy noted in the comment is that draft RG 1.192 (DG-1232)
states OMN-1 ``may be used in lieu of the provisions for stroke time
testing'' versus OMN-1, which states ``it may be used in place of all
provisions.'' After evaluating the comment, the NRC believes both
statements are correct and the same for the following reasons.
The requirements of the ASME OM Code, Subsection ISTC, can be
simplified as having three test requirements:
1. ISTC-3500--``Valve Testing Requirements''
2. ISTC-3600--``Leak Testing Requirements''
3. ISTC-3700--``Position Verification Testing''
Section ISTC-3500 of the ASME OM Code describes valve test
requirements, such as exercise test frequency and obturator movement
verification. Specific instructions for the different valve types can
be found in Section ISTC-5000, ``Specific Testing Requirements,'' of
the ASME OM Code. The ASME OM Code section for specific test
requirements for motor-operated valves (MOVs) is ISTC-5120. The first
specific instruction for an MOV test is ISTC-5121(a), ``Valve Stroke
Testing,'' which states, ``Active valves shall have their stroke times
measured when exercised in accordance with ISTC-3500.'' The specific
instruction for the
[[Page 65785]]
stroke-time test encompasses all the requirements of ISTC-3500. Leak
testing requirement ISTC-3600 remains the same. The position
verification test is not specifically spelled out in the ASME OM Code
Case OMN-1, but credit is given on the basis that OMN-1 requires
diagnostic testing of MOVs to verify that they are set up correctly and
will meet their design basis function.
The comment also stated that all provisions of ISTC are implemented
in OMN-1. This statement is not fully accurate. After a recent industry
valve failure, it has been noted by the ASME OM Code Subgroup committee
on MOVs that the ASME OM Code Case OMN-1 does not directly address the
issue of verifying obturator movement, which is required in Section
ISTC-3530. The subgroup committees for ISTC and MOVs are currently
working on addressing this issue. Also, a review of past NRC documents,
regulatory guides, and safety evaluations were completed. The majority
of the NRC correspondence refers to ASME OM Code requirements for MOVs
as being ``stroke time testing.''
No change has been made to RG 1.192, Revision 1, as a result of
this comment.
Code Case OMN-11
Comment: In DG-1232, delete the first sentence in Condition (2) on
OMN-11 (2006 Addenda). It exceeds the NRC's authority.
In DG-1232, the conditions on OMN-11 (2006 addenda) add an
unnecessary administrative burden.
In DG-1232, in the discussion of OMN-11 (2006 addenda), Condition
(1) should be deleted. This defeats the purpose of alternate
requirements.
In DG-1232, in the discussion of OMN-11 (2006 addenda), Condition
(2) should be deleted. The OMN-11 3(b) rule requires the same treatment
to be applied as OMN-1 3.5(b) by requiring an evaluation of all test
results for every MOV in the group. The OMN-11 3(d) rule requires all
low safety significant components (LSSC) to be tested over a 10-year
period. This requires the same treatment to be applied as OMN-1 3.5(d)
over a 10-year period, which requires testing for all valves in the
group. The OMN-1 3.5(e) simply says the test results for a
representative MOV from the group shall be applied to all MOVs in the
group when doing the section 6 analyses and evaluation. This is the
same rule described within the OMN-11 3(b) requirement that requires
test results from an individual valve within a group to be applied to
all MOVs within the group.
In DG-1232, in the discussion of OMN-11 (2006 addenda), Condition
(3) should be deleted. It is already imposed for OMN-1 (required for
OMN-11).
In DG-1232, in the discussion of OMN-11 (2006 addenda), note 1
should be deleted because it is circular and provides no guidance or
information.
In DG-1232, in the discussion of OMN-11 (2006 addenda), note 2
directs the reader to the wrong edition (2004) for OMN-1. If it
referenced 2006, it would not provide any new information.
In DG-1232, in the discussion of OMN-11 (2006 addenda), note 3
should be incorporated into Table 2 OMN-1 note 2 or deleted. (Comstock-
2.3)
NRC Response: The NRC agrees that the specification of conditions
in Table 2 of RG 1.192 on Code Case OMN-11 in the 2006 Addenda of the
ASME OM Code is not necessary because OMN-1 in the 2006 Addenda has
incorporated the provisions from OMN-11. Therefore, OMN-11 has been
deleted from Table 2 of RG 1.192. A new Note 2 has been included for
OMN-1 in Table 2 of RG 1.192 explaining the incorporation of OMN-11
into OMN-1 such that the use of OMN-11 in the 2006 Addenda is no longer
appropriate. Table 3 of RG 1.192 continues to specify conditions for
the use of OMN-11 in the 2001 Edition, 2003 Addenda, and 2004 Edition
of the OM Code for those superseded versions of OMN-11. In particular,
Condition (1) on OMN-11 indicates that all provisions in OMN-1 must be
satisfied, except those allowed to be relaxed by the risk-informed
provisions in OMN-11. Condition (2) on OMN-11 indicates that only
specific provisions for grouping of MOVs in OMN-1 may be relaxed
through the use of OMN-11. Condition (3) on OMN-11 is repeated from a
similar condition on OMN-1 because OMN-11 has a specific section on
high risk MOVs. Note 1 on OMN-11 in Table 3 of RG 1.192 indicates that
the permission to use allowable risk ranking methodologies applies to
both OMN-1 and OMN-11. There are no additional notes on OMN-11 in Table
3 of RG 1.192.
Code Case OMN-12
Comment: Code Case OMN-12 should be removed from DG-1232 since its
application will always require NRC permission to implement due to the
ASME OM Code for which it applies. The conditions described for the use
of ASME Code Case OMN-12 do not allow it to be applied to any other
ASME OM Code for which it was written (ASME OM Code 1998). In light of
the current 10 CFR 50.55a regulations, this renders the Code Case
unusable for anyone in the USA through the application of RG 1.192. The
extra conditions also make the application of OMN-12 so burdensome,
that no one would be willing to incur the extra expense and
administrative burden associated with implementing this process under
the Inservice Testing Program. (Comstock-2.4)
NRC Response: The NRC disagrees with this comment. The comment
seems to be interpreting that the NRC is endorsing the use of OMN-12
only if the licensee's IST Program is based on the 1998 Code. That is
not the case. The NRC accepts with conditions the use of OMN-12 with
any Code from 1998 up to and including the 2006 Addenda.
No change has been made to the final rule as a result of this
comment.
Table 3--Code Cases That Have Been Superseded by Revised Code Cases
Comment: Table 3 of DG-1232 should be deleted. It serves no useful
purpose. The information is available via other sources. It delays the
rule. (Comstock-2.5)
NRC Response: The NRC disagrees with this comment. Table 3 in RG
1.192 lists those OM Code Cases that have been superseded by revised
Code Cases. Similar tables exist in RGs 1.84 and 1.147 addressing
Section III and Section XI Code Cases respectively. Section 50.55a
allows applicants and licensees to continue to apply superseded Code
Cases for the remainder of an inservice inspection or testing interval.
The ASME procedures require that the latest version of a Code Case be
implemented. If not for the provision in the regulation, licensees
would be required to update their inservice inspection and testing
programs for every Code Case that is revised (i.e., that the licensee
or applicant had previously implemented). Accordingly, any Code and
standard that has been incorporated by reference into Sec. 50.55a and
is still in use must continue to be listed in the regulation.
No change has been made to RG 1.192, Revision 1, as a result of
this comment.
Regulatory Guide 1.193, Revision 4 (DG-1233)
Code Case N-659-2
Comment: In DG-1233, in the discussion of N-659-2, there is a
typographical error on page 7. It should say ``radiography,'' not
``radiology.'' (ASME-5.4.1)
NRC Response: The NRC agrees with this comment.
The NRC corrected the title of Code Case N-659-2 in RG 1.193,
Revision 4.
[[Page 65786]]
N-805
Comment: The U.S. Nuclear Regulatory Commission (NRC) should
consider including in this rulemaking Code Case N-805, ``Alternative to
Class 1 Extended Boundary End of lnterval or Class 2 System Leakage
Testing of the Reactor Vessel Head Flange O-Ring Leak-Detection System
Section XI, Division 1.'' (Inservice Inspection Program Owners Group-
1.1)
NRC Response: The NRC declines to adopt the suggestion to adopt
Code Case N-805 in the final rulemaking and final regulatory guide.
Code Case N-805 was published by the ASME in Supplement 6 to the 2010
Edition which was not considered for inclusion in this rulemaking and
draft regulatory guide. The NRC plans to include Code Case N-805 in
draft Revision 18 to RG 1.147 which is scheduled for public comment in
spring 2015.
No change was made to the final rule as a result of this comment.
IV. NRC Approval of New and Amended ASME Code Cases
This final rule incorporates by reference the latest revisions of
the NRC's RGs that list ASME BPV and OM Code Cases the NRC finds to be
acceptable or ``conditionally acceptable'' (i.e., NRC-specified
conditions). Regulatory Guide 1.84, Revision 36 (ADAMS Accession No.
ML13339A515), supersedes the incorporation by reference of Revision 35;
RG 1.147, Revision 17 (ADAMS Accession No. ML13339A689), supersedes the
incorporation by reference of Revision 16; and RG 1.192, Revision 1
(ADAMS Accession No. ML13340A034), supersedes the incorporation by
reference of Revision 0.
This final rule addresses two categories of ASME Code Cases. The
first category of Code Cases are the new and revised Section III and
Section XI Code Cases listed in Supplements 1 through 10 to the 2007
Edition of the BPV Code, and the OM Code Cases published with the 2002
Addenda through the 2006 Addenda. The second category is the Code Cases
that were not addressed in the final rule published in the Federal
Register on October 5, 2010 (75 FR 61321). The 2010 final rule
addressed the new and revised Section III and Section XI Code Cases
listed in Supplements 2 through 11 to the 2004 Edition and Supplement 0
to the 2007 Edition of BPV Code. Public comments were received during
the proposed rule stage (June 2, 2009; 74 FR 26303) on (Code Cases N-
508-4, N-597-2, N-619, N-648, N-702, and N-748) requesting that the NRC
include certain revised Code Cases in the final guides that were not
listed in the draft guides. The NRC determined that the revised Code
Cases represented changes significant enough to warrant broader public
participation prior to the NRC making a final determination of them.
Accordingly, the NRC requested comment on these Code Cases in the
proposed rule (June 24, 2013; 78 FR 37886). The comment responses shown
earlier include responses to those Code Cases.
The latest editions and addenda of the ASME BPV and OM Codes that
the NRC has approved for use are referenced in Sec. 50.55a. The ASME
also publishes Code Cases that provide alternatives to existing Code
requirements developed and approved by ASME. The final rule
incorporated by reference RGs 1.84, 1.147, and 1.192. The NRC, by
incorporating by reference these three RGs, allows nuclear power plant
licensees and applicants for standard design certifications, standard
design approvals, manufacturing licenses, applicants for OLs, CPs, and
COLs under the regulations that govern license certifications, to use
the Code Cases listed in these RGs as suitable alternatives to the ASME
BPV and OM Codes for the construction, ISI, and IST of nuclear power
plant components. This action is consistent with the provisions of the
National Technology Transfer and Advancement Act of 1995, Public Law
104-113, which encourages Federal regulatory agencies to consider
adopting industry consensus standards as an alternative to de novo
agency development of standards affecting an industry. This action is
also consistent with the NRC's policy of evaluating the latest versions
of consensus standards in terms of their suitability for endorsement by
regulations or regulatory guides.
The NRC follows a three-step process to determine the acceptability
of new and revised Code Cases and the need for regulatory positions on
the uses of these Code Cases. This process was employed in the review
of the Code Cases in Supplements 1 through 10 to the 2007 Edition of
the BPV Code and the 2002 Addenda through the 2006 Addenda of the OM
Code. The Code Cases in these supplements are the subject of this final
rule. First, the ASME develops Code Cases through a consensus
development process, as administered by ANSI, which ensures that the
various technical interests (e.g., utility, manufacturing, insurance,
regulatory) are represented on standards development committees and
that their viewpoints are addressed fairly. This process includes
development of a technical justification in support of each new or
revised Code Case. The ASME committee meetings are open to the public,
and attendees are encouraged to participate. Task groups, working
groups, and subgroups report to a standards committee. The standards
committee is the decisive consensus committee and ensures that the
development process fully complies with the ANSI consensus process. The
NRC actively participates through full involvement in discussions and
technical debates of the task groups, working groups, subgroups, and
standards committee regarding the development of new and revised
standards.
Second, the standards committee transmits to its members a first
consideration letter ballot requesting comment or approval of new and
revised Code Cases. To be approved, Code Cases from the first
consideration letter ballot must receive the following: (1) Approval
votes from at least two thirds of the eligible consensus committee
membership, (2) no disapprovals from the standards committee, and (3)
no substantive comments from ASME oversight committees such as the
Technical Oversight Management Committee (TOMC). The TOMC's duties, in
part, are to oversee various standards committees to ensure technical
adequacy and provide recommendations in the development of Codes and
standards, as required. The Code Cases that are disapproved or receive
substantive comments from the first consideration ballot are reviewed
by the working level group(s) responsible for their development to
consider the comments received. These Code Cases may be approved by the
standards committee on second consideration with an approval vote by at
least two thirds of the eligible consensus committee membership, with
no more than three disapprovals from the consensus committee.
Third, the NRC reviews new and revised Code Cases to determine
their acceptability for incorporation by reference in Sec. 50.55a
through the subject RGs. This rulemaking process, when considered
together with the ANSI process for developing and approving ASME codes
and standards and ASME Code Cases, constitutes the NRC's basis that the
Code Cases (with conditions as necessary) provide reasonable assurance
of adequate protection to public health and safety.
The NRC reviewed the new and revised Code Cases identified in this
final rule and concluded, in accordance with the process previously
described, that the Code Cases are technically
[[Page 65787]]
adequate (with conditions as necessary) and consistent with current NRC
regulations. Therefore, the new and revised Code Cases listed in the
subject RGs are approved for use subject to any specified conditions.
A. ASME Code Cases Approved for Unconditional Use
The NRC determined, in accordance with the process previously
described for review of ASME Code Cases, that each ASME Code Case
listed in Table II is appropriate for incorporation by reference and
has been newly added to the RGs
Table II--Unconditionally Approved Code Cases
------------------------------------------------------------------------
Code case No. Code supplement Code case title
------------------------------------------------------------------------
ASME BPV Code Case, Section III
------------------------------------------------------------------------
N-4-13........................ 5................ Special Type 403
Modified Forgings or
Bars, Section III,
Division 1, Class 1
and CS.
N-570-2....................... 7................ Alternative Rules for
Linear Piping and
Linear Standard
Supports for Classes
1, 2, 3, and MC,
Section III,
Division 1.
N-580-2....................... 4................ Use of Alloy 600 With
Columbium Added,
Section III,
Division 1.
N-655-1....................... 2................ Use of SA-738, Grade
B, for Metal
Containment Vessels,
Class MC, Section
III, Division 1.
N-708......................... 2................ Use of JIS G-4303,
Grades SUS304,
SUS304L, SUS316, and
SUS316L, Section
III, Division 1.
N-759-2....................... 4................ Alternative Rules for
Determining
Allowable External
Pressure and
Comprehensive Stress
for Cylinders,
Cones, Spheres, and
Formed Heads,
Section III,
Division 1.
N-760-2....................... 7................ Welding of Valve
Plugs to Valve Stem
Retainers, Classes
1, 2, and 3, Section
III, Division 1.
N-767......................... 4................ Use of 21 Cr-6Ni-9Mn
(Alloy UNS S21904)
Grade GXM-11
(Conforming to SA
182/SA-182M and SA-
336/SA-336M), Grade
TPXM-11 (Conforming
to SA 312/SA-312M)
and Type XM-11
(Conforming to SA-
666) Material, for
Class 1
Construction,
Section III,
Division 1.
N-774......................... 7................ Use of 13Cr-4Ni
(Alloy UNS S41500)
Grade F6NM Forgings
Weighing in Excess
of 10,000 lb (4,540
kg) and Otherwise
conforming to the
Requirements of SA-
336/SA-336M for
Class 1, 2, and 3
Construction,
Section III,
Division 1.
N-782......................... 9................ Use of Editions,
Addenda, and Cases,
Section III,
Division 1.
N-801......................... 4 (2010 Edition). Rules for Repair of N-
Stamped Class 1, 2,
and 3 Components by
Organization Other
Than the N
Certificate Holder
That Originally
Stamped the
Component Being
Repaired, Section
III, Division 1.
N-802......................... 4 (2010 Edition). Rules for Repair of
Stamped Components
by the N Certificate
Holder That
Originally Stamped
the Component,
Section III,
Division 1.
------------------------------------------------------------------------
ASME BPV Code Case, Section XI
------------------------------------------------------------------------
N-532-5....................... 5................ Alternative
Requirements to
Repair and
Replacement
Documentation
Requirements and
Inservice Summary
Report Preparation
and Submission as
Required by IWA-4000
and IWA-6000,
Section XI, Division
1.
N-716-1....................... 1 (2013 Edition). Alternative Piping
Classification and
Examination
Requirements,
Section XI, Division
1.
N-739-1....................... 1................ Alternative
Qualification
Requirements for
Personnel Performing
Class CC Concrete
and Post-Tensioning
System Visual
Examinations,
Section XI, Division
1.
N-747......................... 9................ Reactor Vessel Head-
to-Flange Weld
Examinations,
Section XI, Division
1.
N-762......................... 1................ Temper Bead Procedure
Qualification
Requirements for
Repair/Replacement
Activities Without
Post Weld Heat
Treatment, Section
XI, Division 1.
N-765......................... 8................ Alternative to
Inspection Interval
Scheduling
Requirements of IWA-
2430, Section XI,
Division 1.
N-769......................... 8................ Roll Expansion of
Class 1 In-Core
Housing Bottom Head
Penetrations in
BWRs, Section XI,
Division 1.
N-773......................... 8................ Alternative
Qualification
Criteria for Eddy
Current Examinations
of Piping Inside
Surfaces, Section
XI, Division 1.
------------------------------------------------------------------------
ASME OM Code Case
------------------------------------------------------------------------
OMN-6......................... 2006 Addenda..... Alternate Rules for
Digital Instruments.
OMN-8......................... 2006 Addenda..... Alternative Rules for
Preservice and
Inservice Testing of
Power-Operated
Valves That Are Used
for System Control
and Have a Safety
Function per OM-10,
ISTC-1.1, or ISTA-
1100.
OMN-14........................ 2004 Addenda..... Alternative Rules for
Valve Testing
Operations and
Maintenance,
Appendix I: BWR CRD
Rupture Disk
Exclusion.
OMN-16........................ 2006 Addenda..... Use of a Pump Curve
for Testing.
------------------------------------------------------------------------
[[Page 65788]]
B. ASME Code Cases Approved for Use With Conditions
The NRC has determined that certain Code Cases, as issued by ASME,
are generally acceptable for use, but that the alternative requirements
specified in those Code Cases must be supplemented to provide an
acceptable level of quality and safety. Accordingly, the NRC proposes
to impose conditions on the use of these Code Cases to modify, limit or
clarify their requirements. For each applicable Code Case, the
conditions would specify the additional activities that must be
performed, the limits on the activities specified in the Code Case,
and/or the supplemental information needed to provide clarity. These
ASME Code Cases are included in Table III of the following: RG 1.84
(DG-1230), RG 1.147 (DG-1231), and RG 1.192 (DG-1232). The NRC's
evaluation of the Code Cases and the reasons for the NRC's conditions
are discussed in the following paragraphs.
Table III--Conditionally Approved Code Cases
----------------------------------------------------------------------------------------------------------------
Code case No. Code supplement Code case title Conditions
----------------------------------------------------------------------------------------------------------------
ASME BPV Code Case, Section III
----------------------------------------------------------------------------------------------------------------
N-60-5........................ Reinstating condition. Material for Core The maximum yield strength of
Support Structures, strain-hardened austenitic
Section III, Division stainless steel shall not
I, Class 1. exceed 90,000 psi in view of
the susceptibility of this
material to environmental
cracking.
N-208-2....................... 4..................... Fatigue Analysis for (1) In Figure A, the words ``No
Precipitation mean stress'' shall be
Hardening Nickel implemented with the
Alloy Bolting understanding that it denotes
Material to ``Maximum mean stress.''
Specification SB-637 (2) In Figure A, [sigma]y shall
N07718 for Class 1 be implemented with the
Construction, Section understanding that it denotes
III, Division 1. [sigma]max.
N-520-2....................... 4..................... Alternative Rules for The Code Case is considered
Renewal of Active or acceptable with one
Expired N-type clarification: an AIA is an
Certificates for Authorized Inspection Agency
Plants Not in Active and the AIA employs the
Construction, Section Authorized Nuclear Inspector
III, Division 1. (ANI).
N-757-1....................... 2..................... Alternative Rules for The design provisions of ASME
Acceptability for Section III, Division 1,
Class 2 and 3 Valves Appendix XIII, shall not be
(DN 25) and Smaller used for Class 3 valves.
with Welded and
Nonwelded End
Connections Other
than Flanges, Section
III, Division 1.
----------------------------------------------------------------------------------------------------------------
ASME BPV Code Case, Section XI
----------------------------------------------------------------------------------------------------------------
N-508-4....................... 8..................... Rotation of Serviced When Section XI requirements are
Snubbers and Pressure used to govern the examination
Retaining Items for and testing of snubbers and the
the Purpose of ISI Code of Record is earlier
Testing, Section XI, than Section XI, 2006 Addenda,
Division 1. Footnote 1 shall not be
applied.
N-561-2....................... 1..................... Alternative (1) Paragraph 5(b): for repairs
Requirements for Wall performed on a wet surface, the
Thickness Restoration overlay is only acceptable
of Class 2 and High until the next refueling
Energy Class 3 Carbon outage.
Steel Piping, Section (2) Paragraph 7(c): if the cause
XI, Division 1. of the degradation has not been
determined, the repair is only
acceptable until the next
refueling outage.
(3) The area where the weld
overlay is to be applied must
be examined using ultrasonic
methods to demonstrate that no
crack-like defects exist.
