Approval of American Society of Mechanical Engineers' Code Cases, 65775-65814 [2014-25491]
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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
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Federal Register / Vol. 79, No. 214 / Wednesday, November 5, 2014 / Rules and Regulations
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
Nuclear Regulatory
Commission.
ACTION: Final rule.
AGENCY:
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
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SUMMARY:
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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
https://www.regulations.gov and search
for Docket ID NRC–2009–0359. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–287–3422;
email: Carol.Gallagher@nrc.gov. 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 publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘ADAMS Public Documents’’ and then
select ‘‘Begin Web-Based ADAMS
Search.’’ For problems with ADAMS,
please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov. 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: Jennifer.Tobin@nrc.gov; or
Wallace Norris, Office of Nuclear
Regulatory Research, telephone: 301–
251–7650; email: Wallace.Norris@
nrc.gov; 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
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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 NRCapproved 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 § 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
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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 § 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,
§ 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 § 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
§ 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.’’
This final rule adds provisions that
allow the NRC to authorize alternatives
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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
Affiliation
Comment submission ADAMS
Accession No.
William Culp ............................................
Saige Stephens .......................................
Richard Swayne ......................................
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Commenter name
Private Citizen ........................................
Private Citizen ........................................
ASME .....................................................
ML13210A143
ML13210A151
ML13253A076
ML13252A286 **
Mark Richter ............................................
Nuclear Energy Institute ........................
ML13259A040
ML13254A080 **
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
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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
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Affected code cases/subject
Proposed Rule.
General.
N–60–5.
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.
Proposed Rule.
regulatory guides that have been incorporated by
reference into 10 CFR 50.55a.
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TABLE I—COMMENT SUBMISSIONS RECEIVED ON THE PROPOSED RULE AND DRAFT REGULATORY GUIDES—Continued
Comment submission ADAMS
Accession No.
Commenter name
Affiliation
Edward Colie ...........................................
Patricia Campbell ....................................
Devin Kelley ............................................
David Helker ............................................
South Carolina Electric and Gas ...........
GE Hitachi Nuclear Energy ....................
AREVA ...................................................
Exelon Generation Company, LLC ........
ML13254A082
ML13259A038
ML13259A039
ML13269A371
Shawn Comstock ....................................
Private Citizen ........................................
ML13182A081
Roy Hall ...................................................
Inservice Inspection Program Owners
Group.
ML13197A239
Affected code cases/subject
Proposed Rule.
1332–6.
N–71–18.
N–60–5.
N–798.
N–800.
N–702.
OMN–1 (2006 Addenda).
OMN–11 (2006 Addenda).
OMN–12 (2004 Edition).
N–805.
** 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
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Proposed Rule
Comment: The commenter developed
a proposed one-page revision to the
overall Codes and standards rule in
§ 50.55a that reflects the commenter’s
view of the current regulatory process
and suggested parsing the details of
§ 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 § 50.55a. (Culp–3)
NRC Response: The main purpose of
this rulemaking is to amend § 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 § 50.55a.
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The NRC is not proposing a major
restructuring or simplification of the
requirements in § 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 § 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 § 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 § 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
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on a generic basis or as part of a plantspecific review process that covers more
than the Codes and standards
mentioned. It does not seem appropriate
for § 50.55a to reference Codes and
standards that have been withdrawn
(e.g., IEEE 279). The content of § 50.55a
represents an archive of once-upon-atime 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
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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 § 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 § 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 § 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 § 50.55a in this rulemaking. The
scope of this rulemaking is to: (1)
Incorporate by reference the three
Regulatory Guides identifying NRCapproved 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 § 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 § 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 § 50.55a to aid
the reader of this regulation. The
hierarchy used in § 50.55a is that which
is used throughout the Code of Federal
Regulations and is dictated by the OFR.
