Incorporation by Reference of American Society of Mechanical Engineers Codes and Code Cases, 32934-32986 [2017-14166]
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32934
Federal Register / Vol. 82, No. 136 / Tuesday, July 18, 2017 / Rules and Regulations
NUCLEAR REGULATORY
COMMISSION
10 CFR Part 50
[NRC–2011–0088]
RIN 3150–AI97
Incorporation by Reference of
American Society of Mechanical
Engineers Codes and Code Cases
Nuclear Regulatory
Commission.
ACTION: Final rule.
AGENCY:
The U.S. Nuclear Regulatory
Commission (NRC) is amending its
regulations to incorporate by reference
recent editions and addenda to the
American Society of Mechanical
Engineers (ASME) Codes for nuclear
power plants and a standard for quality
assurance. The NRC is also
incorporating by reference six ASME
Code Cases. This action is in accordance
with the NRC’s policy to periodically
update the regulations to incorporate by
reference new editions and addenda of
the ASME Codes and is intended to
maintain the safety of nuclear power
plants and to make NRC activities more
effective and efficient.
DATES: This final rule is effective on
August 17, 2017. The incorporation by
reference of certain publications listed
in the regulation is approved by the
Director of the Federal Register as of
August 17, 2017.
ADDRESSES: Please refer to Docket ID
NRC–2011–0088 when contacting the
NRC about the availability of
information for this action. You may
obtain publicly-available information
related to this action by any of the
following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2011–0088. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individuals listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain 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
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SUMMARY:
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email to pdr.resource@nrc.gov. For the
convenience of the reader, instructions
about obtaining materials referenced in
this document are provided in the
‘‘Availability of Documents’’ section.
• 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:
Daniel I. Doyle, Office of Nuclear
Reactor Regulation, telephone: 301–
415–3748, email: Daniel.Doyle@nrc.gov;
or Keith Hoffman, Office of Nuclear
Reactor Regulation, telephone: 301–
415–1294, email: Keith.Hoffman@
nrc.gov. Both are staff of the U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
The NRC is amending its regulations
to incorporate by reference recent
editions and addenda to the ASME
Codes for nuclear power plants and an
ASME standard for quality assurance.
The NRC is also incorporating by
reference six ASME Code Cases.
This final rule is the latest in a series
of rulemakings to amend the NRC’s
regulations to incorporate by reference
revised and updated ASME Codes for
nuclear power plants. The ASME is a
voluntary consensus standards body,
and the ASME Codes are voluntary
consensus standards. The ASME
periodically revises and updates its
codes for nuclear power plants by
issuing new editions and addenda. The
NRC’s use of the ASME Codes is
consistent with applicable requirements
of the National Technology Transfer and
Advancement Act (NTTAA). This
rulemaking is in accordance with the
NRC’s policy to update the regulations
to incorporate by reference those new
editions and addenda. The
incorporation by reference of the new
editions and addenda will maintain the
safety of nuclear power plants, make
NRC activities more effective and
efficient, and allow nuclear power plant
licensees and applicants to take
advantage of the latest ASME Codes.
Additional discussion of voluntary
consensus standards and the NRC’s
compliance with the NTTAA is set forth
in Section XIV of this document,
‘‘Voluntary Consensus Standards.’’
B. Major Provisions
Major provisions of this final rule
include:
• Incorporation by reference of ASME
Codes into the NRC’s regulations and
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delineation of the NRC’s requirements
for the use of these codes, including
conditions.
• Incorporation by reference of
various versions of quality assurance
standard NQA–1 into NRC regulations
and approval for their use.
• Incorporation by reference of six
ASME Code Cases.
C. Costs and Benefits
The NRC prepared a regulatory
analysis (ADAMS Accession No.
ML16130A522) to identify the costs and
benefits associated with this final rule.
The regulatory analysis prepared for this
rulemaking was used to determine if the
rule is cost-effective, overall, and to
help the NRC evaluate potentially costly
conditions placed on specific provisions
of the ASME Codes and Code Cases
which are the subject of this
rulemaking. Therefore, the regulatory
analysis focuses on the marginal
difference in benefits and costs for each
provision of this final rule relative to the
‘‘no action’’ baseline alternative. The
regulatory analysis identified costs and
benefits in a quantitative fashion as well
as in a qualitative fashion. An
uncertainty analysis was performed to
evaluate the effects of uncertainties in
the quantitative estimation of both costs
and benefits, and this analysis showed
the rule alternative is cost effective with
over 99 percent certainty. The standard
deviation of the cost estimate net benefit
is $4.1 million.
TABLE 1—COST-BENEFIT SUMMARY
Objective
Industry .................................
NRC ......................................
Net benefit ............................
Alternative 2—
the rule
alternative
net benefits
(costs)
(million) (Net
present value,
7% discount
rate)
$11.5
3.28
14.7
Table 1 summarizes the costs and
benefits for the alternative of proceeding
with the final rule (Alternative 2) and
shows that the final rule is
quantitatively cost-beneficial with a net
benefit of $14.7 million to both the
industry and the NRC when compared
to the regulatory baseline (Alternative
1). The regulatory analysis shows that
implementing the final rule is
quantitatively cost-effective and an
efficient use of NRC and Industry
resources. Uncertainty analysis shows a
standard deviation of $4.08 million,
resulting in a net benefit range of $8.19
million to $21.6 million. Because the
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Federal Register / Vol. 82, No. 136 / Tuesday, July 18, 2017 / Rules and Regulations
rulemaking alternative is cost-effective,
the rulemaking approach is
recommended.
There are several benefits associated
with this final rule. The new motoroperated valve (MOV) provisions in this
final rule result in over $25 million in
averted costs (7-percent net present
value) due to the removal of quarterly
testing requirements and replacing those
requirements with less frequent
diagnostic and biannual testing
requirements. Additionally, the
provisions in this final rule will result
in averted costs to the NRC and the
industry from relief requests for the
code cases in this final rule, in
particular the ASME OMN–20 Code
Case Time Period Extension provision,
in excess of $5.1 million (7-percent net
present value).
Qualitative factors which were
considered include regulatory stability
and predictability, regulatory efficiency,
and consistency with the NTTAA. Table
50 in the regulatory analysis includes a
discussion of the costs and benefits that
were considered qualitatively.
Considering non-quantified costs and
benefits, the regulatory analysis shows
that the rulemaking is justified because
the number and significance of the nonquantified benefits outweigh the nonquantified costs. Certainly, if the
qualitative benefits (including the safety
benefit, regulatory efficiency, and other
nonquantified benefits) are considered
together with the quantified benefits,
then the benefits would outweigh the
identified quantitative and qualitative
impacts. Therefore, integrating both
quantified and non-quantified costs and
benefits, the benefits of the final rule
outweigh the identified quantitative and
qualitative impacts attributable to the
final rule.
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Table of Contents
I. Background
II. Discussion
A. ASME BPV Code, Section III
B. ASME BPV Code, Section XI
C. OM Code
D. ASME Code Cases
III. Opportunities for Public Participation
IV. NRC Responses to Public Comments
V. Section-by-Section Analysis
VI. Generic Aging Lessons Learned Report
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Plain Writing
XI. Finding of No Significant Impact:
Environmental Assessment
XII. Paperwork Reduction Act Statement
XIII. Congressional Review Act
XIV. Voluntary Consensus Standards
XV. Incorporation by Reference—Reasonable
Availability to Interested Parties
XVI. Availability of Guidance
XVII. Availability of Documents
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I. Background
The ASME develops and publishes
the ASME Boiler and Pressure Vessel
Code (BPV Code), which contains
requirements for the design,
construction, and inservice inspection
(ISI) of nuclear power plant
components; and the OM Code,1 which
contains requirements for inservice
testing (IST) of nuclear power plant
components. Until 2012, the ASME
issued new editions of the ASME BPV
Code every 3 years and addenda to the
editions annually, except in years when
a new edition was issued. Similarly, the
ASME periodically published new
editions and addenda of the OM Code.
Starting in 2012, the ASME decided to
issue editions of its BPV and OM Codes
(no addenda) every 2 years. The new
editions and addenda typically revise
provisions of the ASME BPV and OM
Codes (ASME Codes) to broaden their
applicability, add specific elements to
current provisions, delete specific
provisions, and/or clarify them to
narrow the applicability of the
provision. The revisions to the editions
and addenda of the ASME Codes do not
significantly change philosophy or
approach.
It has been the NRC’s practice to
establish requirements for the design,
construction, operation, ISI
examination, and IST of nuclear power
plants by approving the use of editions
and addenda of the ASME Codes in
§ 50.55a of title 10 of the Code of
Federal Regulations (10 CFR), ‘‘Codes
and standards.’’ The NRC approves and/
or mandates the use of certain parts of
editions and addenda of these ASME
Codes in § 50.55a through the
rulemaking process of ‘‘incorporation by
reference.’’ Upon incorporation by
reference of the ASME Codes into
§ 50.55a, the provisions of the ASME
Codes are legally-binding NRC
requirements as delineated in § 50.55a
and subject to the conditions on certain
specific ASME Code provisions that are
set forth in § 50.55a. The editions and
addenda of the ASME BPV and OM
Codes were last incorporated by
reference into the regulations in a final
rule dated June 21, 2011 (76 FR 36232),
subject to the NRC’s conditions.
The ASME Codes are consensus
standards developed by participants
with broad and varied interests,
including the NRC and licensees of
nuclear power plants. The ASME’s
adoption of new editions of, and
1 The editions and addenda of the ASME Code for
Operation and Maintenance of Nuclear Power
Plants have had different titles from 2005 to 2012
and are referred to collectively in this rule as the
‘‘OM Code.’’
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addenda to, the ASME Codes does not
mean that there is unanimity on every
provision in the ASME Codes. There
may be disagreement among the
technical experts, including NRC
representatives, on the ASME Code
committees and subcommittees,
regarding the acceptability or
desirability of a particular Code
provision included in an ASMEapproved Code edition or addenda. If
the NRC believes that there is a
significant technical or regulatory
concern with a provision in an ASMEapproved Code edition or addenda
being considered for incorporation by
reference, then the NRC will condition
the use of that provision when it
incorporates by reference that ASME
Code edition or addenda. In some cases,
the condition increases the level of
safety afforded by the ASME Code
provision or addresses a regulatory issue
not considered by the ASME. In other
instances, where research data or
experience has shown that certain Code
provisions are unnecessarily
conservative, the condition may provide
that the Code provision need not be
complied with in some or all respects.
The NRC’s conditions are included in
§ 50.55a, typically in paragraph (b) of
that regulation. In a Staff Requirements
Memorandum (SRM) dated September
10, 1999, the Commission indicated that
NRC rulemakings adopting
(incorporating by reference) a voluntary
consensus standard must identify and
justify each part of the standard that is
not adopted. For this rulemaking, the
provisions of the 2009 Addenda, 2010
Edition, 2011 Addenda, and 2013
Edition of Section III, Division 1; and
the 2009 Addenda, 2010 Edition, 2011
Addenda, and 2013 Edition of Section
XI, Division 1, of the ASME BPV Code;
and the 2009 Edition, 2011 Addenda,
and 2012 Edition of the OM Code that
the NRC is not adopting, or partially
adopting, are identified in the
Discussion, Regulatory Analysis, and
Backfitting and Issue Finality sections of
this document. The provisions of those
specific editions and addenda and Code
Cases that are the subject of this
rulemaking that the NRC finds to be
conditionally acceptable, together with
the applicable conditions, are also
identified in the Discussion, Regulatory
Analysis, and Backfitting and Issue
Finality sections of this document.
The ASME Codes are voluntary
consensus standards, and the NRC’s
incorporation by reference of these
Codes is consistent with applicable
requirements of the NTTAA. Additional
discussion on NRC’s compliance with
the NTTAA is set forth in Section XIV
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of this document, ‘‘Voluntary Consensus
Standards.’’
This final rule reflects the NRC’s
redesignation of paragraphs within
§ 50.55a set forth in a final rule dated
November 5, 2014 (79 FR 65776), as
corrected on December 11, 2014 (79 FR
73461). The re-designation of
paragraphs was needed to address the
Office of the Federal Register’s
requirements in 1 CFR part 51 for
incorporation by reference. For
additional information on the November
2014 final rule, please consult the
statement of considerations (preamble)
for that final rule.
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II. Discussion
The NRC regulations incorporate by
reference ASME Codes for nuclear
power plants. The ASME periodically
revises and updates its codes for nuclear
power plants. This final rule is the latest
in a series of rulemakings to amend the
NRC’s regulations to incorporate by
reference revised and updated ASME
Codes for nuclear power plants. The
proposed rule which led to this final
rule was published on September 18,
2015 (80 FR 56820). This rulemaking is
intended to maintain the safety of
nuclear power plants and make NRC
activities more effective and efficient.
The NRC follows a three-step process
to determine acceptability of new
provisions in new editions and addenda
to the Codes and the need for conditions
on the uses of these Codes. This process
was employed in the review of the
Codes that are the subject of this rule.
First, the NRC staff actively participates
with other ASME committee members
with full involvement in discussions
and technical debates in the
development of new and revised Codes.
This includes a technical justification of
each new or revised Code. Second, the
NRC committee representatives discuss
the Codes and technical justifications
with other cognizant NRC staff to ensure
an adequate technical review. Third, the
NRC position on each Code is reviewed
and approved by NRC management as
part of the rule amending § 50.55a to
incorporate by reference new editions
and addenda of the ASME Codes and
conditions on their use. This regulatory
process, when considered together with
the ASME’s own process for developing
and approving the ASME Codes,
provides reasonable assurance that the
NRC approves for use only those new
and revised Code edition and addenda,
with conditions as necessary, that
provide reasonable assurance of
adequate protection to public health and
safety, and that do not have significant
adverse impacts on the environment.
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The NRC is amending its regulations
to incorporate by reference:
• The 2009 Addenda, 2010 Edition,
2011 Addenda, and 2013 Edition of the
ASME BPV Code, Section III, Division 1
and Section XI, Division 1, with
conditions on their use.
• The 2009 Edition, the 2011
Addenda, and the 2012 Edition of
Division 1 of the OM Code, with
conditions on their use.
• ASME Standard NQA–1, ‘‘Quality
Assurance Requirements for Nuclear
Facility Applications,’’ including
several editions and addenda to NQA–
1 from previous years with slightly
varying titles as identified in
§ 50.55a(a)(1)(v). More specifically, the
NRC is incorporating by reference the
1983 Edition through the 1994 Edition,
the 2008 Edition, and the 2009–1a
Addenda to the 2008 Edition of ASME
NQA–1, with conditions on their use.
• ASME BPV Code Case N–513–3,
‘‘Evaluation Criteria for Temporary
Acceptance of Flaws in Moderate
Energy Class 2 or 3 Piping Section XI,
Division 1,’’ Mandatory Appendix I,
‘‘Relations for Fm, Fb, and F for ThroughWall Flaws,’’ Approval Date: January 26,
2009. This Code Case has already been
approved for use by the NRC in
Regulatory Guide (RG) 1.147 (75 FR
61321; October 5, 2010), but is now
being incorporated by reference in order
to adopt a condition on Nonmandatory
Appendix U, which requires the use of
this Code Case appendix.
• ASME BPV Code Case N–729–4,
‘‘Alternative Examination Requirements
for PWR Reactor Vessel Upper Heads
With Nozzles Having Pressure-Retaining
Partial-Penetration Welds Section XI,
Division 1,’’ ASME approval date: June
22, 2012, with conditions on its use.
• ASME BPV Code Case N–770–2,
‘‘Alternative Examination Requirements
and Acceptance Standards for Class 1
PWR Piping and Vessel Nozzle Butt
Welds Fabricated with UNS N06082 or
UNS W86182 Weld Filler Material With
or Without Application of Listed
Mitigation Activities, Section XI,
Division 1,’’ ASME approval date: June
9, 2011, with conditions on its use.
• ASME BPV Code Case N–824,
‘‘Ultrasonic Examination of Cast
Austenitic Piping Welds From the
Outside Surface Section XI, Division 1,’’
ASME approval date: October 16, 2012.
• ASME BPV Code Case N–852,
‘‘Application of the ASME NPT Stamp,
Section III, Division 1; Section III,
Division 2; Section III, Division 3;
Section III, Division 5,’’ Approval Date:
February 9, 2015.
• OM Code Case OMN–20, ‘‘Inservice
Test Frequency.’’
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The current regulations in
§ 50.55a(a)(1)(ii) incorporate by
reference ASME BPV Code, Section XI,
1970 Edition through the 1976 Winter
Addenda; and the 1977 Edition
(Division 1) through the 2008 Addenda
(Division 1), subject to the existing
conditions in § 50.55a(b)(2)(i) through
(xxix). This amendment revises
§ 50.55a(a)(1)(ii) to incorporate by
reference the 2009 Addenda (Division 1)
through the 2013 Edition (Division 1) of
the ASME BPV Code, Section XI. It also
clarifies the wording and adds, removes,
or revises some of the conditions as
explained in this document.
The NRC is revising § 50.55a(a)(1)(iv)
to incorporate by reference the 2009
Edition, 2011 Addenda, and 2012
Edition of Division 1 of the OM Code.
Based on this revision, the NRC
regulations will incorporate by reference
in § 50.55a the 1995 Edition through the
2012 Edition of the OM Code.
The NRC reviewed changes to the
Codes in the editions and addenda of
the Codes identified in this rulemaking,
and published a proposed rule in the
Federal Register setting forth the NRC’s
proposal to incorporate by reference the
ASME Codes, together with proposed
conditions on their use (80 FR 56820;
September 18, 2015). After
consideration of the public comments
received on the proposed rule (public
comments are discussed in Section IV of
this document, ‘‘NRC Responses to
Public Comments’’), the NRC concludes,
in accordance with the process for
review of changes to the Codes, that
each of the editions and addenda of the
Codes, and the 2008 Edition and the
2009–1a Addenda of NQA–1, are
technically adequate, consistent with
current NRC regulations, and approved
for use with specified conditions set
forth in this final rule. Each of the NRC
conditions and the reasons for each
condition are discussed in the following
sections. The discussions are organized
under the applicable ASME Code and
Section.
There is not a separate heading for
ASME quality assurance standard NQA–
1 because there are three separate
discussions of NQA–1—one under the
heading for ASME BPV Code, Section
III, one under the heading for ASME
BPV Code, Section XI, and one under
the heading for OM Code—because
there are three conditions related to
NQA–1, one in each of those areas
(§ 50.55a(b)(1)(iv) for Section III,
§ 50.55a(b)(2)(x) for Section XI, and
§ 50.55a(b)(3)(i) for the OM Code). In
addition, administrative and editorial
changes to various paragraphs of
§ 50.55a are being adopted for accuracy,
clarity, consistency, and general
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administrative convenience. These
editorial changes are not further
discussed in this heading, but are
described in Section V of this
document, ‘‘Section-by-Section
Analysis.’’
Four of the six ASME Code Cases
being incorporated by reference in this
rulemaking (N–729–4, N–770–2, N–824,
and OMN–20) are discussed in Section
II.D of this document, ‘‘ASME Code
Cases.’’ A fifth ASME Code Case, N–
852, is discussed in Section II.A,
‘‘ASME BPV Code, Section III,’’ because
the NRC’s approval of that Code Case
relates to a provision of Section III,
which is addressed in § 50.55a(b)(1)(ix).
The sixth ASME Code Case, N–513–3, is
discussed in Section II.B, ‘‘ASME BPV
Code, Section XI,’’ because the NRC’s
approval of that Code Case relates to a
provision of Section XI, which is
addressed in § 50.55a(b)(2)(xxxiv).
A. ASME BPV Code, Section III
10 CFR 50.55a(a)(1)(i) ASME Boiler and
Pressure Vessel Code, Section III
The NRC is clarifying that Section III
Nonmandatory Appendices are not
incorporated by reference. This
language was originally added in a final
rule published on June 21, 2011 (76 FR
36232); however, it was omitted from
the final rule published on November 5,
2014 (79 FR 65776). The NRC is
correcting the omission by inserting the
parenthetical clause ‘‘(excluding
Nonmandatory Appendices)’’ in
§ 50.55a(a)(1)(i).
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10 CFR 50.55a(b)(1)(ii) Section III
Condition: Weld Leg Dimensions
The NRC is identifying prohibited
subparagraphs and notes for each ASME
BPV Code edition and addenda in
tabular form as opposed to the narrative
form of the existing regulation. No
substantive change to the requirements
is intended by this revision. The NRC
believes that presenting the information
in tabular form will increase the clarity
and understandability of the regulation.
The existing condition in
§ 50.55a(b)(1)(ii) prohibits, for welds
with leg sizes less than 1.09 tn, the use
of certain Code provisions in ASME
BPV Code, Section III, Division 1. The
Code provisions provide stress indices
for welded joints used in the design of
Class 2 and Class 3 piping. The use of
these indices is prohibited for welds
with leg sizes less than 1.09 tn, where tn
is the nominal pipe thickness because
this would result in a weld that would
be weaker than the pipe to which it is
adjoined under these dimensions. The
location of the prohibited provisions
vary in the Code editions and addenda
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from the 1989 Addenda through the
2013 Edition, so in this final rule the
NRC clearly identifies the prohibited
code provisions in the editions and
addenda in a tabular format.
As an editorial matter, this final rule
identifies the prohibited ASME BPV
Code provisions as ‘‘notes,’’ which is
the term used by the ASME, rather than
‘‘footnotes.’’ The NRC is using the
terminology used by the ASME for
clarity.
10 CFR 50.55a(b)(1)(iv) Section III
Condition: Quality Assurance
The NRC is approving for use the
version of NQA–1 referenced in the
2010 Edition, 2011 Addenda, and 2013
Edition of the ASME BPV Code, Section
III, Subsection NCA, Article 7000,
which this rule is also incorporating by
reference. This allows applicants and
licensees to use the 2008 Edition and
the 2009–1a Addenda of NQA–1 when
using the 2010 and later editions and
addenda of Section III.
In the 2010 Edition of ASME BPV
Code, Section III, Subsection NCA,
Article NCA–4000, ‘‘Quality
Assurance,’’ was updated to require NType Certificate Holders to comply with
the requirements of Part 1 of the 2008
Edition and the 2009–1a Addenda of
ASME Standard NQA–1, ‘‘Quality
Assurance Requirements for Nuclear
Facility Applications,’’ as modified and
supplemented in NCA–4120(b) and
NCA–4134. In addition, NCA–4110(b)
was revised to remove the reference to
a specific edition and addenda of ASME
NQA–1, and Table NCA–7100–2,
‘‘Standards and Specifications
Referenced in Division 1,’’ was updated
to require the 2008 Edition and 2009–
1a Addenda of NQA–1 when using the
2010 Edition of Section III. In light of
these changes, the NRC reviewed the
2008 Edition and the 2009–1a Addenda
of NQA–1 and compared it to
previously approved versions of NQA–
1 and found that there were no
significant differences. In addition, the
NRC reviewed the changes to
Subsection NCA that reference the 2008
Edition and 2009–1a Addenda of NQA–
1, compared them to previously
approved versions of Subsection NCA,
and found that there were no significant
differences. Therefore, the NRC has
concluded that these editions and
addenda of NQA–1 are acceptable for
use.
The NRC is revising § 50.55a(b)(1)(iv)
to clarify that an applicant’s or
licensee’s commitments addressing
those areas where NQA–1 either does
not address a requirement in appendix
B to 10 CFR part 50, ‘‘Quality Assurance
Criteria for Nuclear Power Plants and
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32937
Fuel Reprocessing Plants,’’ or is less
stringent than the comparable appendix
B requirement govern the applicant’s or
licensee’s Section III activities. The
clarification is consistent with
§ 50.55a(b)(2)(x) and (b)(3)(i). The NQA–
1 provides the ASME’s method for
establishing and implementing a quality
assurance (QA) program for the design
and construction of nuclear power
plants and fuel reprocessing plants.
However, NQA–1, as modified and
supplemented in NCA–4120(b) and
NCA–4134, does not address some of
the requirements of appendix B to 10
CFR part 50. In some cases, the
provisions of NQA–1 are less stringent
than the comparable appendix B
requirements. Therefore, in order to
meet the requirements of appendix B, an
applicant’s or licensee’s QA program
description must contain commitments
addressing those provisions of appendix
B which are not covered by NQA–1, as
well as provisions that supplement or
replace the NQA–1 provisions where
the appendix B requirement is more
stringent.
Finally, the NRC is removing the
reference in § 50.55a(b)(1)(iv) to
versions of NQA–1 older than the 1994
Edition because the NRC did not receive
any adverse comments from any
applicant or licensee about removing
versions of NQA–1 older than the 1994
Edition from the regulation. The NRC
received only one comment regarding
NQA–1. The comment expressed
support for incorporation by reference
of NQA–1 and did not respond to the
NRC’s request for comment regarding
the removal of references to older
versions of NQA–1.
10 CFR 50.55a(b)(1)(vii) Section III
Condition: Capacity Certification and
Demonstration of Function of
Incompressible-Fluid Pressure-Relief
Valves
The NRC is revising § 50.55a(b)(1)(vii)
so that the existing condition
prohibiting the use of paragraph NB–
7742(a)(2) of the 2006 Addenda through
the 2007 Edition, up to and including
the 2008 Addenda, is extended to
include the editions and addenda up to
the 2013 Edition, which are the subject
of this rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III
Condition: Use of ASME Certification
Marks
The NRC is adding § 50.55a(b)(1)(viii)
to allow licensees to use either the
ASME BPV Code Symbol Stamps of
editions and addenda earlier than the
2011 Addenda to the 2010 Edition of the
ASME BPV Code or the ASME
Certification Marks with the appropriate
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certification designators and class
designators as specified in the 2013
Edition through the latest edition and
addenda incorporated by reference in
§ 50.55a.
The ASME BPV Code requires, in
certain instances, that components be
stamped. The stamp signifies that the
component has been designed,
fabricated, examined and tested, as
specified in the ASME BPV Code. The
stamp also signifies that the required
ASME BPV Code data report forms have
been completed, and the authorized
inspector has inspected the item and
authorized the application of the ASME
BPV Code Symbol Stamp.
The ASME has instituted changes in
the BPV Code to consolidate the
different ASME BPV Code Symbol
Stamps into a common ASME
Certification Mark. This action was
implemented in the 2011 Addenda to
the 2010 Edition of the ASME BPV
Code. As of the end of 2012, ASME no
longer utilizes the ASME BPV Code
Symbol Stamp. Licensees, however,
may not have updated to the edition or
addenda that identifies the use of the
ASME Certification Mark. Nevertheless,
licensees are legally required to
implement the ASME BPV Code Edition
and Addenda identified as their current
code of record. As ASME components
are procured, these components may be
received with the ASME Certification
Mark, while the licensee’s current code
of record may require the component to
have the ASME BPV Code Symbol
Stamp. Installation of a component
under such circumstances would not be
in compliance with the regulations that
the licensees are required to meet.
Both the ASME Certification Mark
and the ASME BPV Code Symbol Stamp
are official ASME methods of certifying
compliance with the Code. Although
these ASME Certification Marks differ
slightly in appearance, they serve the
same purpose of certifying code
compliance by the ASME Certificate
Holder and continue to provide for the
same level of quality assurance for the
application of the ASME Certification
Mark as was required for the application
of the ASME BPV Code Symbol Stamp.
The new ASME Certification Mark
represents a small, non-safety
significant modification of ASME’s
trademark. As such, it does not change
the technical requirements of the Code.
The ASME has confirmed that the
Certification Mark with designator is
equivalent to the corresponding BPV
Code Symbol Stamp. Based on
statements made by ASME in a letter
dated August 17, 2012, the NRC has
concluded that the ASME BPV Code
Symbol Stamps and ASME Certification
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Mark with code-specific designators are
equivalent with respect to their
certification of compliance with the
BPV Code. The NRC discussed this
issue in Regulatory Issue Summary
2013–07, ‘‘NRC Staff Position on the
Use of American Society of Mechanical
Engineers Certification Mark,’’ dated
May 28, 2013.
10 CFR 50.55a(b)(1)(ix) Section III
Condition: NPT Code Symbol Stamps
The NRC is adding § 50.55a(b)(1)(ix)
to allow licensees to use the NPT Code
Symbol Stamp with the letters arranged
horizontally as specified in ASME BPV
Code Case N–852 for the service life of
a component that had the NPT Code
Symbol Stamp applied during the time
period from January 1, 2005, through
December 31, 2015.
Public comments on the use of the
NPT Code Symbol requested that the
NRC accept the NPT Code Symbol
Stamp having the NPT letters arranged
horizontally as an acceptable NPT
Stamp to certify Code compliance for
fabricated items that have already been
stamped prior to receiving a
replacement NPT Code Symbol Stamp
from the ASME. The comments
requested that the NRC include
acceptance of Code Case N–852 in this
final rule for this purpose. Within the
context of its Code rules, ASME asserts
that the NPT Code Symbol Stamp
having the NPT letters arranged
horizontally, although differing slightly
in appearance from the NPT Code
Symbol Stamp as illustrated in Section
III, Table NCA–8100–1 of the ASME
BPV Code, 2010 Edition and earlier
editions and addenda, serves the same
purpose of certifying Code compliance
by the ASME NPT Certificate Holder
with confirmation by the Authorized
Nuclear Inspector and provides the
same level of quality assurance. In
addition, ASME indicated that on or
after January 1, 2016, the ASME will no
longer authorize use of the NPT Code
Symbol Stamp having the NPT letters
arranged horizontally. Accordingly, on
or after January 1, 2016, fabricated items
will only be stamped with the NPT
Code Symbol Stamp as illustrated in
Section III, Table NCA–8100–1 of the
ASME BPV Code, 2010 Edition and
earlier editions and addenda.