(4) Piping with wall thickness
less than the diameter of the
electrode shall be
depressurized before welding.
N-562-2....................... 1..................... Alternative (1) Paragraph 5(b): for repairs
Requirements for Wall performed on a wet surface, the
Thickness Restoration overlay is only acceptable
of Class 3 Moderate until the next refueling
Energy Carbon Steel outage.
Piping, Section XI, (2) Paragraph 7(c): if the cause
Division 1. of the degradation has not been
determined, the repair is only
acceptable until the next
refueling outage.
(3) The area where the weld
overlay is to be applied must
be examined using ultrasonic
methods to demonstrate that no
crack-like defects exist.
(4) Piping with wall thickness
less than the diameter of the
electrode shall be
depressurized before welding.
N-597-2....................... Previously approved Requirements for New condition (6): The
Code Case. NRC had Analytical Evaluation evaluation criteria in Code
proposed one new of Pipe Wall Case N-513-2 may be applied to
condition in response Thinning, Section XI, Code Case N-597-2 for temporary
to public comment on Division 1. acceptance of wall thinning
last rulemaking. (until the next refueling
outage) for moderate-energy
Class 2 and 3 piping. Moderate-
energy piping is defined as
Class 2 and 3 piping whose
maximum operating temperature
does not exceed 200 [deg]F (93
[deg]C) and whose maximum
operating pressure does not
exceed 275 psig (1.9MPa). Code
Case N[dash]597-2 shall not be
used to evaluate through-wall
leakage conditions.
N-606-1....................... Public comment Similar and Dissimilar Prior to welding, an examination
received on Metal Welding Using or verification must be
previously approved Ambient Temperature performed to ensure proper
rule requesting Machine GTAW Temper preparation of the base metal,
revision to Bead Technique for and that the surface is
condition. Condition BWR CRD Housing/Stub properly contoured so that an
was revised. Tube Repairs, Section acceptable weld can be
XI, Division 1. produced. This verification is
to be required in the welding
procedures.
N-619......................... Responding to comment Alternative In lieu of a UT examination,
on previously Requirements for licensees may perform a VT-1
approved Code Case. Nozzle Inner Radius examination in accordance with
Inspections for Class the code of record for the
1 Pressurizer and Inservice Inspection Program
Steam Generator utilizing the allowable flaw
Nozzles, Section XI, length criteria of Table IWB-
Division 1. 3512-1 with limiting
assumptions on the flaw aspect
ratio.
N-648-1....................... Responding to comment Alternative In lieu of a UT examination,
on previously Requirements for licensees may perform a VT-1
approved Code Case. Inner Radius examination in accordance with
Inspections for Class the code of record for the
1 Reactor Vessel Inservice Inspection Program
Nozzles, Section XI, utilizing the allowable flaw
Division 1. length criteria of Table IWB-
3512-1 with limiting
assumptions on the flaw aspect
ratio.
N-661-2....................... 1..................... Alternative (1) Paragraph 5(b): for repairs
Requirements for Wall performed on a wet surface, the
Thickness Restoration overlay is only acceptable
of Classes 2 and 3 until the next refueling
Carbon Steel Piping outage.
for Raw Water (2) Paragraph 7(c): if the cause
Service, Section XI, of the degradation has not been
Division 1. determined, the repair is only
acceptable until the next
refueling outage.
(3) The area where the weld
overlay is to be applied must
be examined using ultrasonic
methods to demonstrate that no
crack-like defects exist.
(4) Piping with wall thickness
less than the diameter of the
electrode shall be
depressurized before welding.
[[Page 65789]]
N-702......................... Responding to comment Alternative The technical basis supporting
on previously Requirements for the implementation of this Code
approved Code Case. Boiling Water Reactor Case is addressed by BWRVIP-
(BWR) Nozzle Inner 108: BWR Vessel and Internals
Radius and Nozzle-to- Project, ``Technical Basis for
Shell Welds, Section the Reduction of Inspection
XI, Division 1. Requirements for the Boiling
Water Reactor Nozzle-to-Vessel
Shell Welds and Nozzle Blend
Radii,'' EPRI Technical Report
1003557, October 2002 (ADAMS
Accession No. ML023330203); and
BWRVIP-241: BWR Vessels and
Internals Project,
``Probabilistic Fracture
Mechanics Evaluation for the
Boiling Water Reactor Nozzle-to-
Vessel Shell Welds and Nozzle
Blend Radii,'' EPRI Technical
Report 1021005, October 2010
(ADAMS Accession No.
ML11119A041). The applicability
of Code Case N-702 must be
shown by demonstrating that the
criteria in Section 5.0 of NRC
Safety Evaluation regarding
BWRVIP-108 dated December 18,
2007 (ADAMS Accession No.
ML073600374), or Section 5.0 of
NRC Safety Evaluation regarding
BWRVIP-241 dated April 19, 2013
(ADAMS Accession No.
ML13071A240), are met. The
evaluation demonstrating the
applicability of the Code Case
shall be reviewed and approved
by the NRC prior to the
application of the Code Case.
----------------------------------------------------------------------------------------------------------------
ASME OM Code Cases
----------------------------------------------------------------------------------------------------------------
OMN-1......................... 2006 Addenda.......... Alternative Rules for Licensees may use Code Case OMN-
Preservice and 1, ``Alternative Rules for
Inservice Testing of Preservice and Inservice
Active Electric Motor- Testing of Certain Electric
Operated Valve Motor-Operated Valve Assemblies
Assemblies in Light- in Light-Water Reactor Power
Water Reactor Power Plants,'' in lieu of the
Plants. provisions for stroke-time
testing in Subsection ISTC of
the 1995 Edition up to and
including the 2006 Addenda of
the ASME OM Code when applied
in conjunction with the
provisions for leakage rate
testing in, as applicable, ISTC
4.3 (1995 Edition with the 1996
and 1997 Addenda) and ISTC-3600
(1998 Edition through the 2006
Addenda). In addition,
licensees who continue to
implement Section XI of the
ASME BPV Code as their Code of
Record may use OMN-1 in lieu of
the provisions for stroke-time
testing specified in Paragraph
4.2.1 of ASME/ANSI OM Part 10
as required by 10 CFR
50.55a(b)(2)(vii) subject to
the conditions in this
regulatory guide. Licensees who
choose to apply OMN-1 must
apply all its provisions.
(1) The adequacy of the
diagnostic test interval for
each motor-operated valve (MOV)
must be evaluated and adjusted
as necessary, but not later
than 5 years or three refueling
outages (whichever is longer)
from initial implementation of
OMN-1.
(2) When extending exercise test
intervals for high risk MOVs
beyond a quarterly frequency,
licensees must ensure that the
potential increase in Core
Damage Frequency (CDF) and risk
associated with the extension
is small and consistent with
the intent of the Commission's
Safety Goal Policy Statement.
(3) When applying risk insights
as part of the implementation
of OMN-1, licensees must
categorize MOVs according to
their safety significance using
the methodology described in
Code Case OMN-3, ``Requirements
for Safety Significance
Categorization of Components
Using Risk Insights for
Inservice Testing of LWR Power
Plants,'' with the conditions
discussed in this regulatory
guide or use other MOV risk
ranking methodologies accepted
by the NRC on a plant specific
or industry-wide basis with the
conditions in the applicable
safety evaluations.
Note 1: As indicated at 64 FR
51370-51386, licensees are
cautioned that, when
implementing OMN 1, the
benefits of performing a
particular test should be
balanced against the potential
adverse effects placed on the
valves or systems caused by
this testing.
Note 2: RG 1.192, Rev. 0,
conditionally accepted Code
Case OMN-11 for use in
conjunction with Code Case OMN-
1. The provisions of Code Case
OMN-11 were acceptably
incorporated into Code Case OMN-
1, 2006 Addenda, including the
conditions in the RG on the use
of Code Case OMN-11. Code Case
OMN-11, 2006 Addenda, is
therefore no longer appropriate
for use. Accordingly,
applicants and licensees
choosing to perform risk-
informed testing of motor-
operated valves (MOVs) as
allowed by RG 1.192 must do so
in accordance with the
applicable provisions of Code
Case OMN-1 together with the
conditions specified for its
use in Table 2 of this
regulatory guide. In accordance
with 10 CFR 50.55a(b)(6)(ii),
applicants and licensees that
have implemented versions of
Code Cases OMN-1 and OMN-11
earlier than the 2006 Addenda
(i.e., with the conditions as
specified in Table 3 of this
RG) may continue to use those
versions through the end of the
current IST interval. If that
applicant or licensee plans to
continue to implement a risk-
informed IST program for its
MOVs in the subsequent IST
interval, then OMN-1, 2006
Addenda, with the conditions
specified in Table 2 of this RG
will need to be implemented.
[[Page 65790]]
OMN-3......................... 2004 Edition.......... Requirements for In addition to those components
Safety Significance identified in ASME IST Program
Categorization of Plan, implementation of Section
Components Using Risk 1, ``Applicability,'' of the
Insights for Code Case must include within
Inservice Testing of the scope of a licensee's risk-
LWR Power Plants. informed IST Program non-ASME
Code Components categorized as
high safety significant
components (HSSCs) that might
not currently be included in
the IST Program Plan.
(2) The decision criteria
discussed in Section 4.4.1,
``Decision Criteria,'' of the
Code Case for evaluating the
acceptability of aggregate risk
effects (i.e., for Core Damage
Frequency [CDF] and Large Early
Release Frequency [LERF]) must
be consistent with the guidance
provided in Regulatory Guide
1.174, ``An Approach for Using
Probabilistic Risk Assessment
in Risk-Informed Decisions on
Plant-Specific Changes to the
Licensing Basis.''
(3) Section 4.4.4, ``Defense in
Depth,'' of the Code Case must
be consistent with the guidance
contained in Sections 2.2.1,
``Defense-in-Depth
Evaluation''; and 2.2.2,
``Safety Margin Evaluation,''
of Regulatory Guide 1.175, ``An
Approach for Plant-Specific,
Risk-Informed Decisionmaking:
Inservice Testing.''
(4) Implementation of Sections
4.5, ``Inservice Testing
Program''; and 4.6,
``Performance Monitoring,'' of
the Code Case must be
consistent with the guidance
pertaining to inservice testing
of pumps and valves provided in
Section 3.2, ``Program
Implementation''; and Section
3.3, ``Performance
Monitoring,'' of Regulatory
Guide 1.175. Testing and
performance monitoring of
individual components must be
performed as specified in the
risk-informed components Code
Cases (e.g., OMN-1, OMN-4, OMN-
7, and OMN-12, as modified by
the conditions discussed in
this regulatory guide).
(5) Implementation of Section
3.2, ``Plant Specific PRA,'' of
the Code Case must be
consistent with the guidance
that the Owner is responsible
for demonstrating and
justifying the technical
adequacy of the probabilistic
risk assessment (PRA) analyses
used as the basis to perform
component risk ranking and for
estimating the aggregate risk
impact. Regulatory Guide 1.200,
``An Approach for Determining
the Technical Adequacy of
Probabilistic Risk Assessment
Results for Risk-Informed
Activities,'' provides guidance
for determining the technical
adequacy of the PRA used in a
risk-informed regulatory
activity. Regulatory Guide
1.201, ``Guidelines for
Categorizing Structures,
Systems, and Components in
Nuclear Power Plants According
to their Safety Significance,''
describes one acceptable method
to categorize the safety
significance of an active
component, including methods to
use when a plant-specific PRA
that meets the appropriate
Regulatory Guide 1.200
capability for specific hazard
group(s) (e.g., seismic and
fire) is not available.
(6) Section 4.2.4,
``Reconciliation,'' paragraph
(b), is not endorsed. The
expert panel may not classify
components that are ranked HSSC
by the results of a qualitative
or quantitative PRA evaluation
(excluding the sensitivity
studies) or the defense-in-
depth assessment to low safety
significant component (LSSC).
(7) Implementation of Section
3.3, ``Living PRA,'' must be
consistent with the following:
(1) To account for potential
changes in failure rates and
other changes that could affect
the PRA, changes to the plant
must be reviewed, and, as
appropriate, the PRA updated;
(2) When the PRA is updated,
the categorization of
structures, systems, and
components must be reviewed and
changed if necessary to remain
consistent with the
categorization process; and (3)
The review of plant changes
must be performed in a timely
manner and must be performed
once every two refueling
outages or as required by 10
CFR 50.71(h)(2) for combined
license holders.
Note 1: The Code Case
methodology for risk ranking
uses two categories of safety
significance. The NRC staff has
determined that this is
acceptable for ranking all
component types. However, the
NRC staff has accepted other
methodologies for risk ranking
MOVs, with certain conditions
that use three categories of
safety significance.
OMN-4......................... 2004 Edition.......... Requirements for Risk (1) Valve opening and closing
Insights for functions must be demonstrated
Inservice Testing of when flow testing or
Check Valves at LWR examination methods
Power Plants. (nonintrusive, or disassembly
and inspection) are used.
(2) The initial interval for
tests and associated
examinations may not exceed two
fuel cycles or 3 years,
whichever is longer; any
extension of this interval may
not exceed one fuel cycle per
extension with the maximum
interval not to exceed 10
years. Trending and evaluation
of existing data must be used
to reduce or extend the time
interval between tests.
(3) If the Appendix II condition
monitoring program is
discontinued, the requirements
of ISTC 4.5.1, ``Exercising
Test Frequency,'' through ISTC
4.5.4, ``Valve Obturator
Movement,'' (1996 and 1997
Addenda) or ISTC 3510, 3520,
3540, and 5221 (1998 Edition
with the 1999 and 2000
Addenda), as applicable, must
be implemented.
Note 1: The conditions with
respect to allowable
methodologies for OMN-3 risk
ranking specified for the use
of OMN-1 also apply to OMN-4.
OMN-9......................... 2004 Edition.......... Use of a Pump Curve (1) When a reference curve may
for Testing. have been affected by repair,
replacement, or routine
servicing of a pump, a new
reference curve must be
determined, or an existing
reference curve must be
reconfirmed, in accordance with
Section 3 of this Code Case.
(2) If it is necessary or
desirable, for some reason
other than that stated in
Section 4 of this Code Case, to
establish an additional
reference curve or set of
curves, these new curves must
be determined in accordance
with Section 3.
[[Page 65791]]
OMN-12........................ 2004 Edition.......... Alternative (1) Paragraph 4.2, ``Inservice
Requirements for Test Requirements,'' of OMN-12
Inservice Testing specifies inservice test
Using Risk Insights requirements for pneumatically
for Pneumatically and and hydraulically operated
Hydraulically valve assemblies categorized as
Operated Valve high safety significant within
Assemblies in Light- the scope of the Code Case. The
Water Reactor Power inservice testing program must
Plants (OM-Code 1998, include a mix of static and
Subsection ISTC). dynamic valve assembly
performance testing. The mix of
valve assembly performance
testing may be altered when
justified by an engineering
evaluation of test data.
(2) Paragraph 4.2.2.3 of OMN 12
specifies the periodic test
requirements for pneumatically
and hydraulically operated
valve assemblies categorized as
high safety significant within
the scope of the code case. The
adequacy of the diagnostic test
interval for each high safety
significant valve assembly must
be evaluated and adjusted as
necessary, but not later than 5
years or three refueling
outages (whichever is longer)
from initial implementation of
OMN-12.
(3) Paragraph 4.2.3, ``Periodic
Valve Assembly Exercising,'' of
OMN 12 specifies periodic
exercising for pneumatically
and hydraulically operated
valve assemblies categorized as
high safety significant within
the scope of the code case.
Consistent with the requirement
in OMN 3 to evaluate the
aggregate change in risk
associated with changes in test
strategies, when extending
exercise test intervals for
high safety significant valve
assemblies beyond a quarterly
frequency, the potential
increase in Core Damage
Frequency (CDF) and risk
associated with the extension
must be evaluated and
determined to be small and
consistent with the intent of
the Commission's Safety Goal
Policy Statement.
(4) Paragraph 4.4.1,
``Acceptance Criteria,'' of OMN
12 specifies that acceptance
criteria must be established
for the analysis of test data
for pneumatically and
hydraulically operated valve
assemblies categorized as high
safety significant within the
scope of the code case. When
establishing these acceptance
criteria, the potential
degradation rate and available
capability margin for each
valve assembly must be
evaluated and determined to
provide assurance that the
valve assemblies are capable of
performing their design basis
functions until the next
scheduled test.
(5) Paragraph 5, ``Low Safety
Significant Valve Assemblies,''
of OMN 12 specifies that the
purpose of its provisions is to
provide a high degree of
confidence that pneumatically
and hydraulically operated
valve assemblies categorized as
low safety significant within
the scope of the code case will
perform their intended safety
function if called upon. The
licensee must have reasonable
confidence that low safety
significant valve assemblies
remain capable of performing
their intended design-basis
safety functions until the next
scheduled test. The test and
evaluation methods may be less
rigorous than those applied to
high safety significant valve
assemblies.
(6) Paragraph 5.1, ``Set Points
and/or Critical Parameters,''
of OMN 12 specifies
requirements and guidance for
establishing set points and
critical parameters of
pneumatically and hydraulically
operated valve assemblies
categorized as low safety
significant within the scope of
the code case. Setpoints for
these valve assemblies must be
based on direct dynamic test
information, a test based
methodology, or grouping with
dynamically tested valves, and
documented according to
Paragraph 5.1.4. The setpoint
justification methods may be
less rigorous than provided for
high risk significant valve
assemblies.
(7) Paragraph 5.4,
``Evaluations,'' of OMN-12,
specifies evaluations to be
performed of pneumatically and
hydraulically operated valve
assemblies categorized as low
safety significant within the
scope of the Code Case. Initial
and periodic diagnostic testing
must be performed to establish
and verify the setpoints of
these valve assemblies to
ensure that they are capable of
performing their design-basis
safety functions. Methods for
testing and establishing test
frequencies may be less
rigorous than applied to high
risk significant valve
assemblies.
(8) Paragraph 5.6, ``Corrective
Action,'' of OMN-12 specifies
that corrective action must be
initiated if the parameters
monitored and evaluated for
pneumatically and hydraulically
operated valve assemblies
categorized as low safety
significant within the scope of
the code case do not meet the
established criteria. Further,
if the valve assembly does not
satisfy its acceptance
criteria, the operability of
the valve assembly must be
evaluated.
Note 1: Licensees are cautioned
that, when implementing OMN-12,
the benefits of performing a
particular test should be
balanced against the potential
adverse effects placed on the
valves or systems caused by
this testing.
Note 2: Paragraph 3.1 of OMN-12
states that ``Valve assemblies
shall be classified as either
high safety significant or low
safety significant in
accordance with Code Case OMN-
3.'' This note as well as Note
2 to OMN-4 have been added to
ensure the consistent
consideration of risk insights.
----------------------------------------------------------------------------------------------------------------
C. ASME Code Cases Not Approved for Use
The ASME Code Cases which are currently issued by ASME but not
approved for generic use by the NRC are listed in RG 1.193, ``ASME Code
Cases Not Approved for Use.'' The Code Cases which are not approved for
use include Code Cases on high-temperature gas cooled reactors; certain
requirements in Section III, Division 2, not endorsed by the NRC,
liquid metal; and submerged spent fuel waste casks. Regulatory Guide
1.193 is not incorporated by reference into Sec. 50.55a. Regulatory
Guide 1.193 is prepared by the NRC as a resource for stakeholders,
allowing
[[Page 65792]]
them to easily identify Code Cases which the NRC has not approved for
use as a generic matter. Listing of a Code Case in RG 1.193 does not
preclude an application or licensee for seeking individual, case-by-
case NRC approval to use a listed Code Case.
V. Petition for Rulemaking (PRM-50-89)
On December 14, 2007, Mr. Raymond West (the petitioner) submitted a
PRM requesting the NRC to amend Sec. 50.55a to allow consideration of
alternatives to the NRC-approved ASME BPV and OM Code Cases. The
petitioner submitted an amended petition on December 19, 2007 (ADAMS
Accession No. ML073600974). The petition was docketed by the NRC as
PRM-50-89. The petitioner requested that the regulations be amended to
provide applicants and licensees a process for requesting NRC approval
of changes or modifications to ASME Code Cases that are listed in the
relevant NRC-approved RGs cited in the current regulations. The
petitioner stated that the current requirements do not allow changes or
modifications to be proposed as alternatives to NRC-approved ASME Code
Cases, and asserted that such changes or modifications should be
allowed as alternatives to NRC Code Cases. Overall, the petitioner
requested that the regulations be amended to allow applicants and
licensees to request authorization of NRC-approved Code Cases with
proposed modifications directly through Sec. 50.55a(a)(3).
The NRC determined that the issues raised in this PRM should be
considered in the NRC's rulemaking process, and the NRC published a FRN
with this determination on April 22, 2009 (74 FR 18303).
The NRC believes that Code Cases often provide alternatives that
have technical merit and, in many instances, are incorporated into
future ASME Code editions. The ASME Code Case process itself
constitutes a method of how an applicant or licensee can seek to obtain
ASME approval for a variation of a previously-approved Code provision.
Section 50.55a(a)(3) currently provides specific approaches for
obtaining NRC authorization of alternatives to ASME Code provisions.
Inasmuch as ASME Code Cases are analogous to ASME Code provisions, it
is not unreasonable to provide an analogous regulatory approach for
obtaining NRC authorization of alternatives to ASME Code Cases.
Therefore, the NRC has included language in Sec. 50.55a(z) (previously
Sec. 50.55a(a)(3)) that would allow applicants and licensees to
request authorization of alternatives for changes to conditions on NRC-
approved ASME Code Cases in current paragraphs (b)(4), (b)(5), and
(b)(6) of Sec. 50.55a. In addition, the NRC is extending the scope of
the petitioner's request for allowing alternatives to NRC-approved Code
Case conditions to allow applicants and licensees to request
authorization of alternatives for changes to conditions on Section III
and XI of the ASME BPV Code and OM Code in current paragraphs (b)(1),
(b)(2), and (b)(3).