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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 § 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 § 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 § 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 § 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
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NRC to consider alternative methods for
endorsing ASME Codes and standards,
such as moving many of the
requirements currently specified in
§ 50.55a into a suitable regulatory guide
that can be referenced within the
regulation. The NRC agrees that the
format and organization of § 50.55a
could be improved, and the NRC may,
in the future, conduct a rulemaking to
restructure and simplify § 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 § 50.55a proposed rule.
No change was made to the final rule
as a result of this comment.
Comment: The proposed revision to
§ 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
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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
§ 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 § 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 NRCapproved 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)
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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 § 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
§ 50.55a because it would aid readers in
navigating the rule. (South Carolina
Electric and Gas–7.2)
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NRC Response: The NRC appreciates
these practical suggestions and agrees
that adding hyperlinks or indentation
would aid the readers in navigating
§ 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 § 50.55a: ‘‘Final
Reorganization of Paragraphs and
Subparagraphs in 10 CFR 50.55a, ‘Codes
and standards’ ’’ (ADAMS Accession
No. ML14015A191) and ‘‘CrossReference Tables’’ (ADAMS Accession
No. ML14211A050—package with two
tables). The NRC is currently
considering developing several
alternatives to improve the format and
organization of § 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
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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)
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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
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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
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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
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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 § 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
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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
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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
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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 °F (93
°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 °F (93 °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
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3 piping. Moderate-energy piping is
defined as Class 2 and 3 piping whose
maximum operating temperature does
not exceed 200 °F (93 °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 § 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 1mil wire standard for qualification of
visual examinations for components
within the scope of these code cases.
Research has shown that characters on
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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 https://
www.nrc.gov/reading-rm/doccollections/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)
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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–
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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 § 50.55a(a)(3)
[§ 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)
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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
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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.
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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.
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65785
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.
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. (Comstock2.4)
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.
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
Regulatory Guide 1.193, Revision 4
(DG–1233)
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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 § 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.
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.
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N–805
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
§ 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
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.
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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
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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 § 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
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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
65787
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 .....................................................
N–570–2 ..........................................
7 .....................................................
N–580–2 ..........................................
N–655–1 ..........................................
4 .....................................................
2 .....................................................
N–708 ..............................................
2 .....................................................
N–759–2 ..........................................
4 .....................................................
N–760–2 ..........................................
7 .....................................................
N–767 ..............................................
4 .....................................................
N–774 ..............................................
7 .....................................................
N–782 ..............................................
N–801 ..............................................
9 .....................................................
4 (2010 Edition) .............................
N–802 ..............................................
4 (2010 Edition) .............................
Special Type 403 Modified Forgings or Bars, Section III, Division 1,
Class 1 and CS.
Alternative Rules for Linear Piping and Linear Standard Supports for
Classes 1, 2, 3, and MC, Section III, Division 1.
Use of Alloy 600 With Columbium Added, Section III, Division 1.
Use of SA–738, Grade B, for Metal Containment Vessels, Class MC,
Section III, Division 1.
Use of JIS G–4303, Grades SUS304, SUS304L, SUS316, and
SUS316L, Section III, Division 1.
Alternative Rules for Determining Allowable External Pressure and
Comprehensive Stress for Cylinders, Cones, Spheres, and Formed
Heads, Section III, Division 1.
Welding of Valve Plugs to Valve Stem Retainers, Classes 1, 2, and
3, Section III, Division 1.
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.
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.
Use of Editions, Addenda, and Cases, Section III, Division 1.
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.
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 .....................................................
N–716–1 ..........................................
1 (2013 Edition) .............................
N–739–1 ..........................................
1 .....................................................
N–747 ..............................................
9 .....................................................
N–762 ..............................................
1 .....................................................
N–765 ..............................................
8 .....................................................
N–769 ..............................................
8 .....................................................
N–773 ..............................................
8 .....................................................
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.
Alternative Piping Classification and Examination Requirements, Section XI, Division 1.
Alternative Qualification Requirements for Personnel Performing
Class CC Concrete and Post-Tensioning System Visual Examinations, Section XI, Division 1.
Reactor Vessel Head-to-Flange Weld Examinations, Section XI, Division 1.