The NRC agrees in general with this
comment, in which the ASME asserts
that the ASME NPT Code Symbol Stamp
with the letters arranged horizontally to
be equivalent to the ‘‘N over PT’’ ASME
NPT Code Symbol Stamp. Therefore,
using either Code Symbol Stamp serves
the same purpose of certifying code
compliance by the ASME Certificate
Holder with confirmation by the
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Authorized Nuclear Inspector and
provides the same level of quality
assurance. The NRC also notes that the
same administrative and technical
requirements in the ASME Code still
apply whether an ASME NPT Code
Symbol Stamp with the letters arranged
horizontally or an ‘‘N over PT’’ ASME
NPT Code Symbol Stamp is applied.
However, since this NPT Code Symbol
Stamp having the NPT letters arranged
horizontally will only be applied onto
fabricated components from the time
period of January 1, 2005, through
December 31, 2015, the time period for
when this NPT Code Symbol Stamp was
applied to the component should be
limited to these dates to prevent
inadvertent fraudulent material.
Therefore, the NRC agrees that the
ASME BPV Code Case N–852 is
acceptable for the service life of the
component that had the NPT Code
Symbol stamp applied from the time
period of January 1, 2005, through
December 31, 2015. In response to this
comment, the NRC added
§ 50.55a(b)(1)(ix) to include a statement
that licensees may use the NPT Code
Symbol Stamp with the letters arranged
horizontally as specified in ASME BPV
Code Case N–852 for the service life of
a component that had the NPT Code
Symbol Stamp applied during the time
period from January 1, 2005, through
December 31, 2015. The NRC is
incorporating by reference ASME BPV
Code Case N–852 in § 50.55a(a)(1)(iii)(F)
because it is referenced in
§ 50.55a(b)(1)(ix).
Although the proposed rule did not
include this Code Case, the NRC has
determined that the incorporation by
reference of this Code Case at the final
rule stage is a logical outgrowth of the
proposed rule. The NRC’s intent to
ensure that § 50.55a identify all ASMEapproved methods for labelling Code
components is apparent from the
statement of considerations for the
proposed rule. See 80 FR 56820
(September 18, 2015) at 56823–56824.
The NRC did not entirely achieve that
purpose, and this resulted in public
comments seeking approval of this Code
Case, which supports the proposition
that the public had a reasonable
opportunity to either propose the
correction, with conditions as the
commenter believes are necessary or
desirable, or to indicate why the
(anticipated) correction should not be
made. Therefore, the NRC concludes
that it may incorporate by reference
ASME BPV Code Case N–852.
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B. ASME BPV Code, Section XI
10 CFR 50.55a(a)(1)(ii) ASME Boiler and
Pressure Vessel Code, Section XI
In the proposed rule, the NRC
proposed a revision to § 50.55a(a)(1)(ii)
that would have clarified that Section XI
Nonmandatory Appendix U of the 2013
Edition of ASME BPV Code, Section XI
was not incorporated by reference and
therefore not approved for use. After
considering public comments, the NRC
has determined that it will not exclude
Appendix U from the incorporation by
reference because it is the integration of
ASME BPV Code Cases N–513–3,
‘‘Evaluation Criteria for Temporary
Acceptance of Flaws in Moderate
Energy Class 2 or 3 Piping Section XI,
Division 1,’’ and N–705, ‘‘Evaluation
Criteria for Temporary Acceptance of
Degradation in Moderate Energy Class 2
or 3 Vessels and Tanks Section XI,
Division 1,’’ into Section XI. The NRC
has approved Code Cases N–513–3 and
N–705 in RG 1.147. However, as
described in the discussion for
§ 50.55a(b)(2)(xxxiv) in Section II.B,
‘‘ASME BPV Code Section XI,’’ the NRC
has found it necessary to adopt two new
conditions to the use of Nonmandatory
Appendix U.
The NRC is adopting two conditions
in the language of
§ 50.55a(a)(1)(ii)(C)(52) and (53) to
address two inconsistencies that were
identified between the NRC’s position
in a proposed rule regarding the
acceptability of ASME Code Cases (81
FR 10780; March 2, 2016) (2016 Code
Case proposed rule) and the proposed
rule for this rulemaking (80 FR 56820;
September 18, 2015). The first
inconsistency is that the NRC’s
proposed conditions on ASME BPV
Code Case N–799, ‘‘Dissimilar Metal
Welds Joining Vessel Nozzles to
Components,’’ in the 2016 Code Case
proposed rule were not reflected in the
2015 proposed rule for this rulemaking,
even though the technical content of
ASME BPV Code Case N–799 has been
incorporated into the 2011 Addenda
and 2013 Edition of ASME BPV Code,
Section XI. The second inconsistency is
that the NRC’s proposed disapproval of
ASME BPV Code Case N–813,
‘‘Alternative Requirements for
Preservice Volumetric and Surface
Examination,’’ in the 2016 Code Case
proposed rule was not reflected in the
2015 proposed rule for this rulemaking,
even though the technical content of
ASME BPV Code Case N–813 has been
incorporated into the 2013 Edition of
the ASME BPV Code, Section XI as
IWB–3112(a)(3) and IWC–3112(a)(3). To
address these two inconsistencies, the
NRC is excluding these ASME BPV
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Code, Section XI items from
incorporation by reference, as reflected
in § 50.55a(a)(1)(ii)(C)(52) and (53) of
the final rule. The NRC plans to
complete the development of the
regulatory approaches for examination
of component-to-component welds for
new construction plants and the
acceptance of preservice flaws by
analytical evaluation for operating
plants and include them in a future
rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI
Condition: Effective Edition and
Addenda of Subsection IWE and
Subsection IWL
The NRC is revising § 50.55a(b)(2)(vi)
to expressly state that licensees that
implemented the expedited examination
of containment during the 5-year period
from September 9, 1996, to September 9,
2001, 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), when
implementing the initial 120-month
inspection interval for the containment
ISI requirements of this section.
The expedited examination involved
the completion of the first set of
examinations of the first or initial 120month containment inspection interval.
It is noted that all of the operating
reactors in the previously stated class
would have gone past their initial 120month inspection interval by 2011. The
change removes the possibility of
misinterpretation of the provision as
requiring plants that do not fall in the
previously stated class, such as reactors
licensed after September 9, 2001, to use
the 1992 Edition with 1992 Addenda or
the 1995 Edition with 1996 Addenda of
Subsection IWE and Subsection IWL,
Section XI for implementing the initial
120-month inspection interval of the
containment ISI program. Applicants
and licensees that do not fall in the
previously stated class must use Code
editions and addenda in accordance
with § 50.55a(g)(4)(i) and (ii),
respectively, for the initial and
successive 120-month containment ISI
intervals.
10 CFR 50.55a(b)(2)(viii) Section XI
Condition: Concrete Containment
Examinations
The NRC is revising
§ 50.55a(b)(2)(viii) by removing the
condition for using the 2007 Edition
with 2009 Addenda through the 2013
Edition of Subsection IWL requiring
compliance with § 50.55a(b)(2)(viii)(E).
To support the removal of the condition,
the NRC is adding new requirements
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32939
governing the performance and
documentation of concrete containment
examinations in § 50.55a(b)(2)(viii)(H)
and (I), which are discussed separately
in the next two headings.
Section 50.55a(b)(2)(viii)(E) is one of
several conditions that apply to the
inservice examination of concrete
containments using Subsection IWL of
various editions and addenda of the
ASME BPV Code, Section XI,
incorporated by reference in
§ 50.55a(a)(1)(ii). The NRC is removing
the condition in § 50.55a(b)(2)(viii)(E)
when applying the 2007 Edition with
2009 Addenda through the 2013 Edition
of Subsection IWL because its intent has
been incorporated into the Code in the
new provision IWL–2512, ‘‘Inaccessible
Areas.’’
10 CFR 50.55a(b)(2)(viii)(H) Concrete
Containment Examinations: Eighth
Provision
The NRC is adding
§ 50.55a(b)(2)(viii)(H) to specify the
information that must be provided in
the ISI Summary Report required by
IWA–6000, when inaccessible concrete
surfaces are evaluated under the new
Code provision IWL–2512. This new
condition replaces the existing
condition in § 50.55a(b)(2)(viii)(E),
when using the 2007 Edition with the
2009 Addenda through the 2013 Edition
of Subsection IWL.
The existing condition in
§ 50.55a(b)(2)(viii)(E) of the current rule
requires that, for Class CC applications,
the licensee shall evaluate the
acceptability of inaccessible areas when
conditions exist in accessible areas that
could indicate the presence of or result
in degradation to such inaccessible
areas, and provide the evaluation
information required by
§ 50.55a(b)(2)(viii)(E)(1), (2), and (3) in
the IWA–6000 ISI Summary Report.
In the 2009 Addenda Subsection IWL,
the ASME revised existing provisions
IWL–1220 and IWL–2510 and added the
new provision IWL–2512 intended to
incorporate the condition in
§ 50.55a(b)(2)(viii)(E) into Subsection
IWL. The IWL–2510, ‘‘Surface
Examination,’’ was restructured into
new paragraphs in IWL–2511,
‘‘Accessible Areas,’’ with almost the
same provisions as the previous IWL–
2510 and IWL–2512, ‘‘Inaccessible
Areas,’’ to be specific to examinations
required for accessible areas, and
differentiate between those and the new
requirements for inaccessible areas. The
inaccessible areas addressed by the new
IWL–2512 are: (1) Concrete surfaces
obstructed by adjacent structures, parts
or appurtenances (e.g., generally abovegrade inaccessible areas); and (2)
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concrete surfaces made inaccessible by
foundation material or backfill (e.g.,
below-grade inaccessible areas).
The revised IWL–2511(a) has a new
requirement that states that, ‘‘If the
Responsible Engineer determines that
observed suspect conditions indicate
the presence of, or could result in,
degradation of inaccessible areas, the
requirements of IWL–2512(a) shall be
met.’’ The new IWL–2512(a) requires
the ‘‘Responsible Engineer’’ to evaluate
suspect conditions and specify the type
and extent of examinations, if any,
required to be performed on
inaccessible surface areas described in
the previous paragraph. The
acceptability of the evaluated
inaccessible area would be determined
either based on the evaluation or based
on the additional examinations, if
determined to be required. The new
IWL–2512(b) further requires a periodic
technical evaluation of below-grade
inaccessible areas of concrete to be
performed to determine and manage its
susceptibility to degradation regardless
of whether suspect conditions exist in
accessible areas that would warrant an
evaluation of inaccessible areas based
on the condition in
§ 50.55a(b)(2)(viii)(E). Therefore, the
revised IWL–2511(a) and new IWL–
2512 code provisions address the
evaluation and acceptability of
inaccessible areas consistent with the
existing condition in
§ 50.55a(b)(2)(viii)(E), with one
exception. The exception is that the new
IWL–2512 provision does not explicitly
require the information specified in
§ 50.55a(b)(2)(viii)(E)(1), (2), and (3) of
the existing condition to be provided in
the IWA–6000 ISI Summary Report.
For these reasons, the NRC is
identifying the information that must be
provided in the ISI Summary Report
required by IWA–6000 when
inaccessible concrete surfaces are
evaluated under the new code provision
IWL–2512. This new condition replaces
the existing condition in
§ 50.55a(b)(2)(viii)(E) when using the
2007 Edition with the 2009 Addenda
through the 2013 Edition of Subsection
IWL. The information required by the
new condition must be provided when
inaccessible concrete areas are
evaluated per IWL–2512(a) for
degradation based on suspect conditions
found in accessible areas, as well as
when periodic technical evaluations of
inaccessible below-grade concrete areas
required by IWL–2512(b) are performed.
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10 CFR 50.55a(b)(2)(viii)(I) Concrete
Containment Examinations: Ninth
Provision
The NRC is adding
§ 50.55a(b)(2)(viii)(I) to place a
condition on the periodic technical
evaluation requirements in the new
IWL–2512(b), for consistency with
NUREG–1801, Revision 2, ‘‘Generic
Aging Lessons Learned (GALL) Report,’’
with regard to aging management of
below-grade containment concrete
surfaces. The new IWL–2512(b)
provision is applicable to inaccessible
below-grade concrete surfaces exposed
to foundation soil, backfill, or
groundwater. This condition would
apply only during the period of
extended operation of a renewed license
under 10 CFR part 54, when using IWL–
2512(b) of the 2007 Edition with 2009
Addenda through the 2013 Edition of
Subsection IWL.
In the 2009 Addenda of Subsection
IWL, the ASME added new Code
provisions, IWL–2512(b) and (c) as well
as a new line item L1.13 in Table IWL–
2500–1, intended to specifically address
aging management concerns with
potentially unidentified degradation of
inaccessible below-grade containment
concrete areas and to be responsive to
actions outlined in the GALL Report
related to aging management of
inaccessible below-grade concrete
surfaces. It is noted that these new Code
provisions are an enhancement to the
requirement of the existing condition in
§ 50.55a(b)(2)(viii)(E) to specifically
address aging management of
inaccessible below-grade containment
concrete areas and is generally
acceptable to the NRC.
The new IWL–2512(b) provides
requirements for systematically
performing a periodic technical
evaluation of concrete surfaces exposed
to foundation soil, backfill, or
groundwater to determine susceptibility
of the concrete to deterioration that
could affect its ability to perform its
intended design function under
conditions anticipated through the
service life of the structure. It requires
the technical evaluation to be performed
and documented at periodic intervals
not to exceed 10 years regardless of
whether conditions exist in accessible
areas that would warrant an evaluation
of inaccessible areas by the existing
condition in § 50.55a(b)(2)(viii)(E),
which the NRC finds reasonable for the
initial 40-year operating license period.
The new IWL–2512(b) further provides
the specific elements, including aging
mechanisms considered, that the
technical evaluation should include, as
well as the definition of an aggressive
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below-grade environment. The new
IWL–2512(c) requires that the
evaluation results of IWL–2512(b) be
used to define and document the
condition monitoring program, if
determined to be required, including
required examinations and frequencies,
to be implemented for the management
of degradation and aging effects of the
below-grade concrete surface areas. If it
is determined that additional
examinations are required, these
examinations of inaccessible belowgrade areas will be implemented in
accordance with new line item L1.13 in
Table IWL–2500–1 under Examination
Category L–A, Concrete, with
acceptance criteria based on IWL–3210.
It should be noted that a technical
evaluation approach, such as in IWL–
2512(b), could be used, and is generally
used, to determine acceptability of a
below-grade inaccessible area to satisfy
the condition in § 50.55a(b)(2)(viii)(E).
The technical evaluation
requirements in IWL–2512(b) assist in
determining the susceptibility to
degradation and manage aging effects of
inaccessible below-grade concrete
surfaces, before the loss of intended
function. The requirements are based
on, and are generally consistent with,
the guidance in the GALL Report, with
the following two exceptions. The first
exception is that IWL–2512(b) requires
the technical evaluation to determine
the susceptibility of the concrete to
degradation and the ability to perform
the intended design function through its
service life at periodic intervals not to
exceed 10 years. The aging management
programs (AMPs) for safety-related
structures (e.g., Structures Monitoring)
in the GALL Report require such
evaluation to be performed at intervals
not to exceed 5 years, which is also
consistent with applicant commitments
during review of license renewal
applications. The second exception is
that IWL–2512(b) requires that
examination of representative samples
of below-grade concrete be performed if
excavated for any reason when an
aggressive below-grade environment is
present. However, the NRC notes that
the AMPs (X1.S6 Structures Monitoring
and X1.S7 Water Control Structures) in
the GALL Report require the same
examination even for a non-aggressive
below-grade environment.
Based on these reasons, the NRC is
adding § 50.55a(b)(2)(viii)(I) to place a
condition on the periodic technical
evaluation requirements in IWL–2512(b)
for consistency with the GALL Report,
when addressing the two exceptions
previously described with respect to
aging management of inaccessible
below-grade concrete components of the
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containment. The new condition
requires that, during the period of
extended operation of a renewed
license, the technical evaluation under
IWL–2512(b) of inaccessible belowgrade concrete surfaces exposed to
foundation soil, backfill, or groundwater
be performed at periodic intervals not to
exceed 5 years, as opposed to the 10year interval in IWL–2512. In addition,
the condition requires the examination
of representative samples of the exposed
portions of the below-grade concrete be
performed when excavated for any
reason as opposed to IWL–2512, which
limits the examination to excavations in
aggressive, below-grade environments.
Since the GALL Report is the technical
basis document for license renewal, this
new condition applies only during the
period of extended operation of a
renewed license under 10 CFR part 54,
when using IWL–2512(b) of the 2007
Edition with 2009 Addenda through the
2013 Edition of Subsection IWL, Section
XI.
10 CFR 50.55a(b)(2)(ix) Section XI
Condition: Metal Containment
Examinations
The NRC is extending the
applicability of the existing conditions
in § 50.55a(b)(2)(ix)(A)(2) and
(b)(2)(ix)(B) and (J), governing
examinations of metal containments and
the liners of concrete containments
under Subsection IWE, to the ASME
BPV Code editions and addenda which
are the subject of this rulemaking (i.e.,
the 2007 Edition with 2009 Addenda
through the 2013 Edition). The last
sentence of § 50.55a(b)(2)(ix) prior to
this final rule stated that the referenced
conditions were applicable only to
addenda, but not to editions, approved
by the NRC after the 2007 Edition of the
ASME BPV Code. To rectify this, the
NRC is revising the last sentence of
§ 50.55a(b)(2)(ix) to refer to the latest
‘‘edition and’’ addenda after the 2007
Edition which are incorporated by
reference into § 50.55a.
The NRC reviewed the Code changes
in Subsection IWE of the 2009 Addenda
through the 2013 Edition of ASME BPV
Code, Section XI, and noted that all of
the changes were editorial or
administrative with the intent to
improve the clarity of the existing
requirements or correct errors by errata.
There were no changes to Subsection
IWE in the Code editions and addenda
that are the subject of this rulemaking
that the NRC believes would require
new regulatory conditions to ensure
safety, nor do the changes to Subsection
IWE address the NRC’s reasons for
adopting the conditions on the use of
Subsection IWE. Accordingly, the NRC
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is extending the applicability of the
existing conditions (by adding the
words ‘‘edition and’’ to § 50.55a(b)(2)(ix)
as discussed) without any change to the
provisions of the conditions.
10 CFR 50.55a(b)(2)(x) Section XI
Condition: Quality Assurance
The NRC is approving for use the
version of NQA–1 referenced in the
2009 Addenda, 2010 Edition, 2011
Addenda, and the 2013 Edition of the
ASME BPV Code, Section XI, Table IWA
1600–1, ‘‘Referenced Standards and
Specifications,’’ which this rule is also
incorporating by reference. This allows,
but does not require, licensees to use the
1994 Edition or the 2008 Edition and
the 2009–1a Addenda of NQA–1 when
using the 2009 Addenda and later
editions and addenda of Section XI.
In the 2013 Edition of ASME BPV
Code, Section XI, Table IWA 1600–1
was updated to allow licensees to use
the 1994 Edition or the 2008 Edition
with the 2009–1a Addenda of NQA–1
when using the 2013 Edition of Section
XI. In the 2010 Edition of ASME BPV
Code, Section XI, IWA–1400, ‘‘Owner’s
Responsibilities,’’ Subparagraph (n)(2)
was updated to reference the NQA–1
Part I, Basic Requirements and
Supplementary Requirements for
Nuclear Facilities. In the 2009 Addenda
of the 2007 Edition of ASME BPV Code,
Section XI, Table IWA–1600–1,
‘‘Referenced Standards and
Specifications,’’ was updated to allow
licensees to use the 1994 Edition of
NQA–1. The NRC reviewed the 2008
Edition and the 2009–1a Addenda of
NQA–1 and compared it to previously
approved versions of NQA–1 and found
that there were no significant
differences. Therefore, the NRC has
concluded that these editions and
addenda of NQA–1 are acceptable for
use.
The NRC is amending
§ 50.55a(b)(2)(x) to clarify that a
licensee’s commitments addressing
those areas where NQA–1 either does
not address a requirements in appendix
B to 10 CFR part 50 or is less stringent
than the comparable appendix B
requirement govern the licensee’s
Section XI activities. The clarification is
consistent with § 50.55a(b)(1)(iv) and
(b)(3)(i). The ASME’s method for
establishing and implementing a QA
program for the design and construction
of nuclear power plants and fuel
reprocessing plants is described in
NQA–1. However, NQA–1 does not
address some of the requirements of
appendix B to 10 CFR part 50. In some
cases, the provisions of NQA–1 are less
stringent than the comparable appendix
B requirements. Therefore, in order to
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meet the requirements of appendix B, a
licensee’s QA program description must
contain commitments addressing those
provisions of appendix B which are not
covered by NQA–1, as well as
provisions that supplement or replace
the NQA–1 provisions where the
appendix B requirement is more
stringent.
Finally, the NRC is removing the
reference in § 50.55a(b)(2)(x) to versions
of NQA–1 older than the 1994 Edition
because the NRC did not receive any
adverse comments from any applicant
or licensee regarding concerns about
removing versions of NQA–1 older than
the 1994 Edition from the regulation.
The NRC received only one comment
regarding NQA–1. The comment
expressed support for incorporation by
reference of NQA–1 and did not
respond to the NRC’s request for
comment regarding the removal of
references to older versions of NQA–1.
10 CFR 50.55a(b)(2)(xii) Section XI
Condition: Underwater Welding
The NRC is revising § 50.55a(b)(2)(xii)
to allow underwater welding on
irradiated materials in accordance with
IWA–4660, ‘‘Underwater Welding,’’ of
Section XI, 1997 Addenda through the
latest edition and addenda incorporated
by reference in § 50.55a(a)(1)(ii). The
conditions for which underwater
welding would be permitted without
prior NRC approval are based on
technical factors, such as neutron
fluence and, for certain material classes,
helium concentration.
The existing condition in
§ 50.55a(b)(2)(xii) does not allow
underwater welding on irradiated
materials by prohibiting the use of
IWA–4660, ‘‘Underwater Welding,’’ of
Section XI, 1997 Addenda through the
latest edition and addenda incorporated
by reference in § 50.55a(a)(1)(ii) on
materials that are irradiated; however,
there are two problems with the
restriction in § 50.55a(b)(2)(xii). First,
the neutron fluence threshold above
which a material is considered to be
irradiated is not defined in
§ 50.55a(b)(2)(xii). Second, studies such
as those documented in Boiling Water
Reactor Vessel and Internals Project
(BWRVIP) Report 1003020 (BWRVIP–
97) have shown that reactor internals
can tolerate some neutron irradiation
without suffering damage to weldability,
as long as the helium concentration in
the material does not exceed a certain
threshold. The NRC completed its
Safety Evaluation of BWRVIP–97 in May
2008 and concluded that
implementation of the guidelines in the
BWRVIP–97 report, with some
modifications as documented in the
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NRC Safety Evaluation dated June 30,
2008, will provide an acceptable
technical basis for the design of weld
repairs based on the helium content of
irradiated reactor vessel internals. The
current version of § 50.55a(b)(2)(xii)
does not define a threshold of helium
concentration below which the material
is considered to be weldable.
The most recent editions of the ASME
BPV Code state in Article IWA–4660
that underwater welding may not be
performed on irradiated materials other
than P-No. 8 materials containing less
than 0.1 atomic parts per million (appm)
measured or calculated helium content
generated through irradiation. Some
editions and addenda of the ASME BPV
Code prior to 2010 state in Article IWA–
4660 that underwater welding may only
be performed in applications not
predicted to exceed a thermal neutron
fluence of 1 × 1017 n/cm2. Other
editions and addenda of the ASME BPV
Code prior to 2010 do not restrict the
underwater welding of irradiated
materials. Therefore, there is
inconsistent treatment among the
various editions and addenda of the
ASME BPV Code on the underwater
welding of irradiated materials.
Current ASME BPV Code and Code
Case requirements for welding on
irradiated materials, other than the
underwater welding requirements
specified in IWA–4660, are inconsistent.
Thresholds for weldability may be
stated in terms of fast neutron fluence,
thermal neutron fluence, or helium
concentration. In some cases, thresholds
are not defined and the Code or Code
Case simply states that consideration
must be given to irradiation effects
when welding. The NRC believes that
thresholds for welding on irradiated
materials should be based on the current
understanding of irradiation damage, as
supported by technical studies (such as
BWRVIP–97) which have been
evaluated by the NRC. In addition, the
NRC believes that these thresholds
should be consistently applied for all
Code and Code Case applications.
During the public comment period for
this rulemaking, a representative of
ASME recommended that
§ 50.55a(b)(2)(xii) be revised such that it
applies only to those editions and
addenda earlier than the 2010 Edition.
The effect of such a revision would be
to allow welding on P-No. 8 materials
containing less than 0.1 appm measured
or calculated helium content generated
through irradiation. However, this
proposed revision would not be
consistent with other ASME BPV Code
or Code Case requirements for welding
on irradiated materials, and this
proposed revision does not address
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standards for welding on material
classes other than P-No. 8. Instead the
NRC is adopting conditions that would
apply to all materials and which can be
consistently applied for all Code and
Code Case applications. The first
condition, § 50.55a(b)(2)(xii)(A), is
based on fast neutron fluence and
applies to ferritic materials. The second
condition, § 50.55a(b)(2)(xii)(B), is based
on helium content and/or thermal
fluence and applies to austenitic
materials. For P-No. 8 austenitic
materials, the evaluation of BWRVIP–97
supports a weldability threshold based
on helium content and thermal fluence.
For austenitic materials other than P-No.
8, there are insufficient data to support
a weldability threshold based on helium
content, and, therefore, the NRC is
adopting a weldability threshold based
on thermal fluence only.
The conditions for which underwater
welding are permitted, as stated in the
revision of § 50.55a(b)(2)(xii), were
determined, in part, based on technical
discussions in a Category 2 public
meeting with industry representatives
held on January 19, 2016. The NRC later
presented the new conditions at a
public meeting held on March 2, 2016.
There were no comments on this change
from the attendees at the March 2, 2016,
public meeting. Summaries of the
January 19 and March 2, 2016, public
meetings are available in ADAMS under
Accession Nos. ML16050A383 and
ML16069A408, respectively.
10 CFR 50.55a(b)(2)(xviii)(D) NDE
Personnel Certification: Fourth
Provision
The NRC is adding
§ 50.55a(b)(2)(xviii)(D) to prohibit
applicants and licensees from using the
ultrasonic examination nondestructive
examination (NDE) personnel
certification requirements in Section XI,
Appendix VII and Subarticle VIII–2200
of the 2011 Addenda and 2013 Edition
of the ASME BPV Code. Paragraph
(b)(2)(xviii) currently includes
conditions on the certification of NDE
personnel. In addition, the new
paragraph will require applicants and
licensees to use the 2010 Edition, Table
VII–4110–1 training hour requirements
for Levels I, II, and III ultrasonic
examination personnel, and the 2010
Edition, Subarticle VIII–2200 of
Appendix VIII prerequisites for
personnel requirements. In the 2011
Addenda and 2013 Edition, the ASME
BPV Code added an accelerated
Appendix VII training process for
certification of ultrasonic examination
personnel based on training and prior
experience, and separated the Appendix
VII training requirements from the
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Appendix VIII qualification
requirements. These new ASME BPV
Code provisions will provide personnel
in training with less experience and
exposure to representative flaws in
representative materials and
configurations common to operating
nuclear power plants, and they would
permit personnel with prior non-nuclear
ultrasonic examination experience to
qualify for examinations in nuclear
power plants without exposure to the
variety of defects, examination
conditions, components, and
regulations common to operating
nuclear power plants.
The impact of reduced training and
nuclear power plant familiarization is
unknown. The ASME BPV Code
supplants training hours and field
experience without a technical basis,
minimum defined training criteria,
process details, or standardization. For
these reasons, the NRC is prohibiting
the use of Appendix VII and Subarticle
VIII–2200 of the 2011 Addenda and
2013 Edition. The NRC is requiring
applicants and licensees using the 2011
Addenda and 2013 Edition to use the
prerequisites for ultrasonic examination
personnel certifications in Table VII–
4110–1 and Subarticle VIII–2200,
Appendix VIII in the 2010 Edition.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB–
2500–1 Examination Requirements:
First Provision
The NRC is revising
§ 50.55a(b)(2)(xxi)(A) to modify the
standard for visual magnification
resolution sensitivity and contrast for
visual examinations performed on
Examination Category B–D components
instead of ultrasonic examinations,
making the rule conform with ASME
BPV Code, Section XI requirements for
VT–1 examinations. The character
recognition rules are used in ASME BPV
Code, Section XI, Table IWA–2211–1 for
VT–1 tests, and are the standard tests
used for resolution and contrast checks
of the VT–1 equipment. This revision
essentially removed a requirement that
was an addition to ASME BPV Code that
required 1-mil wires to be used in
licensees’ Sensitivity, Resolution, and
Contrast Standard targets. In 2004, the
NRC published NUREG/CR–6860, ‘‘An
Assessment of Visual Testing,’’ showing
that a linear target, such as a wire, is not
an effective method for testing the
resolution of a video camera system. In
addition, Boiling Water Reactor Vessel
and Internals Project Report 105696
(BWRVIP–03) was changed to eliminate
a 1⁄2 mil wire from the Sensitivity,
Resolution, and Contrast Standards due
to similar concerns.
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Simple line detection can be a poor
performance standard, allowing
detection of a highly blurred image.