In the final rule, the requirements in former paragraph (a)(3) have
been moved to newly created paragraph (z), making room in this section
for the listing of all standards to be incorporated by reference in
paragraph (a). The reasons for this change is discussed in the
SUPPLEMENTARY INFORMATION in Section VI. Changes addressing the Office
of the Federal Register's Guidelines on Incorporation by Reference.
This final rule resolves and represents the NRC's final action on
PRM-50-89.
VI. Changes Addressing the Office of the Federal Register's Guidelines
on Incorporation by Reference
This final rule includes changes to Sec. Sec. 50.54, 50.55, and
50.55a. These changes were made in accordance with the guidance for
incorporation by reference of multiple standards that are included in
Chapter 6 of the OFR's ``Federal Register Document Drafting Handbook,''
January 2011 Revision. This latest revision of the OFR's guidance
provides several options for incorporating by reference multiple
standards into regulations.
The NRC has incorporated by reference, in a single paragraph, the
multiple standards mentioned in Sec. 50.55a. For the least disruption
to the existing structure of the section, the NRC incorporated by
reference the multiple standards into Sec. 50.55a(a), the first
paragraph of the section. Each national consensus standard that is
being incorporated by reference in Sec. 50.55a has been listed
separately. Accordingly, the regulatory language of Sec. Sec. 50.54,
50.55, and 50.55a has been reorganized by moving existing paragraphs,
creating new paragraphs, and revising introductory and regulatory
texts.
The NRC has made conforming changes to references throughout Sec.
50.55a to reflect this reorganization. A detailed discussion of the
affected paragraphs, other than the aforementioned reference changes,
is provided in Section VIII, ``Paragraph-by-Paragraph Discussion,'' of
this document. The regulatory text of Sec. 50.55a has been set out in
its entirety for the convenience of the reader. The NRC staff has also
developed reader aids to help users understand these changes (see
Section VII of this document).
VII. Addition of Headings to Paragraphs
The NRC has added headings (explanatory titles) to paragraphs and
all lower-level subparagraphs of Sec. 50.55a. These headings are
intended to enhance the readers' ability to identify the paragraphs
(e.g., paragraphs (a), (b), (c)) and subparagraphs with the same
subject matter. The NRC evaluated a range of solutions, including the
creation of new regulations and relocation of existing requirements
from Sec. 50.55a to the new regulations.
Some alternatives the NRC considered were a new regulation adjacent
to Sec. 50.55a (e.g., Sec. Sec. 50.55b, 50.55c, 50.55d), a new
subpart containing a new series of regulations at the end of 10 CFR
part 50 (e.g., subpart B beginning at Sec. 50.200, and continuing with
Sec. Sec. 50.201, 50.202, 50.203), or a new part (designated for Codes
and standards) containing a new series of regulations addressing Codes
and standards approved for incorporation by reference by the OFR. The
relocation of each existing requirement to a new regulation (or set of
regulations) would follow a set of organizing principles established by
the NRC after consideration of public views.
Upon consideration of these alternatives, the NRC decided that
these alternatives should not be adopted--at least not at this time
without further public input--and instead that the NRC should develop
and adopt headings for paragraphs and subparagraphs. The primary reason
for the NRC's decision is external stakeholders' objections to a
previous attempt by the NRC to re-designate paragraphs in Sec. 50.55a
(75 FR 24324; May 4, 2010). As the NRC understands it, many nuclear
power plant licensees' procedures reference specific paragraphs and
subparagraphs of Sec. 50.55a. It would require substantial rewriting
of these procedures and documents to correct the references to the old
(superseded) section, paragraphs and subparagraphs. In addition,
currently-approved design certification rules may require conforming
amendments to be made to correct references to ASME Code provisions on
design (and possibly ISI and IST). As mentioned earlier in the response
to Comment No. 1, the NRC received several public comments but deferred
their consideration to a potential future rulemaking effort for
reorganizing the entire Sec. 50.55a with public input. The current
reorganization of this
[[Page 65793]]
rulemaking is based upon two major issues- consideration of the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations and addition of headings (explanatory titles) to
paragraphs and lower-level subparagraphs of Sec. 50.55a as reader
aids.
A. NRC's Convention for Headings and Subheadings
The NRC has added headings to all first, second, third, fourth, and
some fifth-level paragraphs for certain sections of Sec. 50.55a to add
clarity and a user-friendly method for following sublevel contents
within a regulation. The heading for a fourth-level follows the same
convention, but may designate the provision number only. Fifth-level
paragraphs are only for newly incorporated Code Cases. Each first-level
paragraph (designated using letters [e.g., (a), (b), (c)]) have a
heading that concisely describes the general subject matter addressed
in that paragraph. Each second-level paragraph (designated using
numbers [e.g., (1), (2), (3)] have a heading comprised of a summary of
the first-level paragraph's heading and a semicolon (``;''), followed
by a concise description of the subject matter addressed in the second
paragraph. The heading for a third-level paragraph follows the same
convention (i.e., a heading comprised of a summary level of the higher-
level paragraph's title and a semicolon, followed by a concise
description of the subject matter addressed in that subparagraph). The
heading for a fourth-level paragraph follows the same convention, but
designate the provision number only. The fifth-level paragraph is
applied to only paragraph (a) for incorporation by reference of
approved editions and addenda to the ASME BPV and OM Codes.
B. Reader Aids
The NRC staff has developed a table showing the structure of Sec.
50.55a. This table, ``Final Reorganization of Paragraphs and
Subparagraphs in 10 CFR 50.55a, `Codes and standards''' (ADAMS
Accession No. ML14015A191), is available in a separate document and
outlines the section showing all paragraph designations, including the
new paragraph headings. The NRC staff has also developed cross-
reference tables showing the current designations for Sec. Sec. 50.54,
50.55, and 50.55a regulations and the new designations for these
sections. These tables contain the new headings and a description of
each change and are available in separate documents (ADAMS Accession
No. ML14211A050- package contains two tables).
VIII. Paragraph-by-Paragraph Discussion
Overall Considerations on the Use of ASME Code Cases
This rulemaking has amended Sec. 50.55a to incorporate by
reference RG 1.84, Revision 36, which supersedes Revision 35; RG 1.147,
Revision 17, which supersedes Revision 16; and RG 1.192, Revision 1,
which supersedes Revision 0. The following general guidance applies to
the use of the ASME Code Cases approved in the latest versions of the
RGs that are incorporated by reference into Sec. 50.55a as part of
this rulemaking.
The approval of a Code Case in the NRC RGs constitutes acceptance
of its technical position for applications that are not precluded by
regulatory or other requirements or by the recommendations in these or
other RGs. The applicant and/or licensee are responsible for ensuring
that use of the Code Case does not conflict with regulatory
requirements or licensee commitments. The Code Cases listed in the RGs
are acceptable for use within the limits specified in the Code Cases.
If the RG states an NRC condition on the use of a Code Case, then the
NRC condition supplements and does not supersede any condition(s)
specified in the Code Case, unless otherwise stated in the NRC
condition.
The ASME Code Cases may be revised for many reasons (e.g., to
incorporate operational examination and testing experience and to
update material requirements based on research results). On occasion,
an inaccuracy in an equation is discovered or an examination, as
practiced, is found not to be adequate to detect a newly discovered
degradation mechanism. Hence, when an applicant or a licensee initially
implements a Code Case, Sec. 50.55a requires that the applicant or the
licensee implement the most recent version of that Code Case as listed
in the RGs incorporated by reference. Code Cases superseded by revision
are no longer acceptable for new applications unless otherwise
indicated.
Section III of the ASME BPV Code applies only to new construction
(i.e., the edition and addenda to be used in the construction of a
plant are selected based on the date of the construction permit and are
not changed thereafter, except voluntarily by the applicant or the
licensee). Hence, if a Section III Code Case is implemented by an
applicant or a licensee and a later version of the Code Case is
incorporated by reference into Sec. 50.55a and listed in the RGs, the
applicant or the licensee may use either version of the Code Case
(subject, however, to whatever change requirements apply to its
licensing basis (e.g., Sec. 50.59)).
A licensee's ISI and IST programs must be updated every 10 years to
the latest edition and addenda of Section XI and the OM Code,
respectively, that were incorporated by reference into Sec. 50.55a and
in effect 12 months prior to the start of the next inspection and
testing interval. Licensees who were using a Code Case prior to the
effective date of its revision may continue to use the previous version
for the remainder of the 120-month ISI or IST interval. This relieves
licensees of the burden of having to update their ISI or IST program
each time a Code Case is revised by the ASME and approved for use by
the NRC. Code Cases apply to specific editions and addenda, and Code
Cases may be revised if they are no longer accurate or adequate, so
licensees choosing to continue using a Code Case during the subsequent
ISI or IST interval must implement the latest version incorporated by
reference into Sec. 50.55a and listed in the RGs.
The ASME may annul Code Cases that are no longer required, are
determined to be inaccurate or inadequate, or have been incorporated
into the ASME BPV or OM Codes. If an applicant or a licensee applied a
Code Case before it was listed as annulled, the applicant or the
licensee may continue to use the Code Case until the applicant or the
licensee updates its Construction Code of Record (in the case of an
applicant, updates its application) or until the licensee's 120 month
ISI or IST update interval expires, after which the continued use of
the Code Case is prohibited unless NRC authorization is given under the
current Sec. 50.55a(a)(3). If a Code Case is incorporated by reference
into Sec. 50.55a and later annulled by the ASME because experience has
shown that the design analysis, construction method, examination
method, or testing method is inadequate; the NRC will amend Sec.
50.55a and the relevant RG to remove the approval of the annulled Code
Case. Applicants and licensees should not begin to implement such
annulled Code Cases in advance of the rulemaking.
A Code Case may be revised, for example, to incorporate user
experience. The older or superseded version of the Code Case cannot be
applied by the licensee or applicant for the first time.
If an applicant or a licensee applied a Code Case before it was
listed as superseded, the applicant or the licensee may continue to use
the Code
[[Page 65794]]
Case until the applicant or the licensee updates its Construction Code
of Record (in the case of an applicant, updates its application) or
until the licensee's 120-month ISI or IST update interval expires,
after which the continued use of the Code Case is prohibited unless NRC
authorization is given under new Sec. 50.55a(z). If a Code Case is
incorporated by reference into Sec. 50.55a and later a revised version
is issued by the ASME because experience has shown that the design
analysis, construction method, examination method, or testing method is
inadequate; the NRC will amend Sec. 50.55a and the relevant RG to
remove the approval of the superseded Code Case. Applicants and
licensees should not begin to implement such superseded Code Cases in
advance of the rulemaking.
Incorporation by Reference
The final rule includes changes to Sec. Sec. 50.54, 50.55, and
50.55a. This change brings the NRC's requirements into compliance with
the OFR's revised guidelines for incorporating by reference consensus
standards in regulations.
Section 50.54
In Sec. 50.54, the introductory statement has been revised to
include a reference to Sec. 50.55a. This revision clarifies that
nuclear power plant licensees, as described in the introductory
paragraph of Sec. 50.54, also are subject to the applicable
requirements delineated in Sec. 50.55a. In addition, the NRC revised
the introductory text of this section and added and reserved paragraph
(ii), and added paragraph (jj) to include a condition of every license.
This requirement is currently contained in Sec. 50.55a(a)(1), and no
change to the requirement is intended by the transfer of this
requirement from Sec. 50.55a(a)(1) to Sec. 50.54(jj), except for
clarification of its applicability.
Section 50.55
In Sec. 50.55, the introductory text has been revised to include
references to existing Sec. 50.55a, and paragraphs (g) and (h) have
been added and reserved for future use. Further, existing Sec.
50.55a(a)(1) has been moved to a newly created Sec. 50.55(i) enabling
the removal of the current regulation from the current 50.55a(a)(1). No
change to the requirement is intended by this transfer, except for
clarification of its applicability. The introductory text of Sec.
50.55 has been revised to maintain the existing applicability of the
requirement in the newly created Sec. 50.55(i) to construction permits
for utilization facilities.
Section 50.55a
The introductory text to Sec. 50.55a was relocated to several
other locations. There is no introductory text to Sec. 50.55a in the
new rule. The first sentence in the previous introductory text was
relocated to the first sentence in Sec. 50.55. The remaining sentences
were relocated to Sec. 50.55a(b) (second sentence), Sec. 50.55a(b)(1)
(first sentence), Sec. 50.55a(b)(4) (first sentence), Sec. 50.55a(c)
(second sentence), Sec. 50.55a(d) (second sentence), Sec. 50.55a(e)
(second sentence), Sec. 50.55a(f) (second and third sentences), Sec.
50.55a(g) (second and third sentences), and Sec. 50.55a(h) (second
sentence).
In addition to moving existing paragraphs, creating new paragraphs,
and revising introductory and regulatory texts, the footnotes in Sec.
50.55a have been reorganized to appear in sequential order. The NRC
also has reserved footnote numbers so that the NRC may add a footnote
in a future rulemaking without having to renumber the existing
footnotes.
Paragraph (a): A new paragraph (a) has been created in Sec. 50.55a
to incorporate by reference the multiple standards currently identified
in existing Sec. 50.55a. The heading has been revised to read
``Documents approved for incorporation by reference.''
Paragraph (a)(1): This paragraph, ``American Society of Mechanical
Engineers (ASME),'' has been added to group all ASME sections.
Paragraph (a)(1)(i): This paragraph, ``ASME Boiler and Pressure
Vessel Code, Section III,'' has been added to discuss the availability
of standards referenced in current paragraph (b)(1).
Paragraph (a)(1)(i)(A): This paragraph, ``Rules for Construction of
Nuclear Vessels,'' has been added to group all the individual standards
referenced regarding the subject matter included in current paragraph
(b)(1).
Paragraph (a)(1)(i)(B): This paragraph, ``Rules for Construction of
Nuclear Power Plant Components,'' has been added to group all the
individual standards referenced regarding the subject matter included
in current paragraph (b)(1).
Paragraph (a)(1)(i)(C): This paragraph, ``Division 1 Rules for
Construction of Nuclear Power Plant Components,'' has been added to
group all the individual standards referenced regarding the subject
matter included in current paragraph (b)(1).
Paragraph (a)(1)(i)(D): This paragraph, ``Rules for Construction of
Nuclear Power Plant Components--Division 1,'' has been added to group
all the individual standards referenced regarding the subject matter
included in current paragraph (b)(1).
Paragraph (a)(1)(i)(E): This paragraph, ``Rules for Construction of
Nuclear Facility Components--Division 1,'' has been added to group all
the individual standards referenced regarding the subject matter
included in current paragraph (b)(1).
Paragraph (a)(1)(ii): This paragraph, ``ASME Boiler and Pressure
Vessel Code, Section XI,'' has been added to discuss the availability
of standards referenced in current paragraph (b)(2).
Paragraph (a)(1)(ii)(A): This paragraph, ``Rules for Inservice
Inspection of Nuclear Reactor Coolant Systems,'' has been added to
discuss the availability of individual standards referenced regarding
the subject matter included in current paragraph (b)(2).
Paragraph (a)(1)(ii)(B): This paragraph, ``Rules for Inservice
Inspection of Nuclear Power Plant Components,'' has been added to
discuss the availability of individual standards referenced regarding
the subject matter included in current paragraph (b)(2).
Paragraph (a)(1)(ii)(C): This paragraph, ``Rules for Inservice
Inspection of Nuclear Power Plant Components--Division 1,'' has been
added to discuss the availability of individual standards referenced
regarding the subject matter included in current paragraph (b)(2).
Paragraph (a)(1)(iii): This paragraph, ``ASME Code Cases: Nuclear
Components,'' has been added to discuss the newly approved Code Cases
referenced regarding the subject matter in current paragraph (b).
Paragraph (a)(1)(iii)(A): This paragraph, ``ASME Code Case N-722-
1,'' has been added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b).
Paragraph (a)(1)(iii)(B): This paragraph, ``ASME Code Case N-729-
1,'' has been added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b).
Paragraph (a)(1)(iii)(C): This paragraph, ``ASME Code Case N-770-
1,'' has been added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b).
Paragraph (a)(1)(iv): This paragraph, ``ASME Operation and
Maintenance Code,'' has been added to group all the individual
standards referenced in current paragraph (b).
Paragraph (a)(1)(iv)(A): This paragraph, ``Code for Operation and
[[Page 65795]]
Maintenance of Nuclear Power Plants,'' has been added to group all the
individual standards referenced in current paragraph (b).
Paragraph (a)(1)(iv)(B): This paragraph has been added and reserved
for future use.
Paragraph (a)(2): This paragraph, ``Institute of Electrical and
Electronics Engineers (IEEE) Service Center,'' has been added to list
all IEEE sections.
Paragraph (a)(2)(i): This paragraph, ``IEEE Standard 279--1971,''
has been added to discuss the availability of standards referenced in
current paragraph (h)(2).
Paragraph (a)(2)(ii): This paragraph, ``IEEE Standard 603--1991,''
has been added to discuss the availability of the standard referenced
in current paragraphs (h)(2) and (h)(3).
Paragraph (a)(2)(iii): This paragraph, ``IEEE Standard 603--1991
correction sheet,'' has been added to discuss the availability of the
standard referenced in current paragraphs (h)(2) and (h)(3).
Paragraph (a)(3): This paragraph, ``U.S. Nuclear Regulatory
Commission (NRC) Reproduction and Distribution Services Section,''
lists all RGs being incorporated by reference.
Paragraph (a)(3)(i): This paragraph, ``NRC Regulatory Guide 1.84,
Revision 36,'' has been added to discuss the availability of the
standard.
Paragraph (a)(3)(ii): This paragraph, ``NRC Regulatory Guide 1.147,
Revision 17,'' has been added to discuss the availability of the
standard.
Paragraph (a)(3)(iii): This paragraph, ``NRC Regulatory Guide
1.192, Revision 1,'' has been added to discuss the availability of the
standard.
Paragraph (b): The paragraph heading has been revised to ``Use and
conditions on the use of standards.'' The contents have been moved, in
part, to Sec. 50.55a(a) for compliance with the OFR's revised
guidelines for incorporating by reference consensus standards in
regulations.
Paragraphs (b)(4): Reference to the revision number for RG 1.84 has
been changed from ``Revision 35'' to ``Revision 36.''
Paragraphs (b)(5): Reference to the revision number for RG 1.147
has been changed from ``Revision 16'' to ``Revision 17.''
Paragraphs (b)(6): Reference to the revision number for RG 1.192
has been changed from ``Revision 0'' to ``Revision 1.''
Paragraph (c): Introductory text has been added to the existing
paragraph (c). Explanatory headings have been added for subparagraphs.
Paragraph (d): The new paragraph adds introductory text to
``Quality Group B components,'' as part of the NRC initiative of adding
headings and providing clarity. Explanatory headings have been added
for subparagraphs.
Paragraph (e): The new paragraph adds introductory text to
``Quality Group C components,'' as part of the NRC initiative of adding
headings and providing clarity. Explanatory headings have been added
for subparagraphs.
Paragraph (f): Introductory text has been revised and expanded in
``Inservice testing requirements,'' as part of the NRC initiative of
adding headings and providing clarity. Explanatory headings have been
added for subparagraphs.
Paragraph (g): Introductory text has been revised and expanded in
``Inservice inspection requirements,'' as part of the NRC initiative of
adding headings and providing clarity. Explanatory headings have been
added for subparagraphs.
Paragraphs (b)(5), (f)(2), (f)(3)(iii)(A), (f)(3)(iv)(A),
(f)(4)(ii), (g)(2), (g)(3)(i), (g)(3)(ii), (g)(4)(i), and (g)(4)(ii):
Reference to the revision number for RG 1.147 has been changed from
``Revision 16'' to ``Revision 17.''
Paragraph (h)(1): This paragraph has been designated as reserved
because the informational content from current (h)(1) has been moved to
paragraph (a)(2).
Paragraphs (i)-(y): These paragraphs have been added and reserved
for future use.
Paragraph (z): This paragraph has been added to contain information
that has been relocated from the introductory text of current paragraph
(a)(3) and current subparagraphs (a)(3)(i)-(ii) as a result of the
NRC's compliance with the OFR's revised guidelines for incorporating by
reference consensus standards in regulations. Paragraph (z) has also
been revised to allow applicants and licensees to request alternatives
to the requirements in paragraph (b) of this section.
IX. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the
Commission certifies that this final rule would not impose a
significant economic impact on a substantial number of small entities.
This final rule would affect only the licensing and operation of
nuclear power plants. The companies that own these plants are not
``small entities'' as defined in the Regulatory Flexibility Act or the
size standards established by the NRC (10 CFR 2.810).
X. Regulatory Analysis
The ASME Code Cases listed in the RGs to be incorporated by
reference provide voluntary alternatives to the provisions in the ASME
BPV and OM Codes for design, construction, ISI, and IST of specific
structures, systems, and components used in nuclear power plants.
Implementation of these Code Cases is not required. Licensees and
applicants use NRC-approved ASME Code Cases to reduce unnecessary
regulatory burden or gain additional operational flexibility. It would
be difficult for the NRC to provide these advantages independently of
the ASME Code Case publication process without expending considerable
additional resources. The NRC has prepared a regulatory analysis
addressing the qualitative benefits of the alternatives considered in
this rulemaking and comparing the costs associated with each
alternative (ADAMS Accession No. ML14010A426). Copies of the regulatory
analysis are available to the public as indicated in Section XVIII,
``Availability of Documents,'' of this document.
XI. Backfitting and Issue Finality
The provisions in this final rule would allow licensees and
applicants to voluntarily apply NRC-approved Code Cases, sometimes with
NRC-specified conditions. The approved Code Cases are listed in three
RGs that are incorporated by references into Sec. 50.55a.
An applicant's and/or a licensee's voluntary application of an
approved Code Case does not constitute backfitting, inasmuch as there
is no imposition of a new requirement or new position. Similarly,
voluntary application of an approved Code Case by a 10 CFR part 52
applicant or licensee does not represent NRC imposition of a
requirement or action, which is inconsistent with any issue finality
provision in 10 CFR part 52. For these reasons, the NRC finds that this
final rule does not involve any provisions requiring the preparation of
a backfit analysis or documentation demonstrating that one or more of
the issue finality criteria in 10 CFR part 52 are met.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31883).