Temper Bead Procedure Qualification Requirements for Repair/Replacement Activities Without Post Weld Heat Treatment, Section
XI, Division 1.
Alternative to Inspection Interval Scheduling Requirements of IWA–
2430, Section XI, Division 1.
Roll Expansion of Class 1 In-Core Housing Bottom Head Penetrations in BWRs, Section XI, Division 1.
Alternative Qualification Criteria for Eddy Current Examinations of
Piping Inside Surfaces, Section XI, Division 1.
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ASME OM Code Case
OMN–6 ............................................
OMN–8 ............................................
2006 Addenda ...............................
2006 Addenda ...............................
OMN–14 ..........................................
2004 Addenda ...............................
OMN–16 ..........................................
2006 Addenda ...............................
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Alternate Rules for Digital Instruments.
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.
Alternative Rules for Valve Testing Operations and Maintenance, Appendix I: BWR CRD Rupture Disk Exclusion.
Use of a Pump Curve for Testing.
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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 Support Structures, Section
III, Division I, Class 1.
N–208–2 ........................
4 .......................................................................
N–520–2 ........................
4 .......................................................................
N–757–1 ........................
2 .......................................................................
Fatigue Analysis for Precipitation Hardening
Nickel Alloy Bolting Material to Specification
SB–637 N07718 for Class 1 Construction,
Section III, Division 1.
Alternative Rules for Renewal of Active or Expired N-type Certificates for Plants Not in
Active Construction, Section III, Division 1.
Alternative Rules for Acceptability for Class 2
and 3 Valves (DN 25) and Smaller with
Welded and Nonwelded End Connections
Other than Flanges, Section III, Division 1.
The maximum yield strength of strain-hardened austenitic stainless
steel shall not exceed 90,000 psi in view of the susceptibility of
this material to environmental cracking.
(1) In Figure A, the words ‘‘No mean stress’’ shall be implemented
with the understanding that it denotes ‘‘Maximum mean stress.’’
(2) In Figure A, sy shall be implemented with the understanding that
it denotes smax.
The Code Case is considered acceptable with one clarification: an
AIA is an Authorized Inspection Agency and the AIA employs the
Authorized Nuclear Inspector (ANI).
The design provisions of ASME Section III, Division 1, Appendix XIII,
shall not be used for Class 3 valves.
ASME BPV Code Case, Section XI
8 .......................................................................
N–561–2 ........................
1 .......................................................................
N–562–2 ........................
1 .......................................................................
Alternative Requirements for Wall Thickness
Restoration of Class 3 Moderate Energy
Carbon Steel Piping, Section XI, Division 1.
N–597–2 ........................
Previously approved Code Case. NRC had
proposed one new condition in response to
public comment on last rulemaking.
Requirements for Analytical Evaluation of
Pipe Wall Thinning, Section XI, Division 1.
N–606–1 ........................
Public comment received on previously approved rule requesting revision to condition.
Condition was revised.
N–619 ............................
Responding to comment on previously approved Code Case.
N–648–1 ........................
Responding to comment on previously approved Code Case.
Similar and Dissimilar Metal Welding Using
Ambient Temperature Machine GTAW
Temper Bead Technique for BWR CRD
Housing/Stub Tube Repairs, Section XI, Division 1.
Alternative Requirements for Nozzle Inner Radius Inspections for Class 1 Pressurizer
and Steam Generator Nozzles, Section XI,
Division 1.
Alternative Requirements for Inner Radius Inspections for Class 1 Reactor Vessel Nozzles, Section XI, Division 1.
N–661–2 ........................
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N–508–4 ........................
1 .......................................................................
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Rotation of Serviced Snubbers and Pressure
Retaining Items for the Purpose of Testing,
Section XI, Division 1.
Alternative Requirements for Wall Thickness
Restoration of Class 2 and High Energy
Class 3 Carbon Steel Piping, Section XI,
Division 1.
Alternative Requirements for Wall Thickness
Restoration of Classes 2 and 3 Carbon
Steel Piping for Raw Water Service, Section XI, Division 1.