This does not emulate sharpness quality
recognition for evaluation of weld
discontinuities. The 750 mm (30 mil)
and the even smaller 25 mm (1 mil)
widths should not be used as
performance standards because they do
not determine image sharpness. This
technique only measures the ‘‘visible
minimum’’ for long linear indications,
and does not measure a system’s
resolution or recognition limits. If the
wire, or printed line, has a strong
enough contrast against the background,
then a linear feature well below the
resolution of a system can be detected.
10 CFR 50.55a(b)(2)(xxiii) Section XI
Condition: Evaluation of Thermally Cut
Surfaces
The NRC is revising
§ 50.55a(b)(2)(xxiii) to clarify that this
condition, prohibiting the ASME BPV
Code provisions allowing elimination of
mechanical processing of thermally cut
surfaces under certain circumstances,
only applies to the 2001 Edition through
the 2009 Addenda.
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10 CFR 50.55a(b)(2)(xxx) Section XI
Condition: Steam Generator Preservice
Examinations
In the proposed rule, the NRC
proposed adding § 50.55a(b)(2)(xxx),
with a condition regarding steam
generator preservice examinations. The
NRC received requests for clarification
of the proposed condition, including
elaboration on the kind of preservice
examination that should be performed.
The NRC agrees with the need for this
clarification; however, during the
development of the final rule, the NRC
determined that additional time was
needed to evaluate this proposed
condition. Therefore, to ensure that this
rulemaking is concluded as timely as
possible, the NRC is not including this
condition in this final rule and will
address the need for a condition in a
future rulemaking. The NRC has
concluded that omitting this condition
does not present a health or safety
concern because licensees are currently
performing appropriate steam generator
preservice inspections under existing
programs.
10 CFR 50.55a(b)(2)(xxxi) Section XI
Condition: Mechanical Clamping
Devices
The NRC is adding
§ 50.55a(b)(2)(xxxi) to require the use of
Nonmandatory Appendix W when using
a mechanical clamping device on an
ASME BPV Code Class piping system.
This condition, in part, clearly prohibits
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the use of mechanical clamping devices
on small item Class 1 piping and
portions of piping systems that form the
containment boundary. This condition
also maintains the previously required
design and testing requirements for the
implementation of mechanical clamping
devices on ASME BPV Code Class
piping systems.
In the 2010 Edition of the ASME BPV
Code, a change was made to include
mechanical clamping devices under the
small items exclusion rules of IWA–
4131. Currently in the 2007 Edition/
2008 Addenda of Section XI under
IWA–4133, ‘‘Mechanical Clamping
Devices Used as Piping Pressure
Boundary,’’ mechanical clamping
devices may be used only if they meet
the requirements of Mandatory
Appendix IX of Section XI of the ASME
BPV Code. Article IX–1000 (c) of
Appendix IX prohibits the use of
mechanical clamping devices on (1)
Class 1 piping and (2) portions of a
piping system that form the
containment boundary.
In the 2010 Edition, IWA–4133 was
modified to allow use of IWA–4131.1(c)
for the installation of mechanical
clamping devices. This change allowed
the use of small items exclusion rules in
the installation of mechanical clamping
devices. Subparagraph IWA–4131.1(c)
was added such that mechanical
clamping devices installed on items
classified as ‘‘small items’’ under IWA–
4131, including Class 1 piping and
portions of a piping system that form
the containment boundary, would be
allowed without a repair/replacement
plan, pressure testing, services of an
Authorized Inspection Agency, and
completion of the NIS–2 form. The NRC,
in accordance with the previously
approved IWA–4133 of the 2007
Edition/2008 Addenda of the ASME
BPV Code, does not believe that the
ASME has provided a sufficient
technical basis to support the use of
mechanical clamping devices on Class 1
piping or portions of a piping system
that form the containment boundary as
a permanent repair. Furthermore, the
NRC finds that the ASME has not
provided any basis for the small item
exemption allowing the installation of
mechanical clamps on these
components. In the 2011 Addenda of
the ASME BPV Code, IWA–4131.1(c)
was relocated to IWA–4131.1(d). To add
clarity to the condition, the NRC has
included statements such that the
implementation of these paragraphs is
now prohibited.
In the 2013 Edition, Mandatory
Appendix IX of Section XI of the ASME
BPV Code was changed to
Nonmandatory Appendix W of Section
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XI of the ASME BPV Code. The NRC
found insufficient basis to make this
change, removing the mandatory
requirements for the use of mechanical
clamping devices on ASME BPV Code
Class piping systems. By taking this
action, the ASME BPV Code now allows
mechanical clamping devices to be
installed in various methods through
interpretations of the ASME BPV Code
that do not maintain the requirements
for design and testing of the formerly
mandatory Appendix IX. Therefore, to
clarify the requirement for the
implementation of mechanical clamps
in ASME BPV Code Class systems, the
NRC requires the use of Appendix W of
Section XI when using mechanical
clamping devices, and prohibits the use
of mechanical clamping devices on
small item Class 1 piping and portions
of a piping system that form the
containment boundary, as would
otherwise be permitted under IWA–
4131.1(c) in the 2010 Edition and IWA–
4131.1(d) in the 2011 Addenda through
2013 Edition.
10 CFR 50.55a(b)(2)(xxxii) Section XI
Condition: Summary Report Submittal
The NRC is adding
§ 50.55a(b)(2)(xxxii) to require licensees
using the 2010 Edition and later
editions and addenda of Section XI to
continue to submit Summary Reports as
required in IWA–6240 of the 2009
Addenda.
Prior to the 2010 Edition, Section XI
required the preservice summary report
to be submitted prior to the date of
placement of the unit into commercial
service, and the inservice summary
report to be submitted within 90
calendar days of the completion of each
refueling outage. In the 2010 Edition,
IWA–6240 was revised to state,
‘‘Summary reports shall be submitted to
the enforcement and regulatory
authorities having jurisdiction at the
plant site, if required by these
authorities.’’ This change in the 2010
Edition could lead to confusion as to
whether or not the summary reports
need to be submitted to the NRC, as well
as the time for submitting the reports, if
they were required. The NRC concludes
that summary reports must continue to
be submitted to the NRC in a timely
manner because they provide valuable
information regarding examinations
performed, conditions noted, corrective
actions taken, and the implementation
status of preservice inspection and ISI
programs. Therefore, the NRC is adding
§ 50.55a(b)(2)(xxxii) to ensure that
preservice and inservice summary
reports will continue to be submitted
within the timeframes currently
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10 CFR 50.55a(b)(2)(xxxiv) Section XI
Condition: Nonmandatory Appendix U
10 CFR 50.55a(b)(2)(xxxiii) Section XI
Condition: Risk-Informed Allowable
Pressure
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established in Section XI editions and
addenda prior to the 2010 Edition.
The NRC is adding
§ 50.55a(b)(2)(xxxiv) to require that two
conditions, (A) and (B), be satisfied
when using Nonmandatory Appendix U
of the 2013 Edition of the ASME BPV
Code, Section XI. In the proposed rule,
the NRC had proposed to exclude
Nonmandatory Appendix U from the
incorporation by reference and therefore
not approve it for use. After considering
public comments, the NRC has
incorporated by reference Appendix U
in this final rule because it integrates
ASME BPV Code Cases N–513–3,
‘‘Evaluation Criteria for Temporary
Acceptance of Flaws in Moderate
Energy Class 2 or 3 Piping Section XI,
Division 1,’’ and N–705, ‘‘Evaluation
Criteria for Temporary Acceptance of
Degradation in Moderate Energy Class 2
or 3 Vessels and Tanks Section XI,
Division 1,’’ into Section XI. The NRC
has approved the use of ASME BPV
Code Cases N–513–3 and N–705 in RG
1.147, which allows licensees to use
these code cases without prior
permission from the NRC.
The first condition on the use of
Appendix U is set forth in
§ 50.55a(b)(2)(xxxiv)(A) of this final rule
and requires that an ASME BPV Code
repair or replacement activity
temporarily deferred under the
provisions of Nonmandatory Appendix
U to the 2013 Edition of the ASME BPV
Code, Section XI, must be performed
during the next scheduled outage. This
condition is consistent with the NRC’s
condition on the use of ASME BPV
Code Case N–513–3 in RG 1.147,
Revision 17. Appendix U defines that
the evaluation period is the operational
time for which the temporary
acceptance criteria are satisfied but not
exceeding 26 months from the initial
discovery of the condition. Original
versions of ASME BPV Code Case N–
513 stated, in part, that certain flaws
may be acceptable without performing a
repair/replacement activity for a limited
time, not to exceed the time to the next
scheduled outage. The NRC staff found
that the acceptance of ASME BPV Code
Case N–513 was based on allowing
continued plant operation with a
monitored and evaluated low safety
significant degraded condition for a
limited time until plant shutdown. By
allowing use of this Appendix, this
option is allowed rather than requiring
an unnecessary plant shutdown to
repair the degradation. However, the
NRC believes once the plant is shut
down, the degraded piping must be
repaired.
The NRC is adding
§ 50.55a(b)(2)(xxxiii) to prohibit the use
of Appendix G, Paragraph G–2216, in
the 2011 Addenda and later editions
and addenda of the ASME BPV Code,
Section XI. The 2011 Addenda of the
ASME BPV Code included, for the first
time, a risk-informed methodology to
compute allowable pressure as a
function of inlet temperature for reactor
heat-up and cool-down at rates not to
exceed 100 degrees F/hr (56 degrees C/
hr). This methodology was developed
based upon probabilistic fracture
mechanics (PFM) evaluations that
investigated the likelihood of reactor
pressure vessel (RPV) failure based on
specific heat-up and cool-down
scenarios.
During the ASME’s consideration of
this change, the NRC staff noted that
additional requirements would need to
be placed on the use of this alternative.
For example, the NRC staff indicated
that it would be important for a licensee
who wishes to utilize such a riskinformed methodology for determining
plant-specific pressure-temperature
limits to ensure that the material
condition of its facility is consistent
with assumptions made in the PFM
evaluations that supported the
development of the methodology. One
aspect of this would be evaluating plantspecific ISI data to determine whether
the facility’s RPV flaw distribution was
consistent with the flaw distribution
assumed in the supporting PFM
evaluations. This consideration is
consistent with a similar requirement
established by the NRC in § 50.61a,
‘‘Alternative Fracture Toughness
Requirements for Protection against
Pressurized Thermal Shock Events.’’
The PFM methodology that supports
§ 50.61a is very similar to that which
was used to support ASME BPV Code,
Section XI, Appendix G, Paragraph G–
2216. These concerns with the
Paragraph G–2216 methodology for
computing allowable pressure as a
function of inlet temperature for reactor
heat-up and cooldown were not
addressed by the ASME. Accordingly,
the NRC is prohibiting the use of
Paragraph G–2216 in Appendix G of the
2010 Edition. The continued use of the
deterministic methodology of Section
XI, Appendix G to generate PressureTemperature (P–T) limits remains
acceptable.
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The second condition on the use of
Appendix U is set forth in
§ 50.55a(b)(2)(xxxiv)(B) of this final rule.
This paragraph requires the use of the
mandatory appendix in ASME BPV
Code Case N–513–3 in lieu of the
appendix referenced in paragraph U–
S1–4.2.1(c) of Appendix U (which was
inadvertently omitted from Appendix
U). The NRC is incorporating by
reference the mandatory appendix in
ASME BPV Code Case N–513–3 in
§ 50.55a(a)(1)(iii)(A) because it is
referenced in § 50.55a(b)(2)(xxxiv)(B).
A proposed condition on Disposition
of Flaws in Class 3 Components, which
was located in § 50.55a(b)(2)(xxxiv) of
the proposed rule, is not included in
this final rule based on public
comments that the error has been
corrected by ASME in published
erratum.
10 CFR 50.55a(b)(2)(xxxv) Section XI
Condition: Use of RTT0 in the KIa and KIc
Equations
The NRC is adding
§ 50.55a(b)(2)(xxxv) to specify that when
licensees use the 2013 Edition of the
ASME BPV Code, Section XI, Appendix
A, Paragraph A–4200, if T0 is available,
then RTT0 may be used in place of
RTNDT for applications using the KIc
equation and the associated KIc curve,
but not for applications using the KIa
equation and the associated KIa curve.
Nonmandatory Appendix A provides
a procedure based on linear elastic
fracture mechanics (LEFM) for
determining the acceptability of flaws
that have been detected during inservice
inspections that exceed the allowable
flaw indication standards of IWB–3500.
Sub-article A–4200 provides a
procedure for determining fracture
toughness of the material used in the
LEFM analysis. The NRC staff’s concern
is related to the proposed insertion
regarding an alternative based on the
use of the Master Curve methodology to
determine the nil-ductility transition
reference temperature RTNDT, which is
an important parameter in determining
the fracture toughness of the material.
Specifically, the insertion proposed to
use the Master Curve reference
temperature RTT0, which is defined as
RTT0 = T0 + 35 °F, where T0 is a
material-specific temperature value
determined in accordance with ASTM
E1921, ‘‘Standard Test Method for
Determination of Reference
Temperature, T0, for Ferritic Steels in
the Transition Range,’’ to index (shift)
the fracture toughness KIc curve, based
on the lower bound of static initiation
critical stress intensity factor, as well as
the KIa curve, based on the lower bound
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of crack arrest critical stress intensity
factor.
While use of RTT0 to index the KIc
curve is acceptable, using RTT0 to index
the KIa curve is questionable. This
concern is based on the data analysis in
‘‘A Physics-Based Model for the Crack
Arrest Toughness of Ferritic Steels,’’
written by NRC staff member Mark Kirk
and published in ‘‘Fatigue and Fracture
Mechanics, 33rd Volume, ASTM STP
1417’’ which indicated that the crack
arrest data does not support using RTT0
as RTNDT to index the KIa curve. This is
also confirmed by industry data
disclosed in a presentation, ‘‘Final
Results from the CARINA Project on
Crack Initiation and Arrest of Irradiated
German RPV Steels for Neutron
Fluences in the Upper Bound,’’ by
AREVA at the 26th Symposium on
Effects of Radiation on Nuclear
Materials (June 12–13, 2013,
Indianapolis, Indiana, USA). The NRC
staff recognized that the proposed
insertion is consistent with ASME BPV
Code Case N–629, ‘‘Use of Fracture
Toughness Test Data to Establish
Reference Temperature for Pressure
Retaining Materials,’’ which was
accepted by the NRC without
conditions. In addition to the current
NRC effort, the appropriate ASME BPV
Code committee is in the process of
correcting this issue in a future revision
of Appendix A of Section XI.
With this condition, users of
Appendix A can avoid using an
erroneous fracture toughness KIa value
in their LEFM analysis for determining
the acceptability of a detected flaw in
applicable components. Therefore, the
NRC is adding a condition which
permits the use of RTT0 in place of
RTNDT in applications using the KIc
equation and the associated KIc curve,
but does not permit the use of RTT0 in
place of RTNDT in applications using the
KIa equation and the associated KIa
curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI
Condition: Fracture Toughness of
Irradiated Materials
The NRC is adding
§ 50.55a(b)(2)(xxxvi) to require licensees
using ASME BPV Code, Section XI,
2013 Edition, Appendix A, Paragraph
A–4400, to obtain NRC approval under
§ 50.55a(z) before using irradiated T0
and the associated RTT0 in establishing
fracture toughness of irradiated
materials.
Sub-article A–4400 provides guidance
for considering irradiation effects on
materials. The NRC staff’s concern is
related to use of RTT0 based on
measured T0 of the irradiated materials.
Specifically, the NRC staff has concerns
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over this sentence in the proposed
insertion: ‘‘Measurement of RTT0 of
unirradiated or irradiated materials as
defined in A–4200(b) is permitted,
including use of the procedures given in
ASTM E1921 to obtain direct
measurement of irradiated T0.’’
Permission of measurement of RTT0 of
irradiated materials, without providing
guidelines regarding how to use the
measured parameter in determining the
fracture toughness of the irradiated
materials, may mislead the users of
Appendix A into adopting methodology
that has not been accepted by the NRC.
With this condition, users of Appendix
A can avoid inappropriately using a
fracture toughness KIc value based on
the irradiated T0 and the associated
RTT0 in their LEFM analysis for
determining the acceptability of a
detected flaw in applicable components.
10 CFR 50.55a(g) Inservice and
Preservice Inspection Requirements
The NRC is adding new paragraphs
(g)(2)(i), (ii), and (iii) and revising
current paragraphs (g) introductory text,
(g)(2), (g)(3) introductory text, and
(g)(3)(i), (ii), and (v) to distinguish the
requirements for accessibility and
preservice examination from those for
inservice inspection in § 50.55a(g). In
addition, consistent with other
paragraphs of this section, headings are
added to the subordinate paragraphs of
(g) in order to enhance readability of the
regulation. No substantive change to the
requirements are intended by these
revisions.
C. OM Code
10 CFR 50.55a(b)(3) Conditions on
ASME OM Code
The NRC is revising § 50.55a(b)(3) to
clarify that Subsections ISTA, ISTB,
ISTC, ISTD, ISTE, and ISTF; Mandatory
Appendices I, II, III, and V; and
Nonmandatory Appendices A through H
and J through M of the OM Code are
each incorporated by reference into
§ 50.55a. The NRC is also clarifying that
the OM Code Nonmandatory
Appendices incorporated by reference
into § 50.55a are approved for use, but
are not mandated. The Nonmandatory
Appendices may be used by applicants
and licensees of nuclear power plants,
subject to the conditions in
§ 50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition:
Quality Assurance
The NRC is revising § 50.55a(b)(3)(i)
to allow use of the 1994 Edition, 2008
Edition, and the 2009–1a Addenda of
NQA–1, ‘‘Quality Assurance
Requirements for Nuclear Facility
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Applications.’’ The NRC reviewed these
editions and addenda, compared them
to the previously approved versions of
NQA–1, and found that there were no
significant differences.
The NRC is removing the reference in
§ 50.55a(b)(3)(i) to versions of NQA–1
older than the 1994 Edition, inasmuch
as these versions do not appear to be in
use at any nuclear power plant. The
NRC did not receive any adverse
comments from any applicant or
licensee regarding concerns about
removing versions of NQA–1 older than
the 1994 Edition from the regulation.
The NRC received one comment
regarding NQA–1, supporting
incorporation by reference of NQA–1
but not responding to the NRC’s request
for comment regarding the removal of
references to older versions of NQA–1.
Accordingly, the NRC concludes that
removal of NQA–1 versions older than
the 1994 Edition will not have any
adverse effect on licensees, and the final
rule removes these older versions from
§ 50.55a(b)(3)(i).
10 CFR 50.55a(b)(3)(ii) OM Condition:
Motor-Operated Valve (MOV) Testing
The NRC is revising § 50.55a(b)(3)(ii)
to reflect the new Appendix III,
‘‘Preservice and Inservice Testing of
Active Electric Motor Operated Valve
Assemblies in Light-Water Reactor
Power Plants,’’ of the OM Code, 2009
Edition, 2011 Addenda, and 2012
Edition. Appendix III of the OM Code
establishes provisions for periodic
verification of the design-basis
capability of MOVs within the scope of
the IST program. Appendix III of the
OM Code reflects the incorporation of
OM Code Cases OMN–1, ‘‘Alternative
Rules for Preservice and Inservice
Testing of Active Electric MotorOperated Valve Assemblies in LightWater Reactor Power Plants,’’ and
OMN–11, ‘‘Risk-Informed Testing for
Motor-Operated Valves.’’ The NRC is
adding four new conditions on the use
of Mandatory Appendix III in new
§ 50.55a(b)(3)(ii)(A), (B), (C), and (D) to
address periodic verification of MOV
design-basis capability. These new
conditions are discussed in the next
four sections.
10 CFR 50.55a(b)(3)(ii)(A) MOV
Diagnostic Test Interval (First Condition
on Use of Mandatory Appendix III)
In the proposed rule, the NRC
specified in § 50.55a(b)(3)(ii)(A) that
licensees evaluate the adequacy of the
diagnostic test interval for each MOV
and adjust the interval as necessary, but
not later than 5 years or three refueling
outages (whichever is longer) from
initial implementation of OM Code,
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Appendix III. Paragraph III–3310(b) in
Appendix III includes a provision
stating that if insufficient data exist to
determine the IST interval, then MOV
inservice testing shall be conducted
every two refueling outages or 3 years
(whichever is longer) until sufficient
data exist, from an applicable MOV or
MOV group, to justify a longer IST
interval. As discussed in a final rule
published September 22, 1999 (64 FR
51386), with respect to the use of OM
Code Case OMN–1, the NRC considers
it appropriate to include a modification
requiring licensees to evaluate the
information obtained for each MOV,
during the first 5 years or three refueling
outages (whichever is longer) of the use
of Appendix III to validate assumptions
made in justifying a longer test interval.
In response to public comments, the
NRC revised § 50.55a(b)(3)(ii)(A) to
clarify its intent for licensees to evaluate
the test interval within 5 years or three
refueling outages (whichever is longer)
following implementation of Appendix
III to the OM Code, rather than implying
that every MOV must be tested within
5 years or three refueling outages of the
initial implementation of Appendix III.
For example, the condition allows
grouping of MOVs to share test
information in the evaluation of the
MOV periodic verification intervals
within 5 years or three refueling outages
(whichever is longer) of the
implementation of OM Code, Appendix
III. Therefore, § 50.55a(b)(3)(ii)(A) of this
final rule states that licensees shall
evaluate the adequacy of the diagnostic
test intervals established for MOVs
within the scope of OM Code,
Mandatory Appendix III, not later than
5 years or three refueling outages
(whichever is longer) from initial
implementation of OM Code, Appendix
III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing
Impact on Risk (Second Condition on
Use of Mandatory Appendix III)
The NRC is adding
§ 50.55a(b)(3)(ii)(B) to require that when
using Mandatory Appendix III, licensees
ensure that the potential increase in
core damage frequency (CDF) and large
early release frequency (LERF)
associated with the extension is
acceptably small when extending
exercise test intervals for high risk
MOVs beyond a quarterly frequency. As
discussed in a final rule published
September 22, 1999 (64 FR 51386), with
respect to the use of OM Code Case
OMN–1, the NRC considers it important
for licensees to have sufficient
information from the specific MOV, or
similar MOVs, to demonstrate that
exercising on a refueling outage
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frequency does not significantly affect
component performance. The
information may be obtained by
grouping similar MOVs and establishing
periodic exercising intervals of MOVs in
the group over the refueling interval.
Section 50.55a(b)(3)(ii)(B) requires
that the increase in the overall plant
CDF and LERF resulting from the
extension be acceptably small. As
presented in RG 1.174, ‘‘An Approach
for Using Probabilistic Risk Assessment
[PRA] in Risk-Informed Decisions on
Plant-Specific Changes to the Licensing
Basis,’’ the NRC considers acceptably
small changes to be relative and to
depend on the current plant CDF and
LERF. For plants with total baseline
CDF of 10¥4 per year or less, acceptably
small means CDF increases of up to
10¥5 per year; and for plants with total
baseline CDF greater than 10¥4 per year,
acceptably small means CDF increases
of up to 10¥6 per year. For plants with
total baseline LERF of 10¥5 per year or
less, acceptably small LERF increases
are considered to be up to 10¥6 per
year; and for plants with total baseline
LERF greater than 10¥5 per year,
acceptably small LERF increases are
considered to be up to 10¥7 per year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk
Categorization (Third Condition on Use
of Mandatory Appendix III)
The NRC is adding
§ 50.55a(b)(3)(ii)(C) to require, when
applying Mandatory Appendix III, that
licensees categorize MOVs according to
their safety significance using the
methodology described in OM Code
Case OMN–3, ‘‘Requirements for Safety
Significance Categorization of
Components Using Risk Insights for
Inservice Testing of LWR Power Plants,’’
subject to the conditions discussed in
RG 1.192, or using an MOV risk ranking
methodology accepted by the NRC on a
plant-specific or industry-wide basis in
accordance with the conditions in the
applicable safety evaluation. Paragraph
III–3720 in Appendix III to the OM Code
states that when applying risk insights,
each MOV shall be evaluated and
categorized using a documented risk
ranking methodology. Further,
Appendix III only addresses risk
ranking methodologies that include two
risk categories. In light of the potential
extension of quarterly test intervals for
high risk MOVs and the relaxation of
IST activities for low risk MOVs based
on risk insights, the NRC has
determined that the rule should specify
that plant-specific or industry-wide risk
ranking methodologies must have been
accepted by the NRC through RG 1.192
(which accepts OM Code Case OMN–3
with the specified conditions) or the
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issuance of safety evaluations. As noted
in the response to public comments, the
intent of this condition is to indicate
that when applying Appendix III to the
OM Code, licensees may use either a
two-risk category approach (high or low)
or a three-risk category approach (high,
medium, and low), provided the risk
ranking method has been accepted by
the NRC.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke
Time (Fourth Condition on Use of
Mandatory Appendix III)
The NRC is adding
§ 50.55a(b)(3)(ii)(D) to require that when
a licensee applies Paragraph III–3600,
‘‘MOV Exercising Requirements,’’ of
Appendix III to the OM Code, the
licensee verify that the stroke time of
the MOV satisfies the assumptions in
the plant’s safety analyses. Previous
editions and addenda of the OM Code
specified that the licensee must perform
quarterly MOV stroke time
measurements that could be used to
verify that the MOV stroke time satisfies
the assumptions in the safety analyses
consistent with plant TS. The need for
verification of the MOV stroke time
during periodic exercising is consistent
with the NRC’s lessons learned from the
implementation of OM Code Case
OMN–1. However, Paragraph III–3600 of
Appendix III of the versions of the OM
Code that will be incorporated by
reference in this rulemaking no longer
require the verification of MOV stroke
time during periodic exercising. For this
reason, the NRC is adopting this new
condition, which will effectively retain
the need to verify that the MOV stroke
time during periodic exercising satisfies
the assumptions in the plant’s safety
analyses.
Based on the discussion during the
public webinar on March 2, 2016, the
NRC revised the condition to clarify that
it applies to MOVs referenced in the
plant TS. In particular, the NRC revised
the condition to indicate that when a
licensee applies Paragraph III–3600 of
Appendix III to the OM Code, the
licensee shall verify that the stroke time
of MOVs specified in plant technical
specifications satisfies the assumptions
in the plant’s safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition:
New Reactors
The NRC is adding § 50.55a(b)(3)(iii)
to apply specific conditions for IST
programs applicable to licensees of new
nuclear power plants in addition to the
provisions of the OM Code as
incorporated by reference with
conditions in § 50.55a. Licensees of
‘‘new reactors’’ are, as identified in the
paragraph: (1) Holders of operating
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licenses for nuclear power reactors that
received construction permits under
this part on or after the date 12 months
after August 17, 2017, and (2) holders of
combined licenses (COLs) issued under
10 CFR part 52, whose initial fuel
loading occurs on or after the date 12
months after August 17, 2017. This
implementation schedule for new
reactors is consistent with the NRC
regulations governing inservice testing
in § 50.55a(f)(4)(i).
Commission Papers SECY–90–016,
‘‘Evolutionary Light Water Reactor
(LWR) Certification Issues and Their
Relationship to Current Regulatory
Requirements;’’ SECY–93–087, ‘‘Policy,
Technical, and Licensing Issues
Pertaining to Evolutionary and
Advanced Light-Water Reactor (ALWR)
Designs;’’ SECY–94–084, ‘‘Policy and
Technical Issues Associated with the
Regulatory Treatment of Non-Safety
Systems (RTNSS) in Passive Plant
Designs;’’ and SECY–95–132, ‘‘Policy
and Technical Issues Associated with
the Regulatory Treatment of Non-Safety
Systems (RTNSS) in Passive Plant
Designs (SECY–94–084),’’ discuss IST
programs for new reactors licensed
under 10 CFR part 52.
In recognition of new reactor designs
and lessons learned from nuclear power
plant operating experience, the ASME is
updating the OM Code to incorporate
improved IST provisions for
components used in nuclear power
plants that were issued (or will be
issued) construction permits, or COLs,
on or following January 1, 2000 (defined
in the OM Code as post-2000 plants).
The first phase of the ASME effort
incorporated IST provisions that specify
full flow pump testing and other
clarifications for post-2000 plants in the
OM Code beginning with the 2011
Addenda. The second phase of the
ASME effort incorporated preservice
and inservice inspection and
surveillance provisions for pyrotechnicactuated (squib) valves in the 2012
Edition of the OM Code. The ASME is
considering further modifications to the
OM Code to address additional lessons
learned from valve operating experience
and new reactor issues. As described in
the following paragraphs,
§ 50.55a(b)(3)(iii) will include four
specific conditions which are discussed
in the following paragraphs.
10 CFR 50.55a(b)(3)(iii)(A) PowerOperated Valves
The NRC is adding
§ 50.55a(b)(3)(iii)(A) to require that
licensees within the scope of
§ 50.55a(b)(3)(iii) periodically verify the
capability of power-operated valves
(POVs) to perform their design-basis
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safety functions. While Appendix III to
the OM Code addresses this requirement
for MOVs with the conditions specified
in § 50.55a, applicable applicants and
licensees will need to develop programs
to periodically verify the design-basis
capability of other POVs. The NRC’s
Regulatory Issue Summary 2000–03,
‘‘Resolution of Generic Issue 158:
Performance of Safety-Related PowerOperated Valves Under Design Basis
Conditions,’’ provides attributes for a
successful long-term periodic
verification program for POVs by
incorporating lessons learned from
MOV performance at operating nuclear
power plants and research programs.
Implementation of Appendix III to the
OM Code as accepted in
§ 50.55a(b)(3)(ii) satisfies
§ 50.55a(b)(3)(iii)(A) for MOVs.