[[Page 65796]]
XIII. Finding of No Significant Environmental Impact: Environmental
Assessment
This action stems from the Commission's practice of incorporating
by reference the RGs listing the most recent set of NRC-approved ASME
Code Cases. The purpose of this action is to allow licensees to use the
Code Cases listed in the RGs as alternatives to requirements in the
ASME BPV and OM Codes for the construction, ISI, and IST of nuclear
power plant components. This action is intended to advance the NRC's
strategic goal of ensuring adequate protection of public health and
safety and the environment. It also demonstrates the agency's
commitment to participate in the national consensus standards process
under the National Technology Transfer and Advancement Act of 1995
(NTTAA), Public Law 104-113.
The National Environmental Policy Act of 1969, as amended (NEPA),
requires Federal government agencies to study the impacts of their
``major Federal actions significantly affecting the quality of the
human environment'' and prepare detailed statements on the
environmental impacts of the action and alternatives to the action (42
U.S.C. 4332(C); Sec. 102(C) of NEPA).
The Commission has determined under NEPA, as amended, and the
Commission's regulations in subpart A of 10 CFR part 51, that this rule
would not be a major Federal action significantly affecting the quality
of the human environment. Therefore, an environmental impact statement
is not required.
As alternatives to the ASME Code, NRC-approved Code Cases provide
an equivalent level of safety. Therefore, the probability or
consequences of accidents is not changed. There are also no
significant, non-radiological impacts associated with this action
because no changes would be made affecting non-radiological plant
effluents and because no changes would be made in activities that would
adversely affect the environment. The determination of this
environmental assessment is that there will be no significant offsite
impact to the public from this action.
XIV. Paperwork Reduction Act Statement
This final rule contains new or amended information collection
requirements that are subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). These requirements were approved by the
Office of Management and Budget (OMB), approval number 3150-0011.
The burden to the public for these information collections is
estimated to average a reduction of 80 hours per response, including
the time for reviewing instructions, searching existing data sources,
gathering and maintaining the data needed, and completing and reviewing
the information collection. Send comments on any aspect of these
information collections, including suggestions for further reducing the
burden, to the FOIA, Privacy, and Information Collections Branch (T-5
F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or
by email to [email protected]; and to the Desk Officer,
Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011),
Office of Management and Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
XV. Congressional Review Act
In accordance with the Congressional Review Act of 1996 (5 U.S.C.
801-808), the NRC has determined that this action is not a major rule
and has verified this determination with the Office of Information and
Regulatory Affairs of OMB.
XVI. Voluntary Consensus Standards
Section 12(d)(3) of the NTTAA, Public Law 104-113, and implementing
guidance in OMB Circular A-119 (February 10, 1998), require each
Federal government agency (should it decide that regulation is
necessary) to use a voluntary consensus standard instead of developing
a government-unique standard. An exception to using a voluntary
consensus standard is allowed where the use of such a standard is
inconsistent with applicable law or is otherwise impractical. The NTTAA
requires Federal agencies to use industry consensus standards to the
extent practical; it does not require Federal agencies to endorse a
standard in its entirety. Neither the NTTAA nor OMB Circular A-119
prohibit an agency from adopting a voluntary consensus standard while
taking exception to specific portions of the standard, if those
provisions are deemed to be ``inconsistent with applicable law or
otherwise impractical.'' Furthermore, taking specific exceptions
furthers the Congressional intent of Federal reliance on voluntary
consensus standards because it allows the adoption of substantial
portions of consensus standards without the need to reject the
standards in their entirety because of limited provisions that are not
acceptable to the agency.
In this rulemaking, the NRC is continuing its existing practice of
approving the use of ASME BPV and OM Code Cases, which are ASME-
approved alternatives to compliance with various provisions of the ASME
BPV and OM Codes. The NRC's approval of the ASME Code Cases is
accomplished by amending the NRC's regulations to incorporate by
reference the latest revisions of the following, which are the subject
of this rulemaking, into Sec. 50.55a: RG 1.84, ``Design, Fabrication,
and Materials Code Case Acceptability, ASME Section III,'' Revision 36;
RG 1.147, ``Inservice Inspection Code Case Acceptability, ASME Section
XI, Division 1,'' Revision 17; and RG 1.192, ``Operation and
Maintenance Code Case Acceptability, ASME Code,'' Revision 1. These RGs
list the ASME Code Cases that the NRC has approved for use. The ASME
Code Cases are national consensus standards as defined in the NTTAA and
OMB Circular A-119. The ASME Code Cases constitute voluntary consensus
standards, in which all interested parties (including the NRC and
licensees of nuclear power plants) participate. Therefore, the NRC's
approval of the use of the ASME Code Cases identified in RGs 1.84,
Revision 36; RG 1.147, Revision 17; and RG 1.192, Revision 1, which are
the subject of this rulemaking, is consistent with the overall
objectives of the NTTAA and OMB Circular A-119.
The NRC reviews each Section III, Section XI, and OM Code Case
published by the ASME to ascertain whether it is consistent with the
safe operation of nuclear power plants. The Code Cases found to be
generically acceptable are listed in the RGs that are incorporated by
reference in Sec. 50.55a. The Code Cases found to be unacceptable are
listed in RG 1.193, but licensees may still seek the NRC's approval to
apply these Code Cases through the processes in Sec. 50.55a for
requesting the approval of alternatives or for relief. Code Cases that
the NRC finds to be conditionally acceptable are also listed in RGs
1.84, 1.147, and 1.192, which are the subject of this rulemaking,
together with the conditions that must be used if the Code Case is
applied. The NRC believes that this rule complies with the NTTAA and
OMB Circular A-119 despite these conditions. If the NRC did not
[[Page 65797]]
conditionally accept ASME Code Cases, it would disapprove these Code
Cases entirely. The effect would be that licensees and applicants would
submit a larger number of requests for use of alternatives under the
current Sec. 50.55a(a)(3), requests for relief under Sec. 50.55a(f)
and (g), or requests for exemptions under Sec. Sec. 50.12 and/or 52.7.
For these reasons, the final rule does not conflict with any policy on
agency use of consensus standards specified in OMB Circular A-119.
The NRC did not identify any other voluntary consensus standards
developed by the United States voluntary consensus standards bodies for
use within the United States that the NRC could approve instead of the
ASME Code Cases.
The NRC also did not identify any voluntary consensus standards
developed by multinational voluntary consensus standards bodies for use
on a multinational basis that the NRC could incorporate by reference
instead of the ASME Code Cases. This is because no other multinational
voluntary consensus body would develop alternatives to a voluntary
consensus standard (i.e., either the ASME BPV Code or the ASME OM Code)
for which they did not develop and do not maintain.
In summary, this final rule satisfies the requirements of Section
12(d)(3) of the NTTAA and OMB Circular A-119.
XVII. Availability of Regulatory Guides
Regulatory Guides Being Incorporated by Reference
The NRC is issuing three revisions to existing guides in the
agency's ``Regulatory Guide'' series. This final rule is incorporating
by reference these three RGs into 10 CFR 50.55a.
Revision 36 of RG 1.84, ``Design, Fabrication, and Materials Code
Case Acceptability, ASME Section III,'' is available electronically
under ADAMS Accession No. ML13339A515.
Revision 17 of RG 1.147, ``Inservice Inspection Code Case
Acceptability, ASME Section XI, Division 1,'' is available
electronically under ADAMS Accession No. ML13339A689.
Revision 1 of RG 1.192, ``Operation and Maintenance [OM] Code Case
Acceptability, ASME OM Code,'' is available electronically under ADAMS
Accession No. ML13340A034.
As discussed in Section II of this document, ``Opportunities for
Public Participation,'' these three RGs were issued in draft form for
public comment in June 2013. The NRC staff's responses to the public
comments received are located in Section III of this document, ``Public
Comment Analysis.''
Issuance of Regulatory Guide 1.193
The NRC is issuing a revision to an existing guide in the NRC's
``Regulatory Guide'' series. This RG is not being incorporated by
reference in this final rule.
Revision 4 of RG 1.193, ``ASME Code Cases Not Approved for Use,''
was issued with a temporary identification of Draft Regulatory Guide,
DG-1233. This revision of RG 1.193 includes new information reviewed by
the NRC in ASME BPV Code Section III and Section XI Code Cases listed
in Supplements 1-10 to the 2007 Edition, and the OM Code Cases listed
in the 2002 Addenda through the 2006 Addenda. This is an update to RG
1.193, Revision 3, which included information from Supplements 2-11 to
the 2004 Edition, and Supplement 0 to the 2007 Edition of the BPV Code.
This RG does not approve the use of the Code Cases listed herein.
Licensees may submit a plant-specific request to implement one or more
of the Code Cases listed in this RG. The request must address the NRC's
concerns about the Code Case at issue.
The NRC published DG-1233 in the Federal Register on June 24, 2013
(78 FR 37848), for a 75-day public comment period. The public comment
period closed on September 9, 2013. Public comments on DG-1233 and the
NRC staff responses to the public comments are available in ADAMS under
Accession No. ML14106A577.
XVIII. Availability of Documents
The NRC is making the documents identified in Table IV available to
interested persons through one or more of the following methods, as
indicated. To access documents related to this action, see the
ADDRESSES section of this document.
Table IV--Availability of Documents
------------------------------------------------------------------------
Proposed rule documents ADAMS Accession No.
------------------------------------------------------------------------
Proposed Rule-Regulatory Analysis....... ML103060189
Proposed Rule-Federal Register Notice... ML103060003
Proposed Reorganization of Paragraphs ML12289A121
and Subparagraphs.
Draft RG 1.84, Revision 36 (DG-1230).... ML102590003
Draft RG 1.147, Revision 17 (DG-1231)... ML102590004
Draft RG 1.192, Revision 1 (DG-1232).... ML102600001
------------------------------------------------------------------------
Final rule documents ADAMS Accession No.
------------------------------------------------------------------------
Final Rule-Regulatory Analysis.......... ML14010A426
Final Rule-Federal Register Notice...... ML14008A332
Final Reorganization of Paragraphs and ML14015A191
Subparagraphs.
Cross-Reference Tables (package)........ ML14211A050
RG 1.84, ``Design, Fabrication, and ML13339A515
Materials Code Case Acceptability, ASME
Section III,'' Revision 36.
RG 1.147, ``Inservice Inspection Code ML13339A689
Case Acceptability, ASME Section XI,
Division 1,'' Revision 17.
RG 1.192, ``Operation and Maintenance ML13340A034
Code Case Acceptability, ASME OM
Code,'' Revision 1.
RG 1.193, ``ASME Code Cases Not Approved ML13350A001
for Use,'' Revision 4.
RG 1.200, ``An Approach for Determining ML090410014
the Technical Adequacy of Probabilistic
Risk Assessment Results for Risk-
informed Activities,'' Revision 2.
RG 1.201, ``Guidelines for Categorizing ML061090627
Structures, Systems, and Components in
Nuclear Power Plants According to Their
Safety Significance,'' Revision 1.
2007/12/19--``SECY--Petition for ML073600974
Rulemaking to amend 10 CFR 50.55a--
Rev.1'' submitted by Ray West.
Hatch Plant Report--``Hatch, Units 1 & ML033280037
2, Farley, Units 1 & 2, Vogtle, Units 1
& 2, Safety Evaluation Re. Request to
Use ASME Code Case N-661''.
[[Page 65798]]
EPRI Technical Report--Project No. 704-- ML023330203
BWRVIP-108: BWR Vessel & Internals
Project, Technical Basis for Reduction
of Inspection Requirements for Boiling
Water Reactor Nozzle-to-Vessel Shell
Welds & Nozzle Blend Radii.
Safety Evaluation of Proprietary EPRI ML073600374
Report--BWR Vessel and Internals
Project, Technical Basis for the
Reduction of Inspection Requirements
for the Boiling Water Reactor Nozzle-to-
Vessel Shell Welds and Nozzle Inner
Radius (BWRVIP-108).
Comment Letter--Comment (4) of Bryan A. ML092190138
Erler on Behalf of ASME Supporting
Draft Regulatory Guides DG-1191, DG-
1192, DG-1193, and the Proposed Rule
Incorporating the Final Revisions of
these Regulatory Guides into 10 CFR
50.55a.
SRM-COMNJD-03-0002--Stabilizing the PRA ML033520457
Quality Expectations and Requirements.
SECY-04-0118--Plan for the ML041470505
Implementation of the Commission's
Phased Approach to Probabilistic Risk
Assessment Quality.
SRM-SECY-04-0118--Plan for the ML042800369
Implementation of the Commission's
Phased Approach to Probabilistic Risk
Assessment Quality.
NUREG-0800--Chapter 4, Section 4.5.1, ML070230007
Revision 3, Control Rod Drive
Structural Materials, dated March 2007.
NUREG-0800--Chapter 5, Section 5.2.3, ML063190006
Revision 3, Reactor Coolant Pressure
Boundary Materials, dated March 2007.
NUREG/CR-6943--A Study of Remote Visual ML073110060
Methods to Detect Cracking in Reactor
Components.
------------------------------------------------------------------------
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Radiation protection, Reactor siting
criteria, Reporting and recordkeeping requirements.
For the reasons set forth in the preamble and under the authority
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. The authority citation for part 50 is revised to read as follows:
Authority: Atomic Energy Act secs. 102, 103, 104, 105, 147, 149,
161, 181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133, 2134,
2135, 2167, 2169, 2201, 2231, 2232, 2233, 2236, 2239, 2273, 2282);
Energy Reorganization Act secs. 201, 202, 206 (42 U.S.C. 5841, 5842,
5846); Nuclear Waste Policy Act sec. 306 (42 U.S.C. 10226);
Government Paperwork Elimination Act sec. 1704 (44 U.S.C. 3504
note); Energy Policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 194
(2005). Section 50.7 also issued under Pub. L. 95-601, sec. 10, as
amended by Pub. L. 102-486, sec. 2902 (42 U.S.C. 5851). Section
50.10 also issued under Atomic Energy Act secs. 101, 185 (42 U.S.C.
2131, 2235); National Environmental Protection Act sec. 102 (42
U.S.C. 4332). Sections 50.13, 50.54(d), and 50.103 also issued under
Atomic Energy Act sec. 108 (42 U.S.C. 2138).
Sections 50.23, 50.35, 50.55, and 50.56 also issued under Atomic
Energy Act sec. 185 (42 U.S.C. 2235). Appendix Q also issued under
National Environmental Protection Act sec. 102 (42 U.S.C. 4332).
Sections 50.34 and 50.54 also issued under sec. 204 (42 U.S.C.
5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.
97-415 (42 U.S.C. 2239). Section 50.78 also issued under Atomic
Energy Act sec. 122 (42 U.S.C. 2152). Sections 50.80-50.81 also
issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).
0
2. In Sec. 50.54, revise the introductory text, add reserved paragraph
(ii), and add paragraph (jj) to read as follows:
Sec. 50.54 Conditions of licenses.
The following paragraphs of this section, with the exception of
paragraphs (r) and (gg), and the applicable requirements of 10 CFR
50.55a, are conditions in every nuclear power reactor operating license
issued under this part. The following paragraphs with the exception of
paragraph (r), (s), and (u) of this section are conditions in every
combined license issued under part 52 of this chapter, provided,
however, that paragraphs (i) introductory text, (i)(1), (j), (k), (l),
(m), (n), (q), (w), (x), (y), (z), and (hh) of this section are only
applicable after the Commission makes the finding under Sec. 52.103(g)
of this chapter.
* * * * *
(ii) [Reserved]
(jj) Structures, systems, and components subject to the codes and
standards in 10 CFR 50.55a must be designed, fabricated, erected,
constructed, tested, and inspected to quality standards commensurate
with the importance of the safety function to be performed.
0
3. In Sec. 50.55, revise the introductory text, add reserved
paragraphs (g) and (h), and add paragraph (i) to read as follows:
Sec. 50.55 Conditions of construction permits, early site permits,
combined licenses, and manufacturing licenses.
Each construction permit for a utilization facility is subject to
the following terms and conditions and the applicable requirements of
Sec. 50.55a; each construction permit for a production facility is
subject to the following terms and conditions with the exception of
paragraph (i); each early site permit is subject to the terms and
conditions in paragraph (f) of this section; each manufacturing license
is subject to the terms and conditions in paragraphs (e), (f), and (i)
of this section and the applicable requirements of Sec. 50.55a; and
each combined license is subject to the terms and conditions in
paragraphs (e), (f), and (i) of this section and the applicable
requirements of Sec. 50.55a until the date that the Commission makes
the finding under Sec. 52.103(g) of this chapter:
* * * * *
(g) [Reserved]
(h) [Reserved]
(i) Structures, systems, and components subject to the codes and
standards in 10 CFR 50.55a must be designed, fabricated, erected,
constructed, tested, and inspected to quality standards commensurate
with the importance of the safety function to be performed.
0
4. Revise Sec. 50.55a to read as follows:
Sec. 50.55a Codes and standards.
(a) Documents approved for incorporation by reference. The
standards listed in this paragraph have been approved for incorporation
by reference by the Director of the Federal Register pursuant to 5
U.S.C. 552(a) and 1 CFR part 51. The standards are available for
inspection at the NRC Technical Library, 11545 Rockville Pike,
Rockville, Maryland 20852; telephone: 301-415-6239; or at the National
Archives and Records Administration (NARA). For information on the
availability of this material at NARA, call 202-741-6030 or go to
http://www.archives.gov/federal-register/cfr/ibr-locations.html.
(1) American Society of Mechanical Engineers (ASME), Three Park
Avenue, New York, NY 10016; telephone:
[[Page 65799]]
1-800-843-2763; http://www.asme.org/Codes/.
(i) ASME Boiler and Pressure Vessel Code, Section III. The editions
and addenda for Section III of the ASME Boiler and Pressure Vessel Code
are listed below, but limited to those provisions identified in
paragraph (b)(1) of this section.
(A) ``Rules for Construction of Nuclear Vessels:''
(1) 1963 Edition,
(2) Summer 1964 Addenda,
(3) Winter 1964 Addenda,
(4) 1965 Edition,
(5) 1965 Summer Addenda,
(6) 1965 Winter Addenda,
(7) 1966 Summer Addenda,
(8) 1966 Winter Addenda,
(9) 1967 Summer Addenda,
(10) 1967 Winter Addenda,
(11) 1968 Edition,
(12) 1968 Summer Addenda,
(13)1968 Winter Addenda,
(14) 1969 Summer Addenda,
(15) 1969 Winter Addenda,
(16) 1970 Summer Addenda, and
(17) 1970 Winter Addenda.
(B) ``Rules for Construction of Nuclear Power Plant Components:''
(1) 1971 Edition,
(2) 1971 Summer Addenda,
(3) 1971 Winter Addenda,
(4) 1972 Summer Addenda,
(5) 1972 Winter Addenda,
(6) 1973 Summer Addenda, and
(7) 1973 Winter Addenda.
(C) ``Division 1 Rules for Construction of Nuclear Power Plant
Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda,
(4) 1975 Summer Addenda,
(5) 1975 Winter Addenda,
(6) 1976 Summer Addenda, and
(7) 1976 Winter Addenda;
(D) ``Rules for Construction of Nuclear Power Plant Components--
Division 1'';
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Summer Addenda,
(10) 1980 Winter Addenda,
(11) 1981 Summer Addenda,
(12) 1981 Winter Addenda,
(13) 1982 Summer Addenda,
(14) 1982 Winter Addenda,
(15) 1983 Edition,
(16) 1983 Summer Addenda,
(17) 1983 Winter Addenda,
(18) 1984 Summer Addenda,
(19) 1984 Winter Addenda,
(20) 1985 Summer Addenda,
(21) 1985 Winter Addenda,
(22) 1986 Edition,
(23) 1986 Addenda,
(24) 1987 Addenda,
(25) 1988 Addenda,
(26) 1989 Edition,
(27) 1989 Addenda,
(28) 1990 Addenda,
(29) 1991 Addenda,
(30) 1992 Edition,
(31) 1992 Addenda,
(32) 1993 Addenda,
(33) 1994 Addenda,
(34) 1995 Edition,
(35) 1995 Addenda,
(36) 1996 Addenda, and
(37) 1997 Addenda.
(E) ``Rules for Construction of Nuclear Facility Components--
Division 1:''
(1) 1998 Edition,
(2) 1998 Addenda,
(3) 1999 Addenda,
(4) 2000 Addenda,
(5) 2001 Edition,
(6) 2001 Addenda,
(7) 2002 Addenda,
(8) 2003 Addenda,
(9) 2004 Edition,
(10) 2005 Addenda,
(11) 2006 Addenda,
(12) 2007 Edition, and
(13) 2008 Addenda.
(ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions
and addenda for Section XI of the ASME Boiler and Pressure Vessel Code
are listed below, but limited to those provisions identified in
paragraph (b)(2) of this section.
(A) ``Rules for Inservice Inspection of Nuclear Reactor Coolant
Systems:''
(1) 1970 Edition,
(2) 1971 Edition,
(3) 1971 Summer Addenda,
(4) 1971 Winter Addenda,
(5) 1972 Summer Addenda,
(6) 1972 Winter Addenda,
(7) 1973 Summer Addenda, and
(8) 1973 Winter Addenda.
(B) ``Rules for Inservice Inspection of Nuclear Power Plant
Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda, and
(4) 1975 Summer Addenda.