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When Section XI requirements are used to govern the examination
and testing of snubbers and the ISI Code of Record is earlier than
Section XI, 2006 Addenda, Footnote 1 shall not be applied.
(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.
(2) Paragraph 7(c): if the cause 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.
(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.
(2) Paragraph 7(c): if the cause 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.
New condition (6): The evaluation criteria in Code Case N–513–2
may be applied to Code Case N–597–2 for 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 °F (93 °C) and whose maximum operating pressure does not exceed 275 psig (1.9MPa). Code Case N-597–2
shall not be used to evaluate through-wall leakage conditions.
Prior to welding, an examination or verification must be performed to
ensure proper preparation of the base metal, and that the surface
is properly contoured so that an acceptable weld can be produced.
This verification is to be required in the welding procedures.
In lieu of a UT examination, licensees may perform a VT–1 examination in accordance with the code of record for the Inservice Inspection Program utilizing the allowable flaw length criteria of Table
IWB–3512–1 with limiting assumptions on the flaw aspect ratio.
In lieu of a UT examination, licensees may perform a VT–1 examination in accordance with the code of record for the Inservice Inspection Program utilizing the allowable flaw length criteria of Table
IWB–3512–1 with limiting assumptions on the flaw aspect ratio.
(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.
(2) Paragraph 7(c): if the cause 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.
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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued
Code case No.
Code supplement
Code case title
Conditions
N–702 ............................
Responding to comment on previously approved Code Case.
Alternative Requirements for Boiling Water
Reactor (BWR) Nozzle Inner Radius and
Nozzle-to-Shell Welds, Section XI, Division
1.
The technical basis supporting the implementation of this Code Case
is addressed by BWRVIP–108: 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 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
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OMN–1 ..........................
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Alternative Rules for Preservice and Inservice
Testing of Active Electric Motor-Operated
Valve Assemblies in Light-Water Reactor
Power Plants.
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Licensees may use Code Case OMN–1, ‘‘Alternative Rules for
Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants,’’ in
lieu of the 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.
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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued
Code supplement
Code case title
Conditions
OMN–3 ..........................
2004 Edition .....................................................
Requirements for Safety Significance Categorization of Components Using Risk Insights for Inservice Testing of LWR Power
Plants.
OMN–4 ..........................
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Code case No.
2004 Edition .....................................................
Requirements for Risk Insights for Inservice
Testing of Check Valves at LWR Power
Plants.
OMN–9 ..........................
2004 Edition .....................................................
Use of a Pump Curve for Testing ...................
In addition to those components identified in ASME IST Program
Plan, implementation of Section 1, ‘‘Applicability,’’ of the Code
Case must include within the scope of a licensee’s risk-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, RiskInformed 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.
(1) Valve opening and closing functions must be demonstrated when
flow testing or examination methods (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.
(1) When a reference curve may 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.
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TABLE III—CONDITIONALLY APPROVED CODE CASES—Continued
Code supplement
Code case title
Conditions
OMN–12 ........................
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Code case No.
2004 Edition .....................................................
Alternative Requirements for Inservice Testing
Using Risk Insights for Pneumatically and
Hydraulically Operated Valve Assemblies in
Light-Water Reactor Power Plants (OMCode 1998, Subsection ISTC).
(1) Paragraph 4.2, ‘‘Inservice Test Requirements,’’ of OMN–12 specifies inservice test requirements for pneumatically and hydraulically
operated valve assemblies categorized as high safety significant
within the scope of the Code Case. The inservice testing program
must include a mix of static and 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
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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
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the NRC, liquid metal; and submerged
spent fuel waste casks. Regulatory
Guide 1.193 is not incorporated by
reference into § 50.55a. Regulatory
Guide 1.193 is prepared by the NRC as
a resource for stakeholders, allowing
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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 § 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
§ 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 § 50.55a(z)
(previously § 50.55a(a)(3)) that would
allow applicants and licensees to
request authorization of alternatives for
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changes to conditions on NRC-approved
ASME Code Cases in current paragraphs
(b)(4), (b)(5), and (b)(6) of § 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
§§ 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
§ 50.55a. For the least disruption to the
existing structure of the section, the
NRC incorporated by reference the
multiple standards into § 50.55a(a), the
first paragraph of the section. Each
national consensus standard that is
being incorporated by reference in
§ 50.55a has been listed separately.