Section 50.55a(b)(3)(iii)(A) is
consistent with the Commission policy
for new reactors summarized in an NRC
Staff Memorandum, ‘‘Consolidation of
SECY–94–084 and SECY–95–132,’’
dated July 24, 1995, that (a) the design
capability of safety-related POVs should
be demonstrated by a qualification test
prior to installation; (b) prior to initial
startup, POV capability under designbasis differential pressure and flow
should be verified by a pre-operational
test; and (c) during the operational
phase, POV capability under designbasis differential pressure and flow
should be verified periodically through
a program similar to that developed for
MOVs in Generic Letter 89–10, ‘‘SafetyRelated Motor-Operated Valve Testing
and Surveillance,’’ dated June 28, 1989.2
The condition in § 50.55a(b)(3)(iii)(A)
specifies with the same level of detail as
the condition in § 50.55a(b)(3)(ii) that
nuclear power plant licensees must
establish a program to ensure the
continued capability of MOVs in
performing their design-basis safety
functions. When establishing the MOV
periodic verification condition, the NRC
provided guidance in the final rule
published September 22, 1999 (64 FR
51370), for licensees to develop
acceptable programs that would satisfy
the MOV periodic verification
condition. Similarly, the NRC staff is
providing guidance herein for new
reactor applicants and licensees to
develop acceptable programs to
periodically verify the capability of
2 The NRC issued seven supplements to provide
guidance for the implementation of the MOV testing
program requested in Generic Letter 89–10. The
supplements to Generic Letter 89–10 did not
modify the substance of the MOV testing program
requested in Generic Letter 89–10 to provide
reasonable assurance in the capability of safetyrelated MOVs to perform their design-basis
functions.
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POVs to perform their design-basis
safety functions.
In NUREG–2124, ‘‘Final Safety
Evaluation Report [FSER] Related to the
Combined Licenses for Vogtle Electric
Generating Plant, Units 3 and 4,’’ the
NRC staff found the provisions
established by the COL applicant for
Vogtle Units 3 and 4 in its Final Safety
Analysis Report (FSAR), Revision 5,
Section 3.9.6.2.2, ‘‘Valve Testing,’’ to
periodically verify the capability of
POVs (such as air-operated valves
(AOVs), solenoid-operated valves
(SOVs), and hydraulic-operated valves
(HOVs)) to perform their design-basis
safety functions to be acceptable. In
particular, the Vogtle Units 3 and 4
FSAR specifies that:
Power-operated valves other than active
MOVs are exercised quarterly in accordance
with OM ISTC, unless justification is
provided in the inservice testing program for
testing these valves at other than Code
mandated frequencies. Although the design
basis capability of power-operated valves is
verified as part of the design and
qualification process, power-operated valves
that perform an active safety function are
tested again after installation in the plant, as
required, to ensure valve setup is acceptable
to perform their required functions,
consistent with valve qualification. These
tests, which are typically performed under
static (no flow or pressure) conditions, also
document the ‘‘baseline’’ performance of the
valves to support maintenance and trending
programs. During the testing, critical
parameters needed to ensure proper valve
setup are measured. Depending on the valve
and actuator type, these parameters may
include seat load, running torque or thrust,
valve travel, actuator spring rate, bench set
and regulator supply pressure. Uncertainties
associated with performance of these tests
and use of the test results (including those
associated with measurement equipment and
potential degradation mechanisms) are
addressed appropriately. Uncertainties may
be considered in the specification of
acceptable valve setup parameters or in the
interpretation of the test results (or a
combination of both). Uncertainties affecting
both valve function and structural limits are
addressed. Additional testing is performed as
part of the air-operated valve (AOV) program,
which includes the key elements for an AOV
Program as identified in the JOG AOV
program document, Joint Owners Group Air
Operated Valve Program Document, Revision
1, December 13, 2000 (References 203 and
204) [JOG AOV Program Document, Revision
1, December 13, 2000 (ADAMS Accession
No. ML010950310), and NRC comment letter
dated October 8, 1999, to Nuclear Energy
Institute (ADAMS Accession No.
ML020360077)]. The AOV program
incorporates the attributes for a successful
power-operated valve long-term periodic
verification program, as discussed in
Regulatory Issue Summary 2000–03,
Resolution of Generic Safety Issue 158:
Performance of Safety-Related PowerOperated Valves Under Design Basis
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Conditions, by incorporating lessons learned
from previous nuclear power plant
operations and research programs as they
apply to the periodic testing of air- and other
power-operated valves included in the IST
program.
For example, key lessons learned
addressed in the AOV program include:
• Valves are categorized according to their
safety significance and risk ranking.
• Setpoints for AOVs are defined based on
current vendor information or valve
qualification diagnostic testing, such that the
valve is capable of performing its designbasis function(s).
• Periodic static testing is performed, at a
minimum on high risk (high safety
significance) valves, to identify potential
degradation, unless those valves are
periodically cycled during normal plant
operation, under conditions that meet or
exceed the worst case operating conditions
within the licensing basis of the plant for the
valve, which would provide adequate
periodic demonstration of AOV capability. If
required based on valve qualification or
operating experience, periodic dynamic
testing is performed to re-verify the
capability of the valve to perform its required
functions.
• Sufficient diagnostics are used to collect
relevant data (e.g., valve stem thrust and
torque, fluid pressure and temperature,
stroke time, operating and/or control air
pressure, etc.) to verify the valve meets the
functional requirements of the qualification
specification.
• Test frequency is specified, and is
evaluated each refueling outage based on
data trends as a result of testing. Frequency
for periodic testing is in accordance with
References 203 and 204, with a minimum of
5 years (or 3 refueling cycles) of data
collected and evaluated before extending test
intervals.
• Post-maintenance procedures include
appropriate instructions and criteria to
ensure baseline testing is re-performed as
necessary when maintenance on the valve,
repair or replacement, have the potential to
affect valve functional performance.
• Guidance is included to address lessons
learned from other valve programs specific to
the AOV program.
• Documentation from AOV testing,
including maintenance records and records
from the corrective action program are
retained and periodically evaluated as a part
of the AOV program.
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*
*
*
*
*
The attributes of the AOV testing program
described above, to the extent that they apply
to and can be implemented on other safetyrelated power-operated valves, such as
electro-hydraulic operated valves, are
applied to those other power-operated
valves.’’ (Vogtle Electric Generating Plant,
Units 3 and 4, Updated Final Safety Analysis
Report (UFSAR), Section 3.9.6, ‘‘Inservice
Testing of Pumps and Valves’’)
Applicable applicants and licensees
may follow the method described in the
Vogtle Units 3 and 4 FSAR in satisfying
§ 50.55a(b)(3)(iii)(A), or may establish a
different method, subject to evaluation
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by the NRC during the licensing process
or inspections.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
The NRC is adding
§ 50.55a(b)(3)(iii)(B) to require that
licensees within the scope of
§ 50.55a(b)(3)(iii) perform bi-directional
testing of check valves within the IST
program where practicable. Nuclear
power plant operating experience has
revealed that testing check valves in
only the flow direction can result in
significant degradation, such as a
missing valve disc, not being identified
by the test. Nonmandatory Appendix M,
‘‘Design Guidance for Nuclear Power
Plant Systems and Component Testing,’’
to OM Code, 2011 Addenda and 2012
Edition, includes guidance for the
design of new reactors to enable bidirectional testing of check valves. New
reactor designs will provide the
capability for licensees of new nuclear
power plants to perform bi-directional
testing of check valves within the IST
program. Bi-directional testing of check
valves in new reactors, as required by
§ 50.55a(b)(3)(iii)(B), could be
accomplished by valve-specific testing
or condition monitoring activities in
accordance with Appendix II to the OM
Code as accepted in § 50.55a. The NRC
is specifying this provision for bidirectional testing of check valves for
new reactors in § 50.55a(b)(3)(iii)(B) to
emphasize that new reactors should
include the capability for bi-directional
testing of check valves as part of their
initial design.
10 CFR 50.55a(b)(3)(iii)(C) FlowInduced Vibration
In the proposed rule, the NRC
proposed adding § 50.55a(b)(3)(iii)(C) to
require that licensees subject to
§ 50.55a(b)(3)(iii) monitor flow-induced
vibration (FIV) from hydrodynamic
loads and acoustic resonance during
preservice testing and inservice testing
to identify potential adverse flow effects
that might impact components within
the scope of the IST program.
Nuclear power plant operating
experience has revealed the potential for
adverse flow effects from vibration
caused by hydrodynamic loads and
acoustic resonance on components in
the reactor coolant, steam, and
feedwater systems. Therefore, the
licensee will be required to address
potential adverse flow effects on safetyrelated pumps, valves, and dynamic
restraints within the IST program in the
reactor coolant, steam, and feedwater
systems from hydraulic loading and
acoustic resonance during plant
operation. In response to public
comments, the NRC revised
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§ 50.55a(b)(3)(iii)(C) to clarify its intent
that FIV monitoring of components may
be conducted during preservice testing
or inservice testing. This requirement
will confirm that piping, components,
restraints, and supports have been
designed and installed to withstand the
dynamic effects of steady-state FIV and
anticipated operational transient
conditions. As part of preservice testing
activities, the initial test program may
be used to verify that safety-related
piping and components are properly
installed and supported such that
vibrations caused by steady-state or
dynamic effects do not result in
excessive stress or fatigue in safetyrelated plant systems.
In the Vogtle Units 3 and 4 FSER, the
NRC staff found the provisions
established by the COL applicant for
Vogtle Units 3 and 4 in its FSAR,
Revision 5, Section 3.9, ‘‘Mechanical
Systems and Components,’’ Section
14.2.9, ‘‘Preoperational Test
Descriptions,’’ and Section 14.2.10,
‘‘Startup Test Procedures,’’ with
incorporation by reference of
corresponding sections of the AP1000
Design Control Document (DCD), to
monitor FIV from hydrodynamic loads
and acoustic resonance during
preservice testing or inservice testing to
be acceptable. In particular, the NRC
staff stated in the Vogtle Units 3 and 4
FSER:
AP1000 DCD Tier 2, Section 3.9.2,
‘‘Dynamic Testing and Analysis,’’ describes
tests to confirm that piping, components,
restraints, and supports have been designed
to withstand the dynamic effects of steadystate FIV and anticipated operational
transient conditions. Section 14.2.9.1.7,
‘‘Expansion, Vibration and Dynamic Effects
Testing,’’ in AP1000 DCD Tier 2, Chapter 14,
‘‘Initial Test Program,’’ states that the
purpose of the expansion, vibration and
dynamic effects testing is to verify that
safety-related, high energy piping and
components are properly installed and
supported such that, in addition to other
factors, vibrations caused by steady-state or
dynamic effects do not result in excessive
stress or fatigue to safety-related plant
systems. Nuclear power plant operating
experience has revealed the potential for
adverse flow effects from vibration caused by
hydrodynamic loads and acoustic resonance
on reactor coolant, steam, and feedwater
systems. . . . In its response, SNC [Vogtle
Units 3 and 4 COL applicant] stated that it
intended to use the overall Initial Test
Program to demonstrate that the plant has
been constructed as designed and the
systems perform consistent with design
requirements. SNC referenced the provisions
in the AP1000 DCD for vibration monitoring
and testing to be implemented at VEGP. For
example, the applicant notes that AP1000
DCD Tier 2, Section 3.9.2.1, ‘‘Piping
Vibration, Thermal Expansion and Dynamic
Effects,’’ specifies that the preoperational test
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program for ASME BPV Code, Section III,
Class 1, 2, and 3 piping systems simulates
actual operating modes to demonstrate that
components comprising these systems meet
functional design requirements and that
piping vibrations are within acceptable
levels. SNC indicates that the planned
vibration testing program described in
AP1000 DCD Tier 2, Sections 14.2.9 and
14.2.10, with the preservice and IST
programs described in AP1000 DCD Tier 2,
Sections 3.9.3.4.4 and 3.9.6, will confirm
component installation in accordance with
design requirements, and address the effects
of steady-state (flow-induced) and transient
vibration to ensure the operability of valves
and dynamic restraints in the IST Program.
The NRC staff considers the response by SNC
clarifies its application of the provisions in
the AP1000 DCD to ensure that potential
adverse flow effects will be addressed at
VEGP. Therefore, the NRC staff considers
Standard Content Open Item 3.9–5 to be
resolved for the VEGP COL application.’’
(NUREG–2124, ‘‘Final Safety Evaluation
Report Related to the Combined Licenses for
Vogtle Electric Generating Plant, Units 3 and
4,’’ Section 3.9.6, ‘‘Inservice Testing of
Pumps and Valves (Related to RG 1.206,
Section C.III.1, Chapter 3, C.I.3.9.6,
‘Functional Design, Qualification, and
Inservice Testing Programs for Pumps,
Valves, and Dynamic Restraints’)’’).
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As clarified in the final rule in
response to public comments, a licensee
may monitor components for adverse
FIV effects during preservice testing or
IST activities.
Applicable applicants and licensees
may either apply the methods described
in the Vogtle Units 3 and 4 FSAR in
satisfying § 50.55a(b)(3)(iii)(C) or
develop their own plant-specific
methods to satisfy § 50.55a(b)(3)(iii)(C)
for NRC review during the licensing
process.
10 CFR 50.55a(b)(3)(iii)(D) High-Risk
Non-Safety Systems
The NRC is adding
§ 50.55a(b)(3)(iii)(D) to require that
licensees within the scope of
§ 50.55a(b)(3)(iii) establish a program to
assess the operational readiness of
pumps, valves, and dynamic restraints
within the scope of the Regulatory
Treatment of Non-Safety Systems
(RTNSS) for applicable reactor designs.
As of the time of this final rule, these
are designs which have been certified in
a design certification rule under 10 CFR
part 52. In SECY–94–084 and SECY–95–
132, the Commission discusses RTNSS
policy and technical issues associated
with passive plant designs. Some new
nuclear power plants have advanced
light-water reactor (ALWR) designs that
use passive safety systems that rely on
natural forces, such as density
differences, gravity, and stored energy to
supply safety injection water and to
provide reactor core and containment
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cooling. Active systems in passive
ALWR designs are categorized as nonsafety systems with limited exceptions.
Active systems in passive ALWR
designs provide the first line of defense
to reduce challenges to the passive
systems in the event of a transient at the
nuclear power plant. Active systems
that provide a defense-in-depth function
in passive ALWR designs need not meet
all of the acceptance criteria for safetyrelated systems. However, there should
be a high level of confidence that these
active systems will be available and
reliable when needed. The combined
activities to provide confidence in the
capability of these active systems in
passive ALWR designs to perform their
functions important to safety are
referred to as the RTNSS program. In the
NRC Staff Memorandum,
‘‘Consolidation of SECY–94–084 and
SECY–95–132,’’ dated July 24, 1995, the
NRC staff provided a consolidated list of
the approved policy and technical
positions associated with RTNSS
equipment in passive plant designs
discussed in SECY–94–084 and SECY–
95–132. This new paragraph specifies
the need for licensees to assess the
operational readiness of RTNSS pumps,
valves, and dynamic restraints.
The July 24, 1995, staff memorandum
summarizes the Commission policy
positions related to inservice testing of
RTNSS pumps and valves as follows:
The staff also concluded that additional
inservice testing requirements may be
necessary for certain pumps and valves in
passive plant designs. The unique passive
plant design relies significantly on passive
safety systems, but also depends on nonsafety systems (which are traditionally safetyrelated systems in current light-water
reactors) to prevent challenges to passive
systems. Therefore, the reliable performance
of individual components is a very
significant factor in enhancing the safety of
passive plant design. The staff recommends
that the following provisions be applied to
passive ALWR plants to ensure reliable
component performance.
1. Important non-safety-related
components are not required to meet criteria
similar to safety-grade criteria. However, the
non-safety-related piping systems with
functions that have been identified as being
important by the RTNSS process should be
designed to accommodate testing of pumps
and valves to assure that the components
meet their intended functions. Specific
positions on the inservice testing
requirements for those components will be
determined as a part of the staff’s review of
plant-specific implementation of the
regulatory treatment of non-safety systems for
passive reactor designs.
2. . . . The vendors for advanced passive
reactors, for which the final designs are not
complete, have sufficient time to include
provisions in their piping system designs to
allow testing at power. Quarterly testing is
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32949
the base testing frequency in the Code and
the original intent of the Code. Furthermore,
the COL holder may need to test more
frequently than during cold shutdowns or at
every refueling outage to ensure that the
reliable performance of components is
commensurate with the importance of the
safety functions to be performed and with
system reliability goals. Therefore, to the
extent practicable, the passive ALWR piping
systems should be designed to accommodate
the applicable Code requirements for the
quarterly testing of valves. However, design
configuration changes to accommodate Coderequired quarterly testing should be done
only if the benefits of the test outweigh the
potential risk.
3. The passive system designs should
incorporate provisions (1) to permit all
critical check valves to be tested for
performance, to the extent practicable, in
both forward- and reverse-flow directions,
although the demonstration of a non-safety
direction test need not be as rigorous as the
corresponding safety direction test, and (2) to
verify the movement of each check valve’s
obturator during inservice testing by
observing a direct instrumentation indication
of the valve position such as a position
indicator or by using nonintrusive test
methods.
4. . . . Similarly, to the extent practicable,
the design of non-safety-related piping
systems with functions under design-basis
condition that have been identified as being
important by the RTNSS process should
incorporate provisions to periodically test
power-operated valves in the system during
operations to assure that the valves meet
their intended functions under design-basis
conditions.
5. . . . Mispositioning may occur through
actions taken locally (manual or electrical), at
a motor control center, or in the control
room, and includes deliberate changes of
valve position to perform surveillance
testing. The staff will determine if and the
extent to which this concept should be
applied to MOVs in important non-safetyrelated systems when the staff reviews the
implementation of the regulatory treatment of
non-safety systems.’’ (NRC Staff
Memorandum, ‘‘Consolidation of SECY–94–
084 and SECY–95–132,’’ July 24, 1995, pages
26–28).
Consistent with the Commission
policy for RTNSS equipment,
§ 50.55a(b)(3)(iii)(D) specifies that new
reactor licensees shall assess the
operational readiness of pumps, valves,
and dynamic restraints within the
RTNSS scope. This regulatory
requirement will allow licensees
flexibility in developing programs to
assess operational readiness of RTNSS
components that satisfy the Commission
policy. Guidance on the implementation
of the Commission policy for RTNSS
equipment is set forth in NRC
Inspection Procedure 73758, ‘‘Part 52,
Functional Design and Qualification,
and Preservice and Inservice Testing
Programs for Pumps, Valves and
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Dynamic Restraints,’’ dated April 19,
2013.
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10 CFR 50.55a(b)(3)(iv) OM Condition:
Check Valves (Appendix II)
The NRC is revising § 50.55a(b)(3)(iv)
to address Appendix II, ‘‘Check Valve
Condition Monitoring Program,’’
provided in the 2003 Addenda through
the 2012 Edition of the OM Code. In the
proposed rule, the NRC proposed a
condition in § 50.55a(b)(3)(iv) to provide
assurance that the valve or group of
valves is capable of performing its
intended function(s) over the entire
interval. Public comments indicated
that the proposed condition could be
misinterpreted. Therefore, the NRC
revised the proposed condition to
clarify that the implementation of
Appendix II must include periodic
sampling of the check valves over the
maximum interval allowed by
Appendix II for the check valve
condition monitoring program. A new
table was added to the paragraph to
specify the maximum intervals between
check valve condition monitoring
activities when applying interval
extensions.
The conditions currently specified for
the use of Appendix II, 1995 Edition
with the 1996 and 1997 Addenda, and
1998 Edition through the 2002
Addenda, of the OM Code remain
unchanged by this final rule.
10 CFR 50.55a(b)(3)(vii) OM Condition:
Subsection ISTB
The NRC is adding a new condition,
§ 50.55a(b)(3)(vii), to prohibit the use of
Subsection ISTB, ‘‘Inservice Testing of
Pumps in Light-Water Reactor Nuclear
Power Plants,’’ in the 2011 Addenda of
the OM Code. In the 2011 Addenda to
the OM Code, the upper end of the
‘‘Acceptable Range’’ and the ‘‘Required
Action Range’’ for flow and differential
or discharge pressure for comprehensive
pump testing in Subsection ISTB was
raised to higher values. The NRC staff
on the OM Code committee accepted the
proposed increase of the upper end of
the ‘‘Acceptable Range’’ and ‘‘Required
Action Range’’ with the planned
addition of a requirement for a pump
periodic verification test program in the
OM Code. However, the 2011 Addenda
to the OM Code did not include the
requirement for a pump periodic
verification test program. Since then,
the 2012 Edition of the OM Code has
incorporated Mandatory Appendix V,
‘‘Pump Periodic Verification Test
Program,’’ which supports the changes
to the acceptable and required action
ranges for comprehensive pump testing
in Subsection ISTB. Therefore, the new
§ 50.55a(b)(3)(vii) prohibits the use of
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Subsection ISTB in the 2011 Addenda
of the OM Code. Licensees will be
allowed to apply Subsection ISTB with
the revised acceptable and required
action ranges in the 2012 Edition of the
OM Code as incorporated by reference
in § 50.55a.
10 CFR 50.55a(b)(3)(viii) OM Condition:
Subsection ISTE
The NRC is adding § 50.55a(b)(3)(viii)
to specify that licensees who wish to
implement Subsection ISTE, ‘‘RiskInformed Inservice Testing of
Components in Light-Water Reactor
Nuclear Power Plants,’’ of the OM Code,
2009 Edition, 2011 Addenda, and 2012
Edition, must request and obtain NRC
approval in accordance with § 50.55a(z)
to apply Subsection ISTE on a plantspecific basis as a risk-informed
alternative to the applicable IST
requirements in the OM Code.
In the 2009 Edition of the OM Code,
the ASME included new Subsection
ISTE that describes a voluntary riskinformed approach in developing an IST
program for pumps and valves at
nuclear power plants. If a licensee
chooses to implement this risk-informed
IST approach, Subsection ISTE
indicates that all requirements in
Subsection ISTA, ‘‘General
Requirements,’’ Subsection ISTB, and
Subsection ISTC, ‘‘Inservice Testing of
Valves in Light-Water Reactor Nuclear
Power Plants,’’ of the OM Code continue
to apply, except those identified in
Subsection ISTE. The ASME selected
risk-informed guidance from OM Code
Cases OMN–1, OMN–3, OMN–4,
‘‘Requirements for Risk Insights for
Inservice Testing of Check Valves at
LWR Power Plants,’’ OMN–7,
‘‘Alternative Requirements for Pump
Testing,’’ OMN–11, and OMN–12,
‘‘Alternative Requirements for Inservice
Testing Using Risk Insights for
Pneumatically and Hydraulically
Operated Valve Assemblies in LightWater Reactor Power Plants,’’ in
preparing Subsection ISTE of the OM
Code.
During development of Subsection
ISTE, the NRC staff participating on the
OM Code committees indicated that the
conditions specified in RG 1.192 for the
use of the applicable OM Code Cases
need to be considered when evaluating
the acceptability of the implementation
of Subsection ISTE. In addition, the
NRC staff noted that several aspects of
Subsection ISTE will need to be
addressed on a case-by-case basis when
determining the acceptability of its
implementation. Therefore, the new
condition in § 50.55a(b)(3)(viii) requires
that licensees who wish to implement
Subsection ISTE of the OM Code must
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request approval from the NRC to apply
Subsection ISTE on a plant-specific
basis as a risk-informed alternative to
the applicable IST requirements in the
OM Code.
Nuclear power plant applicants for
construction permits under 10 CFR part
50, or combined licenses for
construction and operation under 10
CFR part 52, may describe their
proposed implementation of the riskinformed IST approach specified in
Subsection ISTE of the OM Code for
NRC review in their applications.
10 CFR 50.55a(b)(3)(ix) OM Condition:
Subsection ISTF
The NRC is adding a condition on the
use of Subsection ISTF in
§ 50.55a(b)(3)(ix). First, the condition
states that Subsection ISTF, 2011
Addenda, is prohibited for use. Second,
the condition specifies that licensees
applying Subsection ISTF, ‘‘Inservice
Testing of Pumps in Light-Water Reactor
Nuclear Power Plants—Post-2000
Plants,’’ in the 2012 Edition of the OM
Code shall satisfy the requirements of
Mandatory Appendix V, ‘‘Pump
Periodic Verification Test Program,’’ of
the OM Code, 2012 Edition.
As previously discussed regarding the
new condition in § 50.55a(b)(3)(vii), the
upper end of the ‘‘Acceptable Range’’
and the ‘‘Required Action Range’’ for
flow and differential or discharge
pressure for comprehensive pump
testing in Subsection ISTB in the OM
Code was raised to higher values in
combination with the incorporation of
Mandatory Appendix V, ‘‘Pump
Periodic Verification Test Program.’’
However, the 2011 Addenda of the OM
Code does not include Appendix V. In
addition, Subsection ISTF in the 2011
Addenda and 2012 Edition of the OM
Code does not include a requirement for
a pump periodic verification test
program. Therefore, the new condition
in § 50.55a(b)(3)(ix) requires that the
provisions of Appendix V be applied
when implementing Subsection ISTF of
the 2012 Edition of the OM Code to
support the application of the upper end
of the Acceptable Range and the
Required Action Range for flow and
differential or discharge pressure for
inservice pump testing in Subsection
ISTF.
10 CFR 50.55a(b)(3)(xi) OM Condition:
Valve Position Indication
The NRC is adding § 50.55a(b)(3)(xi)
to emphasize the provisions in OM
Code, 2012 Edition, Subsection ISTC–
3700, ‘‘Position Verification Testing,’’ to
verify that valve obturator position is
accurately indicated. Subsection ISTC–
3700 of the OM Code requires that
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valves with remote position indicators
shall be observed locally at least once
every 2 years to verify that valve
operation is accurately indicated.
Subsection ISTC–3700 states that where
practicable, this local observation
should be supplemented by other
indications, such as the use of flow
meters or other suitable instrumentation
to verify obturator position. Subsection
ISTC–3700 also states that where local
observation is not possible, other
indications shall be used for verification
of valve operation. Nuclear power plant
operating experience has revealed that
reliance on indicating lights and stem
travel are not sufficient to satisfy the
requirement in ISTC–3700 to verify that
valve operation is accurately indicated.
Appendix A, ‘‘General Design Criteria
for Nuclear Power Plants,’’ to 10 CFR
part 50 requires that where generally
recognized codes and standards are
used, they shall be identified and
evaluated to determine their
applicability, adequacy, and sufficiency,
and shall be supplemented or modified
as necessary to assure a quality product
in keeping with the required safety
function. This new condition specifies
that when implementing OM Code,
Subsection ISTC–3700, licensees shall
verify that valve operation is accurately
indicated by supplementing valve
position indicating lights with other
indications, such as flow meters or other
suitable instrumentation, to provide
assurance of proper obturator position.
The OM Code specifies obturator
movement verification in order to detect
certain internal valve failure modes
consistent with the definition of
‘exercising’ found in ISTA–2000,
‘‘Definitions,’’ (i.e., demonstration that
the moving parts of a component
function). Verification of the ability of
an obturator to change or maintain
position is an essential element of valve
operational readiness determination,
which is a fundamental aspect of the
OM Code.
The NRC initially emphasized the
ASME OM Code requirement for valve
position indication in 1995 in the
original issuance of NUREG–1482,
‘‘Guidelines for Inservice Testing at
Nuclear Power Plants,’’ paragraph 4.2.5.
The NRC’s position is further elaborated
in NUREG–1482 (Revision 2),
‘‘Guidelines for Inservice Testing at
Nuclear Power Plants: Inservice Testing
of Pumps and Valves and Inservice
Examination and Testing of Dynamic
Restraints (Snubbers) at Nuclear Power
Plants,’’ paragraph 4.2.7. As discussed
in NUREG–1482 (Revision 2), ISTC–
3700 allows flexibility to licensees in
verifying that operation of valves with
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remote position indicators is accurately
indicated. For example, NUREG–1482
refers to various methods to verify valve
operation, such as nonintrusive
techniques, flow initiation or absence of
flow, leak testing, and pressure testing.
The extent of verification necessary for
valve operation to satisfy ISTC–3700
will depend on the type of valve, the
sophistication of the diagnostic
equipment used in testing the valve,
possible failure modes of the valve, and
the operating history of the valve and
similar valve types. To satisfy ISTC–
3700, the licensee is responsible for
developing and implementing a method
to provide reasonable assurance that
valve operation is accurately indicated.
The NRC is requiring this condition
for the implementation of the 2012
Edition of the OM Code for the 120month IST interval in order to allow
additional time for licensees to comply
with this condition.
10 CFR 50.55a(f): Preservice and
Inservice Testing Requirements
The NRC is revising the introductory
text of § 50.55a(f) to indicate that
systems and components must meet the
requirements for ‘‘preservice and
inservice testing’’ in the applicable
ASME Codes and that both activities are
referred to as ‘‘inservice testing’’ in the
remainder of paragraph (f). The change
clarifies that the OM Code includes
provisions for preservice testing of
components as part of its overall
provisions for IST programs. No
expansion of IST program scope was
intended by this clarification.
In the proposed rule, the NRC
included references to the OM Code in
§ 50.55a(f)(3)(iii)(A), Class 1 Pumps and
Valves: First Provision;
§ 50.55a(f)(3)(iii)(B), Class 1 Pumps and
Valves: Second Provision;
§ 50.55a(f)(3)(iv)(A), Class 2 and 3
Pumps and Valves: First Provision; and
§ 50.55a(f)(3)(iv)(B): Second Provision;
to align the regulatory language with the
current ASME OM Code used for IST
programs. Because § 50.55a(f)(3)(iii) and
(iv) specifically reference Class 1, 2, or
3 pumps and valves, the proposed
changes to these paragraphs referencing
the OM Code are unnecessary and have
not been adopted in this final rule.