(C) ``Rules for Inservice Inspection of Nuclear Power Plant
Components--Division 1:''
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Winter Addenda,
(10) 1981 Summer Addenda,
(11) 1981 Winter Addenda,
(12) 1982 Summer Addenda,
(13) 1982 Winter Addenda,
(14) 1983 Edition,
(15) 1983 Summer Addenda,
(16) 1983 Winter Addenda,
(17) 1984 Summer Addenda,
(18) 1984 Winter Addenda,
(19) 1985 Summer Addenda,
(20) 1985 Winter Addenda,
(21) 1986 Edition,
(22) 1986 Addenda,
(23) 1987 Addenda,
(24) 1988 Addenda,
(25) 1989 Edition,
(26) 1989 Addenda,
(27) 1990 Addenda,
(28) 1991 Addenda,
(29) 1992 Edition,
(30) 1992 Addenda,
(31) 1993 Addenda,
(32) 1994 Addenda,
(33) 1995 Edition,
(34) 1995 Addenda,
(35) 1996 Addenda,
(36) 1997 Addenda,
(37) 1998 Edition,
(38) 1998 Addenda,
(39) 1999 Addenda,
(40) 2000 Addenda,
(41) 2001 Edition,
(42) 2001 Addenda,
(43) 2002 Addenda,
(44) 2003 Addenda,
(45) 2004 Edition,
(46) 2005 Addenda,
(47) 2006 Addenda,
(48) 2007 Edition, and
(49) 2008 Addenda.
(iii) ASME Code Cases: Nuclear Components--(A) ASME Code Case N-
722-1. ASME Code Case N-722-1, ``Additional Examinations for PWR
Pressure Retaining Welds in Class 1 Components Fabricated with Alloy
600/82/182 Materials, Section XI, Division 1'' (Approval Date: January
26, 2009), with the conditions in paragraph (g)(6)(ii)(E) of this
section.
(B) ASME Code Case N-729-1. ASME Code Case N-729-1, ``Alternative
Examination Requirements for PWR Reactor Vessel Upper Heads With
Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section
XI, Division 1'' (Approval Date: March 28, 2006), with the conditions
in paragraph (g)(6)(ii)(D) of this section.
(C) ASME Code Case N-770-1. ASME Code Case N-770-1, ``Additional
Examinations for PWR Pressure Retaining Welds in Class 1 Components
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1''
(Approval Date: December 25, 2009), with the conditions in paragraph
(g)(6)(ii)(F) of this section.
(iv) ASME Operation and Maintenance Code. The editions and addenda
for the ASME Code for Operation and Maintenance of Nuclear
[[Page 65800]]
Power Plants are listed below, but limited to those provisions
identified in paragraph (b)(3) of this section.
(A) ``Code for Operation and Maintenance of Nuclear Power Plants:''
(1) 1995 Edition,
(2) 1996 Addenda,
(3) 1997 Addenda,
(4) 1998 Edition,
(5) 1999 Addenda,
(6) 2000 Addenda,
(7) 2001 Edition,
(8) 2002 Addenda,
(9) 2003 Addenda,
(10) 2004 Edition,
(11) 2005 Addenda, and
(12) 2006 Addenda.
(B) [Reserved]
(2) Institute of Electrical and Electronics Engineers (IEEE)
Service Center, 445 Hoes Lane, Piscataway, NJ 08855; telephone: 1-800-
678-4333; http://ieeexplore.ieee.org.
(i) IEEE standard 279-1971. (IEEE Std 279-1971), ``Criteria for
Protection Systems for Nuclear Power Generating Stations'' (Approval
Date: June 3, 1971), referenced in paragraph (h)(2) of this section.
(ii) IEEE Standard 603-1991. (IEEE Std 603-1991), ``Standard
Criteria for Safety Systems for Nuclear Power Generating Stations''
(Approval Date: June 27, 1991), referenced in paragraphs (h)(2) and (3)
of this section. All other standards that are referenced in IEEE Std
603-1991 are not approved for incorporation by reference.
(iii) IEEE standard 603-1991, correction sheet. (IEEE Std 603-1991
correction sheet), ``Standard Criteria for Safety Systems for Nuclear
Power Generating Stations, Correction Sheet, Issued January 30, 1995,
'' referenced in paragraphs (h)(2) and (3) of this section. (Copies of
this correction sheet may be purchased from Thomson Reuters, 3916
Ranchero Dr., Ann Arbor, MI 48108; http://www.techstreet.com.)
(3) U.S. Nuclear Regulatory Commission (NRC) Public Document Room,
11555 Rockville Pike, Rockville, Maryland 20852; telephone: 1-800-397-
4209; email: [email protected]; http://www.nrc.gov/reading-rm/doc-collections/reg-guides/.
(i) NRC Regulatory Guide 1.84, Revision 36. NRC Regulatory Guide
1.84, Revision 36, ``Design, Fabrication, and Materials Code Case
Acceptability, ASME Section III,'' dated August 2014, with the
requirements in paragraph (b)(4) of this section.
(ii) NRC Regulatory Guide 1.147, Revision 17. NRC Regulatory Guide
1.147, Revision 17, ``Inservice Inspection Code Case Acceptability,
ASME Section XI, Division 1,'' dated August 2014, which lists ASME Code
Cases that the NRC has approved in accordance with the requirements in
paragraph (b)(5) of this section.
(iii) NRC Regulatory Guide 1.192, Revision 1. NRC Regulatory Guide
1.192, Revision 1, ``Operation and Maintenance Code Case Acceptability,
ASME OM Code,'' dated August 2014, which lists ASME Code Cases that the
NRC has approved in accordance with the requirements in paragraph
(b)(6) of this section.
(b) Use and conditions on the use of standards. Systems and
components of boiling and pressurized water-cooled nuclear power
reactors must meet the requirements of the ASME Boiler and Pressure
Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance
of Nuclear Power Plants (OM Code) as specified in this paragraph. Each
combined license for a utilization facility is subject to the following
conditions.
(1) Conditions on ASME BPV Code Section III. Each manufacturing
license, standard design approval, and design certification under part
52 of this chapter is subject to the following conditions. As used in
this section, references to Section III refer to Section III of the
ASME Boiler and Pressure Vessel Code and include the 1963 Edition
through 1973 Winter Addenda and the 1974 Edition (Division 1) through
the 2008 Addenda (Division 1), subject to the following conditions:
(i) Section III condition: Section III materials. When applying the
1992 Edition of Section III, applicants or licensees must apply the
1992 Edition with the 1992 Addenda of Section II of the ASME Boiler and
Pressure Vessel Code.
(ii) Section III condition: Weld leg dimensions. When applying the
1989 Addenda through the latest edition and addenda, applicants or
licensees may not apply subparagraphs NB-3683.4(c)(1) and NB-
3683.4(c)(2) or Footnote 11 from the 1989 Addenda through the 2003
Addenda, or Footnote 13 from the 2004 Edition through the 2008 Addenda
to Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg size
less than 1.09 tn.
(iii) Section III condition: Seismic design of piping. Applicants
or licensees may use Subarticles NB-3200, NB-3600, NC-3600, and ND-3600
for seismic design of piping, up to and including the 1993 Addenda,
subject to the condition specified in paragraph (b)(1)(ii) of this
section. Applicants or licensees may not use these subarticles for
seismic design of piping in the 1994 Addenda through the 2005 Addenda
incorporated by reference in paragraph (a)(1) of this section, except
that Subarticle NB-3200 in the 2004 Edition through the 2008 Addenda
may be used by applicants and licensees, subject to the condition in
paragraph (b)(1)(iii)(A) of this section. Applicants or licensees may
use Subarticles NB-3600, NC-3600, and ND-3600 for the seismic design of
piping in the 2006 Addenda through the 2008 Addenda, subject to the
conditions of this paragraph corresponding to those subarticles.
(A) Seismic design of piping: First provision. When applying Note
(1) of Figure NB-3222-1 for Level B service limits, the calculation of
Pb stresses must include reversing dynamic loads (including
inertia earthquake effects) if evaluation of these loads is required by
NB-3223(b).
(B) Seismic design of piping: Second provision. For Class 1 piping,
the material and Do/t requirements of NB-3656(b) must be met
for all Service Limits when the Service Limits include reversing
dynamic loads, and the alternative rules for reversing dynamic loads
are used.
(iv) Section III condition: Quality assurance. When applying
editions and addenda later than the 1989 Edition of Section III, the
requirements of NQA-1, ``Quality Assurance Requirements for Nuclear
Facilities,'' 1986 Edition through the 1994 Edition, are acceptable for
use, provided that the edition and addenda of NQA-1 specified in NCA-
4000 is used in conjunction with the administrative, quality, and
technical provisions contained in the edition and addenda of Section
III being used.
(v) Section III condition: Independence of inspection. Applicants
or licensees may not apply NCA-4134.10(a) of Section III, 1995 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1) of this section.
(vi) Section III condition: Subsection NH. The provisions in
Subsection NH, ``Class 1 Components in Elevated Temperature Service,''
1995 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1) of this section, may only be used for the
design and construction of Type 316 stainless steel pressurizer heater
sleeves where service conditions do not cause the components to reach
temperatures exceeding 900[emsp14][deg]F.
(vii) Section III condition: Capacity certification and
demonstration of function of incompressible-fluid pressure-relief
valves. When applying the 2006 Addenda through the 2007 Edition up to
and including the 2008 Addenda, applicants and licensees may use
paragraph NB-7742, except that paragraph NB-7742(a)(2) may not be used.
For a valve design of a single size
[[Page 65801]]
to be certified over a range of set pressures, the demonstration of
function tests under paragraph NB-7742 must be conducted as prescribed
in NB-7732.2 on two valves covering the minimum set pressure for the
design and the maximum set pressure that can be accommodated at the
demonstration facility selected for the test.
(2) Conditions on ASME BPV Code Section XI. As used in this
section, references to Section XI refer to Section XI, Division 1, of
the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition
through the 1976 Winter Addenda and the 1977 Edition through the 2007
Edition with the 2008 Addenda, subject to the following conditions:
(i) [Reserved]
(ii) Section XI condition: Pressure-retaining welds in ASME Code
Class 1 piping (applies to Table IWB-2500 and IWB-2500-1 and Category
B-J). If the facility's application for a construction permit was
docketed prior to July 1, 1978, the extent of examination for Code
Class 1 pipe welds may be determined by the requirements of Table IWB-
2500 and Table IWB-2600 Category B-J of Section XI of the ASME BPV Code
in the 1974 Edition and Addenda through the Summer 1975 Addenda or
other requirements the NRC may adopt.
(iii) [Reserved]
(iv) [Reserved]
(v) [Reserved]
(vi) Section XI condition: Effective edition and addenda of
Subsection IWE and Subsection IWL. Applicants or licensees may use
either the 1992 Edition with the 1992 Addenda or the 1995 Edition with
the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned
by the requirements in paragraphs (b)(2)(viii) and (ix) of this
section, when implementing the initial 120-month inspection interval
for the containment inservice inspection requirements of this section.
Successive 120-month interval updates must be implemented in accordance
with paragraph (g)(4)(ii) of this section.
(vii) Section XI condition: Section XI references to OM Part 4, OM
Part 6, and OM Part 10 (Table IWA-1600-1). When using Table IWA-1600-1,
``Referenced Standards and Specifications,'' in the Section XI,
Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified
``Revision Date or Indicator'' for ASME/ANSI OM part 4, ASME/ANSI part
6, and ASME/ANSI part 10 must be the OMa-1988 Addenda to the OM-1987
Edition. These requirements have been incorporated into the OM Code,
which is incorporated by reference in paragraph (a)(1)(iv) of this
section.
(viii) Section XI condition: Concrete containment examinations.
Applicants or licensees applying Subsection IWL, 1992 Edition with the
1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through (E) of this
section. Applicants or licensees applying Subsection IWL, 1995 Edition
with the 1996 Addenda, must apply paragraphs (b)(2)(viii)(A),
(b)(2)(viii)(D)(3), and (b)(2)(viii)(E) of this section. Applicants or
licensees applying Subsection IWL, 1998 Edition through the 2000
Addenda, must apply paragraphs (b)(2)(viii)(E) and (F) of this section.
Applicants or licensees applying Subsection IWL, 2001 Edition through
the 2004 Edition, up to and including the 2006 Addenda, must apply
paragraphs (b)(2)(viii)(E) through (G) of this section. Applicants or
licensees applying Subsection IWL, 2007 Edition through the latest
edition and addenda incorporated by reference in paragraph (a)(1)(ii)
of this section, must apply paragraph (b)(2)(viii)(E) of this section.
(A) Concrete containment examinations: First provision. Grease caps
that are accessible must be visually examined to detect grease leakage
or grease cap deformations. Grease caps must be removed for this
examination when there is evidence of grease cap deformation that
indicates deterioration of anchorage hardware.
(B) Concrete containment examinations: Second provision. When
evaluation of consecutive surveillances of prestressing forces for the
same tendon or tendons in a group indicates a trend of prestress loss
such that the tendon force(s) would be less than the minimum design
prestress requirements before the next inspection interval, an
evaluation must be performed and reported in the Engineering Evaluation
Report as prescribed in IWL-3300.
(C) Concrete containment examinations: Third provision. When the
elongation corresponding to a specific load (adjusted for effective
wires or strands) during retensioning of tendons differs by more than
10 percent from that recorded during the last measurement, an
evaluation must be performed to determine whether the difference is
related to wire failures or slip of wires in anchorage. A difference of
more than 10 percent must be identified in the ISI Summary Report
required by IWA-6000.
(D) Concrete containment examinations: Fourth provision. The
applicant or licensee must report the following conditions, if they
occur, in the ISI Summary Report required by IWA-6000:
(1) The sampled sheathing filler grease contains chemically
combined water exceeding 10 percent by weight or the presence of free
water;
(2) The absolute difference between the amount removed and the
amount replaced exceeds 10 percent of the tendon net duct volume; and
(3) Grease leakage is detected during general visual examination of
the containment surface.
(E) Concrete containment examinations: Fifth provision. For Class
CC applications, the applicant or licensee must evaluate the
acceptability of inaccessible areas when conditions exist in accessible
areas that could indicate the presence of or the result in degradation
to such inaccessible areas. For each inaccessible area identified, the
applicant or licensee must provide the following in the ISI Summary
Report required by IWA-6000:
(1) A description of the type and estimated extent of degradation,
and the conditions that led to the degradation;
(2) An evaluation of each area, and the result of the evaluation;
and
(3) A description of necessary corrective actions.
(F) Concrete containment examinations: Sixth provision. Personnel
that examine containment concrete surfaces and tendon hardware, wires,
or strands must meet the qualification provisions in IWA-2300. The
``owner-defined'' personnel qualification provisions in IWL-2310(d) are
not approved for use.
(G) Concrete containment examinations: Seventh provision. Corrosion
protection material must be restored following concrete containment
post-tensioning system repair and replacement activities in accordance
with the quality assurance program requirements specified in IWA-1400.
(ix) Section XI condition: Metal containment examinations.
Applicants or licensees applying Subsection IWE, 1992 Edition with the
1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy
the requirements of paragraphs (b)(2)(ix)(A) through (E) of this
section. Applicants or licensees applying Subsection IWE, 1998 Edition
through the 2001 Edition with the 2003 Addenda, must satisfy the
requirements of paragraphs (b)(2)(ix)(A) and (B) and (b)(2)(ix)(F)
through (I) of this section. Applicants or licensees applying
Subsection IWE, 2004 Edition, up to and including the 2005 Addenda,
must satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and
(b)(2)(ix)(F) through (H) of this section. Applicants or licensees
applying Subsection IWE, 2004 Edition with the 2006 Addenda, must
satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and
(b)(2)(ix)(B) of this section. Applicants
[[Page 65802]]
or licensees applying Subsection IWE, 2007 Edition through the latest
addenda incorporated by reference in paragraph (a)(1)(ii) of this
section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2)
and (b)(2)(ix)(B) and (J) of this section.
(A) Metal containment examinations: First provision. For Class MC
applications, the following apply to inaccessible areas.
(1) The applicant or licensee must evaluate the acceptability of
inaccessible areas when conditions exist in accessible areas that could
indicate the presence of or could result in degradation to such
inaccessible areas.
(2) For each inaccessible area identified for evaluation, the
applicant or licensee must provide the following in the ISI Summary
Report as required by IWA-6000:
(i) A description of the type and estimated extent of degradation,
and the conditions that led to the degradation;
(ii) An evaluation of each area, and the result of the evaluation;
and
(iii) A description of necessary corrective actions.
(B) Metal containment examinations: Second provision. When
performing remotely the visual examinations required by Subsection IWE,
the maximum direct examination distance specified in Table IWA-2210-1
may be extended and the minimum illumination requirements specified in
Table IWA-2210-1 may be decreased provided that the conditions or
indications for which the visual examination is performed can be
detected at the chosen distance and illumination.
(C) Metal containment examinations: Third provision. The
examinations specified in Examination Category E-B, Pressure Retaining
Welds, and Examination Category E-F, Pressure Retaining Dissimilar
Metal Welds, are optional.
(D) Metal containment examinations: Fourth provision. This
paragraph (b)(2)(ix)(D) may be used as an alternative to the
requirements of IWE-2430.
(1) If the examinations reveal flaws or areas of degradation
exceeding the acceptance standards of Table IWE-3410-1, an evaluation
must be performed to determine whether additional component
examinations are required. For each flaw or area of degradation
identified that exceeds acceptance standards, the applicant or licensee
must provide the following in the ISI Summary Report required by IWA-
6000:
(i) A description of each flaw or area, including the extent of
degradation, and the conditions that led to the degradation;
(ii) The acceptability of each flaw or area and the need for
additional examinations to verify that similar degradation does not
exist in similar components; and
(iii) A description of necessary corrective actions.
(2) The number and type of additional examinations to ensure
detection of similar degradation in similar components.
(E) Metal containment examinations: Fifth provision. A general
visual examination as required by Subsection IWE must be performed once
each period.
(F) Metal containment examinations: Sixth provision. VT-1 and VT-3
examinations must be conducted in accordance with IWA-2200. Personnel
conducting examinations in accordance with the VT-1 or VT-3 examination
method must be qualified in accordance with IWA-2300. The ``owner-
defined'' personnel qualification provisions in IWE-2330(a) for
personnel that conduct VT-1 and VT-3 examinations are not approved for
use.
(G) Metal containment examinations: Seventh provision. The VT-3
examination method must be used to conduct the examinations in Items
E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method
must be used to conduct the examination in Item E4.11 of Table IWE-
2500-1. An examination of the pressure-retaining bolted connections in
Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must
be conducted once each interval. The ``owner-defined'' visual
examination provisions in IWE-2310(a) are not approved for use for VT-1
and VT-3 examinations.
(H) Metal containment examinations: Eighth provision. Containment
bolted connections that are disassembled during the scheduled
performance of the examinations in Item E1.11 of Table IWE-2500-1 must
be examined using the VT-3 examination method. Flaws or degradation
identified during the performance of a VT-3 examination must be
examined in accordance with the VT-1 examination method. The criteria
in the material specification or IWB-3517.1 must be used to evaluate
containment bolting flaws or degradation. As an alternative to
performing VT-3 examinations of containment bolted connections that are
disassembled during the scheduled performance of Item E1.11, VT-3
examinations of containment bolted connections may be conducted
whenever containment bolted connections are disassembled for any
reason.
(I) Metal containment examinations: Ninth provision. The ultrasonic
examination acceptance standard specified in IWE-3511.3 for Class MC
pressure-retaining components must also be applied to metallic liners
of Class CC pressure-retaining components.
(J) Metal containment examinations: Tenth provision. In general, a
repair/replacement activity such as replacing a large containment
penetration, cutting a large construction opening in the containment
pressure boundary to replace steam generators, reactor vessel heads,
pressurizers, or other major equipment; or other similar modification
is considered a major containment modification. When applying IWE-5000
to Class MC pressure-retaining components, any major containment
modification or repair/replacement must be followed by a Type A test to
provide assurance of both containment structural integrity and
leaktight integrity prior to returning to service, in accordance with
10 CFR part 50, Appendix J, Option A or Option B on which the
applicant's or licensee's Containment Leak-Rate Testing Program is
based. When applying IWE-5000, if a Type A, B, or C Test is performed,
the test pressure and acceptance standard for the test must be in
accordance with 10 CFR part 50, Appendix J.
(x) Section XI condition: Quality assurance. When applying Section
XI editions and addenda later than the 1989 Edition, the requirements
of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,''
1979 Addenda through the 1989 Edition, are acceptable as permitted by
IWA-1400 of Section XI, if the licensee uses its 10 CFR part 50,
Appendix B, quality assurance program, in conjunction with Section XI
requirements. Commitments contained in the licensee's quality assurance
program description that are more stringent than those contained in
NQA-1 must govern Section XI activities. Further, where NQA-1 and
Section XI do not address the commitments contained in the licensee's
Appendix B quality assurance program description, the commitments must
be applied to Section XI activities.
(xi) [Reserved]
(xii) Section XI condition: Underwater welding. The provisions in
IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through
the latest edition and addenda incorporated by reference in paragraph
(a)(1)(ii) of this section, are not approved for use on irradiated
material.
(xiii) [Reserved]
[[Page 65803]]
(xiv) Section XI condition: Appendix VIII personnel qualification.
All personnel qualified for performing ultrasonic examinations in
accordance with Appendix VIII must receive 8 hours of annual hands-on
training on specimens that contain cracks. Licensees applying the 1999
Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section may use the annual
practice requirements in VII-4240 of Appendix VII of Section XI in
place of the 8 hours of annual hands-on training provided that the
supplemental practice is performed on material or welds that contain
cracks, or by analyzing prerecorded data from material or welds that
contain cracks. In either case, training must be completed no earlier
than 6 months prior to performing ultrasonic examinations at a
licensee's facility.
(xv) Section XI condition: Appendix VIII specimen set and
qualification requirements. Licensees using Appendix VIII in the 1995
Edition through the 2001 Edition of the ASME Boiler and Pressure Vessel
Code may elect to comply with all of the provisions in paragraphs
(b)(2)(xv)(A) through (M) of this section, except for paragraph
(b)(2)(xv)(F) of this section, which may be used at the licensee's
option. Licensees using editions and addenda after 2001 Edition through
the 2006 Addenda must use the 2001 Edition of Appendix VIII and may
elect to comply with all of the provisions in paragraphs (b)(2)(xv)(A)
through (M) of this section, except for paragraph (b)(2)(xv)(F) of this
section, which may be used at the licensee's option.