Accordingly, the regulatory language of
§§ 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
§ 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-byParagraph Discussion,’’ of this
document. The regulatory text of
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§ 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 § 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
§ 50.55a to the new regulations.
Some alternatives the NRC considered
were a new regulation adjacent to
§ 50.55a (e.g., §§ 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 § 50.200,
and continuing with §§ 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 redesignate paragraphs in § 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 § 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 § 50.55a with public input. The
current reorganization of this
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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 § 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 § 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 firstlevel 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.
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B. Reader Aids
The NRC staff has developed a table
showing the structure of § 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 §§ 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.
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ML14211A050- package contains two
tables).
VIII. Paragraph-by-Paragraph
Discussion
Overall Considerations on the Use of
ASME Code Cases
This rulemaking has amended
§ 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 § 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,
§ 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
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65793
by reference into § 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., § 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
§ 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
§ 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 § 50.55a(a)(3). If
a Code Case is incorporated by reference
into § 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 § 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
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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 120month ISI or IST update interval
expires, after which the continued use
of the Code Case is prohibited unless
NRC authorization is given under new
§ 50.55a(z). If a Code Case is
incorporated by reference into § 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
§ 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
§§ 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 § 50.54, the introductory statement
has been revised to include a reference
to § 50.55a. This revision clarifies that
nuclear power plant licensees, as
described in the introductory paragraph
of § 50.54, also are subject to the
applicable requirements delineated in
§ 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
§ 50.55a(a)(1), and no change to the
requirement is intended by the transfer
of this requirement from § 50.55a(a)(1)
to § 50.54(jj), except for clarification of
its applicability.
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Section 50.55
In § 50.55, the introductory text has
been revised to include references to
existing § 50.55a, and paragraphs (g) and
(h) have been added and reserved for
future use. Further, existing
§ 50.55a(a)(1) has been moved to a
newly created § 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
§ 50.55 has been revised to maintain the
existing applicability of the requirement
in the newly created § 50.55(i) to
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construction permits for utilization
facilities.
Section 50.55a
The introductory text to § 50.55a was
relocated to several other locations.
There is no introductory text to § 50.55a
in the new rule. The first sentence in the
previous introductory text was relocated
to the first sentence in § 50.55. The
remaining sentences were relocated to
§ 50.55a(b) (second sentence),
§ 50.55a(b)(1) (first sentence),
§ 50.55a(b)(4) (first sentence), § 50.55a(c)
(second sentence), § 50.55a(d) (second
sentence), § 50.55a(e) (second sentence),
§ 50.55a(f) (second and third sentences),
§ 50.55a(g) (second and third sentences),
and § 50.55a(h) (second sentence).
In addition to moving existing
paragraphs, creating new paragraphs,
and revising introductory and regulatory
texts, the footnotes in § 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 § 50.55a to
incorporate by reference the multiple
standards currently identified in
existing § 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
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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
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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 § 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
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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).
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65795
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 § 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).
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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 nonradiological 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.
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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.
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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 INFOCOLLECTS.RESOURCE@
NRC.GOV; 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 governmentunique 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
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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 ASMEapproved 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 § 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 § 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 § 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
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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
§ 50.55a(a)(3), requests for relief under
§ 50.55a(f) and (g), or requests for
exemptions under §§ 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
65797
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
ADAMS
Accession No.
Proposed rule documents
Proposed Rule–Regulatory Analysis ................................................................................................................................................
Proposed Rule–Federal Register Notice ........................................................................................................................................
Proposed Reorganization of Paragraphs and Subparagraphs ........................................................................................................