10 CFR 50.55a(f)(4) Inservice Testing
Standards Requirement for Operating
Plants
The NRC is revising § 50.55a(f)(4) to
clarify that this paragraph is applicable
to pumps and valves that are within the
scope of the OM Code. This revision
aligns the scope of pumps and valves for
inservice testing with the scope defined
in the OM Code.
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Public comments on the alignment of
the IST program scope in § 50.55a(f)(4)
indicated that the nuclear industry is
addressing the requirements in 10 CFR
part 50, appendices A and B, to
establish an IST program for safetyrelated pumps and valves that are not
classified as ASME BPV Code Class 1,
2, or 3 components through either the
OM Code provisions or augmented IST
programs. For example, one public
commenter indicated that generally,
augmented IST programs are designed
to meet the OM Code where practicable,
but relief requests are not required when
alternate testing is necessary. The NRC
regulations in § 50.55a address the
concept of augmented IST programs for
pumps and valves at nuclear power
plants. For example, § 50.55a(f)(6)(ii),
‘‘Augmented IST requirements,’’
indicates that the licensee may follow
an augmented IST program for pumps
and valves for which the NRC deems
that added assurance of operational
readiness is necessary. The NRC finds
that an augmented IST program as
addressed in § 50.55a(f)(6)(ii) is
acceptable for safety-related pumps and
valves that are not classified as ASME
BPV Code Class 1, 2, or 3 components.
Public commenters were concerned
that the alignment of the scope of the
OM Code and § 50.55a would cause a
potential paperwork burden for the
submittal of relief or alternative requests
for safety-related pumps and valves that
are not classified as ASME BPV Code
Class 1, 2, or 3 components. In response
to these comments, the NRC included a
provision in § 50.55a(f)(4) that the IST
requirements for pumps and valves that
are within the scope of the OM Code but
are not classified as ASME BPV Code
Class 1, Class 2, or Class 3 may be
satisfied as an augmented IST program
in accordance with § 50.55a(f)(6)(ii)
without requesting relief under
§ 50.55a(f)(5) or alternatives under
§ 50.55a(z). This use of an augmented
IST program may be acceptable
provided the basis for deviations from
the OM Code, as incorporated by
reference in this section, demonstrates
an acceptable level of quality and safety,
or that implementing the Code
provisions would result in hardship or
unusual difficulty without a
compensating increase in the level of
quality and safety, where documented
and available for NRC review. This
additional provision avoids the
potential paperwork burden for the
submittal of relief or alternative requests
by allowing the licensee to maintain the
documentation demonstrating an
acceptable level of quality and safety on
site for NRC review, as appropriate. The
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documentation and availability of the
basis for deviations from the OM Code
for NRC review are acceptable for
pumps and valves within the scope of
the OM Code but not classified as ASME
BPV Code Class 1, 2, or 3, based on their
lower safety significance in comparison
to ASME BPV Code Class 1, 2, and 3
pumps and valves.
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(g)(4) Inservice Inspection
Standards Requirement for Operating
Plants
The NRC recognizes that updating an
Appendix VIII program is a complex
and time-consuming process. The NRC
also recognizes that licensees would
face the possibility of needing to
maintain multiple Appendix VIII
programs if units were to update their
ISI programs on different dates.
Maintaining certifications to multiple
Appendix VIII programs would be very
complicated, while not improving the
effectiveness of the programs. Based on
public comments, and to assist licensees
in updating and coordinating their ISI
programs, the NRC is adding two
options to the regulations. First, the
NRC is revising § 50.55a(g)(4)(i) and (ii)
to clarify that a licensee whose ISI
interval commences during the 12- to
18-month period after the approval date
of this final rule, may delay the update
of their Appendix VIII program by up to
18 months after the approval date of this
final rule. This will provide licensees
with enough time to incorporate the
changes for the new Appendix VIII
program. Second, the NRC is adding the
option for licensees to update their ISI
program to use the latest edition and
addenda of Appendix VIII incorporated
by reference in § 50.55a(a)(1) at any time
in the licensee’s ten-year interval.
Licensees can normally update their ISI
programs using all or portions of newer
versions of ASME BPV Code Section XI
under § 50.55a(g)(4)(iv), subject to NRC
review and approval. While some
requests to use portions of ASME BPV
Code Section XI require a detailed
review by the NRC, a licensee asking to
use the entire latest incorporated-byreference version of Appendix VIII
would certainly be approved by the
NRC staff in this process. This provision
will, therefore, allow licensees to use
the latest incorporated version of
Appendix VIII, as long as it is coupled
with the same edition and addenda of
Appendix I, without the NRC review
and approval process. This will allow
licensees to coordinate their ISI
programs and use the latest approved
version of Appendix VIII without the
delay imposed by submitting a relief
request under § 50.55a (g)(4)(iv).
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D. ASME Code Cases
Administrative Changes to References in
§ 50.55a to NRC Regulatory Guides
Identifying ASME Code Cases Approved
for Use by the NRC
The NRC is removing the revision
number of the three RGs currently
approved by the Office of the Federal
Register for incorporation by reference
throughout the substantive provisions of
§ 50.55a addressing the ASME Code
Cases, i.e., paragraphs (b) through (g).
The revision numbers for the RGs
approved for incorporation by reference
(currently, RG 1.84, RG 1.147, and RG
1.192) will be retained in
§ 50.55a(a)(3)(i) through (iii), where the
RGs are listed by full title, including
revision number. These changes
simplify the regulatory language
containing cross-references to these RGs
and reduce the possibility of NRC error
in preparing future amendments to
§ 50.55a with respect to these RGs.
These changes are administrative in
nature and do not change substantive
requirements with respect to the RGs
and the Code Cases listed in the RGs.
Administrative Changes To Comply
With Requirements for Incorporation by
Reference
The NRC is revising § 50.55a(a)(1)(iii)
to maintain the ASME Code Cases in
alphanumeric order.
Organization of NRC’s Discussion of the
Six ASME Code Cases Incorporated by
Reference in This Final Rule
The discussions under the following
headings address four of the six ASME
Code Cases being incorporated by
reference in this rulemaking (N–729–4,
N–770–2, N–824, and OMN–20). A fifth
ASME Code Case, N–852, is discussed
in Section II.A, ‘‘ASME BPV Code,
Section III,’’ because the NRC’s approval
of that Code Case relates to a provision
of Section III, which is addressed in
§ 50.55a(b)(1)(ix). The sixth ASME Code
Case, N–513–3, is discussed in Section
II.B, ‘‘ASME BPV Code, Section XI,’’
because the NRC’s approval of that Code
Case relates to a provision of Section XI,
which is addressed in
§ 50.55a(b)(2)(xxxiv).
ASME BPV Code Case N–729–4
On September 10, 2008, the NRC
issued a final rule to update § 50.55a to
the 2004 Edition of the ASME BPV Code
(73 FR 52730). As part of the final rule,
§ 50.55a(g)(6)(ii)(D) implemented an
augmented ISI program for the
examination of pressurized water
reactor RPV upper head penetration
nozzles and associated partial
penetration welds. The program
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required the implementation of ASME
BPV Code Case N–729–1, with certain
conditions.
The application of ASME BPV Code
Case N–729–1 was necessary because
the inspections required by the 2004
Edition of the ASME BPV Code, Section
XI were not written to address
degradation of the RPV upper head
penetration nozzles and associated
welds by primary water stress corrosion
cracking (PWSCC). The safety
consequences of inadequate inspections
can be significant. The NRC’s
determination that the ASME BPV Code
required inspections are inadequate is
based upon operating experience and
analysis. The absence of an effective
inspection regime could, over time,
result in unacceptable circumferential
cracking, or the degradation of the RPV
upper head or other reactor coolant
system components by leakage assisted
corrosion. These degradation
mechanisms increase the probability of
a loss-of-coolant accident.
Examination frequencies and methods
for RPV upper head penetration nozzles
and welds are provided in ASME BPV
Code Case N–729–1. The use of code
cases is voluntary, so these provisions
were developed, in part, with the
expectation that the NRC would
incorporate the code case by reference
into the CFR. Therefore, the NRC
adopted rule language in
§ 50.55a(g)(6)(ii)(D) requiring
implementation of ASME BPV Code
Case N–729–1, with conditions, in order
to enhance the examination
requirements in the ASME BPV Code,
Section XI for RPV upper head
penetration nozzles and welds. The
examinations conducted in accordance
with ASME BPV Code Case N–729–1
provide reasonable assurance that
ASME BPV Code allowable limits will
not be exceeded and that PWSCC will
not lead to failure of the RPV upper
head penetration nozzles or welds.
However, the NRC concluded that
certain conditions were needed in
implementing the examinations in
ASME BPV Code Case N–729–1. These
conditions are set forth in
§ 50.55a(g)(6)(ii)(D).
On June 22, 2012, the ASME
approved the fourth revision of ASME
BPV Code Case N–729 (N–729–4). This
revision changed certain requirements
based on a consensus review of
inspection techniques and frequencies.
These changes were deemed necessary
by the ASME to supersede the previous
requirements under N–729–1 to
establish an effective long-term
inspection program for the RPV upper
head penetration nozzles and associated
welds in pressurized water reactors. The
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major changes included incorporation of
previous NRC conditions in the CFR.
Minor changes were also made to
address editorial issues, to correct
figures or to add clarity.
The NRC is updating the requirements
of § 50.55a(g)(6)(ii)(D) to require
licensees to implement ASME BPV
Code Case N–729–4, with conditions.
One existing condition on ASME BPV
Code Case N–729–1 has been modified,
four existing conditions are being
deleted in this final rule, one existing
condition is being redesignated without
substantive change, and two new
conditions—in § 50.55a(g)(6)(ii)(D)(3)
and (4)—are adopted in this final rule in
order to address the changes in ASME
BPV Code Case N–729–4. The NRC’s
revisions to the conditions are discussed
under the next three headings. As
discussed earlier, this final rule
incorporates by reference ASME BPV
Code Case N–729–4 into
§ 50.55a(a)(1)(iii)(C).
10 CFR 50.55a(g)(6)(ii)(D)(1)
Implementation
The NRC is revising
§ 50.55a(g)(6)(ii)(D)(1) to change the
version of ASME BPV Code Case N–729
from N–729–1 to N–729–4 for the
reasons previously set forth. Due to the
incorporation of N–729–4, the date to
establish applicability for licensed
pressurized water reactors will be
changed to the effective date of this final
rule.
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6)
(Removed)
The NRC is removing the existing
conditions in § 50.55a(g)(6)(ii)(D)(2)
through (5) and redesignating the
condition currently in
§ 50.55a(g)(6)(ii)(D)(6) as
§ 50.55a(g)(6)(ii)(D)(2) without any
substantive change. The existing
conditions in § 50.55a(g)(6)(ii)(D)(2)
through (5) have all been incorporated
either verbatim or more conservatively
in the revisions to ASME BPV Code
Case N–729, up to version N–729–4.
Therefore, there is no reason to retain
these conditions in § 50.55a.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal
Visual Frequency (New Condition)
The NRC is adopting a new condition
in § 50.55a(g)(6)(ii)(D)(3) to modify the
option in ASME BPV Code Case N–729–
4 to extend bare metal visual
inspections of the RPV upper head
surface beyond the frequency listed in
Table 1 of the Code Case. Previously,
upper heads aged with less than eight
effective degradation years were
considered to have a low probability of
initiating PWSCC, the cracking
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mechanism of concern. This ranking of
effective degradation years was based on
a simple time at temperature
correlation. All of the upper heads
within this category, with the exception
of new heads using Alloy 600
penetration nozzles, were considered to
have lower susceptibility to cracking
due to the upper heads being at or near
the cold leg operating temperature of the
reactor coolant system. Therefore, these
plants were referred to as having ‘‘cold
heads.’’ All of the upper heads that had
experienced cracking prior to 2006 were
near the hot leg operating temperature
of the reactor coolant system, which
validated the time at temperature
model.
In 2006, one of the 21 ‘‘cold head’’
plants identified two indications within
a penetration nozzle and the associated
partial penetration weld. Then, between
2006 and 2013, five of the 21 ‘‘cold
head’’ plants identified multiple
indications within fifteen different
penetration nozzles and the associated
partial penetration welds. None of these
indications caused leakage, and
volumetric examination of the
penetration nozzles showed that no
flaws in the nozzle material had grown
through-wall; however, this increasing
trend creates a reasonable safety
concern.
Recent operational experience has
shown that the volumetric inspection of
penetration nozzles, at the current
inspection frequency, is adequate to
identify indications in the nozzle
material prior to leakage; however,
volumetric examinations cannot be
performed on the partial penetration
welds. Therefore, given the additional
cracking identified at cold leg
temperatures, the NRC staff has
concerns about the adequacy of the
partial penetration weld examinations.
Leakage from a partial penetration
weld into the annulus between the
nozzle and head material can cause
corrosion of the low alloy steel head.
While initially limited in leak rate, due
to limited surface area of the weld being
in contact with the annulus region,
corrosion of the vessel head material
can expose more of the weld surface to
the annulus, allowing a greater leak rate.
Since an indication in the weld cannot
be identified by a volumetric inspection,
a postulated crack through the weld,
just about to cause leakage, could exist
as a plant performed its last volumetric
and/or bare metal visual examination of
the upper head material. This gives the
crack years to breach the surface and
leak prior to the next scheduled visual
examination.
Only a surface examination of the
wetted surface of the partial penetration
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weld can reliably detect flaws in the
weld. Unfortunately, this examination
cannot size the flaws in the weld, and,
if performed manually, requires
significant radiological dose to examine
all of the partial penetration welds on
the upper head. As such, the available
techniques are only able to detect a flaw
after it has caused leakage. These
techniques are a bare metal visual
examination or a volumetric leak path
assessment performed on the frequency
of the volumetric examination.
Volumetric leak path examinations
are only done during outages when a
volumetric examination of the nozzle is
performed. Therefore, under the current
requirements allowed by Note 4 of
ASME BPV Code Case N–729–4, leakage
from a crack in the weld of a ‘‘cold
head’’ plant could start and continue to
grow for the 5 years between the
required bare metal visual examinations
to detect leakage through the partial
penetration weld.
Given the additional cracking
identified at cold leg temperatures of
upper head penetration nozzles and
associated welds, the NRC finds limited
basis to continue to categorize these
‘‘cold head’’ plants as having a low
susceptibility to crack initiation. The
NRC is increasing the frequency of the
bare metal visual examinations of ‘‘cold
heads’’ to identify potential leakage as
soon as reasonably possible due to the
volumetric examination limitations.
Therefore, the NRC is conditioning Note
4 of ASME BPV Code Case N–729–4 to
require a bare metal visual exam during
each outage in which a volumetric exam
is not performed. The NRC also will
allow ‘‘cold head’’ plants to extend their
bare metal visual inspection frequency
from once each refueling outage, as
stated in Table 1 of N–729–1, to once
every 5 years, but only if the licensee
performed a wetted surface examination
of all of the partial penetration welds
during the previous volumetric
examination. Applying the conditioned
bare metal visual inspection frequency
or a volumetric examination each outage
will allow licensees to identify any
potential leakage through the partial
penetration welds prior to significant
degradation of the low alloy steel head
material, thereby providing reasonable
assurance of the structural integrity of
the reactor coolant pressure boundary.
These issues, including the
operational experience, the fact that
volumetric examination is not available
to interrogate the partial penetration
welds, and potential regulatory options,
were discussed publicly at multiple
ASME BPV Code meetings, at the
annual Materials Programs Technical
Information Exchange public meeting
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held at the NRC Headquarters in June
2013, and at the 2013 NRC Regulatory
Information Conference.
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10 CFR 50.55a(g)(6)(ii)(D)(4) Surface
Exam Acceptance Criteria (New
Condition)
The NRC is adopting a new condition
in § 50.55a(g)(6)(ii)(D)(4) to define
surface examination acceptance criteria.
Paragraph –3132(b) of ASME BPV Code
Case N–729–4 sets forth the acceptance
criteria for surface examinations. In
general, throughout Section XI of the
ASME BPV Code, the acceptance
criteria for surface examinations default
to Section III, Paragraph NB–5352,
‘‘Acceptance Standards.’’ Typically, for
rounded indications, the indication was
only unacceptable if it was greater than
3⁄16-inch in size. The NRC requested that
the code case authors include a
requirement that any size rounded
indication causing nozzle leakage is
unacceptable due to operating
experience identifying PWSCC under
rounded indications less than 3⁄16-inch
in size.
Recently, the ASME BPV Code
Committee approved an interpretation
of the language in Paragraph –3132(b),
which implied that any size rounded
indication is acceptable unless there is
relevant indication of nozzle leakage,
even those greater than 3⁄16-inch. The
NRC does not agree with the
interpretation and maintains its original
position on rounded indications that
any size rounded indication is
unacceptable if there is an indication of
leakage. Since the adoption of ASME
BPV Code Case N–729–1 into
§ 50.55a(g)(6)(ii)(D), all licensees have
used the NRC’s position in
implementing Paragraph –3132(b), even
after the recent ASME BPV Code
Committee interpretation approval over
NRC objection.
Therefore, in order to ensure
compliance with the previous and
ongoing requirement, the NRC is
revising condition
§ 50.55a(g)(6)(ii)(D)(4) to include clarity
within the acceptance criteria for
surface examinations. The current
edition requirements of NB–5352 of
ASME BPV Code, Section III for the
licensee’s ongoing 10-year inservice
inspection interval shall be met.
ASME BPV Code Case N–770–2
On June 21, 2011 (76 FR 36232), the
NRC issued a final rule, which included
§ 50.55a(g)(6)(ii)(F) that requires the
implementation of ASME BPV Code
Case N–770–1, ‘‘Alternative
Examination Requirements and
Acceptance Standards for Class 1 PWR
Piping and Vessel Nozzle Butt Welds
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Fabricated with UNS N06082 or UNS
N86182 Weld Filler Material With or
Without Application of Listed
Mitigation Activities,’’ with certain
conditions.
On June 9, 2011, the ASME approved
the second revision of ASME BPV Code
Case N–770 (N–770–2). The major
changes from N–770–1 to N–770–2
included establishing new ASME BPV
Code Case, Table 1, inspection item
classifications for optimized weld
overlays and allowing alternatives when
complete inspection coverage cannot be
met. Minor changes were also made to
address editorial issues, to correct
figures, or to add clarity. The NRC
found that the updates and
improvements in N–770–2 are sufficient
to update § 50.55a(g)(6)(ii)(F).
The NRC, therefore, is updating the
requirements of § 50.55a(g)(6)(ii)(F) to
require licensees to implement ASME
BPV Code Case N–770–2, with
conditions. The NRC conditions have
been modified to address the changes in
ASME BPV Code Case N–770–2 and to
ensure that this regulatory framework
will provide adequate protection of
public health and safety. The following
sections discuss each of the NRC’s
changes to the conditions on ASME BPV
Code Case N–770–2. As discussed
earlier, this final rule incorporates by
reference ASME BPV Code Case N–770–
2 into § 50.55a(a)(1)(iii)(D).
10 CFR 50.55a(g)(6)(ii)(F)(1)
Implementation
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(1) to change the
version of ASME BPV Code Case N–770
from N–770–1 to N–770–2 and to
require its implementation, with
conditions, to incorporate the updates
and improvements contained in N–770–
2. The NRC will allow licensees to begin
using N–770–2 on the effective date of
this rule.
10 CFR 50.55a(g)(6)(ii)(F)(2)
Categorization
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(2) to provide
clarification regarding categorization of
each Alloy 82/182 butt weld, mitigated
or not, under N–770–2. This paragraph
also clarifies the NRC’s position that
Paragraph –1100(e) shall not be used to
exempt welds that rely on Alloy 82/182
for structural integrity from more
frequent ISI schedules until the NRC has
reviewed and authorized an alternative
categorization for the weld.
Additionally, the NRC will change the
inspection item categories for full
structural weld overlays from C to C–1
and F to F–1 due to reclassification
under ASME BPV Code Case N–770–2.
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10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline
Examinations
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(3) to clarify the
baseline examination requirements by
stating that previously-conducted
examinations, in order to count as
baseline examinations, must meet the
requirements of ASME BPV Code Case
N–770–2, as conditioned in this section.
The 2011 rule required the use of ASME
BPV Code Section XI Appendix VIII
qualifications for baseline examinations,
which is stricter than N–770–2 and does
not provide requirements for optimized
weld overlays. The revision also
updates the deadline for baseline
examination requirements, since the
January 20, 2012, deadline from the
previous rule has passed. Finally, upon
implementation of this rule, if a licensee
is currently in an outage, then the
baseline inspection requirement can be
met by performing the inspections in
accordance with the previous regulatory
requirements of § 50.55a(g)(6)(ii)(F), in
lieu of the examination requirements of
Paragraphs –2500(a) or –2500(b) of
ASME BPV Code Case N–770–2.
10 CFR 50.55a(g)(6)(ii)(F)(4)
Examination Coverage
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(4) to define
examination coverage for
circumferential flaws and to prohibit the
use of Paragraph –2500(d) of ASME BPV
Code Case N–770–2 which, in some
circumstances, allows unacceptably low
examination coverage. Paragraph
–2500(d) of N–770–2 would allow the
reduction of circumferential volumetric
examination coverage with analytical
evaluation. Paragraph –2500(c) was
previously prohibited from use, and it
continues to be prohibited. The NRC is
establishing an essentially 100 percent
volumetric examination coverage
requirement, including greater than 90
percent of the required volumetric
examination coverage, for
circumferential flaws to provide
reasonable assurance of structural
integrity of all ASME BPV Code Class 1
butt welds susceptible to PWSCC.
Therefore, the NRC is adopting a
condition prohibiting the use of
Paragraphs –2500(c) and –2500(d). A
licensee may request approval for use of
these paragraphs under 10 CFR
50.55a(z).
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay
Inspection Frequency
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(5) to add the
explanatory heading, ‘‘Inlay/onlay
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inspection frequency,’’ and to make
minor editorial corrections.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting
Requirements
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(6) to add the
explanatory heading, ‘‘Reporting
requirements.’’
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(g)(6)(ii)(F)(7)
Defining ‘‘t’’
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(7) to add the
explanatory heading, ‘‘Defining ‘t’.’’
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized
Weld Overlay Examination
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(8) to add the
explanatory heading, ‘‘Optimized weld
overlay examination,’’ and to maintain
the requirement for the timing of the
initial inservice examination of
optimized weld overlays.
Uncracked welds mitigated with
optimized weld overlays were recategorized by ASME BPV Code Case
N–770–2 from Inspection Item D to
Inspection Item C–2; however, the
initial inspection requirement was not
incorporated into the Code Case for
Inspection Item C–2. The NRC has
determined that uncracked welds
mitigated with an optimized weld
overlay must have an initial inservice
examination no sooner than the third
refueling outage and no later than 10
years following the application of the
weld overlay to identify unacceptable
crack growth. Optimized weld overlays
establish compressive stress on the
inner half thickness of the weld, but the
outer half thickness may also be under
tensile stress. The requirement for an
initial inservice examination no sooner
than the third refueling outage and no
later than 10 years following the
application of the weld overlay is based
on the design of optimized weld
overlays, which require the outer
quarter thickness of the susceptible
material to provide structural integrity
for the weld. Therefore, the NRC is
continuing adoption of the condition,
which requires the initial inservice
examination of uncracked welds
mitigated by optimized weld overlay
(i.e., the welds which are subject to
Inspection Item C–2 of ASME BPV Code
Case N–770–2) within the specified
timeframe.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(9) to add the
explanatory heading, ‘‘Deferral,’’ and to
address changes in ASME BPV Code
Case N–770–2 which allow the deferral
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of the first inservice examination of
uncracked welds mitigated with
optimized weld overlays, Inspection
Item C–2.
Previously, under N–770–1, the initial
inservice examination of these welds
was not allowed to be deferred.
Allowing deferral of the initial inservice
examination in accordance with N–770–
2 could, in certain circumstances, allow
the initial inservice examination to be
performed up to 20 years after
installation. Therefore, the NRC is
adopting a condition which would
preclude the deferral of the initial
inservice examination of uncracked
welds mitigated by optimized weld
overlays.
10 CFR 50.55a(g)(6)(ii)(F)(10)
Examination Technique
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(10) to add the
explanatory heading, ‘‘Examination
technique,’’ and to address changes in
ASME BPV Code Case N–770–2. Note
14(a) of Table 1 of ASME BPV Code
Case N–770–2 provides the previously
required full examination requirement
for optimized weld overlays. The
language of ASME BPV Code Case N–
770–2, however, does not require the
implementation of the full examination
requirements of Note 14(a) of Table 1, if
possible, before implementing the
reduced examination coverage
requirements of Note 14(b) of Table 1 or
Note (b) of Figure 5(a). The NRC agrees
that reduced examination coverage is
the best alternative if the full
examination cannot be met; however,
the full examination requirement should
be implemented, if possible, before the
option of reduced examination coverage
is allowed. Therefore, the NRC is
modifying the current condition in
§ 50.55a(g)(6)(ii)(F)(10) to allow the use
of Note 14(b) of Table 1 and Note (b) of
Figure 5(a) of ASME BPV Code Case N–
770–2 only after the determination that
the requirements of Note 14(a) of Table
1 of ASME BPV Code Case N–770–2
cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast
Stainless Steel
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(11) to address
examination requirements through cast
stainless steel materials by requiring the
use of Appendix VIII qualifications to
meet the inspection requirements of
Paragraph –2500(a) of ASME BPV Code
Case N–770–2. The requirements for
volumetric examination of butt welds
through cast stainless steel materials are
currently being developed as
Supplement 9 to the ASME BPV Code,
Section XI, Appendix VIII. In
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accordance with Appendix VIII for
supplements that have not been
developed, the requirements of
Appendix III apply. Appendix III
requirements are not equivalent to
Appendix VIII requirements. For the
volumetric examination of ASME BPV
Code Class 1 welds, the NRC has
established the requirement for
examination qualification under the
Appendix VIII. Therefore, the NRC is
adopting a condition requiring the use
of Appendix VIII qualifications to meet
the inspection requirements of
Paragraph –2500(a) of ASME BPV Code
Case N–770–2 by January 1, 2022.
The development of a sufficient
number of mockups would be required
to establish an Appendix VIII program
for examination of ASME BPV Code
Class 1 piping and vessel nozzle butt
welds through cast stainless steel
materials. The NRC recognizes that
significant time and resources are
required to create mockups and to allow
for qualification of equipment,
procedures and personnel. Therefore,
the NRC is requiring licensees to use
these Appendix VIII qualifications no
later than their first scheduled weld
examinations involving cast stainless
steel materials occurring after January 1,
2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress
Improvement Inspection Coverage
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(12) to clarify the
examination coverage requirements
allowed under Appendix I of ASME
BPV Code Case N–770–2 for butt welds
joining cast stainless steel material.
Under current ASME BPV Code, Section
XI, Appendix VIII requirements, the
volumetric examination of butt welds
through cast stainless steel materials is
under Supplement 9. Supplement 9
rules are still being developed by the
ASME BPV Code. Therefore, it is
currently impossible to meet the
requirement of Paragraph I.5.1 for butt
welds joining cast stainless steel
material.
The material of concern is the weld
material susceptible to PWSCC
adjoining the cast stainless steel
material. Appendix VIII qualified
procedures are available to perform the
inspection of the susceptible weld
material, but they are not qualified to
inspect the cast stainless steel materials.
Therefore, the NRC is adopting a
condition changing the inspection
volume for stress-improved dissimilar
metal welds with cast stainless steel
from the ASME BPV Code Section XI
requirements to ‘‘the maximum extent
practical including 100 percent of the
susceptible material volume.’’ This will
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remain applicable until an Appendix
VIII qualified procedure for the
inspection through cast stainless steel
materials is available in accordance
with the new condition in
§ 50.55a(g)(6)(ii)(F)(11).
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded
Ultrasonic Examination
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(13) to require the
encoding of ultrasonic volumetric
examinations of Inspection Items A–1,
A–2, B, E, F–2, J, and K in Table 1 of
N–770–2. A human performance gap
has been found between some ultrasonic
testing procedures, as demonstrated
during ASME BPV Code, Section XI,
Appendix VIII qualification versus as
applied in the field.
The human factors that contributed to
the licensee-performed examinations
which failed to identify significant flaws
at North Anna Power Station, Unit 1 in
2012 (Licensee Event Report 50–338/
2012–001–00) and at Diablo Canyon
Nuclear Power Plant in 2013 (Relief
Request REP–1 U2, Revision 2) can be
avoided by the use of encoded
ultrasonic examinations. Encoded
ultrasonic examinations electronically
store both the positional and ultrasonic
information from the inspections.
Encoded examinations allow for the
inspector to evaluate the data and
search for indications outside of a time
limited environment to assure that the
inspection was conducted properly and
to allow for sufficient time to analyze
the data. Additionally, the encoded
examination would allow for an
independent review of the data by other
inspectors or an independent third
party. Finally, the encoded examination
could be compared to previous and/or
future encoded examinations to
determine if flaws are present and flaw
growth rates. Therefore, the NRC is
adopting a condition requiring the use
of encoding for ultrasonic volumetric
examinations of non-mitigated or
cracked mitigated dissimilar metal butt
welds in the reactor coolant pressure
boundary which are within the scope of
ASME BPV Code Case N–770–2.