(A) Specimen set and qualification: First provision. When applying
Supplements 2, 3, and 10 to Appendix VIII, the following examination
coverage criteria requirements must be used:
(1) Piping must be examined in two axial directions, and when
examination in the circumferential direction is required, the
circumferential examination must be performed in two directions,
provided access is available. Dissimilar metal welds must be examined
axially and circumferentially.
(2) Where examination from both sides is not possible, full
coverage credit may be claimed from a single side for ferritic welds.
Where examination from both sides is not possible on austenitic welds
or dissimilar metal welds, full coverage credit from a single side may
be claimed only after completing a successful single-sided Appendix
VIII demonstration using flaws on the opposite side of the weld.
Dissimilar metal weld qualifications must be demonstrated from the
austenitic side of the weld, and the qualification may be expanded for
austenitic welds with no austenitic sides using a separate add-on
performance demonstration. Dissimilar metal welds may be examined from
either side of the weld.
(B) Specimen set and qualification: Second provision. The following
conditions must be used in addition to the requirements of Supplement 4
to Appendix VIII:
(1) Paragraph 3.1, Detection acceptance criteria--Personnel are
qualified for detection if the results of the performance demonstration
satisfy the detection requirements of ASME Section XI, Appendix VIII,
Table VIII-S4-1, and no flaw greater than 0.25 inch through-wall
dimension is missed.
(2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the
50 percent of allowable flaw size, as defined in IWB-3500, need not be
included as detection flaws. For procedures applied from the inside
surface, use the minimum thickness specified in the scope of the
procedure to calculate a/t. For procedures applied from the outside
surface, the actual thickness of the test specimen is to be used to
calculate a/t.
(C) Specimen set and qualification: Third provision. When applying
Supplement 4 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.15 inch RMS must be used in
lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a
length sizing requirement of 0.75 inch RMS must be used in lieu of the
requirement in Subparagraph 3.2(b).
(2) In lieu of the location acceptance criteria requirements of
Subparagraph 2.1(b), a flaw will be considered detected when reported
within 1.0 inch or 10 percent of the metal path to the flaw, whichever
is greater, of its true location in the X and Y directions.
(3) In lieu of the flaw type requirements of Subparagraph
1.1(e)(1), a minimum of 70 percent of the flaws in the detection and
sizing tests must be cracks. Notches, if used, must be limited by the
following:
(i) Notches must be limited to the case where examinations are
performed from the clad surface.
(ii) Notches must be semielliptical with a tip width of less than
or equal to 0.010 inches.
(iii) Notches must be perpendicular to the surface within 2 degrees.
(4) In lieu of the detection test matrix requirements in paragraphs
1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain
a representative distribution of flaw orientations, sizes, and
locations.
(D) Specimen set and qualification: Fourth provision. The following
conditions must be used in addition to the requirements of Supplement 6
to Appendix VIII:
(1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are
qualified for detection if:
(i) No surface connected flaw greater than 0.25 inch through-wall
has been missed.
(ii) No embedded flaw greater than 0.50 inch through-wall has been
missed.
(2) Paragraph 3.1, Detection Acceptance Criteria--For procedure
qualification, all flaws within the scope of the procedure are
detected.
(3) Paragraph 1.1(b) for detection and sizing test flaws and
locations--Flaws smaller than the 50 percent of allowable flaw size, as
defined in IWB-3500, need not be included as detection flaws. Flaws
that are less than the allowable flaw size, as defined in IWB-3500, may
be used as detection and sizing flaws.
(4) Notches are not permitted.
(E) Specimen set and qualification: Fifth provision. When applying
Supplement 6 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.25 inch RMS must be used in
lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and
3.2(c)(3).
(2) In lieu of the location acceptance criteria requirements in
Subparagraph 2.1(b), a flaw will be considered detected when reported
within 1.0 inch or 10 percent of the metal path to the flaw, whichever
is greater, of its true location in the X and Y directions.
(3) In lieu of the length sizing criteria requirements of
Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch
RMS must be used.
(4) In lieu of the detection specimen requirements in Subparagraph
1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The
remaining flaws may be cracks or fabrication type flaws, such as slag
and lack of fusion. The use of notches is not allowed.
(5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test
matrix, personnel demonstration test sets must contain a representative
distribution of flaw orientations, sizes, and locations.
(F) Specimen set and qualification: Sixth provision. The following
conditions may be used for personnel qualification for combined
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII
qualification. Licensees choosing to apply this combined qualification
must apply all of the provisions of Supplements 4 and 6 including the
following conditions:
(1) For detection and sizing, the total number of flaws must be at
least 10. A
[[Page 65804]]
minimum of 5 flaws must be from Supplement 4, and a minimum of 50
percent of the flaws must be from Supplement 6. At least 50 percent of
the flaws in any sizing must be cracks. Notches are not acceptable for
Supplement 6.
(2) Examination personnel are qualified for detection and length
sizing when the results of any combined performance demonstration
satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
(3) Examination personnel are qualified for depth sizing when
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws
are sized within the respective acceptance criteria of those
supplements.
(G) Specimen set and qualification: Seventh provision. When
applying Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII,
or combined Supplement 4 and Supplement 6 qualification, the following
additional conditions must be used, and examination coverage must
include:
(1) The clad-to-base-metal-interface, including a minimum of 15
percent T (measured from the clad-to-base-metal-interface), must be
examined from four orthogonal directions using procedures and personnel
qualified in accordance with Supplement 4 to Appendix VIII.
(2) If the clad-to-base-metal-interface procedure demonstrates
detectability of flaws with a tilt angle relative to the weld
centerline of at least 45 degrees, the remainder of the examination
volume is considered fully examined if coverage is obtained in one
parallel and one perpendicular direction. This must be accomplished
using a procedure and personnel qualified for single-side examination
in accordance with Supplement 6. Subsequent examinations of this volume
may be performed using examination techniques qualified for a tilt
angle of at least 10 degrees.
(3) The examination volume not addressed by paragraph
(b)(2)(xv)(G)(1) of this section is considered fully examined if
coverage is obtained in one parallel and one perpendicular direction,
using a procedure and personnel qualified for single sided examination
when the conditions in paragraph (b)(2)(xv)(G)(2) are met.
(H) Specimen set and qualification: Eighth provision. When applying
Supplement 5 to Appendix VIII, at least 50 percent of the flaws in the
demonstration test set must be cracks and the maximum misorientation
must be demonstrated with cracks. Flaws in nozzles with bore diameters
equal to or less than 4 inches may be notches.
(I) Specimen set and qualification: Ninth provision. When applying
Supplement 5, Paragraph (a), to Appendix VIII, the number of false
calls allowed must be D/10, with a maximum of 3, where D is the
diameter of the nozzle.
(J) [Reserved]
(K) Specimen set and qualification: Eleventh provision. When
performing nozzle-to-vessel weld examinations, the following conditions
must be used when the requirements contained in Supplement 7 to
Appendix VIII are applied for nozzle-to-vessel welds in conjunction
with Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or
combined Supplement 4 and Supplement 6 qualification.
(1) For examination of nozzle-to-vessel welds conducted from the
bore, the following conditions are required to qualify the procedures,
equipment, and personnel:
(i) For detection, a minimum of four flaws in one or more full-
scale nozzle mock-ups must be added to the test set. The specimens must
comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for
flaw locations specified in Table VIII S6-1. Flaws may be notches,
fabrication flaws, or cracks. Seventy-five (75) percent of the flaws
must be cracks or fabrication flaws. Flaw locations and orientations
must be selected from the choices shown in paragraph (b)(2)(xv)(K)(4)
of this section, Table VIII-S7-1--Modified, with the exception that
flaws in the outer eighty-five (85) percent of the weld need not be
perpendicular to the weld. There may be no more than two flaws from
each category, and at least one subsurface flaw must be included.
(ii) For length sizing, a minimum of four flaws as in paragraph
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set.
The length sizing results must be added to the results of combined
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The
combined results must meet the acceptance standards contained in
paragraph (b)(2)(xv)(E)(3) of this section.
(iii) For depth sizing, a minimum of four flaws as in paragraph
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set.
Their depths must be distributed over the ranges of Supplement 4,
Paragraph 1.1, to Appendix VIII, for the inner 15 percent of the wall
thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the
remainder of the wall thickness. The depth sizing results must be
combined with the sizing results from Supplement 4 to Appendix VIII for
the inner 15 percent and to Supplement 6 to Appendix VIII for the
remainder of the wall thickness. The combined results must meet the
depth sizing acceptance criteria contained in paragraphs
(b)(2)(xv)(C)(1), (b)(2)(xv)(E)(1), and (b)(2)(xv)(F)(3) of this
section.
(2) For examination of reactor pressure vessel nozzle-to-vessel
welds conducted from the inside of the vessel, the following conditions
are required:
(i) The clad-to-base-metal-interface and the adjacent examination
volume to a minimum depth of 15 percent T (measured from the clad-to-
base-metal-interface) must be examined from four orthogonal directions
using a procedure and personnel qualified in accordance with Supplement
4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and (C)
of this section.
(ii) When the examination volume defined in paragraph
(b)(2)(xv)(K)(2)(i) of this section cannot be effectively examined in
all four directions, the examination must be augmented by examination
from the nozzle bore using a procedure and personnel qualified in
accordance with paragraph (b)(2)(xv)(K)(1) of this section.
(iii) The remainder of the examination volume not covered by
paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of
paragraphs (b)(2)(xv)(K)(2)(i) and (ii) of this section, must be
examined from the nozzle bore using a procedure and personnel qualified
in accordance with paragraph (b)(2)(xv)(K)(1) of this section, or from
the vessel shell using a procedure and personnel qualified for single
sided examination in accordance with Supplement 6 to Appendix VIII, as
conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.
(3) For examination of reactor pressure vessel nozzle-to-shell
welds conducted from the outside of the vessel, the following
conditions are required:
(i) The clad-to-base-metal-interface and the adjacent metal to a
depth of 15 percent T (measured from the clad-to-base-metal-interface)
must be examined from one radial and two opposing circumferential
directions using a procedure and personnel qualified in accordance with
Supplement 4 to Appendix VIII, as conditioned by paragraphs
(b)(2)(xv)(B) and (C) of this section, for examinations performed in
the radial direction, and Supplement 5 to Appendix VIII, as conditioned
by paragraph (b)(2)(xv)(J) of this section, for examinations performed
in the circumferential direction.
[[Page 65805]]
(ii) The examination volume not addressed by paragraph
(b)(2)(xv)(K)(3)(i) of this section must be examined in a minimum of
one radial direction using a procedure and personnel qualified for
single sided examination in accordance with Supplement 6 to Appendix
VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this
section.
(4) Table VIII-S7-1, ``Flaw Locations and Orientations,''
Supplement 7 to Appendix VIII, is conditioned as follows:
Table VIII--S7-1--Modified
[Flaw locations and orientations]
----------------------------------------------------------------------------------------------------------------
Parallel to weld Perpendicular to weld
----------------------------------------------------------------------------------------------------------------
Inner 15 percent............................................ X X
Outside Diameter Surface.................................... X ........................
Subsurface.................................................. X ........................
----------------------------------------------------------------------------------------------------------------
(L) Specimen set and qualification: Twelfth provision. As a
condition to the requirements of Supplement 8, Subparagraph 1.1(c), to
Appendix VIII, notches may be located within one diameter of each end
of the bolt or stud.
(M) Specimen set and qualification: Thirteenth provision. When
implementing Supplement 12 to Appendix VIII, only the provisions
related to the coordinated implementation of Supplement 3 to Supplement
2 performance demonstrations are to be applied.
(xvi) Section XI condition: Appendix VIII single side ferritic
vessel and piping and stainless steel piping examinations. When
applying editions and addenda prior to the 2007 Edition of Section XI,
the following conditions apply.
(A) Ferritic and stainless steel piping examinations: First
provision. Examinations performed from one side of a ferritic vessel
weld must be conducted with equipment, procedures, and personnel that
have demonstrated proficiency with single side examinations. To
demonstrate equivalency to two sided examinations, the demonstration
must be performed to the requirements of Appendix VIII, as conditioned
by this paragraph and paragraphs (b)(2)(xv)(B) through (G) of this
section, on specimens containing flaws with non-optimum sound energy
reflecting characteristics or flaws similar to those in the vessel
being examined.
(B) Ferritic and stainless steel piping examinations: Second
provision. Examinations performed from one side of a ferritic or
stainless steel pipe weld must be conducted with equipment, procedures,
and personnel that have demonstrated proficiency with single side
examinations. To demonstrate equivalency to two sided examinations, the
demonstration must be performed to the requirements of Appendix VIII,
as conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this
section.
(xvii) Section XI condition: Reconciliation of quality
requirements. When purchasing replacement items, in addition to the
reconciliation provisions of IWA-4200, 1995 Addenda through 1998
Edition, the replacement items must be purchased, to the extent
necessary, in accordance with the licensee's quality assurance program
description required by 10 CFR 50.34(b)(6)(ii).
(xviii) Section XI condition: NDE personnel certification. (A) NDE
personnel certification: First provision. Level I and II nondestructive
examination personnel must be recertified on a 3-year interval in lieu
of the 5-year interval specified in the 1997 Addenda and 1998 Edition
of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 1999 Addenda
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(ii) of this section.
(B) NDE personnel certification: Second provision. When applying
editions and addenda prior to the 2007 Edition of Section XI, paragraph
IWA-2316 may only be used to qualify personnel that observe leakage
during system leakage and hydrostatic tests conducted in accordance
with IWA 5211(a) and (b).
(C) NDE personnel certification: Third provision. When applying
editions and addenda prior to the 2005 Addenda of Section XI,
licensee's qualifying visual examination personnel for VT-3 visual
examination under paragraph IWA-2317 of Section XI must demonstrate the
proficiency of the training by administering an initial qualification
examination and administering subsequent examinations on a 3-year
interval.
(xix) Section XI condition: Substitution of alternative methods.
The provisions for substituting alternative examination methods, a
combination of methods, or newly developed techniques in the 1997
Addenda of IWA-2240 must be applied when using the 1998 Edition through
the 2004 Edition of Section XI of the ASME BPV Code. The provisions in
IWA-4520(c), 1997 Addenda through the 2004 Edition, allowing the
substitution of alternative methods, a combination of methods, or newly
developed techniques for the methods specified in the Construction
Code, are not approved for use. The provisions in IWA-4520(b)(2) and
IWA-4521 of the 2008 Addenda through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section,
allowing the substitution of ultrasonic examination for radiographic
examination specified in the Construction Code, are not approved for
use.
(xx) Section XI condition: System leakage tests--(A) System leakage
tests: First provision. When performing system leakage tests in
accordance with IWA-5213(a), 1997 through 2002 Addenda, the licensee
must maintain a 10-minute hold time after test pressure has been
reached for Class 2 and Class 3 components that are not in use during
normal operating conditions. No hold time is required for the remaining
Class 2 and Class 3 components provided that the system has been in
operation for at least 4 hours for insulated components or 10 minutes
for uninsulated components.
(B) System leakage tests: Second provision. The NDE provision in
IWA-4540(a)(2) of the 2002 Addenda of Section XI must be applied when
performing system leakage tests after repair and replacement activities
performed by welding or brazing on a pressure retaining boundary using
the 2003 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section.
(xxi) Section XI condition: Table IWB-2500-1 examination
requirements. (A) Table IWB-2500-1 examination requirements: First
provision. The provisions of Table IWB 2500-1, Examination Category B-
D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60
(Inspection
[[Page 65806]]
Program A) and Items B3.120 and B3.140 (Inspection Program B) of the
1998 Edition must be applied when using the 1999 Addenda through the
latest edition and addenda incorporated by reference in paragraph
(a)(1)(ii) of this section. A visual examination with magnification
that has a resolution sensitivity to detect a 1-mil width wire or
crack, utilizing the allowable flaw length criteria in Table IWB-3512-
1, 1997 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section, with a limiting
assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed
instead of an ultrasonic examination.
(B) [Reserved]
(xxii) Section XI condition: Surface examination. The use of the
provision in IWA-2220, ``Surface Examination,'' of Section XI, 2001
Edition through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section, that allows use of
an ultrasonic examination method is prohibited.
(xxiii) Section XI condition: Evaluation of thermally cut surfaces.
The use of the provisions for eliminating mechanical processing of
thermally cut surfaces in IWA-4461.4.2 of Section XI, 2001 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(ii) of this section, is prohibited.
(xxiv) Section XI condition: Incorporation of the performance
demonstration initiative and addition of ultrasonic examination
criteria. The use of Appendix VIII and the supplements to Appendix VIII
and Article I-3000 of Section XI of the ASME BPV Code, 2002 Addenda
through the 2006 Addenda, is prohibited.
(xxv) Section XI condition: Mitigation of defects by modification.
The use of the provisions in IWA-4340, ``Mitigation of Defects by
Modification,'' Section XI, 2001 Edition through the latest edition and
addenda incorporated by reference in paragraph (a)(1)(ii) of this
section are prohibited.
(xxvi) Section XI condition: Pressure testing Class 1, 2 and 3
mechanical joints. The repair and replacement activity provisions in
IWA-4540(c) of the 1998 Edition of Section XI for pressure testing
Class 1, 2, and 3 mechanical joints must be applied when using the 2001
Edition through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section.
(xxvii) Section XI condition: Removal of insulation. When
performing visual examination in accordance with IWA-5242 of Section XI
of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-
5241 of the 2007 Edition through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section,
insulation must be removed from 17-4 PH or 410 stainless steel studs or
bolts aged at a temperature below 1100[emsp14][deg]F or having a
Rockwell Method C hardness value above 30, and from A-286 stainless
steel studs or bolts preloaded to 100,000 pounds per square inch or
higher.
(xxviii) Section XI condition: Analysis of flaws. Licensees using
ASME BPV Code, Section XI, Appendix A, must use the following
conditions when implementing Equation (2) in A-4300(b)(1):
For R < 0, [Delta]KI depends on the crack depth (a),
and the flow stress ([sigma]f). The flow stress is
defined by [sigma]f = 1/2([sigma]ys +
[sigma]ult), where [sigma]ys is the yield
strength and [sigma]ult is the ultimate tensile strength
in units ksi (MPa) and (a) is in units in. (mm). For -2 <= R <= 0
and Kmax- Kmin <= 0.8 x 1.12
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI =
Kmax. For R < -2 and Kmax- Kmin <=
0.8 x 1.12 [sigma]f[radic]([pi]a), S = 1 and
[Delta]KI = (1 - R) Kmax/3. For R < 0 and
Kmax - Kmin > 0.8 x 1.12
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI =
Kmax-Kmin.
(xxix) Section XI condition: Nonmandatory Appendix R. Nonmandatory
Appendix R, ``Risk-Informed Inspection Requirements for Piping,'' of
Section XI, 2005 Addenda through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section, may
not be implemented without prior NRC authorization of the proposed
alternative in accordance with paragraph (z) of this section.
(3) Conditions on ASME OM Code. As used in this section, references
to the OM Code refer to the ASME Code for Operation and Maintenance of
Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory
Appendices I and II, and Nonmandatory Appendices A through H and J,
including the 1995 Edition through the 2006 Addenda, subject to the
following conditions:
(i) OM condition: Quality assurance. When applying editions and
addenda of the OM Code, the requirements of NQA-1, ``Quality Assurance
Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as
permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-
1500 of the 1998 Edition through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(iv) of this section,
provided the licensee uses its 10 CFR part 50, Appendix B, quality
assurance program in conjunction with the OM Code requirements.
Commitments contained in the licensee's quality assurance program
description that are more stringent than those contained in NQA-1
govern OM Code activities. If NQA-1 and the OM Code do not address the
commitments contained in the licensee's Appendix B quality assurance
program description, the commitments must be applied to OM Code
activities.
(ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees
must comply with the provisions for MOV testing in OM Code ISTC 4.2,
1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(iv) of this section, and must establish a program to
ensure that motor-operated valves continue to be capable of performing
their design basis safety functions.
(iii) [Reserved]
(iv) OM condition: Check valves (Appendix II). Licensees applying
Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM
Code, 1995 Edition with the 1996 and 1997 Addenda, must satisfy the
requirements of (b)(3)(iv)(A) through (C) of this section. Licensees
applying Appendix II, 1998 Edition through the 2002 Addenda, must
satisfy the requirements of (b)(3)(iv)(A), (B), and (D) of this
section.
(A) Check valves: First provision. Valve opening and closing
functions must be demonstrated when flow testing or examination methods
(nonintrusive, or disassembly and inspection) are used;
(B) Check valves: Second provision. The initial interval for tests
and associated examinations may not exceed two fuel cycles or 3 years,
whichever is longer; any extension of this interval may not exceed one
fuel cycle per extension with the maximum interval not to exceed 10
years. Trending and evaluation of existing data must be used to reduce
or extend the time interval between tests.
(C) Check valves: Third provision. If the Appendix II condition
monitoring program is discontinued, then the requirements of ISTC 4.5.1
through 4.5.4 must be implemented.
(D) Check valves: Fourth provision. The applicable provisions of
subsection ISTC must be implemented if the Appendix II condition
monitoring program is discontinued.
(v) OM condition: Snubbers ISTD. Article IWF-5000, ``Inservice
Inspection Requirements for Snubbers,'' of the ASME BPV Code, Section
XI, must be used when performing inservice inspection examinations and
tests of snubbers at nuclear power plants, except as conditioned in
paragraphs (b)(3)(v)(A) and (B) of this section.
[[Page 65807]]
(A) Snubbers: First provision. Licensees may use Subsection ISTD,
``Preservice and Inservice Examination and Testing of Dynamic
Restraints (Snubbers) in Light-Water Reactor Power Plants,'' ASME OM
Code, 1995 Edition through the latest edition and addenda incorporated
by reference in paragraph (a)(1)(iv) of this section, in place of the
requirements for snubbers in the editions and addenda up to the 2005
Addenda of the ASME BPV Code, Section XI, IWF-5200(a) and (b) and IWF-
5300(a) and (b), by making appropriate changes to their technical
specifications or licensee-controlled documents. Preservice and
inservice examinations must be performed using the VT-3 visual
examination method described in IWA-2213.