Draft RG 1.84, Revision 36 (DG–1230) ...........................................................................................................................................
Draft RG 1.147, Revision 17 (DG–1231) .........................................................................................................................................
Draft RG 1.192, Revision 1 (DG–1232) ...........................................................................................................................................
ADAMS
Accession No.
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Final rule documents
Final Rule–Regulatory Analysis ........................................................................................................................................................
Final Rule–Federal Register Notice ................................................................................................................................................
Final Reorganization of Paragraphs and Subparagraphs ................................................................................................................
Cross-Reference Tables (package) ..................................................................................................................................................
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 .......................................
RG 1.192, ‘‘Operation and Maintenance Code Case Acceptability, ASME OM Code,’’ Revision 1 ...............................................
RG 1.193, ‘‘ASME Code Cases Not Approved for Use,’’ Revision 4 ..............................................................................................
RG 1.200, ‘‘An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed
Activities,’’ Revision 2.
RG 1.201, ‘‘Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their
Safety Significance,’’ Revision 1.
2007/12/19—‘‘SECY—Petition for Rulemaking to amend 10 CFR 50.55a—Rev.1’’ submitted by Ray West ................................
Hatch Plant Report—‘‘Hatch, Units 1 & 2, Farley, Units 1 & 2, Vogtle, Units 1 & 2, Safety Evaluation Re. Request to Use
ASME Code Case N–661’’.
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ADAMS
Accession No.
Final rule documents
EPRI Technical Report—Project No. 704—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 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. 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 Quality Expectations and Requirements .............................................................
SECY–04–0118—Plan for the Implementation of the Commission’s Phased Approach to Probabilistic Risk Assessment Quality.
SRM–SECY–04–0118—Plan for the Implementation of the Commission’s Phased Approach to Probabilistic Risk Assessment
Quality.
NUREG–0800—Chapter 4, Section 4.5.1, Revision 3, Control Rod Drive Structural Materials, dated March 2007 ......................
NUREG–0800—Chapter 5, Section 5.2.3, Revision 3, Reactor Coolant Pressure Boundary Materials, dated March 2007 .........
NUREG/CR–6943—A Study of Remote Visual 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
1. The authority citation for part 50 is
revised to read as follows:
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■
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
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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).
2. In § 50.54, revise the introductory
text, add reserved paragraph (ii), and
add paragraph (jj) to read as follows:
■
§ 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 § 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.
■ 3. In § 50.55, revise the introductory
text, add reserved paragraphs (g) and
(h), and add paragraph (i) to read as
follows:
§ 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 § 50.55a;
each construction permit for a
production facility is subject to the
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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
§ 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
§ 50.55a until the date that the
Commission makes the finding under
§ 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.
■ 4. Revise § 50.55a to read as follows:
§ 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 https://
www.archives.gov/federal-register/cfr/
ibr-locations.html.
(1) American Society of Mechanical
Engineers (ASME), Three Park Avenue,
New York, NY 10016; telephone:
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1–800–843–2763; https://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,
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(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,
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65799
(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
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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;
https://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;
https://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: pdr.resource@nrc.gov;
https://www.nrc.gov/reading-rm/doccollections/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
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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
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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 °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
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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: Pressureretaining 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.
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(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
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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
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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
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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 ‘‘ownerdefined’’ 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
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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]
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(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
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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.
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(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
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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-metalinterface), 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
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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-tovessel 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
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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-metalinterface) 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-tobase-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.
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(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.
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(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
X
X
X
X
........................................
........................................
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Inner 15 percent ......................................................................................................................
Outside Diameter Surface .......................................................................................................
Subsurface ...............................................................................................................................
(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.
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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
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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
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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
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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 °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, DKI depends on the crack depth
(a), and the flow stress (sf). The flow stress
is defined by sf = 1/2(sys + sult), where sys
is the yield strength and sult 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 × 1.12 sf√(pa), S = 1 and DKI =
Kmax. For R < ¥2 and Kmax¥ Kmin ≤ 0.8 × 1.12
sf√(pa), S = 1 and DKI = (1 ¥ R) Kmax/3. For
R < 0 and Kmax ¥ Kmin > 0.8 × 1.12 sf√(pa),
S = 1 and DKI = Kmax¥Kmin.