ASME BPV Code Case N–824
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10 CFR 50.55a(b)(2)(xxxvii) Section XI
Condition: ASME BPV Code Case N–824
The NRC is adding
§ 50.55a(b)(2)(xxxvii) to allow licensees
to use the provisions of ASME BPV
Code Case N–824, ‘‘Ultrasonic
Examination of Cast Austenitic Piping
Welds From the Outside Surface Section
XI, Division 1,’’ subject to four
conditions in § 50.55a(b)(2)(xxxvii)(A)
through (D), when implementing
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inservice examinations in accordance
with the ASME BPV Code, Section XI
requirements.
During the construction of nuclear
power plants, it was recognized that the
grain structure of cast austenitic
stainless steel (CASS) could prevent
effective ultrasonic inspections of
piping welds where one or both sides of
the welds were constructed of CASS.
The high strength and toughness of
CASS (prior to thermal embrittlement)
made it desirable as a building material
despite this known inspection issue.
This choice of construction materials
has rendered many pressure boundary
components without a means to reliably
inspect them volumetrically. While
there is no operational experience of a
CASS component failing, as part of the
reactor pressure boundary, inservice
volumetric inspection of these
components is necessary to provide
reasonable assurance of their structural
integrity.
The current regulatory requirements
for the examination of CASS, provided
in § 50.55a, do not provide sufficient
guidance to assure that the CASS
components are being inspected
adequately. To illustrate that ASME
BPV Code does not provide adequate
guidance, ASME BPV Code, Section XI,
Appendix III, Supplement 1 states,
‘‘Cast materials may preclude
meaningful examinations because of
geometry and attenuation variables.’’
For this reason, over the past several
decades, licensees have been unable to
perform effective inspections of welds
joining CASS components. To allow for
continued operation of their plants,
licensees submitted hundreds of
requests for relief from the ASME BPV
Code requirements for inservice
inspection of CASS components to the
NRC, resulting in a significant
regulatory burden.
The recent advances in inspection
technology are driving renewed work at
ASME BPV Code meetings to produce
Section XI, Appendix VIII, Supplement
9 to resolve the CASS inspection issue,
but it will be years before these code
updates will be published, as well as
additional time to qualify and approve
procedures for use in the field. Until
then, licensees would still use the
requirements of ASME BPV Code
Section XI, Appendix III, Supplement 1,
which states that inspection of CASS
materials meeting the ASME BPV Code
requirements may not be meaningful.
Consequently, less effective
examinations would continue to be used
in the field, and more relief requests
would be generated between now and
the implementation of Supplement 9.
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The NRC commissioned a research
program to determine the effectiveness
of the new technologies for inspections
of CASS components in an effort to
resolve some of the known inspection
issues. The result of this work is
published in NUREG/CR–6933,
‘‘Assessment of Crack Detection in
Heavy-Walled Cast Stainless Steel
Piping Welds Using Advanced LowFrequency Ultrasonic Methods’’, March
2007, and NUREG/CR–7122, ‘‘An
Evaluation of Ultrasonic Phased Array
Testing for Cast Austenitic Stainless
Steel Pressurizer Surge Line Piping
Welds,’’ March 2012. Based on the
improvements in ultrasonic inspection
technology and techniques for CASS
components, the ASME approved BPV
Code Case N–824 (N–824) on October
16, 2012, which describes how to
develop a procedure capable of
meaningfully inspecting welds in CASS
components.
Effective examinations of CASS
components require the use of lower
frequencies and larger transducers than
are typically used for ultrasonic
inspections of piping welds and would
require licensees to modify their
inspection procedures. The NRC
recognizes that requiring the use of
spatial encoding will limit the full
implementation of ASME BPV Code
Case N–824, as spatial encoding is not
practical for many weld configurations.
At this time, the use of ASME BPV
Code Case N–824, as conditioned, is the
most effective known method for
adequately examining welds with one or
more CASS components. With the use
of ASME BPV Code Case N–824, as
conditioned, licensees will be able to
take full credit for completion of the
§ 50.55a required inservice volumetric
inspection of welds involving CASS
components. The implementation of
ASME BPV Code Case N–824, as
conditioned, will have the dual effect of
improving the rigor of required
volumetric inspections and reducing the
number of uninspectable Class 1 and
Class 2 pressure retaining welds.
The NRC concludes that
incorporation of ASME BPV Code Case
N–824, subject to the four conditions in
§ 50.55a(b)(2)(xxxvii)(A) through (D),
will significantly improve the flaw
detection capability of ultrasonic
inspection of CASS components until
Supplement 9 is implemented, thereby
providing reasonable assurance of leak
tightness and structural integrity.
Additionally, it will reduce the
regulatory burden on licensees and
allow licensees to submit fewer relief
requests for welds in CASS materials.
The four conditions on the use of ASME
BPV Code Case N–824,
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§ 50.55a(b)(2)(xxxvii)(A) through (D), are
discussed in the next four headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) (First
Condition on Use of ASME BPV Code
Case
N–824)
The NRC, based upon NUREG/CR–
6933 and NUREG/CR–7122, has
determined that inspections of CASS
materials are very challenging, and
sufficient technical basis exists to
condition the code case to bring the
code case into agreement with the
NUREG/CR reports. The NUREG/CR
reports also show that CASS materials
produce high levels of coherent noise.
The noise signals can be confusing and
mask flaw indications. Use of encoded
inspection data allows the inspector to
mitigate this problem through the ability
to electronically manipulate the data,
which allows for discrimination
between coherent noise and flaw
indications. The NRC found that
encoding CASS inspection data
provides significant detection benefits.
Therefore, the NRC is adding a
condition in § 50.55a(b)(2)(xxxvii)(A) to
require the use of encoded data when
utilizing N–824 for the examination of
CASS components.
10 CFR 50.55a(b)(2)(xxxvii)(B) (Second
Condition on Use of ASME BPV Code
CaseN–824)
The use of dual element phased-array
search units showed the most promise
in obtaining meaningful responses from
flaws. For this reason, the NRC is
adding a condition in
§ 50.55a(b)(2)(xxxvii)(B) to require the
use of dual, transmit-receive, refracted
longitudinal wave, multi-element
phased array search units when
utilizing N–824 for the examination of
CASS components.
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(b)(2)(xxxvii)(C) (Third
Condition on Use of ASME BPV Code
CaseN–824)
The optimum inspection frequencies
for examining CASS components of
various thicknesses are described in
NUREG/CR–6933 and NUREG/CR–7122.
For this reason, the NRC is adding a
condition in § 50.55a(b)(2)(xxxvii)(C) to
require that ultrasonic examinations
performed to implement ASME BPV
Code Case N–824 on piping greater than
1.6 inches (41 mm) thick shall use a
phased array search unit with a center
frequency of 500 kHz with a tolerance
of + /¥ 20 percent.
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10 CFR 50.55a(b)(2)(xxxvii)(D) (Fourth
Condition on Use of ASME BPV Code
CaseN–824)
NUREG/CR–6933 shows that the grain
structure of CASS can reduce the
effectiveness of some inspection angles.
For this reason, the NRC is adding a
condition in § 50.55a(b)(2)(xxxvii)(D) to
require that ultrasonic examinations
performed to implement ASME BPV
Code Case N–824 shall use a phased
array search unit which produces angles
including, but not limited to, 30 to 55
degrees with a maximum increment of
5 degrees.
OM Code Case OMN–20
10 CFR 50.55a(b)(3)(x) OM Condition:
ASME OM Code Case OMN–20
The NRC is adding § 50.55a(b)(3)(x) to
allow licensees to implement OM Code
Case OMN–20, ‘‘Inservice Test
Frequency,’’ in the OM Code, 2012
Edition, for the editions and addenda of
the OM Code that are listed in
§ 50.55a(a)(1)(iv) as being approved for
incorporation by reference. As a
conforming change, § 50.55a(a)(1)(iii)(G)
is being added to incorporate by
reference OM Code Case OMN–20 into
§ 50.55a.
Surveillance Requirement (SR) 3.0.3
from TS 5.5.6, ‘‘Inservice Testing
Program,’’ allows licensees to apply a
delay period before declaring the SR for
TS equipment ‘‘not met’’ when the
licensee inadvertently exceeds or misses
the time limit for performing TS
surveillance. Licensees have been
applying SR 3.0.3 to inservice tests. The
NRC has determined that licensees
cannot use TS 5.5.6 to apply SR 3.0.3 to
inservice tests under § 50.55a(f) that are
not associated with a TS surveillance.
To invoke SR 3.0.3, the licensee shall
first discover that a TS surveillance was
not performed at its specified frequency.
Therefore, the delay period that SR 3.0.3
provides does not apply to non-TS
support components tested under
§ 50.55a(f). The OM Code does not
provide for any inservice test frequency
reductions or extensions. In order to
provide inservice test frequency
reductions or extensions that can no
longer be provided by SR 3.0.3 from TS
5.5.6, the ASME has developed OM
Code Case OMN–20. The NRC has
reviewed OM Code Case OMN–20 and
has found it acceptable for use. The
NRC determined that OM Code Case
OMN–20 may be safely used for all
licensees using editions and addenda of
the OM Code that are listed in
§ 50.55a(a)(1)(iv). The NRC will include
OM Code Case OMN–20 in the next
revision of RG 1.192, at which time a
conforming change will be made to
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delete both this paragraph and
§ 50.55a(a)(1)(iii)(G).
III. Opportunities for Public
Participation
The proposed rule was published on
September 18, 2015, for a 75-day
comment period (80 FR 56820). The
public comment period closed on
December 2, 2015.
After the close of the public comment
period, the NRC held a public meeting
on March 2, 2016, to discuss the
proposed rule, to answer questions on
specific provisions of the proposed rule,
and to discuss public comments
received on the proposed rule in order
to enhance the NRC’s understanding of
the comments. The public meeting
summary is available in ADAMS under
Accession No. ML16069A408.
IV. NRC Responses to Public Comments
The NRC received 27 letters and
emails in response to the opportunity
for public comment on the proposed
rule. These comment submissions were
submitted by the following commenters
(listed in order of receipt):
1. Private citizen, Edward Cavey
2. Private citizen, Dale Matthews
3. Private citizen, Ron Clow
4. ASME
5. Iddeal Solutions, LLC
6. Electric Power Research Institute (EPRI)
7. Private citizen, William Taylor
8. ASME
9. Private citizen, Dan Nowakowski
10. Wolf Creek Nuclear Operating
Corporation
11. Northern States Power Company—
Minnesota
12. FirstEnergy Nuclear Operating Company
13. PSEG Nuclear
14. Dominion Resources Services, Inc.
15. Private citizen, Terence Chan
16. Nuclear Energy Institute
17. EPRI
18. Duke Energy
19. Private Citizen, William Taylor
20. Dominion Engineering, Inc.
21. Tennessee Valley Authority
22. Southern Nuclear Operating Company
23. Prairie Island Nuclear Plant
24. Inservice Test Owners Group
25. Exelon Generation Company
26. EPRI
27. EPRI
In general, the comments:
• Suggested revising or rewording
conditions to make them clearer.
• Supported incorporation of Code
Cases N–729–4, N–770–2, N–824, or
OMN–20 into § 50.55a.
• Supported the proposed changes to
add or remove conditions.
• Opposed proposed conditions.
• Supplied additional information for
NRC consideration.
• Proposed rewriting or renumbering
of paragraphs.
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• Asked questions or requested
information from the NRC.
Due to the large number of comments
received and the length of the NRC’s
responses, this document summarizes
the NRC’s response to comments in
areas of particular interest to
stakeholders that prompted the NRC to
make changes in this final rule from
what was proposed. A discussion of all
comments and complete NRC responses
are presented in a separate document,
‘‘2017 Final Rule (10 CFR 50.55a)
American Society of Mechanical
Engineers Codes and Code Cases:
Analysis of Public Comments,’’
(ADAMS Accession No. ML16130A531).
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10 CFR 50.55a(a)(1)(ii), (b)(2);
Nonmandatory Appendix U
Public commenters were concerned
that the NRC was proposing to exclude
incorporating by reference
Nonmandatory Appendix U because
Nonmandatory Appendix U is the
incorporation of the provisions of ASME
BPV Code Cases N–513–3 and N–705,
without any technical changes, into the
Section XI Code. The NRC agrees with
this comment, in that ASME BPV Code
Cases N–513–3 and N–705 have been
approved in RG 1.147. Based on these
comments, the NRC has removed the
proposed exclusion of Nonmandatory
Appendix U from this final rule.
However, the NRC has found it
necessary to apply two new conditions
in § 50.55a(b)(2)(xxxiv)(A) and (B) to
Nonmandatory Appendix U. The first
condition provides regulatory
consistency with the approval of the
code cases in RG 1.147. The second
condition requires the use of an
Appendix from ASME BPV Code Case
N–513–3 that was unintentionally
omitted from Appendix U. The NRC
discussed these changes at the March 2,
2016, public meeting, and the NRC
considered the public feedback from
that meeting when developing this final
rule.
10 CFR 50.55a(b)(2)(xii), Underwater
Welding
Public commenters were concerned
that the proposed rule continued to
prohibit the use of underwater welding
in § 50.55a(b)(2)(xii), when changes
were made to address this condition in
the 2010 Edition of Section XI. The NRC
agrees that the condition should be
modified to address the changes in the
Code. After consideration of the public
comments, the NRC noted other
inconsistencies for addressing welding
on irradiated materials that appear in
the Code and in some Code Cases.
Section 50.55a(b)(2)(xii) of this final
rule reflects a change to include two
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conditions that provide consistency for
welding of irradiated materials. The
NRC discussed these changes at the
March 2, 2016, public meeting, and the
NRC considered the public feedback
from that meeting when developing this
final rule.
10 CFR 50.55a(b)(2)(xxxi), Mechanical
Clamping Devices
Public commenters were concerned
that the wording of the proposed
condition in § 50.55a(b)(2)(xxxi) was
unclear and that citing the specific
paragraphs of Section XI to which the
NRC is taking exception would be
clearer. The NRC agrees. To clarify the
requirement for the implementation of
mechanical clamps, the condition was
changed to require the use of Appendix
W of Section XI when using mechanical
clamps. Additionally, use of IWA–
4131.1(c) of the 2010 Edition of Section
XI and IWA–4131.1(d) of the 2011
Addenda of the 2010 Edition and later
versions of Section XI is prohibited.
Identifying these specific subparagraphs
was deemed necessary, as they may
have caused confusion with the
intended purpose of the original
proposed condition in maintaining the
previous regulatory requirements for
mechanical clamping devices. Section
50.55a(b)(2)(xxxi) of this final rule
reflects this change.
10 CFR 50.55a(b)(2)(xxxvii), ASME BPV
Code Case N–824
Public commenters had concerns with
conditions proposed on ASME BPV
Code Case N–824, ‘‘Ultrasonic
Examination of Cast Austenitic Piping
Welds From the Outside Surface Section
XI, Division 1,’’ in
§ 50.55a(b)(2)(xxxvii)(A) through (E).
There were concerns that the conditions
would limit the use of Code Case N–824
and that some conditions did not have
a sufficient technical basis. The NRC
partially agreed with the comments
requesting the removal and modification
of some conditions in
§ 50.55a(b)(2)(xxxvii) restricting the
frequencies and angles usable on some
cast austenitic welds. Based on the
public comments, one condition was
removed entirely and two others were
modified. Section
50.55a(b)(2)(xxxvii)(A) through (D) of
this final rule contain the modified and
reduced conditions on the use of ASME
BPV Code Case N–824. The NRC
discussed these changes at the March 2,
2016, public meeting, and the NRC
considered the public feedback from
that meeting when developing this final
rule.
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10 CFR 50.55a(b)(3)(xi), OM Condition:
Valve Position Indication
Public commenters raised concerns
regarding the proposed condition in
§ 50.55a(b)(3)(xi) to emphasize the OM
Code provisions in Subsection ISTC–
3700, ‘‘Position Verification Testing,’’ to
verify that valve operation is accurately
indicated. Public commenters indicated
that because of the significance of
implementing the condition, some
licensees might need time to revise or
create procedures to govern the
implementation of this condition.
Public commenters also suggested that
the condition be limited to active
valves. The NRC partially agrees and
partially disagrees with these
comments. The NRC agrees that
additional time to implement the
condition regarding valve position
verification is appropriate. Therefore,
the NRC has revised the condition to
indicate that it will be effective with
implementation of the 2012 Edition of
the OM Code. The NRC staff does not
agree with the suggestion to limit the
condition to active valves because the
OM Code requires that passive valves
undergo periodic verification of position
indication.
V. Section-by-Section Analysis
Administrative Changes
The NRC is removing the revision
number of the three RGs currently
approved by the Office of the Federal
Register for incorporation by reference
throughout the substantive provisions of
§ 50.55a addressing the ASME Code
Cases, i.e., paragraphs (b) through (g).
The revision numbers for the RGs
approved for incorporation by reference
(currently, RG 1.84, RG 1.147, and RG
1.192) will be retained in
§ 50.55a(a)(3)(i) through (iii), where the
RGs are listed by full title, including
revision number. That paragraph
identifies the specific materials which
the Office of the Federal Register has
approved for incorporation by reference,
as required by Office of the Federal
Register requirements in 1 CFR 51.9.
Readers would need to refer to
§ 50.55a(a) to determine the specific
revision of the relevant RG that is
approved for incorporation by reference
by the Office of the Federal Register.
These changes are administrative in
nature and do not change substantive
requirements with respect to the RGs
and the Code Cases listed in the RGs.
10 CFR 50.55a(a) Documents Approved
for Incorporation by Reference
The NRC is revising the incorporation
by reference language to update the
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contact information for the NRC
Technical Library.
10 CFR 50.55a(a)(1)(i) ASME Boiler and
Pressure Vessel Code, Section III
The NRC is revising § 50.55a(a)(1)(i)
to clarify that Section III Nonmandatory
Appendices of the listed editions and
addenda are excluded from the
incorporation by reference. The
exclusion was originally added in a
final rule published on June 21, 2011
(76 FR 36232); however, it was
erroneously omitted from the final rule
published on November 5, 2014 (79 FR
65776). The NRC is correcting the
omission in this final rule by inserting
‘‘(excluding Nonmandatory
Appendices)’’ in § 50.55a(a)(1)(i). The
NRC is relocating the definition of the
term ‘‘BPV Code,’’ which is used
throughout the section, from § 50.55a(b)
to § 50.55a(a)(1)(i).
10 CFR 50.55a(a)(1)(i)(E) ‘‘Rules for
Construction of Nuclear Facility
Components—Division 1’’
The NRC is revising
§ 50.55a(a)(1)(i)(E) to add ASME BPV
Code, Section III 2009 Addenda, 2010
Edition, 2011 Addenda, and 2013
Edition.
10 CFR 50.55a(a)(1)(ii) ASME Boiler and
Pressure Vessel Code, Section XI
The NRC is revising § 50.55a(a)(1)(ii)
to include two minor editorial changes:
to replace ‘‘Boiler and Pressure Vessel
Code’’ with ‘‘BPV Code’’ and to replace
‘‘limited to’’ with ‘‘limited by.’’
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10 CFR 50.55a(a)(1)(ii)(C)(52) and (53)
‘‘Rules for Inservice Inspection of
Nuclear Power Plant Components—
Division 1’’
The NRC is revising
§ 50.55a(a)(1)(ii)(C)(52) and (53) to add
ASME BPV Code, Section XI 2009
Addenda, 2010 Edition, 2011 Addenda,
and 2013 Edition. The examination
requirements for Examination Category
B–F, Item Numbers B5.11 and B5.71,
Nozzle-to-Component Butt Welds in the
2011 Addenda and the 2013 Edition of
ASME BPV Code, Section XI are
expressly excluded from the
incorporation by reference in
§ 50.55a(a)(1)(ii)(C)(52) and, therefore,
not approved for use. Similarly, the
requirements of IWB–3112(a)(3) and
IWC–3112(a)(3) in the 2013 Edition of
ASME BPV Code, Section XI are
expressly excluded from the
incorporation by reference in
§ 50.55a(a)(1)(ii)(C)(53) and are not
approved for use.
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10 CFR 50.55a(a)(1)(iii)(A) ASME BPV
Code Case N–513–3 Mandatory
Appendix I
The NRC is revising
§ 50.55a(a)(1)(iii)(A) to include
information for a new standard that is
being incorporated by reference,
entitled, ‘‘ASME BPV Code Case N–
513–3 Mandatory Appendix I.’’
10 CFR 50.55a(a)(1)(iii)(B) ASME BPV
Code Case N–722–1
The NRC is revising
§ 50.55a(a)(1)(iii)(B) to maintain
alphanumeric order for the ASME Code
Cases listed in § 50.55a(a)(1)(iii). ASME
BPV Code Case N–722–1 was previously
approved for incorporation by reference.
10 CFR 50.55a(a)(1)(iii)(C) ASME BPV
Code Case N–729–4
The NRC is revising
§ 50.55a(a)(1)(iii)(C) to add the title
‘‘ASME BPV Code Case N–729–4,’’ and
include information for the standard
that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(D) ASME BPV
Code Case N–770–2
The NRC is adding
§ 50.55a(a)(1)(iii)(D) to add the title
‘‘ASME BPV Code Case N–770–2,’’ and
include information for the standard
that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(E) ASME BPV
Code Case N–824
The NRC is adding
§ 50.55a(a)(1)(iii)(E) to include
information for a new standard that is
being incorporated by reference,
entitled, ‘‘ASME BPV Code Case N–
824.’’
10 CFR 50.55a(a)(1)(iii)(F) ASME BPV
Code Case N–852
The NRC is adding
§ 50.55a(a)(1)(iii)(F) to include
information for a new standard that is
being incorporated by reference,
entitled, ‘‘ASME BPV Code Case N–
852.’’
10 CFR 50.55a(a)(1)(iii)(G) ASME OM
Code Case OMN–20
The NRC is adding
§ 50.55a(a)(1)(iii)(G) to include
information for a new standard that is
being incorporated by reference,
entitled, ‘‘ASME OM Code Case OMN–
20.’’
10 CFR 50.55a(a)(1)(iv) ASME Operation
and Maintenance Code
The NRC is revising § 50.55a(a)(1)(iv)
to correct the title of the OM Code and
to relocate the definition of the term
‘‘OM Code,’’ which is used throughout
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the section, from § 50.55a(b) to
§ 50.55a(a)(1)(iv).
10 CFR 50.55a(a)(1)(iv)(B) ‘‘Operation
and Maintenance of Nuclear Power
Plants, Division 1: Section IST Rules for
Inservice Testing of Light-Water Reactor
Power Plants’’
The NRC is adding new
§ 50.55a(a)(1)(iv)(B) to include ASME
OM Code 2009 Edition and 2011
Addenda.
10 CFR 50.55a(a)(1)(iv)(C) ‘‘Operation
and Maintenance of Nuclear Power
Plants, Division 1: OM Code: Section
IST’’
The NRC is adding new
§ 50.55a(a)(1)(iv)(C) to include ASME
OM Code 2012 Edition.
10 CFR 50.55a(a)(1)(v) ASME Quality
Assurance Requirements
The NRC is adding new
§ 50.55a(a)(1)(v) to include information
regarding NQA–1 standards and add the
title ‘‘ASME Quality Assurance
Requirements’’ for ASME NQA–1 Code
as part of NRC titling convention.
10 CFR 50.55a(b) Use and Conditions on
the Use of Standards
The NRC is revising § 50.55a(b) to
correct the title of the ASME OM Code.
10 CFR 50.55a(b)(1) Conditions on
ASME BPV Code Section III
The NRC is revising § 50.55a(b)(1) to
reflect the latest edition incorporated by
reference, the 2013 Edition.
10 CFR 50.55a(b)(1)(ii) Section III
Condition: Weld Leg Dimensions
The NRC is revising § 50.55a(b)(1)(ii)
to clarify rule language and add Table
I, which clarifies prohibited Section III
provisions for welds with leg size less
than 1.09 tn in tabular form.
10 CFR 50.55a(b)(1)(iv) Section III
Condition: Quality Assurance
The NRC is revising § 50.55a(b)(1)(iv)
to clarify that it allows, but does not
require, applicants and licensees to use
the 2008 Edition through the 2009–1a
Addenda of NQA–1 when applying the
2010 Edition and later editions of the
ASME BPV Code, Section III, up to the
2013 Edition. Applicants and licensees
are required to meet appendix B of 10
CFR part 50, and NQA–1 is one way of
meeting portions of appendix B. An
applicant or licensee may select any
version of NQA–1 that has been
approved for use in § 50.55a, but they
must also use the administrative,
quality, and technical provisions
contained in the version of NCA–4000
referencing that Edition or Addenda of
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NQA–1 selected by the applicant or
licensee.
NQA–1 provides a method for
establishing and implementing a QA
program for the design and construction
of nuclear power plants and fuel
reprocessing plants; however, NQA–1,
as modified and supplemented by NCA–
4000, does not meet all of the
requirements of appendix B to 10 CFR
part 50.
Section 50.55a(b)(1)(iv) clarifies that
applicants and licensees using NQA–1
are also required to meet appendix B to
10 CFR part 50 and the commitments
contained in their QA program
descriptions. To meet the requirements
of appendix B, when using NQA–1
during the design and construction
phase, applicants and licensees must
address, in their quality program
description, those areas where NQA–1
is insufficient to meet appendix B.
Additional guidance and regulatory
positions on how to meet appendix B
when using NQA–1 are provided in RG
1.28, ‘‘Quality Assurance Program
Criteria (Design and Construction).’’
10 CFR 50.55a(b)(1)(vii) Section III
Condition: Capacity Certification and
Demonstration of Function of
Incompressible-Fluid Pressure-Relief
Valves
The NRC is revising § 50.55a(b)(1)(vii)
to reflect the editions and addenda of
the ASME BPV Code incorporated by
reference in this rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III
Condition: Use of ASME Certification
Marks
The NRC is adding § 50.55a(b)(1)(viii)
to allow licensees to use either the
ASME BPV Code Symbol Stamp or
ASME Certification Mark with the
appropriate certification designator and
class designator as specified in the 2013
Edition through the latest edition and
addenda incorporated by reference in
§ 50.55a.
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10 CFR 50.55a(b)(1)(ix) Section III
Condition: NPT Code Symbol Stamps
The NRC is adding § 50.55a(b)(1)(ix)
to allow licensees to use the NPT Code
Symbol Stamp with the letters arranged
horizontally as specified in ASME BPV
Code Case N–852 for the service life of
a component that had the NPT Code
Symbol Stamp applied during the time
period from January 1, 2005, through
December 31, 2015.
10 CFR 50.55a(b)(2) Conditions on
ASME BPV Code, Section XI
The NRC is revising § 50.55a(b)(2) to
reflect the editions and addenda of the
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ASME BPV Code incorporated by
reference in this rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI
Condition: Effective Edition and
Addenda of Subsection IWE and
Subsection IWL
The NRC is revising § 50.55a(b)(2)(vi)
to clarify that the provision applies only
to the class of licensees of operating
reactors that were required by previous
versions of § 50.55a to develop and
implement a containment ISI program
in accordance with Subsection IWE and
Subsection IWL, and complete an
expedited examination of containment
during the 5-year period from
September 9, 1996 to September 9,
2001.
10 CFR 50.55a(b)(2)(viii) Section XI
Condition: Concrete Containment
Examinations
The NRC is revising
§ 50.55a(b)(2)(viii) by removing the
condition for using the 2007 Edition
with 2009 Addenda through the 2013
Edition of Subsection IWL requiring
compliance with § 50.55a(b)(2)(viii)(E).
To support the removal of the condition,
the NRC is adding new requirements
governing the performance and
documentation of concrete containment
examinations in § 50.55a(b)(2)(viii)(H)
and (I), which are discussed separately
in the next two headings.
10 CFR 50.55a(b)(2)(viii)(H) Concrete
Containment Examinations: Eighth
Provision
The NRC is adding
§ 50.55a(b)(2)(viii)(H) to require
licensees to provide the applicable
information specified in paragraphs
(b)(2)(viii)(E)(1), (2), and (3) of this
section in the ISI Summary Report
required by IWA–6000 for each
inaccessible concrete surface area
evaluated under the new code provision
IWL–2512 of the 2009 Addenda up to
and including the 2013 Edition.
10 CFR 50.55a(b)(2)(viii)(I) Concrete
Containment Examinations: Ninth
Provision
The NRC is adding
§ 50.55a(b)(2)(viii)(I) to provide a new
condition requiring the technical
evaluation required by IWL–2512(b) of
the 2009 Addenda up to and including
the 2013 Edition of inaccessible belowgrade concrete surfaces exposed to
foundation soil, backfill, or groundwater
be performed at periodic intervals not to
exceed 5 years. In addition, the licensee
must examine representative samples of
the exposed portions of the below-grade
concrete, when such below-grade
concrete is excavated for any reason.
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The condition applies only to holders of
renewed licenses under 10 CFR part 54
during the period of extended operation
(i.e., beyond the expiration date of the
original 40-year license) of a renewed
license when using IWL–2512(b) of the
2007 Edition with 2009 Addenda
through the latest edition and addenda
in § 50.55a(a)(1)(ii)—the 2013 Edition
under this final rule.
10 CFR 50.55a(b)(2)(ix) Section XI
Condition: Metal Containment
Examinations
The NRC is revising § 50.55a(b)(2)(ix)
to continue to apply the existing
conditions in § 50.55a(b)(2)(ix)(A)(2)
and (b)(2)(ix)(B) and (J) with respect to
the metal containment examination
requirements in Subsection IWE up to
and including the 2013 Edition (and all
future editions and addenda of the
ASME BPV Code which the NRC
incorporates by reference into § 50.55a).