(B) Snubbers: Second provision. Licensees must comply with the
provisions for examining and testing snubbers in Subsection ISTD of the
ASME OM Code and make appropriate changes to their technical
specifications or licensee-controlled documents when using the 2006
Addenda and later editions and addenda of Section XI of the ASME BPV
Code.
(vi) OM condition: Exercise interval for manual valves. Manual
valves must be exercised on a 2-year interval rather than the 5-year
interval specified in paragraph ISTC-3540 of the 1999 through the 2005
Addenda of the ASME OM Code, provided that adverse conditions do not
require more frequent testing.
(4) Conditions on Design, Fabrication, and Materials Code Cases.
Each manufacturing license, standard design approval, and design
certification application under part 52 of this chapter is subject to
the following conditions. Licensees may apply the ASME BPV Code Cases
listed in NRC Regulatory Guide 1.84, Revision 36, without prior NRC
approval, subject to the following conditions:
(i) Design, Fabrication, and Materials Code Case condition:
Applying Code Cases. When an applicant or licensee initially applies a
listed Code Case, the applicant or licensee must apply the most recent
version of that Code Case incorporated by reference in paragraph (a) of
this section.
(ii) Design, Fabrication, and Materials Code Case condition:
Applying different revisions of Code Cases. If an applicant or licensee
has previously applied a Code Case and a later version of the Code Case
is incorporated by reference in paragraph (a) of this section, the
applicant or licensee may continue to apply the previous version of the
Code Case as authorized or may apply the later version of the Code
Case, including any NRC-specified conditions placed on its use, until
it updates its Code of Record for the component being constructed.
(iii) Design, Fabrication, and Materials Code Case condition:
Applying annulled Code Cases. Application of an annulled Code Case is
prohibited unless an applicant or licensee applied the listed Code Case
prior to it being listed as annulled in Regulatory Guide 1.84. If an
applicant or licensee has applied a listed Code Case that is later
listed as annulled in Regulatory Guide 1.84, the applicant or licensee
may continue to apply the Code Case until it updates its Code of Record
for the component being constructed.
(5) Conditions on inservice inspection Code Cases. Licensees may
apply the ASME BPV Code Cases listed in Regulatory Guide 1.147,
Revision 17, without prior NRC approval, subject to the following:
(i) ISI Code Case condition: Applying Code Cases. When a licensee
initially applies a listed Code Case, the licensee must apply the most
recent version of that Code Case incorporated by reference in paragraph
(a) of this section.
(ii) ISI Code Case condition: Applying different revisions of Code
Cases. If a licensee has previously applied a Code Case and a later
version of the Code Case is incorporated by reference in paragraph (a)
of this section, the licensee may continue to apply, to the end of the
current 120-month interval, the previous version of the Code Case, as
authorized, or may apply the later version of the Code Case, including
any NRC-specified conditions placed on its use. Licensees who choose to
continue use of the Code Case during subsequent 120-month ISI program
intervals will be required to implement the latest version incorporated
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of
Regulatory Guide 1.147, Revision 17.
(iii) ISI Code Case condition: Applying annulled Code Cases.
Application of an annulled Code Case is prohibited unless a licensee
previously applied the listed Code Case prior to it being listed as
annulled in Regulatory Guide 1.147. If a licensee has applied a listed
Code Case that is later listed as annulled in Regulatory Guide 1.147,
the licensee may continue to apply the Code Case to the end of the
current 120-month interval.
(6) Conditions on Operation and Maintenance of Nuclear Power Plants
Code Cases. Licensees may apply the ASME Operation and Maintenance Code
Cases listed in Regulatory Guide 1.192, Revision 1, without prior NRC
approval, subject to the following:
(i) OM Code Case condition: Applying Code Cases. When a licensee
initially applies a listed Code Case, the licensee must apply the most
recent version of that Code Case incorporated by reference in paragraph
(a) of this section.
(ii) OM Code Case condition: Applying different revisions of Code
Cases. If a licensee has previously applied a Code Case and a later
version of the Code Case is incorporated by reference in paragraph (a)
of this section, the licensee may continue to apply, to the end of the
current 120-month interval, the previous version of the Code Case, as
authorized, or may apply the later version of the Code Case, including
any NRC-specified conditions placed on its use. Licensees who choose to
continue use of the Code Case during subsequent 120-month ISI program
intervals will be required to implement the latest version incorporated
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of
Regulatory Guide 1.192, Revision 1.
(iii) OM Code Case condition: Applying annulled Code Cases.
Application of an annulled Code Case is prohibited unless a licensee
previously applied the listed Code Case prior to it being listed as
annulled in Regulatory Guide 1.192. If a licensee has applied a listed
Code Case that is later listed as annulled in Regulatory Guide 1.192,
the licensee may continue to apply the Code Case to the end of the
current 120-month interval.
(c) Reactor coolant pressure boundary. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code as specified in this paragraph.
Each manufacturing license, standard design approval, and design
certification application under part 52 of this chapter and each
combined license for a utilization facility is subject to the following
conditions:
(1) Standards requirement for reactor coolant pressure boundary
components. Components that are part of the reactor coolant pressure
boundary must meet the requirements for Class 1 components in Section
III \1,4\ of the ASME BPV Code, except as provided in paragraphs (c)(2)
through (4) of this section.
(2) Exceptions to reactor coolant pressure boundary standards
requirement. Components that are connected to the reactor coolant
system and are part of the reactor coolant pressure boundary as defined
in Sec. 50.2 need not meet the requirements of paragraph (c)(1) of
this section, provided that:
[[Page 65808]]
(i) Exceptions: Shutdown and cooling capability. In the event of
postulated failure of the component during normal reactor operation,
the reactor can be shut down and cooled down in an orderly manner,
assuming makeup is provided by the reactor coolant makeup system; or
(ii) Exceptions: Isolation capability. The component is or can be
isolated from the reactor coolant system by two valves in series (both
closed, both open, or one closed and the other open). Each open valve
must be capable of automatic actuation and, assuming the other valve is
open, its closure time must be such that, in the event of postulated
failure of the component during normal reactor operation, each valve
remains operable and the reactor can be shut down and cooled down in an
orderly manner, assuming makeup is provided by the reactor coolant
makeup system only.
(3) Applicable Code and Code Cases and conditions on their use. The
Code edition, addenda, and optional ASME Code Cases to be applied to
components of the reactor coolant pressure boundary must be determined
by the provisions of paragraph NCA-1140, Subsection NCA of Section III
of the ASME BPV Code, subject to the following conditions:
(i) Reactor coolant pressure boundary condition: Code edition and
addenda. The edition and addenda applied to a component must be those
that are incorporated by reference in paragraph (a)(1)(i) of this
section;
(ii) Reactor coolant pressure boundary condition: Earliest edition
and addenda for pressure vessel. The ASME Code provisions applied to
the pressure vessel may be dated no earlier than the summer 1972
Addenda of the 1971 Edition;
(iii) Reactor coolant pressure boundary condition: Earliest edition
and addenda for piping, pumps, and valves. The ASME Code provisions
applied to piping, pumps, and valves may be dated no earlier than the
Winter 1972 Addenda of the 1971 Edition; and
(iv) Reactor coolant pressure boundary condition: Use of Code
Cases. The optional Code Cases applied to a component must be those
listed in NRC Regulatory Guide 1.84 that is incorporated by reference
in paragraph (a)(3)(i) of this section.
(4) Standards requirement for components in older plants. For a
nuclear power plant whose construction permit was issued prior to May
14, 1984, the applicable Code edition and addenda for a component of
the reactor coolant pressure boundary continue to be that Code edition
and addenda that were required by Commission regulations for such a
component at the time of issuance of the construction permit.
(d) Quality Group B components. Systems and components of boiling
and pressurized water-cooled nuclear power reactors must meet the
requirements of the ASME BPV Code as specified in this paragraph. Each
manufacturing license, standard design approval, and design
certification application under part 52 of this chapter, and each
combined license for a utilization facility is subject to the following
conditions:
(1) Standards requirement for Quality Group B components. For a
nuclear power plant whose application for a construction permit under
this part, or a combined license or manufacturing license under part 52
of this chapter, docketed after May 14, 1984, or for an application for
a standard design approval or a standard design certification docketed
after May 14, 1984, components classified Quality Group B \7\ must meet
the requirements for Class 2 Components in Section III of the ASME BPV
Code.
(2) Quality Group B: Applicable Code and Code Cases and conditions
on their use. The Code edition, addenda, and optional ASME Code Cases
to be applied to the systems and components identified in paragraph
(d)(1) of this section must be determined by the rules of paragraph
NCA-1140, Subsection NCA of Section III of the ASME BPV Code, subject
to the following conditions:
(i) Quality Group B condition: Code edition and addenda. The
edition and addenda must be those that are incorporated by reference in
paragraph (a)(1)(i) of this section;
(ii) Quality Group B condition: Earliest edition and addenda for
components. The ASME Code provisions applied to the systems and
components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group B condition: Use of Code Cases. The optional
Code Cases must be those listed in NRC Regulatory Guide 1.84 that is
incorporated by reference in paragraph (a)(3)(i) of this section.
(e) Quality Group C components. Systems and components of boiling
and pressurized water-cooled nuclear power reactors must meet the
requirements of the ASME BPV Code as specified in this paragraph. Each
manufacturing license, standard design approval, and design
certification application under part 52 of this chapter and each
combined license for a utilization facility is subject to the following
conditions.
(1) Standards requirement for Quality Group C components. For a
nuclear power plant whose application for a construction permit under
this part, or a combined license or manufacturing license under part 52
of this chapter, docketed after May 14, 1984, or for an application for
a standard design approval or a standard design certification docketed
after May 14, 1984, components classified Quality Group C \9\ must meet
the requirements for Class 3 components in Section III of the ASME BPV
Code.
(2) Quality Group C applicable Code and Code Cases and conditions
on their use. The Code edition, addenda, and optional ASME Code Cases
to be applied to the systems and components identified in paragraph
(e)(1) of this section must be determined by the rules of paragraph
NCA-1140, subsection NCA of Section III of the ASME BPV Code, subject
to the following conditions:
(i) Quality Group C condition: Code edition and addenda. The
edition and addenda must be those incorporated by reference in
paragraph (a)(1)(i) of this section;
(ii) Quality Group C condition: Earliest edition and addenda for
components. The ASME Code provisions applied to the systems and
components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group C condition: Use of Code Cases. The optional
Code Cases must be those listed in NRC Regulatory Guide 1.84 that is
incorporated by reference in paragraph (a)(3)(i) of this section.
(f) Inservice testing requirements. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code and ASME Code for Operation and
Maintenance of Nuclear Power Plants as specified in this paragraph.
Each operating license for a boiling or pressurized water-cooled
nuclear facility is subject to the following conditions. Each combined
license for a boiling or pressurized water-cooled nuclear facility is
subject to the following conditions, but the conditions in paragraphs
(f)(4) through (6) of this section must be met only after the
Commission makes the finding under Sec. 52.103(g) of this chapter.
Requirements for inservice inspection of Class 1, Class 2, Class 3,
Class MC, and Class CC components (including their supports) are
located in Sec. 50.55a(g).
(1) Inservice testing requirements for older plants (pre-1971 CPs).
For a boiling or pressurized water-cooled nuclear power facility whose
construction permit was issued prior to January 1, 1971, pumps and
valves must meet the test requirements of paragraphs (f)(4) and (5) of
this section to the extent
[[Page 65809]]
practical. Pumps and valves that are part of the reactor coolant
pressure boundary must meet the requirements applicable to components
that are classified as ASME Code Class 1. Other pumps and valves that
perform a function to shut down the reactor or maintain the reactor in
a safe shutdown condition, mitigate the consequences of an accident, or
provide overpressure protection for safety-related systems (in meeting
the requirements of the 1986 Edition, or later, of the BPV or OM Code)
must meet the test requirements applicable to components that are
classified as ASME Code Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice
testing in plants with CPs issued between 1971 and 1974. For a boiling
or pressurized water-cooled nuclear power facility whose construction
permit was issued on or after January 1, 1971, but before July 1, 1974,
pumps and valves that are classified as ASME Code Class 1 and Class 2
must be designed and provided with access to enable the performance of
inservice tests for operational readiness set forth in editions and
addenda of Section XI of the ASME BPV incorporated by reference in
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases
listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide
1.192, Revision 1, that are incorporated by reference in paragraphs
(a)(3)(ii) and (iii) of this section, respectively) in effect 6 months
before the date of issuance of the construction permit. The pumps and
valves may meet the inservice test requirements set forth in subsequent
editions of this Code and addenda that are incorporated by reference in
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases
listed in NRC Regulatory Guide 1.147, Revision 17; or Regulatory Guide
1.192, Revision 1, that are incorporated by reference in paragraphs
(a)(3)(ii) and (iii) of this section, respectively), subject to the
applicable conditions listed therein.
(3) Design and accessibility requirements for performing inservice
testing in plants with CPs issued after 1974. For a boiling or
pressurized water-cooled nuclear power facility whose construction
permit under this part or design approval, design certification,
combined license, or manufacturing license under part 52 of this
chapter was issued on or after July 1, 1974:
(i)-(ii) [Reserved]
(iii) IST design and accessibility requirements: Class 1 pumps and
valves. (A) Class 1 pumps and valves: First provision. In facilities
whose construction permit was issued before November 22, 1999, pumps
and valves that are classified as ASME Code Class 1 must be designed
and provided with access to enable the performance of inservice testing
of the pumps and valves for assessing operational readiness set forth
in the editions and addenda of Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1)(ii) of this section (or
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147,
Revision 17, or Regulatory Guide 1.192, Revision 1, that are
incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this
section, respectively) applied to the construction of the particular
pump or valve or the summer 1973 Addenda, whichever is later.
(B) Class 1 pumps and valves: Second provision. In facilities whose
construction permit under this part, or design certification, design
approval, combined license, or manufacturing license under part 52 of
this chapter, issued on or after November 22, 1999, pumps and valves
that are classified as ASME Code Class 1 must be designed and provided
with access to enable the performance of inservice testing of the pumps
and valves for assessing operational readiness set forth in editions
and addenda of the ASME OM Code (or the optional ASME Code Cases listed
in NRC Regulatory Guide 1.192, Revision 1, that are incorporated by
reference in paragraph (a)(3)(iii) of this section), incorporated by
reference in paragraph (a)(1)(iv) of this section at the time the
construction permit, combined license, manufacturing license, design
certification, or design approval is issued.
(iv) IST design and accessibility requirements: Class 2 and 3 pumps
and valves. (A) Class 2 and 3 pumps and valves: First provision. In
facilities whose construction permit was issued before November 22,
1999, pumps and valves that are classified as ASME Code Class 2 and
Class 3 must be designed and be provided with access to enable the
performance of inservice testing of the pumps and valves for assessing
operational readiness set forth in the editions and addenda of Section
XI of the ASME BPV Code incorporated by reference in paragraph
(a)(1)(ii) of this section (or the optional ASME Code Cases listed in
NRC Regulatory Guide 1.147, Revision 17, that are incorporated by
reference in paragraph (a)(3)(ii) of this section) applied to the
construction of the particular pump or valve or the Summer 1973
Addenda, whichever is later.
(B) Class 2 and 3 pumps and valves: Second provision. In facilities
whose construction permit under this part, or design certification,
design approval, combined license, or manufacturing license under part
52 of this chapter, issued on or after November 22, 1999, pumps and
valves that are classified as ASME Code Class 2 and 3 must be designed
and provided with access to enable the performance of inservice testing
of the pumps and valves for assessing operational readiness set forth
in editions and addenda of the ASME OM Code (or the optional ASME OM
Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that are
incorporated by reference in paragraph (a)(3)(iii) of this section),
incorporated by reference in paragraph (a)(1)(iv) of this section at
the time the construction permit, combined license, or design
certification is issued.
(v) IST design and accessibility requirements: Meeting later IST
requirements. All pumps and valves may meet the test requirements set
forth in subsequent editions of codes and addenda or portions thereof
that are incorporated by reference in paragraph (a) of this section,
subject to the conditions listed in paragraph (b) of this section.
(4) Inservice testing standards requirement for operating plants.
Throughout the service life of a boiling or pressurized water-cooled
nuclear power facility, pumps and valves that are classified as ASME
Code Class 1, Class 2, and Class 3 must meet the inservice test
requirements (except design and access provisions) set forth in the
ASME OM Code and addenda that become effective subsequent to editions
and addenda specified in paragraphs (f)(2) and (3) of this section and
that are incorporated by reference in paragraph (a)(1)(iv) of this
section, to the extent practical within the limitations of design,
geometry, and materials of construction of the components.
(i) Applicable IST Code: Initial 120-month interval. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during the initial 120-month
interval must comply with the requirements in the latest edition and
addenda of the OM Code incorporated by reference in paragraph
(a)(1)(iv) of this section on the date 12 months before the date of
issuance of the operating license under this part, or 12 months before
the date scheduled for initial loading of fuel under a combined license
under part 52 of this chapter (or the optional ASME Code Cases listed
in NRC Regulatory Guide 1.192, Revision 1, that is incorporated by
reference in paragraph (a)(3)(iii) of this section,
[[Page 65810]]
subject to the conditions listed in paragraph (b) of this section).
(ii) Applicable IST Code: Successive 120-month intervals. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during successive 120-month
intervals must comply with the requirements of the latest edition and
addenda of the OM Code incorporated by reference in paragraph
(a)(1)(iv) of this section 12 months before the start of the 120-month
interval (or the optional ASME Code Cases listed in NRC Regulatory
Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that
are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of
this section, respectively), subject to the conditions listed in
paragraph (b) of this section.
(iii) [Reserved]
(iv) Applicable IST Code: Use of later Code editions and addenda.
Inservice tests of pumps and valves may meet the requirements set forth
in subsequent editions and addenda that are incorporated by reference
in paragraph (a)(1)(iv) of this section, subject to the conditions
listed in paragraph (b) of this section, and subject to NRC approval.
Portions of editions or addenda may be used, provided that all related
requirements of the respective editions or addenda are met.
(5) Requirements for updating IST programs--(i) IST program update:
Applicable IST Code editions and addenda. The inservice test program
for a boiling or pressurized water-cooled nuclear power facility must
be revised by the licensee, as necessary, to meet the requirements of
paragraph (f)(4) of this section.
(ii) IST program update: Conflicting IST Code requirements with
technical specifications. If a revised inservice test program for a
facility conflicts with the technical specifications for the facility,
the licensee must apply to the Commission for amendment of the
technical specifications to conform the technical specifications to the
revised program. The licensee must submit this application, as
specified in Sec. 50.4, at least 6 months before the start of the
period during which the provisions become applicable, as determined by
paragraph (f)(4) of this section.
(iii) IST program update: Notification of impractical IST Code
requirements. If the licensee has determined that conformance with
certain Code requirements is impractical for its facility, the licensee
must notify the Commission and submit, as specified in Sec. 50.4,
information to support the determination.
(iv) IST program update: Schedule for completing impracticality
determinations. Where a pump or valve test requirement by the Code or
addenda is determined to be impractical by the licensee and is not
included in the revised inservice test program (as permitted by
paragraph (f)(4) of this section), the basis for this determination
must be submitted for NRC review and approval not later than 12 months
after the expiration of the initial 120-month interval of operation
from the start of facility commercial operation and each subsequent
120-month interval of operation during which the test is determined to
be impractical.
(6) Actions by the Commission for evaluating impractical and
augmented IST Code requirements--(i) Impractical IST requirements:
Granting of relief. The Commission will evaluate determinations under
paragraph (f)(5) of this section that code requirements are
impractical. The Commission may grant relief and may impose such
alternative requirements as it determines are authorized by law, will
not endanger life or property or the common defense and security, and
are otherwise in the public interest, giving due consideration to the
burden upon the licensee that could result if the requirements were
imposed on the facility.
(ii) Augmented IST requirements. The Commission may require the
licensee to follow an augmented inservice test program for pumps and
valves for which the Commission deems that added assurance of
operational readiness is necessary.
(g) Inservice inspection requirements. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code as specified in this paragraph.
Each operating license for a boiling or pressurized water-cooled
nuclear facility is subject to the following conditions. Each combined
license for a boiling or pressurized water-cooled nuclear facility is
subject to the following conditions, but the conditions in paragraphs
(g)(4) through (6) of this section must be met only after the
Commission makes the finding under Sec. 52.103(g) of this chapter.
Requirements for inservice testing of Class 1, Class 2, and Class 3
pumps and valves are located in Sec. 50.55a(f).
(1) Inservice inspection requirements for older plants (pre-1971
CPs). For a boiling or pressurized water-cooled nuclear power facility
whose construction permit was issued before January 1, 1971, components
(including supports) must meet the requirements of paragraphs (g)(4)
and (g)(5) of this section to the extent practical. Components that are
part of the reactor coolant pressure boundary and their supports must
meet the requirements applicable to components that are classified as
ASME Code Class 1. Other safety-related pressure vessels, piping, pumps
and valves, and their supports must meet the requirements applicable to
components that are classified as ASME Code Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice
inspection in plants with CPs issued between 1971 and 1974. For a
boiling or pressurized water-cooled nuclear power facility whose
construction permit was issued on or after January 1, 1971, but before
July 1, 1974, components (including supports) that are classified as
ASME Code Class 1 and Class 2 must be designed and be provided with
access to enable the performance of inservice examination of such
components (including supports) and must meet the preservice
examination requirements set forth in editions and addenda of Section
III or Section XI of the ASME BPV Code incorporated by reference in
paragraph (a)(1) of this section (or the optional ASME Code Cases
listed in NRC Regulatory Guide 1.147, Revision 17, that are
incorporated by reference in paragraph (a)(3)(ii) of this section) in
effect 6 months before the date of issuance of the construction permit.
The components (including supports) may meet the requirements set forth
in subsequent editions and addenda of this Code that are incorporated
by reference in paragraph (a) of this section (or the optional ASME
Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are
incorporated by reference in paragraph (a)(3)(ii) of this section),
subject to the applicable limitations and modifications.