(xxix) Section XI condition:
Nonmandatory Appendix R.
Nonmandatory Appendix R, ‘‘RiskInformed 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
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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.
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(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
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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 120month 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:
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(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 120month 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 120month 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 § 50.2
need not meet the requirements of
paragraph (c)(1) of this section,
provided that:
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(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
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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
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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 § 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 § 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
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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 watercooled 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
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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
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65809
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 120month interval. Inservice tests to verify
operational readiness of pumps and
valves, whose function is required for
safety, conducted during the initial 120month 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,
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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 § 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
§ 50.4, information to support the
determination.
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(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 § 52.103(g) of this chapter.
Requirements for inservice testing of
Class 1, Class 2, and Class 3 pumps and
valves are located in § 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.
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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
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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
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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 120month interval. Inservice examination
of components and system pressure
tests conducted during the initial 120month 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 120month 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 18month 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
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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 posttensioning 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
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for amendment of the technical
specifications to conform the technical
specifications to the revised program.
The licensee must submit this
application, as specified in § 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 § 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]
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(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
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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 throughwall 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,
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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
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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 (MSIPTM) may be
categorized as Inspection Items D or E,
as appropriate, provided the criteria in
Appendix I of the Code Case have been
met. For ISI frequencies, all other butt
welds that rely on Alloy 82/182 for
structural integrity must be categorized
as Inspection Items A–1, A–2 or B until
the NRC staff has reviewed the
mitigation and authorized an alternative
Code Case Inspection Item for the
mitigated weld, or until an alternative
Code Case Inspection Item is used based
on conformance with an ASME
mitigation Code Case endorsed in
Regulatory Guide 1.147 with conditions,
if applicable, and incorporated by
reference in this section.
(3) Baseline examinations for welds in
Table 1, Inspection Items A–1, A–2, and
B, must be completed by the end of the
next refueling outage after January 20,
2012. Previous examinations of these
welds can be credited for baseline
examinations if they were performed
within the re-inspection period for the
weld item in Table 1 using Section XI,
Appendix VIII, requirements and met
the Code required examination volume
of essentially 100 percent. Other
previous examinations that do not meet
these requirements can be used to meet
the baseline examination requirement,
provided NRC approval of alternative
inspection requirements in accordance
with paragraphs (z)(1) or (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.
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65813
(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.
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(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
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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
§ 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
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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 § 50.34(a)
and § 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
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]]>Agencies
[Federal Register Volume 79, Number 214 (Wednesday, November 5, 2014)]
[Rules and Regulations]
[Pages 65775-65814]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-25491]
[[Page 65775]]
Vol. 79
Wednesday,
No. 214
November 5, 2014
Part II
Nuclear Regulatory Commission
-----------------------------------------------------------------------
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]]
-----------------------------------------------------------------------
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.
-----------------------------------------------------------------------
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 https://www.regulations.gov and search for Docket ID NRC-2009-0359. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. 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 https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-Based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. 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: Jennifer.Tobin@nrc.gov; or
Wallace Norris, Office of Nuclear Regulatory Research, telephone: 301-
251-7650; email: Wallace.Norris@nrc.gov; 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.''
---------------------------------------------------------------------------
\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.
---------------------------------------------------------------------------
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
----------------------------------------------------------------------------------------------------------------
Comment submission ADAMS Affected code cases/
Commenter name Affiliation Accession No. subject
----------------------------------------------------------------------------------------------------------------
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 **
[[Page 65778]]
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.
----------------------------------------------------------------------------------------------------------------
** 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 https://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 INFOCOLLECTS.RESOURCE@NRC.GOV; 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
https://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; https://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; https://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; https://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: pdr.resource@nrc.gov; https://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