The NRC is accomplishing this by
adding the words ‘‘edition and’’ to the
last sentence in § 50.55a(b)(2)(ix).
10 CFR 50.55a(b)(2)(ix)(D) Metal
Containment Examinations: Fourth
Provision
The NRC is revising the rule text in
§ 50.55a(b)(2)(ix)(D) to improve clarity.
Section 50.55a(b)(2)(ix)(D) introductory
text and (b)(2)(ix)(D)(1) are combined.
The information required to be included
in the ISI Summary report is now all on
the same paragraph level. No
substantive change to the requirements
is intended by this revision.
10 CFR 50.55a(b)(2)(x) Section XI
Condition: Quality Assurance
The NRC is revising § 50.55a(b)(2)(x)
to clarify that it allows, but does not
require, licensees to use the 1994
Edition or the 2008 Edition through the
2009–1a Addenda of NQA–1 when
applying the 2009 Addenda and later
editions and addenda of the ASME BPV
Code, Section XI, up to the 2013
Edition. Licensees are required to meet
appendix B of 10 CFR part 50, and
NQA–1 is one way of meeting portions
of appendix B. A licensee may select
any version of NQA–1 that has been
approved for use in § 50.55a.
NQA–1 provides a method for
establishing and implementing a QA
program for the design and construction
of nuclear power plants and fuel
reprocessing plants; however, NQA–1
does not meet all of the requirements of
appendix B to 10 CFR part 50. Section
50.55a(b)(2)(x) clarifies that licensees
using NQA–1 are also required to meet
appendix B to 10 CFR part 50 and the
commitments contained in their QA
program descriptions. To meet the
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requirements of appendix B, when using
NQA–1 during ISI phase, licensees must
address, in their quality program
description, those areas where NQA–1
is insufficient to meet appendix B.
Additional guidance and regulatory
positions on how to meet appendix B
when using NQA–1 are provided in RG
1.28.
10 CFR 50.55a(b)(2)(xii) Section XI
Condition: Underwater Welding
The NRC is revising § 50.55a(b)(2)(xii)
to allow underwater welding on
irradiated materials in accordance with
IWA–4660 under certain conditions.
Licensees are allowed to perform
welding on irradiated materials if
certain neutron fluence criteria and, for
certain material classes, helium
concentration criteria are not exceeded.
If these criteria are exceeded, the
licensee is prohibited from performing
welding on irradiated materials unless
the licensee obtains NRC approval in
accordance with § 50.55a(z).
10 CFR 50.55a(b)(2)(xviii)(D) NDE
Personnel Certification: Fourth
Provision
The NRC is adding
§ 50.55a(b)(2)(xviii)(D) to provide a new
condition prohibiting the use of
Appendix VII and Subarticle VIII–2200
of the 2011 Addenda and 2013 Edition
of Section XI of the ASME BPV Code.
Licensees are required to implement
Appendix VII and Subarticle VIII–2200
of the 2010 Edition of Section XI.
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB–
2500–1 Examination Requirements:
First Provision
The NRC is revising
§ 50.55a(b)(2)(xxi)(A) to modify the
standard for visual magnification
resolution sensitivity and contrast for
visual examinations performed on
Examination Category B–D components
instead of ultrasonic examinations. A
visual examination with magnification
that has a resolution sensitivity to
resolve 0.044 inch (1.1 mm) lower case
characters without an ascender or
descender (e.g., a, e, n, v), utilizing the
allowable flaw length criteria in Table
IWB–3512–1, 1997 Addenda through
the latest edition and addenda
incorporated by reference in
§ 50.55a(a)(1)(ii), with a limiting
assumption on the flaw aspect ratio (i.e.,
a/l = 0.5), may be performed instead of
an ultrasonic examination. This revision
removes a requirement that was in
addition to the ASME BPV Code that
required 1-mil wires to be used in
licensees’ Sensitivity, Resolution, and
Contrast Standard targets.
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10 CFR 50.55a(b)(2)(xxiii) Section XI
Condition: Evaluation of Thermally Cut
Surfaces
The NRC is revising
§ 50.55a(b)(2)(xxiii) to modify the
applicability of the condition. The
condition will only apply to the 2001
Edition through the 2009 Addenda
IWA–4461.4, which was revised in the
2010 Edition to remove paragraph IWA–
4461.4.2, which permitted an
application specific evaluation of
thermally cut surfaces in lieu of a
thermal metal removal process
qualification.
10 CFR 50.55a(b)(2)(xxxi) Section XI
Condition: Mechanical Clamping
Devices
The NRC is adding
§ 50.55a(b)(2)(xxxi) to provide a new
condition maintaining the requirement
to use Appendix IX, now renumbered as
Appendix W, when installing a
mechanical clamping device on an
ASME BPV Code Class piping system.
Additionally, the condition prohibits
the use of mechanical clamping devices
in accordance with the changes made to
IWA–4131.1(c) in the 2010 Edition and
IWA–4131.1(d) in the 2011 Addenda
through 2013 Edition on small item
Class 1 piping and portions of a piping
system that form the containment
boundary.
10 CFR 50.55a(b)(2)(xxxii) Section XI
Condition: Summary Report Submittal
The NRC is adding
§ 50.55a(b)(2)(xxxii) to provide a new
condition requiring licensees using the
2010 Edition or later editions and
addenda of Section XI to follow the
requirements of IWA–6240 of the 2009
Addenda of Section XI for the submittal
of Preservice and Inservice Summary
Reports. The condition also describes
the timing of the submission of the
Summary Reports by referencing the
specific Section XI paragraph IWA–
6240(b) in the regulation.
10 CFR 50.55a(b)(2)(xxxiii) Section XI
Condition: Risk-Informed Allowable
Pressure
The NRC is adding
§ 50.55a(b)(2)(xxxiii) to provide a new
condition to prohibit the use of
Appendix G, Paragraph G–2216, in the
2011 Addenda and later editions and
addenda of the ASME BPV Code,
Section XI.
10 CFR 50.55a(b)(2)(xxxiv) Section XI
Condition: Nonmandatory Appendix U
The NRC is adding
§ 50.55a(b)(2)(xxxiv)(A) and (B) to
require that two conditions be satisfied
when using Nonmandatory Appendix U
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of the 2013 Edition of the ASME BPV
Code, Section XI. Paragraph
(b)(2)(xxxiv)(A) requires that an ASME
BPV Code repair or replacement activity
temporarily deferred under the
provisions of Nonmandatory Appendix
U to the 2013 Edition of the ASME BPV
Code, Section XI, shall be performed
during the next scheduled refueling
outage. Paragraph (b)(2)(xxxiv)(B)
requires the use of the mandatory
appendix in ASME BPV Code Case N–
513–3, in lieu of the appendix
referenced in paragraph U–S1–4.2.1(c)
of Appendix U, which was
inadvertently omitted from Appendix U.
10 CFR 50.55a(b)(2)(xxxv) Section XI
Condition: Use of RTT0 in the KIa and KIc
Equations
The NRC is adding
§ 50.55a(b)(2)(xxxv) to provide a new
condition to specify that when licensees
use ASME BPV Code, Section XI, 2013
Edition, Appendix A, paragraph A–
4200, if T0 is available, then RTT0 may
be used in place of RTNDT for
applications using the KIc equation and
the associated KIc curve, but not for
applications using the KIa equation and
the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI
Condition: Fracture Toughness of
Irradiated Materials
The NRC is adding
§ 50.55a(b)(2)(xxxvi) to provide a new
condition requiring licensees using
ASME BPV Code, Section XI, 2013
Edition, Appendix A, paragraph A–
4400, to obtain NRC approval under
§ 50.55a(z) before using irradiated T0
and the associated RTT0 in establishing
fracture toughness of irradiated
materials.
10 CFR 50.55a(b)(2)(xxxvii) Section XI
Condition: ASME BPV Code Case N–824
The NRC is adding
§ 50.55a(b)(2)(xxxvii) to provide a new
provision that allows licensees to
implement ASME BPV Code Case N–
824, ‘‘Ultrasonic Examination of Cast
Austenitic Piping Welds From the
Outside Surface Section XI, Division 1,’’
subject to four conditions in paragraphs
(b)(2)(xxxvii)(A) through (D). Each of
these paragraphs are discussed in the
following headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) Section
XI Condition: ASME BPV Code Case N–
824
The NRC is adding
§ 50.55a(b)(2)(xxxvii)(A) to add a new
condition that requires ultrasonic
examinations performed to implement
ASME BPV Code Case N–824 to be
spatially encoded.
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10 CFR 50.55a(b)(2)(xxxvii)(B) Section
XI Condition: ASME BPV Code Case N–
824
The NRC is adding
§ 50.55a(b)(2)(xxxvii)(B) to add a new
condition that requires that ultrasonic
examinations performed to implement
ASME BPV Code Case N–824 shall use
dual, transmit-receive, refracted
longitudinal wave, multi-element
phased array search units instead of the
requirements of Paragraph 1(c)(1)(–a) of
N–824.
10 CFR 50.55a(b)(2)(xxxvii)(C) Section
XI Condition: ASME BPV Code Case N–
824
The NRC is adding
§ 50.55a(b)(2)(xxxvii)(C) to add a new
condition that requires that ultrasonic
examinations performed to implement
ASME BPV Code Case N–824 on piping
greater than 1.6 inches (41 mm) thick
shall use a phased array search unit
with a center frequency of 500 kHz with
a tolerance of + /¥ 20 percent instead
of the requirements of Paragraph
1(c)(1)(–c)(–2).
sradovich on DSK3GMQ082PROD with RULES2
10 CFR 50.55a(b)(2)(xxxvii)(D) Section
XI Condition: ASME BPV Code Case N–
824
The NRC is adding
§ 50.55a(b)(2)(xxxvii)(D) to add a new
condition that requires that ultrasonic
examinations performed to implement
ASME BPV Code Case N–824 shall use
a phased array search unit which
produces angles including, but not
limited to, 30 to 55 degrees with a
maximum increment of 5 degrees
instead of the requirements of Paragraph
1(c)(1)(–d).
10 CFR 50.55a(b)(3) Conditions on
ASME OM Code
The NRC is revising § 50.55a(b)(3) to
clarify that Subsections ISTA, ISTB,
ISTC, ISTD, ISTE, and ISTF; Mandatory
Appendices I, II, III, and V; and
Nonmandatory Appendices A through H
and J through M of the OM Code are
each incorporated by reference into
§ 50.55a. The NRC is also clarifying that
the OM Code Nonmandatory
Appendices incorporated by reference
into § 50.55a are approved for use, but
are not mandated. The Nonmandatory
Appendices may be used by applicants
and licensees of nuclear power plants,
subject to the conditions in
§ 50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition:
Quality Assurance
The NRC is revising § 50.55a(b)(3)(i)
to allow licensees to use the 1994
Edition, 2008 Edition, and 2009–1a
Addenda of NQA–1 when using the
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1995 Edition through the 2012 Edition
of the OM Code. Licensees are required
to meet appendix B to 10 CFR part 50,
and NQA–1 is one way of meeting
portions of appendix B.
NQA–1 provides a method for
establishing and implementing a QA
program for the design and construction
of nuclear power plants and fuel
reprocessing plants; however, NQA–1
does not meet all of the requirements of
appendix B to 10 CFR part 50. Section
50.55a(b)(3)(i) clarifies that licensees
using NQA–1 are also required to meet
appendix B to 10 CFR part 50 and the
commitments contained in their QA
program descriptions. To meet the
requirements of appendix B, licensees
must address, in their quality program
description, those areas where NQA–1
is insufficient to meet appendix B.
Additional guidance and regulatory
positions on how to meet appendix B
when using NQA–1 are provided in RG
1.28.
10 CFR 50.55a(b)(3)(ii) OM Condition:
Motor-Operated Valve (MOV) Testing
The NRC is revising § 50.55a(b)(3)(ii)
to set forth four conditions on the use
of mandatory Appendix III, ‘‘Preservice
and Inservice Testing of Active Electric
Motor Operated Valve Assemblies in
Light-Water Reactor Power Plants,’’ in
the OM Code, 2009 Edition, 2011
Addenda, and 2012 Edition. The four
conditions, which are set forth in
paragraphs (b)(3)(ii)(A) through (D), are
discussed in the next four headings.
10 CFR 50.55a(b)(3)(ii)(A) MOV
Diagnostic Test Interval
The NRC is adding
§ 50.55a(b)(3)(ii)(A) to require that
licensees evaluate the adequacy of the
diagnostic test intervals established for
MOVs within the scope of OM Code,
Appendix III, not later than 5 years or
three refueling outages (whichever is
longer) from initial implementation of
OM Code, Appendix III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing
Impact on Risk
The NRC is adding
§ 50.55a(b)(3)(ii)(B) to require that
licensees ensure that the potential
increase in CDF and LERF associated
with the extension is acceptably small
when extending exercise test intervals
for high risk MOVs beyond a quarterly
frequency. As specified in RG 1.192,
when extending exercise test intervals
for high risk MOVs beyond a quarterly
frequency, licensees must ensure that
the potential increase in CDF and risk
associated with the extension is small
and consistent with the intent of the
Commission’s Safety Goal Policy
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Statement. As discussed earlier in
Section II, the NRC provides guidance
in RG 1.174 that acceptably small
changes are relative and depend on the
current plant CDF and LERF. For plants
with total baseline CDF of 10¥4 per year
or less, acceptably small means CDF
increases of up to 10¥5 per year; and for
plants with total baseline CDF greater
than 10¥4 per year, acceptably small
means CDF increases of up to 10¥6 per
year. For plants with total baseline
LERF of 10¥5 per year or less,
acceptably small LERF increases are
considered to be up to 10¥6 per year;
and for plants with total baseline LERF
greater than 10¥5 per year, acceptably
small LERF increases are considered to
be up to 10¥7 per year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk
Categorization
The NRC is adding
§ 50.55a(b)(3)(ii)(C) to require, when
applying Appendix III to the OM Code,
that licensees categorize MOVs
according to their safety significance
using the methodology described in OM
Code Case OMN–3 subject to the
conditions discussed in RG 1.192, or
using an MOV risk ranking methodology
accepted by the NRC on a plant-specific
or industry-wide basis in accordance
with the conditions in the applicable
safety evaluation.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke
Time
The NRC is adding
§ 50.55a(b)(3)(ii)(D) to require, when
applying Paragraph III–3600, ‘‘MOV
Exercising Requirements,’’ of Appendix
III to the OM Code, licensees shall verify
that the stroke time of MOVs specified
in plant technical specifications satisfies
the assumptions in the plant’s safety
analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition:
New Reactors
The NRC is adding § 50.55a(b)(3)(iii)
to specify that, in addition to complying
with the provisions in the OM Code as
required with the conditions specified
in § 50.55a(b)(3), holders of operating
licenses for nuclear power reactors that
received construction permits under
this part on or after the date 12 months
after August 17, 2017, and holders of
COLs issued under 10 CFR part 52,
whose initial fuel loading occurs on or
after the date 12 months after August 17,
2017, shall also comply with four
condition on power-operated valves,
check valves, flow-induced vibration,
and operational readiness of high-risk
non-safety systems, to the extent
applicable. These four conditions,
which are set forth in
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§ 50.55a(b)(3)(iii)(A), (B), (C), and (D),
are discussed in the next four headings.
10 CFR 50.55a(b)(3)(iii)(A) PowerOperated Valves (First Condition on
New Reactors)
The NRC is adding
§ 50.55a(b)(3)(iii)(A) to require that
licensees subject to § 50.55a(b)(3)(iii)
periodically verify the capability of
power-operated valves (POVs) to
perform their design-basis safety
functions.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
(Second Condition on New Reactors)
The NRC is adding
§ 50.55a(b)(3)(iii)(B) to require that
licensees subject to § 50.55a(b)(3)(iii)
perform bi-directional testing of check
valves within the IST program where
practicable.
10 CFR 50.55a(b)(3)(iii)(C) FlowInduced Vibration (Third Condition on
New Reactors)
The NRC is adding
§ 50.55a(b)(3)(iii)(C) to require that
licensees subject to § 50.55a(b)(3)(iii)
monitor flow-induced vibration (FIV)
from hydrodynamic loads and acoustic
resonance during preservice testing or
inservice testing to identify potential
adverse flow effects that might impact
components within the scope of the IST
program.
10 CFR 50.55a(b)(3)(iii)(D) High Risk
Non-Safety Systems (Fourth Condition
on New Reactors)
The NRC is adding
§ 50.55a(b)(3)(iii)(D) to require that
licensees subject to § 50.55a(b)(3)(iii)
establish a program to assess the
operational readiness of pumps, valves,
and dynamic restraints within the scope
of the Regulatory Treatment of NonSafety Systems for applicable reactor
designs. As of the time of this final rule,
these are designs which have been
certified in a design certification rule
under 10 CFR part 52. This final rule
refers to these RTNSS components using
the term, ‘‘high risk non-safety
systems.’’
As noted by the public commenters,
ASME is preparing guidance for new
reactor licensees to use in developing
programs for the treatment of RTNSS
equipment. The NRC staff is
participating on the OM Code
committees to assist in developing
guidance for the treatment of RTNSS
equipment that is consistent with
Commission policy. Guidance on the
implementation of the Commission
policy for RTNSS equipment is set forth
in NRC Inspection Procedure 73758,
‘‘Part 52, Functional Design and
Qualification, and Preservice and
Inservice Testing Programs for Pumps,
Valves and Dynamic Restraints,’’ dated
April 19, 2013.
10 CFR 50.55a(b)(3)(iv) OM Condition:
Check Valves (Appendix II)
The NRC is revising § 50.55a(b)(3)(iv)
to extend the existing conditions on the
use of Appendix II to the new Editions
and Addenda which are the subject of
this rulemaking. These conditions are
that: (i) Trending and evaluation shall
support the determination that the valve
or group of valves is capable of
performing its intended function(s) over
the entire interval; and (ii) at least one
of the Appendix II condition monitoring
activities for a valve group shall be
performed on each valve of the group at
approximate equal intervals not to
exceed the maximum interval shown in
the following table:
MAXIMUM INTERVALS FOR USE WHEN APPLYING INTERVAL EXTENSIONS
Maximum
interval
between
activities of
member valves
in the groups
(years)
Group size
≥4 .....................................................................................................................................................................
3 .......................................................................................................................................................................
2 .......................................................................................................................................................................
1 .......................................................................................................................................................................
The conditions currently specified for
the use of Appendix II, 1995 Edition
with the 1996 and 1997 Addenda, and
1998 Edition through the 2002
Addenda, of the OM Code remain the
same in this final rule.
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10 CFR 50.55a(b)(3)(vii) OM Condition:
Subsection ISTB
The NRC is adding § 50.55a(b)(3)(vii)
to prohibit the use of Subsection ISTB
in the 2011 Addenda to the OM Code.
10 CFR 50.55a(b)(3)(viii) OM Condition:
Subsection ISTE
The NRC is adding § 50.55a(b)(3)(viii)
to specify that licensees who wish to
implement Subsection ISTE, ‘‘RiskInformed Inservice Testing of
Components in Light-Water Reactor
Nuclear Power Plants,’’ of the OM Code,
2009 Edition, 2011 Addenda, and 2012
Edition, must first request and obtain
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NRC approval in accordance with
§ 50.55a(z) to apply Subsection ISTE on
a plant-specific basis as a risk-informed
alternative to the applicable IST
requirements in the OM Code.
The NRC will evaluate § 50.55a(z)
requests for approval to implement
Subsection ISTE in accordance with the
following considerations. These
considerations are consistent with the
guidance provided in RG 1.174.
1. Scope of Risk-Informed IST Program
Subsection ISTE–1100,
‘‘Applicability,’’ establishes the
component safety categorization
methodology and process for dividing
the population of pumps and valves, as
identified in the IST Program Plan, into
high safety significant component
(HSSC) and low safety significant
component (LSSC) categories. When
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Maximum
interval
between
activities of
each valve
in the group
(years)
4.5
4.5
6
Not applicable
16
12
12
10
establishing a risk-informed IST
program, the licensee should address a
wide range of components important to
safety at the nuclear power plant that
includes both safety-related and
nonsafety-related components. These
components might extend beyond the
scope of the OM Code.
2. Risk-Ranking Methodology
The licensee should specify, in its
request for authorization to implement a
risk-informed IST program, the
methodology to be applied in risk
ranking its components. ISTE–4000,
‘‘Specific Component Categorization
Requirements,’’ incorporates OM Code
Case OMN–3 for the categorization of
pumps and valves in developing a riskinformed IST program. The OMN–3
Code Case methodology for risk ranking
uses two categories of safety
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significance. The NRC staff has also
accepted other methodologies for risk
ranking that use three categories of
safety significance.
3. Safety Significance Categorization
The licensee should categorize
components according to their safety
significance based on the methodology
described in Subsection ISTE with the
applicable conditions on the use of OM
Code Case OMN–3 specified in RG
1.192, or use other risk ranking
methodologies accepted by the NRC on
a plant-specific or industry-wide basis
with applicable conditions specified by
the NRC for their acceptance. The
licensee should address the seven
conditions in RG 1.192 for the use of
OM Code Case OMN–3, as appropriate,
in developing the risk-informed IST
program described in Subsection ISTE.
With respect to the provisions in
Subsection ISTE, these conditions are:
(a) The implementation of ISTE–1100
should include within the scope of a
licensee’s risk-informed IST program
non-ASME OM Code pumps and valves
categorized as HSSCs that might not
currently be included in the IST
program at the nuclear power plant.
(b) The decision criteria discussed in
ISTE–4410, ‘‘Decision Criteria,’’ and
Nonmandatory Appendix L,
‘‘Acceptance Guidelines,’’ of the OM
Code for evaluating the acceptability of
aggregate risk effects (i.e., for CDF and
LERF) should be consistent with the
guidance provided in RG 1.174, ‘‘An
Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the
Licensing Basis.’’
(c) The implementation of ISTE–4440,
‘‘Defense in Depth,’’ should be
consistent with the guidance contained
in Section 2.2.1, ‘‘Defense-in-Depth
Evaluation,’’ and Section 2.2.2, ‘‘Safety
Margin Evaluation,’’ of RG 1.175, ‘‘An
Approach for Plant-Specific, RiskInformed Decisionmaking: Inservice
Testing.’’
(d) The implementation of ISTE–4500,
‘‘Inservice Testing Program,’’ and ISTE–
6100, ‘‘Performance Monitoring,’’
should be consistent with the guidance
contained in Section 3.2, ‘‘Program
Implementation,’’ and Section 3.3,
‘‘Performance Monitoring,’’ of RG 1.175.
(e) The implementation of ISTE–3210,
‘‘Plant-Specific PRA,’’ should be
consistent with the guidance that the
Owner is responsible for demonstrating
and justifying the technical adequacy of
the PRA analyses used as the basis to
perform component risk ranking and for
estimating the aggregate risk impact. For
example, RG 1.200, ‘‘An Approach for
Determining the Technical Adequacy of
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Probabilistic Risk Assessment Results
for Risk-Informed Activities,’’ and RG
1.201, ‘‘Guidelines for Categorizing
Structures, Systems, and Components in
Nuclear Power Plants According to their
Safety Significance,’’ provide guidance
for PRA technical adequacy and
component risk ranking.
(f) The implementation of ISTE–4240,
‘‘Reconciliation,’’ should specify that
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 defensein-depth assessment to LSSC.
(g) The implementation of ISTE–3220,
‘‘Living PRA,’’ should be consistent
with the following: (i) 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;
(ii) 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 (iii) the review of the plant
changes must be performed in a timely
manner and must be performed once
every two refueling outages, or as
required by § 50.71(h)(2) for COL
holders.
4. Pump Testing
Subsection ISTE–5100, ‘‘Pumps,’’
incorporates OM Code Case OMN–7 for
risk-informed testing of pumps
categorized as LSSCs. Subsection ISTE–
5100 allows the interval for Group A
and Group B testing of LSSC pumps
specified in Subsection ISTB of the OM
Code to be extended from the current 3month interval to intervals of 6 months
or 2 years. Subsection ISTE–5100
eliminates the requirement in
Subsection ISTB to perform
comprehensive pump testing for LSSC
pumps. Table ISTE–5121–1, ‘‘LSSC
Pump Testing,’’ specifies that pump
operation may be required more
frequently than the specified test
frequency (6 months) to meet vendor
recommendations. Subsection ISTE–
4500, ‘‘Inservice Testing Program,’’
specifies in ISTE–4510, ‘‘Maximum
Testing Interval,’’ that the maximum
testing interval shall be based on the
more limiting of (a) the results of the
aggregate risk, or (b) the performance
history of the component. ISTE–5130,
‘‘Maximum Test Interval—Pre-2000
Plants,’’ specifies that the most limiting
interval for LSSC pump testing shall be
determined from ISTE–4510 and ISTE–
5120, ‘‘Low Safety Significant Pump
Testing.’’ The ASME developed the
comprehensive pump test requirements
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in the OM Code to address weaknesses
in the Code requirements to assess the
operational readiness of pumps to
perform their design-basis safety
function. Therefore, the licensee should
ensure that testing under Subsection
ISTE will provide assurance of the
operational readiness of pumps in each
safety significant categorization to
perform their design-basis safety
function as described in RGs 1.174 and
1.175.
5. Motor-Operated Valve Testing
Subsection ISTE–5300, ‘‘Motor
Operated Valve Assemblies,’’ provides a
risk-informed IST approach instead of
the IST requirements for MOVs in
Mandatory Appendix III to the OM
Code. The ASME prepared Appendix III
to the OM Code to replace the
requirement for quarterly stroke-time
testing of MOVs with a program of
periodic exercising and diagnostic
testing to address lessons learned from
nuclear power plant operating
experience and industry and regulatory
research programs for MOV
performance. Subsection ISTC of the
OM Code specifies the implementation
of Appendix III for periodic exercising
and diagnostic testing of MOVs to
replace quarterly stroke-time testing
previously required for MOVs.
Appendix III incorporates provisions
that allow a risk-informed IST approach
for MOVs as described in OM Code
Cases OMN–1 and OMN–11. Subsection
ISTE–5300 is not consistent with the
provisions for the risk-informed IST
program for MOVs specified in
Appendix III to the OM Code (and Code
Cases OMN–1 and 11). Therefore,
licensees who wish to implement
Subsection ISTE should address the
provisions in paragraph III–3700, ‘‘RiskInformed MOV Inservice Testing,’’ of
Appendix III to the OM Code as
incorporated by reference in § 50.55a,
with the applicable conditions, instead
of ISTE–5300.
6. Pneumatically and Hydraulically
Operated Valve Testing
Subsection ISTE–5400,
‘‘Pneumatically and Hydraulically
Operated Valves,’’ specifies that
licensees test their AOVs and HOVs in
accordance with Appendix IV to the OM
Code. Subsection ISTE–5400 indicates
that Appendix IV is in the course of
preparation. The NRC staff will need to
review Appendix IV prior to accepting
its use as part of Subsection ISTE.
Therefore, licensees who wish to
implement Subsection ISTE should
describe the planned IST provisions for
AOVs and HOVs in its request for
approval to implement Subsection ISTE.
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7. Pump Periodic Verification Test
Subsection ISTE does not include a
requirement to implement the pump
periodic verification test program
specified in Mandatory Appendix V to
the OM Code, 2012 Edition. Therefore,
licensee should address the
consideration of a pump periodic
verification test program in its riskinformed IST program, proposed as part
of the authorization request to
implement Subsection ISTE.
10 CFR 50.55a(b)(3)(ix) OM Condition:
Subsection ISTF
The NRC is adding § 50.55a(b)(3)(ix)
to specify that licensees applying
Subsection ISTF, ‘‘Inservice Testing of
Pumps in Light-Water Reactor Nuclear
Power Plants—Post-2000 Plants,’’ in the
2012 Edition of the OM Code shall
satisfy the requirements of Mandatory
Appendix V, ‘‘Pump Periodic
Verification Test Program,’’ of the OM
Code, 2012 Edition. The paragraph also
states that Subsection ISTF, 2011
Addenda, is not acceptable for use.
10 CFR 50.55a(b)(3)(x) OM Condition:
ASME OM Code Case OMN–20
The NRC is adding § 50.55a(b)(3)(x) to
allow licensees to implement OM Code
Case OMN–20, ‘‘Inservice Test
Frequency,’’ in the OM Code, 2012
Edition, for the editions and addenda of
the OM Code that are listed in
§ 50.55a(a)(1)(iv).
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10 CFR 50.55a(b)(3)(xi) OM Condition:
Valve Position Indication
The NRC is adding § 50.55a(b)(3)(xi)
to emphasize the provisions in the OM
Code, 2012 Edition, Subsection ISTC–
3700, ‘‘Position Verification Testing,’’ to
verify that valve obturator position is
accurately indicated. The OM Code,
Subsection ISTC–3700 requires valves
with remote position indicators shall be
observed locally at least once every 2
years to verify that valve operation is
accurately indicated. Licensees will be
required to implement the condition
when adopting the 2012 Edition of the
OM Code as their Code of Record for the
applicable 120-month IST interval.
10 CFR 50.55a(f) Preservice and
Inservice Testing Requirements
The NRC is revising the heading for
§ 50.55a(f) and clarifying that the OM
Code includes provisions for preservice
testing of components as part of its
overall provisions for IST programs.