(3) Design and accessibility requirements for performing inservice
inspection in plants with CPs issued after 1974. For a boiling or
pressurized water-cooled nuclear power facility, whose construction
permit under this part, or design certification, design approval,
combined license, or manufacturing license under part 52 of this
chapter, was issued on or after July 1, 1974, the following are
required:
(i) ISI design and accessibility requirements: Class 1 components
and supports. Components (including supports) that are classified as
ASME Code Class 1 must be designed and be provided with access to
enable the performance of inservice examination of these components and
must meet the preservice examination requirements set forth in the
editions and addenda of Section III or Section XI of the ASME
[[Page 65811]]
BPV Code incorporated by reference in paragraph (a)(1) of this section
(or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147,
Revision 17, that are incorporated by reference in paragraph (a)(3)(ii)
of this section) applied to the construction of the particular
component.
(ii) ISI design and accessibility requirements: Class 2 and 3
components and supports. Components that are classified as ASME Code
Class 2 and Class 3 and supports for components that are classified as
ASME Code Class 1, Class 2, and Class 3 must be designed and provided
with access to enable the performance of inservice examination of these
components and must meet the preservice examination requirements set
forth in the editions and addenda of Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1)(ii) of this section (or
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147,
Revision 17, that are incorporated by reference in paragraph (a)(3)(ii)
of this section) applied to the construction of the particular
component.
(iii)-(iv) [Reserved]
(v) ISI design and accessibility requirements: Meeting later ISI
requirements. All components (including supports) may meet the
requirements set forth in subsequent editions of codes and addenda or
portions thereof that are incorporated by reference in paragraph (a) of
this section, subject to the conditions listed therein.
(4) Inservice inspection standards requirement for operating
plants. Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) that are
classified as ASME Code Class 1, Class 2, and Class 3 must meet the
requirements, except design and access provisions and preservice
examination requirements, set forth in Section XI of editions and
addenda of the ASME BPV Code (or ASME OM Code for snubber examination
and testing) that become effective subsequent to editions specified in
paragraphs (g)(2) and (3) of this section and that are incorporated by
reference in paragraph (a)(1)(ii) or (iv) for snubber examination and
testing of this section, to the extent practical within the limitations
of design, geometry, and materials of construction of the components.
Components that are classified as Class MC pressure retaining
components and their integral attachments, and components that are
classified as Class CC pressure retaining components and their integral
attachments, must meet the requirements, except design and access
provisions and preservice examination requirements, set forth in
Section XI of the ASME BPV Code and addenda that are incorporated by
reference in paragraph (a)(1)(ii) of this section, subject to the
condition listed in paragraph (b)(2)(vi) of this section and the
conditions listed in paragraphs (b)(2)(viii) and (ix) of this section,
to the extent practical within the limitation of design, geometry, and
materials of construction of the components.
(i) Applicable ISI Code: Initial 120-month interval. Inservice
examination of components and system pressure tests conducted during
the initial 120-month inspection interval must comply with the
requirements in the latest edition and addenda of the Code incorporated
by reference in paragraph (a) of this section on the date 12 months
before the date of issuance of the operating license under this part,
or 12 months before the date scheduled for initial loading of fuel
under a combined license under part 52 of this chapter (or the optional
ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when
using Section XI, or Regulatory Guide 1.192, Revision 1, when using the
OM Code, that are incorporated by reference in paragraphs (a)(3)(ii)
and (iii) of this section, respectively), subject to the conditions
listed in paragraph (b) of this section.
(ii) Applicable ISI Code: Successive 120-month intervals. Inservice
examination of components and system pressure tests conducted during
successive 120-month inspection intervals must comply with the
requirements of the latest edition and addenda of the Code incorporated
by reference in paragraph (a) of this section 12 months before the
start of the 120-month inspection interval (or the optional ASME Code
Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using
Section XI, or Regulatory Guide 1.192, Revision 1, when using the OM
Code, that are incorporated by reference in paragraphs (a)(3)(ii) and
(iii) of this section), subject to the conditions listed in paragraph
(b) of this section. However, a licensee whose inservice inspection
interval commences during the 12 through 18-month period after July 21,
2011, may delay the update of their Appendix VIII program by up to 18
months after July 21, 2011.
(iii) Applicable ISI Code: Optional surface examination
requirement. When applying editions and addenda prior to the 2003
Addenda of Section XI of the ASME BPV Code, licensees may, but are not
required to, perform the surface examinations of high-pressure safety
injection systems specified in Table IWB-2500-1, Examination Category
B-J, Item Numbers B9.20, B9.21, and B9.22.
(iv) Applicable ISI Code: Use of subsequent Code editions and
addenda. Inservice examination of components and system pressure tests
may meet the requirements set forth in subsequent editions and addenda
that are incorporated by reference in paragraph (a) of this section,
subject to the conditions listed in paragraph (b) of this section, and
subject to Commission approval. Portions of editions or addenda may be
used, provided that all related requirements of the respective editions
or addenda are met.
(v) Applicable ISI Code: Metal and concrete containments. For a
boiling or pressurized water-cooled nuclear power facility whose
construction permit under this part or combined license under part 52
of this chapter was issued after January 1, 1956, the following are
required:
(A) Metal and concrete containments: First provision. Metal
containment pressure retaining components and their integral
attachments must meet the inservice inspection, repair, and replacement
requirements applicable to components that are classified as ASME Code
Class MC;
(B) Metal and concrete containments: Second provision. Metallic
shell and penetration liners that are pressure retaining components and
their integral attachments in concrete containments must meet the
inservice inspection, repair, and replacement requirements applicable
to components that are classified as ASME Code Class MC; and
(C) Metal and concrete containments: Third provision. Concrete
containment pressure retaining components and their integral
attachments, and the post-tensioning systems of concrete containments,
must meet the inservice inspections, repair, and replacement
requirements applicable to components that are classified as ASME Code
Class CC.
(5) Requirements for updating ISI programs--(i) ISI program update:
Applicable ISI Code editions and addenda. The inservice inspection
program for a boiling or pressurized water-cooled nuclear power
facility must be revised by the licensee, as necessary, to meet the
requirements of paragraph (g)(4) of this section.
(ii) ISI program update: Conflicting ISI Code requirements with
technical specifications. If a revised inservice inspection program for
a facility conflicts with the technical specifications for the
facility, the licensee must apply to the Commission
[[Page 65812]]
for amendment of the technical specifications to conform the technical
specifications to the revised program. The licensee must submit this
application, as specified in Sec. 50.4, at least six months before the
start of the period during which the provisions become applicable, as
determined by paragraph (g)(4) of this section.
(iii) ISI program update: Notification of impractical ISI Code
requirements. If the licensee has determined that conformance with a
Code requirement is impractical for its facility the licensee must
notify the NRC and submit, as specified in Sec. 50.4, information to
support the determinations. Determinations of impracticality in
accordance with this section must be based on the demonstrated
limitations experienced when attempting to comply with the Code
requirements during the inservice inspection interval for which the
request is being submitted. Requests for relief made in accordance with
this section must be submitted to the NRC no later than 12 months after
the expiration of the initial or subsequent 120-month inspection
interval for which relief is sought.
(iv) ISI program update: Schedule for completing impracticality
determinations. Where the licensee determines that an examination
required by Code edition or addenda is impractical, the basis for this
determination must be submitted for NRC review and approval not later
than 12 months after the expiration of the initial or subsequent 120-
month inspection interval for which relief is sought.
(6) Actions by the Commission for evaluating impractical and
augmented ISI Code requirements--(i) Impractical ISI requirements:
Granting of relief. The Commission will evaluate determinations under
paragraph (g)(5) of this section that code requirements are
impractical. The Commission may grant such relief and may impose such
alternative requirements as it determines are authorized by law, will
not endanger life or property or the common defense and security, and
are otherwise in the public interest giving due consideration to the
burden upon the licensee that could result if the requirements were
imposed on the facility.
(ii) Augmented ISI program. The Commission may require the licensee
to follow an augmented inservice inspection program for systems and
components for which the Commission deems that added assurance of
structural reliability is necessary.
(A) [Reserved]
(B) Augmented ISI requirements: Submitting containment ISI
programs. Licensees do not have to submit to the NRC for approval of
their containment inservice inspection programs that were developed to
satisfy the requirements of Subsection IWE and Subsection IWL with
specified conditions. The program elements and the required
documentation must be maintained on site for audit.
(C) Augmented ISI requirements: Implementation of Appendix VIII to
Section XI. (1) Appendix VIII and the supplements to Appendix VIII to
Section XI, Division 1, 1995 Edition with the 1996 Addenda of the ASME
BPV Code must be implemented in accordance with the following schedule:
Appendix VIII and Supplements 1, 2, 3, and 8--May 22, 2000; Supplements
4 and 6--November 22, 2000; Supplement 11--November 22, 2001; and
Supplements 5, 7, and 10--November 22, 2002.
(2) Licensees implementing the 1989 Edition and earlier editions
and addenda of IWA-2232 of Section XI, Division 1, of the ASME BPV Code
must implement the 1995 Edition with the 1996 Addenda of Appendix VIII
and the supplements to Appendix VIII of Section XI, Division 1, of the
ASME BPV Code.
(D) Augmented ISI requirements: Reactor vessel head inspections--
(1) All licensees of pressurized water reactors must augment their
inservice inspection program with ASME Code Case N-729-1, subject to
the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of
this section. Licensees of existing operating reactors as of September
10, 2008, must implement their augmented inservice inspection program
by December 31, 2008. Once a licensee implements this requirement, the
First Revised NRC Order EA-03-009 no longer applies to that licensee
and shall be deemed to be withdrawn.
(2) Note 9 of ASME Code Case N-729-1 must not be implemented.
(3) Instead of the specified ``examination method'' requirements
for volumetric and surface examinations in Note 6 of Table 1 of Code
Case N-729-1, the licensee must perform volumetric and/or surface
examination of essentially 100 percent of the required volume or
equivalent surfaces of the nozzle tube, as identified by Figure 2 of
ASME Code Case N-729-1. A demonstrated volumetric or surface leak path
assessment through all J-groove welds must be performed. If a surface
examination is being substituted for a volumetric examination on a
portion of a penetration nozzle that is below the toe of the J-groove
weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface
examination must be of the inside and outside wetted surface of the
penetration nozzle not examined volumetrically.
(4) By September 1, 2009, ultrasonic examinations must be performed
using personnel, procedures, and equipment that have been qualified by
blind demonstration on representative mockups using a methodology that
meets the conditions specified in paragraphs (g)(6)(ii)(D)(4)(i)
through (iv), instead of the qualification requirements of Paragraph -
2500 of ASME Code Case N-729-1. References herein to Section XI,
Appendix VIII, must be to the 2004 Edition with no addenda of the ASME
BPV Code.
(i) The specimen set must have an applicable thickness
qualification range of +25 percent to -40 percent for nominal depth
through-wall thickness. The specimen set must include geometric and
material conditions that normally require discrimination from primary
water stress corrosion cracking (PWSCC) flaws.
(ii) The specimen set must have a minimum of ten (10) flaws that
provide an acoustic response similar to PWSCC indications. All flaws
must be greater than 10 percent of the nominal pipe wall thickness. A
minimum of 20 percent of the total flaws must initiate from the inside
surface and 20 percent from the outside surface. At least 20 percent of
the flaws must be in the depth ranges of 10-30 percent through-wall
thickness and at least 20 percent within a depth range of 31-50 percent
through-wall thickness. At least 20 percent and no more than 60 percent
of the flaws must be oriented axially.
(iii) Procedures must identify the equipment and essential
variables and settings used for the qualification, in accordance with
Subarticle VIII-2100 of Section XI, Appendix VIII. The procedure must
be requalified when an essential variable is changed outside the
demonstration range as defined by Subarticle VIII-3130 of Section XI,
Appendix VIII, and as allowed by Articles VIII-4100, VIII-4200, and
VIII-4300 of Section XI, Appendix VIII. Procedure qualification must
include the equivalent of at least three personnel performance
demonstration test sets. Procedure qualification requires at least one
successful personnel performance demonstration.
(iv) Personnel performance demonstration test acceptance criteria
must meet the personnel performance demonstration detection test
acceptance criteria of Table VIII--S10-1 of Section XI, Appendix VIII,
Supplement 10. Examination procedures, equipment,
[[Page 65813]]
and personnel are qualified for depth sizing and length sizing when the
RMS error, as defined by Subarticle VIII-3120 of Section XI, Appendix
VIII, of the flaw depth measurements, as compared to the true flaw
depths, do not exceed \1/8\ inch (3 mm) and the root mean square (RMS)
error of the flaw length measurements, as compared to the true flaw
lengths, do not exceed \3/8\ inch (10 mm), respectively.
(5) If flaws attributed to PWSCC have been identified, whether
acceptable or not for continued service under Paragraphs -3130 or -3140
of ASME Code Case N-729-1, the re-inspection interval must be each
refueling outage instead of the re-inspection intervals required by
Table 1, Note (8), of ASME Code Case N-729-1.
(6) Appendix I of ASME Code Case N-729-1 must not be implemented
without prior NRC approval.
(E) Augmented ISI requirements: Reactor coolant pressure boundary
visual inspections \10\--(1) All licensees of pressurized water
reactors must augment their inservice inspection program by
implementing ASME Code Case N-722-1, subject to the conditions
specified in paragraphs (g)(6)(ii)(E)(2) through (4) of this section.
The inspection requirements of ASME Code Case N-722-1 do not apply to
components with pressure retaining welds fabricated with Alloy 600/82/
182 materials that have been mitigated by weld overlay or stress
improvement.
(2) If a visual examination determines that leakage is occurring
from a specific item listed in Table 1 of ASME Code Case N-722-1 that
is not exempted by the ASME Code, Section XI, IWB-1220(b)(1),
additional actions must be performed to characterize the location,
orientation, and length of a crack or cracks in Alloy 600 nozzle
wrought material and location, orientation, and length of a crack or
cracks in Alloy 82/182 butt welds. Alternatively, licensees may replace
the Alloy 600/82/182 materials in all the components under the item
number of the leaking component.
(3) If the actions in paragraph (g)(6)(ii)(E)(2) of this section
determine that a flaw is circumferentially oriented and potentially a
result of primary water stress corrosion cracking, licensees must
perform non-visual NDE inspections of components that fall under that
ASME Code Case N-722-1 item number. The number of components inspected
must equal or exceed the number of components found to be leaking under
that item number. If circumferential cracking is identified in the
sample, non-visual NDE must be performed in the remaining components
under that item number.
(4) If ultrasonic examinations of butt welds are used to meet the
NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (3) of this section,
they must be performed using the appropriate supplement of Section XI,
Appendix VIII, of the ASME BPV Code.
(F) Augmented ISI requirements: Examination requirements for Class
1 piping and nozzle dissimilar-metal butt welds--(1) Licensees of
existing, operating pressurized-water reactors as of July 21, 2011,
must implement the requirements of ASME Code Case N-770-1, subject to
the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (10) of
this section, by the first refueling outage after August 22, 2011.
(2) Full structural weld overlays authorized by the NRC staff may
be categorized as Inspection Items C or F, as appropriate. Welds that
have been mitigated by the Mechanical Stress Improvement Process
(MSIP\TM\) may be categorized as Inspection Items D or E, as
appropriate, provided the criteria in Appendix I of the Code Case have
been met. For ISI frequencies, all other butt welds that rely on Alloy
82/182 for structural integrity must be categorized as Inspection Items
A-1, A-2 or B until the NRC staff has reviewed the mitigation and
authorized an alternative Code Case Inspection Item for the mitigated
weld, or until an alternative Code Case Inspection Item is used based
on conformance with an ASME mitigation Code Case endorsed in Regulatory
Guide 1.147 with conditions, if applicable, and incorporated by
reference in this section.
(3) Baseline examinations for welds in Table 1, Inspection Items A-
1, A-2, and B, must be completed by the end of the next refueling
outage after January 20, 2012. Previous examinations of these welds can
be credited for baseline examinations if they were performed within the
re-inspection period for the weld item in Table 1 using Section XI,
Appendix VIII, requirements and met the Code required examination
volume of essentially 100 percent. Other previous examinations that do
not meet these requirements can be used to meet the baseline
examination requirement, provided NRC approval of alternative
inspection requirements in accordance with paragraphs (z)(1) or (2) of
this section is granted prior to the end of the next refueling outage
after January 20, 2012.
(4) The axial examination coverage requirements of Paragraph--
2500(c) may not be considered to be satisfied unless essentially 100
percent coverage is achieved.
(5) All hot-leg operating temperature welds in Inspection Items G,
H, J, and K must be inspected each inspection interval. A 25 percent
sample of Inspection Items G, H, J, and K cold-leg operating
temperature welds must be inspected whenever the core barrel is removed
(unless it has already been inspected within the past 10 years) or 20
years, whichever is less.
(6) For any mitigated weld whose volumetric examination detects
growth of existing flaws in the required examination volume that exceed
the previous IWB-3600 flaw evaluations or new flaws, a report
summarizing the evaluation, along with inputs, methodologies,
assumptions, and causes of the new flaw or flaw growth is to be
provided to the NRC prior to the weld being placed in service other
than modes 5 or 6.
(7) For Inspection Items G, H, J, and K, when applying the
acceptance standards of ASME BPV Code, Section XI, IWB-3514, for planar
flaws contained within the inlay or onlay, the thickness ``t'' in IWB-
3514 is the thickness of the inlay or onlay. For planar flaws in the
balance of the dissimilar metal weld examination volume, the thickness
``t'' in IWB-3514 is the combined thickness of the inlay or onlay and
the dissimilar metal weld.
(8) Welds mitigated by optimized weld overlays in Inspection Items
D and E are not permitted to be placed into a population to be examined
on a sample basis and must be examined once each inspection interval.
(9) Replace the first two sentences of Extent and Frequency of
Examination for Inspection Item D in Table 1 of Code Case N-770-1 with,
``Examine all welds no sooner than the third refueling outage and no
later than 10 years following stress improvement application.'' Replace
the first two sentences of Note (11)(b)(2) in Code Case N-770-1 with,
``The first examination following weld inlay, onlay, weld overlay, or
stress improvement for Inspection Items D through K must be performed
as specified.''
(10) General Note (b) to Figure 5(a) of Code Case N-770-1
pertaining to alternative examination volume for optimized weld
overlays may not be applied unless NRC approval is authorized under
paragraphs (z)(1) or (2) of this section.
(h) Protection and safety systems. Protection systems of nuclear
power reactors of all types must meet the requirements specified in
this paragraph. Each combined license for a utilization facility is
subject to the following conditions.
[[Page 65814]]
(1) [Reserved]
(2) Protection systems. For nuclear power plants with construction
permits issued after January 1, 1971, but before May 13, 1999,
protection systems must meet the requirements stated in either IEEE
Std. 279, ``Criteria for Protection Systems for Nuclear Power
Generating Stations,'' or in IEEE Std. 603-1991, ``Criteria for Safety
Systems for Nuclear Power Generating Stations,'' and the correction
sheet dated January 30, 1995. For nuclear power plants with
construction permits issued before January 1, 1971, protection systems
must be consistent with their licensing basis or may meet the
requirements of IEEE Std. 603-1991 and the correction sheet dated
January 30, 1995.
(3) Safety systems. Applications filed on or after May 13, 1999,
for construction permits and operating licenses under this part, and
for design approvals, design certifications, and combined licenses
under part 52 of this chapter, must meet the requirements for safety
systems in IEEE Std. 603-1991 and the correction sheet dated January
30, 1995.
(i)-(y) [Reserved]
(z) Alternatives to codes and standards requirements. Alternatives
to the requirements of paragraphs (b) through (h) of this section or
portions thereof may be used when authorized by the Director, Office of
Nuclear Reactor Regulation, or Director, Office of New Reactors, as
appropriate. A proposed alternative must be submitted and authorized
prior to implementation. The applicant or licensee must demonstrate
that:
(1) Acceptable level of quality and safety. The proposed
alternative would provide an acceptable level of quality and safety; or
(2) Hardship without a compensating increase in quality and safety.
Compliance with the specified requirements of this section would result
in hardship or unusual difficulty without a compensating increase in
the level of quality and safety. Footnotes to Sec. 50.55a:
\1\ USAS and ASME Code addenda issued prior to the winter 1977
Addenda are considered to be ``in effect'' or ``effective'' 6 months
after their date of issuance and after they are incorporated by
reference in paragraph (a) of this section. Addenda to the ASME Code
issued after the summer 1977 Addenda are considered to be ``in
effect'' or ``effective'' after the date of publication of the
addenda and after they are incorporated by reference in paragraph
(a) of this section.
2-3 [Reserved].
\4\ For ASME Code editions and addenda issued prior to the
winter 1977 Addenda, the Code edition and addenda applicable to the
component is governed by the order or contract date for the
component, not the contract date for the nuclear energy system. For
the winter 1977 Addenda and subsequent editions and addenda the
method for determining the applicable Code editions and addenda is
contained in Paragraph NCA 1140 of Section III of the ASME Code.
5-6 [Reserved].
\7\ Guidance for quality group classifications of components
that are to be included in the safety analysis reports pursuant to
Sec. 50.34(a) and Sec. 50.34(b) may be found in Regulatory Guide
1.26, ``Quality Group Classifications and Standards for Water-,
Steam-, and Radiological-Waste-Containing Components of Nuclear
Power Plants,'' and in Section 3.2.2 of NUREG-0800, ``Standard
Review Plan for Review of Safety Analysis Reports for Nuclear Power
Plants.''
8-9 [Reserved].
\10\ For inspections to be conducted once per interval, the
inspections must be performed in accordance with the schedule in
Section XI, paragraph IWB-2400, except for plants with inservice
inspection programs based on a Section XI edition or addenda prior
to the 1994 Addenda. For plants with inservice inspection programs
based on a Section XI edition or addenda prior to the 1994 Addenda,
the inspection must be performed in accordance with the schedule in
Section XI, paragraph IWB-2400, of the 1994 Addenda.
Dated at Rockville, Maryland, this 11th day of August 2014.
For the Nuclear Regulatory Commission.
Daniel H. Dorman,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2014-25491 Filed 11-4-14; 8:45 am]
BILLING CODE 7590-01-P