10 CFR 50.55a(f)(4) Inservice Testing
Standards Requirement for Operating
Plants
The NRC is revising § 50.55a(f)(4) to
ensure that the paragraph is applicable
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to pumps and valves that are within the
scope of the OM Code. The NRC is also
including an additional provision in
§ 50.55a(f)(4) stating that the IST
requirements for pumps and valves that
are within the scope of the OM Code but
are not classified as ASME BPV Code
Class 1, Class 2, or Class 3 may be
satisfied as an augmented IST program,
in accordance with § 50.55a(f)(6)(ii),
without requesting relief under
§ 50.55a(f)(5) or alternatives under
§ 50.55a(z). This use of an augmented
IST program may be acceptable
provided the basis for deviations from
the OM Code, as incorporated by
reference in this section, demonstrates
an acceptable level of quality and safety,
or that implementing the Code
provisions would result in hardship or
unusual difficulty without a
compensating increase in the level of
quality and safety, where documented
and available for NRC review. These
changes align the scope of pumps and
valves for inservice testing with the
scope defined in the OM Code without
imposing an unnecessary paperwork
burden on nuclear power plant
licensees for the submittal of relief and
alternative requests for pumps and
valves within the scope of the OM Code
but not classified as ASME BPV Code
Class 1, Class 2, or Class 3 components.
10 CFR 50.55a(g) Preservice and
Inservice Inspection Requirements
The NRC is revising the heading in
§ 50.55a(g), adding new paragraphs
(g)(2)(i), (ii), and (iii), and revising
current paragraphs (g) introductory text,
(g)(2), (g)(3) introductory text, and
(g)(3)(i), (ii), and (v) to distinguish the
requirements for accessibility,
preservice examination, and inservice
inspection. No substantive change to the
requirements is intended by these
revisions.
10 CFR 50.55a(g)(4) Inservice Inspection
Standards Requirement for Operating
Plants
The NRC is revising § 50.55a(g)(4)(ii)
to add an implementation period of 18months for licensees whose ISI interval
commences during the 12 through 18month period after the publication of
this final rule. The NRC is also revising
§ 50.55a(g)(4)(i) and (ii) to add a
provision allowing licensees to adopt
the latest version of Appendix VIII of
the ASME BPV Code edition or addenda
listed in § 50.55a(a)(1) at any time in the
licensee’s 120-month ISI interval.
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10 CFR 50.55a(g)(6)(ii)(D) Augmented
ISI Requirements: Reactor Vessel Head
Inspections
The NRC is revising
§ 50.55a(g)(6)(ii)(D) to reflect the NRC’s
approval of ASME BPV Code Case N–
729–4, which supersedes the NRC’s
earlier approval of ASME BPV Code
Case N–729–1. The revisions include
changes to the conditions governing the
use of the Code Case to reflect the
change from N–729–1 to N–729–4. The
effect of these changes is to require
licensees to implement an augmented
ISI program for the examination of the
pressurized water reactor RPV upper
head penetrations. The following
discussions provide a more detailed
discussion of the revisions to
§ 50.55a(g)(6)(ii)(D).
10 CFR 50.55a(g)(6)(ii)(D)(1)
Implementation
The NRC is revising
§ 50.55a(g)(6)(ii)(D)(1) to require
licensees to implement an augmented
ISI program for the examination of the
pressurized water reactor RPV upper
head penetrations meeting ASME BPV
Code Case N–729–4 instead of the
previously approved requirements to
use ASME BPV Code Case N–729–1, as
conditioned by the NRC.
Removal of Existing Conditions in 10
CFR 50.55a(g)(6)(ii)(D)(2) Through (5)
The NRC is removing the existing
conditions in § 50.55a(g)(6)(ii)(D)(2)
through (5) and redesignating the
existing condition in
§ 50.55a(g)(6)(ii)(D)(6) as
§ 50.55a(g)(6)(ii)(D)(2).
10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I
Use
The NRC is revising the existing
condition in § 50.55a(g)(6)(ii)(D)(6),
which is redesignated as
§ 50.55a(g)(6)(ii)(D)(2) in this final rule,
to require NRC approval prior to
implementing Appendix I of ASME BPV
Code Case N–729–4.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal
Visual Frequency
The NRC is adding a new condition
in § 50.55a(g)(6)(ii)(D)(3) which requires
cold head plants with less than eight
effective degradation years (EDY<8)
without PWSCC flaws to perform a bare
metal visual examination (VE) each
outage a volumetric exam is not
performed and allows these plants to
extend the bare metal visual inspection
frequency from once each refueling
outage, as stated in Table 1 of N–729–
4, to once every 5 years, only if the
licensee performed a wetted surface
examination of all of the partial
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penetration welds during the previous
volumetric examination. In addition,
this new condition clarifies that a bare
metal visual examination is not required
during refueling outages when a
volumetric or surface examination is
performed of the partial penetration
welds.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface
Exam Acceptance Criteria
The NRC is adding a new condition
in § 50.55a(g)(6)(ii)(D)(4) clarifying that
rounded indications found by surface
examinations of the partial-penetration
or associated fillet welds in accordance
with N–729–4 must meet the acceptance
criteria for surface examinations of
paragraph NB–5352 of ASME 2013
Edition of Section III for the licensee’s
ongoing 10-year ISI interval.
10 CFR 50.55a(g)(6)(ii)(F)(1)
Implementation
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(1) to require
licensees to implement an augmented
ISI program for the examination of
ASME Class 1 piping and nozzle butt
welds meeting ASME BPV Code Case
N–770–2 instead of the previously
approved ASME BPV Code Case N–770–
1.
Furthermore, the NRC is revising
§ 50.55a(g)(6)(ii)(F)(1) to update the date
of applicability for pressurized water
reactors, to note the change to
implement ASME BPV Code Case N–
770–2 instead of N–770–1, and to reflect
the number of conditions which must be
applied.
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10 CFR 50.55a(g)(6)(ii)(F)(2)
Categorization
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(2) to clarify the
requirements for licensees to establish
the initial categorization of each weld
and modify the wording to reflect the
ASME BPV Code Case N–770–2 change
in the inspection item category for full
structural weld overlays (C to C–1 and
F to F–1). Additionally, the NRC is
adding a sentence which clarifies the
NRC position that Paragraph –1100(e) of
ASME BPV Code Case N–770–2 shall
not be used to exempt welds that rely
on Alloy 82/182 for structural integrity
from any requirement of
§ 50.55a(g)(6)(ii)(F).
10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline
Examinations
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(3) to clarify the
current requirement in this paragraph to
complete baseline examinations by
stating that previously-conducted
examinations, in order to count as
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baseline examinations, must meet the
requirements of ASME BPV Code Case
N–770–2, as conditioned in this section.
Additionally, this condition clarifies
that the examination coverage
requirements, for a licensee to count
previous inspections as baseline
examinations, must meet the
examination coverage requirements
described in Paragraphs –2500(a) or
–2500(b) of ASME BPV Code Case N–
770–2, as conditioned by the NRC in
this section. Upon implementation of
this rule, if a licensee is currently in an
outage, then the baseline inspection
requirement can be met by performing
the inspections in accordance with the
previous regulatory requirements of
§ 50.55a(g)(6)(ii)(F), in lieu of the
examination requirements of Paragraphs
–2500(a) or –2500(b) of ASME BPV
Code Case N–770–2.
10 CFR 50.55a(g)(6)(ii)(F)(4)
Examination Coverage
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(4) to clarify that
when licensees are implementing
paragraph –2500(a) of ASME BPV Code
Case N–770–2, essentially 100 percent
of the required volumetric examination
coverage shall be obtained, including
greater than 90 percent volumetric
examination coverage is obtained for
circumferential flaws, to continue the
restriction on the licensee’s use of
Paragraph –2500(c) and to continue the
restriction that the use of new Paragraph
–2500(d) of ASME BPV Code Case N–
770–2 is not allowed without prior NRC
review and approval in accordance with
§ 50.55a(z), as it would permit a
reduction in volumetric examination
coverage for circumferential flaws.
However, a licensee may request
approval for use of these paragraphs
under § 50.55a(z), and the NRC may
approve the request if technically
justified.
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay
Inspection Frequency
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(5) to add an
explanatory heading, ‘‘Inlay/onlay
inspection frequency,’’ and to make
minor editorial corrections without
substantive changes in the requirement.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting
Requirements
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(6) to add an
explanatory heading, ‘‘Reporting
requirements.’’
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10 CFR 50.55a(g)(6)(ii)(F)(7) Defining
‘‘t’’
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(7) to add an
explanatory heading, ‘‘Defining ‘t’.’’
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized
Weld Overlay Examination
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(8) to add an
explanatory heading, ‘‘Optimized weld
overlay examination,’’ and to continue
the current condition located in
§ 50.55a(g)(6)(ii)(F)(9) which requires
that the initial examination of optimized
weld overlays (i.e., Inspection Item C–2
of ASME BPV Code Case N–770–2) be
performed between the third refueling
outage and no later than 10 years after
application of the overlay and delete the
other current examination requirements
for optimized weld overlay examination
frequency, as these requirements were
included in the revision from N–770–1
to N–770–2.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(9) to add an
explanatory heading, ‘‘Deferral,’’ and to
modify the current condition to
continue denial of the deferral of the
initial inservice examination of
uncracked welds mitigated by
optimized weld overlays. These welds
shall continue to have their initial
inservice examinations as prescribed in
N–770–1 within 10 years of the
application of the optimized weld
overlay and not allow deferral of this
initial examination. Subsequent
inservice examinations may be deferred
as allowed by N–770–2. Additionally,
the modified condition will delete the
current condition on examination
requirements for the deferral of welds
mitigated by inlay, onlay, stress
improvement and optimized weld
overlay, as these requirements were,
with one exception (i.e., optimized weld
overlay), included in the revision from
N–770–1 to N–770–2.
10 CFR 50.55a(g)(6)(ii)(F)(10)
Examination Technique
The NRC is revising
§ 50.55a(g)(6)(ii)(F)(10) to add an
explanatory heading, ‘‘Examination
technique,’’ and to modify the current
condition to allow the previously
prohibited alternate examination
requirements of Note (b) of Figure 5(a)
of ASME BPV Code Cases N–770–1 and
N–770–2 and the same requirements in
Note 14(b) of Table 1 of ASME BPV
Code Case N–770–2 for optimized weld
overlays only if the full examination
requirements of Note 14(a) of Table 1 of
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ASME BPV Code Case N–770–2 cannot
be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast
Stainless Steel
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(11) to provide a new
condition requiring licensees to
establish a Section XI, Appendix VIII,
qualification requirement for ultrasonic
inspection of cast stainless steel and
through cast stainless steel to meet the
examination requirements of Paragraph
–2500(a) of ASME BPV Code Case N–
770–2 by January 1, 2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress
Improvement Inspection Coverage
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(12) to provide a new
condition that would allow licensees to
implement a stress improvement
mitigation technique for items
containing cast stainless steel that
would meet the requirements of
Appendix I of ASME BPV Code Case N–
770–2, if the required examination
volume can be examined by Appendix
VIII procedures to the maximum extent
practical including 100 percent of the
susceptible material volume.
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded
Ultrasonic Examination
The NRC is adding
§ 50.55a(g)(6)(ii)(F)(13) to provide a new
condition requiring licensees to perform
encoded examinations of 100 percent of
the required inspection volume when
required to perform volumetric
examinations of all non-mitigated and
cracked mitigated butt welds in the
reactor coolant pressure boundary in
accordance with ASME BPV Code Case
N–770–2.
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VI. Generic Aging Lessons Learned
Report
Background
In December 2010, the NRC issued
NUREG–1801, Revision 2, for applicants
to use in preparing their license renewal
applications. The GALL Report provides
aging management programs (AMPs)
that the NRC staff has concluded are
sufficient for aging management in
accordance with the license renewal
rule, as required in § 54.21(a)(3). In
addition, NUREG–1800, Revision 2,
‘‘Standard Review Plan for Review of
License Renewal Applications for
Nuclear Power Plants,’’ was issued in
December 2010 to ensure the quality
and uniformity of NRC staff reviews of
license renewal applications and to
present a well-defined basis on which
the NRC staff evaluates the applicant’s
aging management programs and
activities. In April 2011, the NRC issued
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NUREG–1950, ‘‘Disposition of Public
Comments and Technical Bases for
Changes in the License Renewal
Guidance Documents NUREG–1801 and
NUREG–1800,’’ which describes the
technical bases for the changes in
Revision 2 of the GALL Report and
Revision 2 of the SRP for review of
license renewal applications. Revision 2
of the GALL Report, in Sections XI.M1,
XI.S1, XI.S2, and XI.S3, describes the
evaluation and technical bases for
determining the sufficiency of ASME
BPV Code Subsections IWB, IWC, IWD,
IWE, IWF, and IWL for managing aging
during the period of extended operation.
In addition, many other AMPs in the
GALL Report rely, in part but to a lesser
degree, on the requirements specified in
the ASME BPV Code, Section XI.
Revision 2 of the GALL Report also
states that the 1995 Edition through the
2004 Edition of the ASME BPV Code,
Section XI, Subsections IWB, IWC, IWD,
IWE, IWF, and IWL, as modified and
limited by § 50.55a, were found to be
acceptable editions and addenda for
complying with the requirements of
§ 54.21(a)(3), unless specifically noted
in certain sections of the GALL Report.
The GALL Report further states that the
future Federal Register notices that
amend § 50.55a will discuss the
acceptability of editions and addenda
more recent than the 2004 edition for
their applicability to license renewal.
In a final rule issued on June 21, 2011
(76 FR 36232), subsequent to Revision 2
of the GALL Report, the NRC found that
the 2004 Edition with the 2005
Addenda through the 2007 Edition with
the 2008 Addenda of Section XI of the
ASME BPV Code, Subsections IWB,
IWC, IWD, IWE, IWF, and IWL, as
subject to the conditions in § 50.55a, are
acceptable for the AMPs in the GALL
Report and the conclusions of the GALL
Report remain valid with the
augmentations specifically noted in the
GALL Report.
Evaluation With Respect to Aging
Management
As part of this rulemaking, the NRC
evaluated whether those AMPs in
Revision 2 of the GALL Report which
rely upon Subsections IWB, IWC, IWD,
IWE, IWF, and IWL of Section XI in the
editions and addenda of the ASME BPV
Code incorporated by reference into
§ 50.55a, continue to be acceptable if the
AMP relies upon the versions of these
Subsections in the 2007 Edition with
the 2009 Addenda through the 2013
Edition. The NRC finds that the 2007
Edition with the 2009 Addenda through
the 2013 Edition of Section XI of the
ASME BPV Code, Subsections IWB,
IWC, IWD, IWE, IWF, and IWL, as
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subject to the conditions of this rule, are
acceptable for the AMPs in the GALL
Report and the conclusions of the GALL
Report remain valid with the
augmentations specifically noted in the
GALL Report. Accordingly, an applicant
for license renewal may use, in its plantspecific license renewal application,
Subsections IWB, IWC, IWD, IWE, IWF,
and IWL of Section XI of the 2007
Edition with the 2009 Addenda through
the 2013 Edition of the ASME BPV
Code, as subject to the conditions in this
rule, without additional justification.
Similarly, a licensee approved for
license renewal that relied on the GALL
AMPs may use Subsections IWB, IWC,
IWD, IWE, IWF, and IWL of Section XI
of the 2007 Edition with the 2009
Addenda through the 2013 Edition of
the ASME BPV Code. However, a
licensee must assess and follow
applicable NRC requirements with
regard to changes to its licensing basis.
Some of the AMPs in the GALL Report
recommend augmentation of certain
Code requirements in order to ensure
adequate aging management for license
renewal. The technical and regulatory
aspects of the AMPs for which
augmentations are recommended also
apply if the editions or addenda from
the 2007 Edition with the 2009
Addenda through the 2013 Edition of
Section XI of the ASME BPV Code are
used to meet the requirements of
§ 54.21(a)(3). The NRC staff evaluated
the changes in the 2007 Edition with the
2009 Addenda through the 2013 Edition
of Section XI of the ASME BPV Code to
determine if the augmentations
described in the GALL Report remain
necessary. The NRC staff’s evaluation
has concluded that the augmentations
described in the GALL Report are
necessary to ensure adequate aging
management. For example, Table IWB–
2500–1, in the 2007 Edition with the
2009 Addenda of ASME BPV Code,
Section XI, Subsection IWB, requires
surface examination of ASME BPV Code
Class 1 branch pipe connection welds
less than nominal pipe size (NPS) 4
under Examination Category B–J.
However, the NRC staff finds that
volumetric or opportunistic destructive
examination, rather than surface
examination, is necessary to adequately
detect and manage the aging effect due
to stress corrosion cracking or thermal,
mechanical and vibratory loadings in
the components for the period of
extended operation. Therefore, GALL
Report Section XI.M35, ‘‘One-Time
Inspection of ASME BPV Code Class 1
Small-Bore Piping,’’ includes the
augmentation of the requirements in
ASME BPV Code, Section XI,
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Subsection IWB to perform a one-time
inspection of a sample of ASME BPV
Code Class 1 piping less than NPS 4 and
greater than or equal to NPS 1 using
volumetric or opportunistic destructive
examination. The GALL Report
addresses this augmentation to confirm
that there is no need to manage agerelated degradation through periodic
volumetric inspections or that an
existing AMP (for example, Water
Chemistry AMP) is effective to manage
the aging effect due to stress corrosion
cracking or thermal, mechanical and
vibratory loadings for the period of
extended operation. A license renewal
applicant may either augment its AMPs
as described in the GALL Report, or
propose alternatives for the NRC to
review as part of the applicant’s plantspecific justification for its AMPs.
VII. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act
(5 U.S.C. 605(b)), the NRC certifies that
this rule does not have a significant
economic impact on a substantial
number of small entities. This final rule
affects only the licensing and operation
of nuclear power plants. The companies
that own these plants do not fall within
the scope of the definition of ‘‘small
entities’’ set forth in the Regulatory
Flexibility Act or the size standards
established by the NRC (§ 2.810).
VIII. Regulatory Analysis
The NRC has prepared a final
regulatory analysis on this regulation.
The analysis examines the costs and
benefits of the alternatives considered
by the NRC. The regulatory analysis is
available as indicated in the
‘‘Availability of Documents’’ section of
this document.
IX. Backfitting and Issue Finality
sradovich on DSK3GMQ082PROD with RULES2
Introduction
The NRC’s Backfit Rule in § 50.109
states that the NRC shall require the
backfitting of a facility only when it
finds the action to be justified under
specific standards stated in the rule.
Section 50.109(a)(1) defines backfitting
as the modification of or addition to
systems, structures, components, or
design of a facility; the design approval
or manufacturing license for a facility;
or the procedures or organization
required to design, construct, or operate
a facility. Any of these modifications or
additions may result from a new or
amended provision in the NRC’s rules
or the imposition of a regulatory
position interpreting the NRC’s rules
that is either new or different from a
previously applicable NRC position
after issuance of the construction permit
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or the operating license or the design
approval.
Section 50.55a requires nuclear power
plant licensees to:
• Construct ASME BPV Code Class 1,
2, and 3 components in accordance with
the rules provided in Section III,
Division 1, of the ASME BPV Code
(‘‘Section III’’).
• Inspect Class 1, 2, 3, Class MC, and
Class CC components in accordance
with the rules provided in Section XI,
Division 1, of the ASME BPV Code
(‘‘Section XI’’).
• Test Class 1, 2, and 3 pumps,
valves, and dynamic restraints
(snubbers) in accordance with the rules
provided in the OM Code.
This final rule is incorporating by
reference the 2009 Addenda, 2010
Edition, 2011 Addenda, and the 2013
Edition of the ASME BPV Code, Section
III, Division 1 and ASME BPV Code,
Section XI, Division 1, including NQA–
1 (with conditions on its use), as well
as the 2009 Edition and 2011 Addenda
and 2012 Edition of the OM Code and
Code Cases N–770–2 and N–729–4.
The ASME BPV and OM Codes are
national consensus standards developed
by participants with broad and varied
interests, in which all interested parties
(including the NRC and utilities)
participate. A consensus process
involving a wide range of stakeholders
is consistent with the NTTAA,
inasmuch as the NRC has determined
that there are sound regulatory reasons
for establishing regulatory requirements
for design, maintenance, ISI, and IST by
rulemaking. The process also facilitates
early stakeholder consideration of
backfitting issues. Therefore, the NRC
believes that the NRC need not address
backfitting with respect to the NRC’s
general practice of incorporating by
reference updated ASME Codes.
Overall Backfitting Considerations:
Section III of the ASME BPV Code
Incorporation by reference of more
recent editions and addenda of Section
III of the ASME BPV Code does not
affect a plant that has received a
construction permit or an operating
license or a design that has been
approved. This is because the edition
and addenda to be used in constructing
a plant are, under § 50.55a, determined
based on the date of the construction
permit, and are not changed thereafter,
except voluntarily by the licensee. The
incorporation by reference of more
recent editions and addenda of Section
III ordinarily applies only to applicants
after the effective date of a final rule
incorporating these new editions and
addenda. Therefore, incorporation by
reference of a more recent edition and
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addenda of Section III does not
constitute ‘‘backfitting’’ as defined in
§ 50.109(a)(1).
Overall Backfitting Considerations:
Section XI of the ASME BPV Code and
the OM Code
Incorporation by reference of more
recent editions and addenda of Section
XI of the ASME BPV Code and the OM
Code affects the ISI and IST programs of
operating reactors. However, the Backfit
Rule generally does not apply to
incorporation by reference of later
editions and addenda of the ASME BPV
Code (Section XI) and OM Code. As
previously mentioned, the NRC’s
longstanding regulatory practice has
been to incorporate later versions of the
ASME Codes into § 50.55a. Under
§ 50.55a, licensees shall revise their ISI
and IST programs every 120 months to
the latest edition and addenda of
Section XI of the ASME BPV Code and
the OM Code incorporated by reference
into § 50.55a 12 months before the start
of a new 120-month ISI and IST
interval. Therefore, when the NRC
approves and requires the use of a later
version of the Code for ISI and IST, it
is implementing this longstanding
regulatory practice and requirement.
Other circumstances where the NRC
does not apply the Backfit Rule to the
approval and requirement to use later
Code editions and addenda are as
follows:
1. When the NRC takes exception to
a later ASME BPV Code or OM Code
provision but merely retains the current
existing requirement, prohibits the use
of the later Code provision, limits the
use of the later Code provision, or
supplements the provisions in a later
Code. The Backfit Rule does not apply
because the NRC is not imposing new
requirements. However, the NRC
explains any such exceptions to the
Code in the statement of considerations
and regulatory analysis for the rule.
2. When an NRC exception relaxes an
existing ASME BPV Code or OM Code
provision but does not prohibit a
licensee from using the existing Code
provision. The Backfit Rule does not
apply because the NRC is not imposing
new requirements.
3. The NRC’s consideration of
backfitting for modifications and
limitations imposed during previous
routine updates of § 50.55a have
established a precedent for determining
the kinds of modifications or limitations
which should be considered backfitting,
or require a backfit analysis (e.g., final
rule dated September 10, 2008 (73 FR
52730), and a correction dated October
2, 2008 (73 FR 57235)). The
consideration of backfitting and issue
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finality with respect to the
modifications and limitations in this
rulemaking are consistent with the
consideration and application of
backfitting and issue finality
requirements to analogous
modifications and limitations in
previous § 50.55a rulemakings.
The incorporation by reference and
adoption of a requirement mandating
the use of a later ASME BPV Code or
OM Code may constitute backfitting in
some circumstances. In these cases, the
NRC would perform a backfit analysis or
documented evaluation in accordance
with § 50.109. These include the
following:
1. When the NRC endorses a later
provision of the ASME BPV Code or OM
Code that takes a substantially different
direction from the existing
requirements, the action is treated as a
backfit (e.g., 61 FR 41303 (August 8,
1996)).
2. When the NRC requires
implementation of a later ASME BPV
Code or OM Code provision on an
expedited basis, the action is treated as
a backfit. This applies when
implementation is required sooner than
it would be required if the NRC simply
endorsed the Code without any
expedited language (e.g., 64 FR 51370
(September 22, 1999)).
3. When the NRC takes an exception
to an ASME BPV Code or OM Code
provision and imposes a requirement
that is substantially different from the
existing requirement as well as
substantially different from the later
Code (e.g., 67 FR 60529 (September 26,
2002)).
sradovich on DSK3GMQ082PROD with RULES2
Detailed Backfitting Discussion:
Changes Beyond Those Necessary To
Incorporate by Reference the New ASME
BPV and OM Code Provisions
This section discusses the backfitting
considerations for all the changes to
§ 50.55a that go beyond the minimum
changes necessary and required to adopt
the new ASME Code Addenda into
§ 50.55a.
ASME BPV Code, Section III
1. Revise § 50.55a(b)(1)(ii), ‘‘Weld leg
dimensions,’’ to clarify rule language
and add Table I, which clarifies
prohibited Section III provisions for
welds with leg sizes less than 1.09 tn in
tabular form. This change does not alter
the original intent of this requirement
and, therefore, does not impose a new
requirement. Therefore, this change is
not a backfit.
2. Revise § 50.55a(b)(1)(iv), ‘‘Quality
assurance,’’ to require that when
applying editions and addenda later
than the 1989 Edition of Section III, the
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requirements of NQA–1, 1994 Edition,
2008 Edition, and the 2009–1a Addenda
are acceptable for use, provided that the
edition and addenda of NQA–1
specified in either NCA–4000 or NCA–
7000 is used in conjunction with the
administrative, quality, and technical
provisions contained in the edition and
addenda of Section III being used. This
revision clarifies the current
requirements, and is considered to be
consistent with the meaning and intent
of the current requirements, and
therefore is not considered to result in
a change in requirements. Therefore,
this change is not a backfit.
3. Add a new condition as
§ 50.55a(b)(1)(viii), ‘‘Use of ASME
Certification Marks,’’ to allow licensees
to use either the ASME BPV Code
Symbol Stamp or ASME Certification
Mark with the appropriate certification
designator and class designator as
specified in the 2013 Edition through
the latest edition and addenda
incorporated by reference in § 50.55a.
This condition does not result in a
change in requirements previously
approved in the Code and, therefore, is
not a backfit.
ASME BPV Code, Section XI
1. Revise § 50.55a(b)(2)(vi), ‘‘Effective
edition and addenda of Subsection IWE
and Subsection IWL,’’ to clarify that the
provision applies only to the class of
licensees of operating reactors that were
required by previous versions of
§ 50.55a to develop, implement a
containment ISI program in accordance
with Subsection IWE and Subsection
IWL, and complete an expedited
examination of containment during the
5-year period from September 9, 1996,
to September 9, 2001. This revision
clarifies the current requirements, is
considered to be consistent with the
meaning and intent of the current
requirements, and is not considered to
result in a change in requirements.
Therefore, this change is not a backfit.
2. Revise § 50.55a(b)(2)(viii),
‘‘Concrete containment examinations,’’
so that when using the 2007 Edition
with 2009 Addenda through the 2013
Edition of Subsection IWL, the
conditions in § 50.55a(b)(2)(viii)(E) do
not apply, but the new conditions in
§ 50.55a(b)(2)(viii)(H) and (I) do apply.
This revision does not require
§ 50.55a(b)(2)(viii)(E) to be used when
following the 2007 Edition with 2009
Addenda through the 2013 Edition of
Subsection IWL because most of its
requirements have been included in
IWL–2512, ‘‘Inaccessible Areas.’’
Therefore, this change is not a backfit
because the requirements have not
changed. The revision to add the
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condition in § 50.55a(b)(2)(viii)(H)
captures the reporting requirements of
the current § 50.55a(b)(2)(viii)(E) which
were not included in IWL–2512.
Therefore, this change is not a backfit
because the requirements have not
changed. The revision to add the
condition in § 50.55a(b)(2)(viii)(I)
addresses a new code provision in IWL–
2512(b) for evaluation of below-grade
concrete surfaces during the period of
extended operation of a renewed
license. The condition assures
consistency with the GALL Report,
Revision 2, and applies to plants going
forward using the 2007 Edition with
2009 Addenda through the 2013 Edition
of Subsection IWL. The requirements
remain unchanged from the
recommendations in the GALL Report
and, therefore, this change is not a
backfit.
3. Revise § 50.55a(b)(2)(ix), ‘‘Metal
containment examinations,’’ to extend
the applicability of the existing
conditions in § 50.55a(b)(2)(ix)(A)(2)
and (b)(2)(ix)(B) and (J) to the 2007
Edition with 2009 Addenda through the
2013 Edition of Subsection IWE. This
condition does not result in a change to
current requirements, and is therefore
not a backfit.
4. Revise § 50.55a(b)(2)(x), ‘‘Quality
assurance,’’ to require that when
applying the editions and addenda later
than the 1989 Edition of ASME BPV
Code, Section XI, the requirements of
NQA–1, 1994 Edition, the 2008 Edition,
and the 2009–1a Addenda specified in
either IWA–1400 or Table IWA 1600–1,
‘‘Referenced Standards and
Specifications,’’ of that edition and
addenda of Section XI are acceptable for
use, provided the licensee uses its
appendix B to 10 CFR part 50 QA
program in conjunction with Section XI
requirements. This revision clarifies the
current requirements, which the NRC
considers to be consistent with the
meaning and intent of the current
requirements. Therefore, the NRC does
not consider the clarification to be a
change in requirements. Therefore, this
change is not a backfit.
5. Revise § 50.55a(b)(2)(xii),
‘‘Underwater welding,’’ to allow
underwater welding on irradiated
materials under certain conditions. The
revision eliminates the prohibition on
welding on irradiated materials.
Therefore, this change is not a backfit.
6. Add a new condition as
§ 50.55a(b)(2)(xviii)(D), ‘‘NDE personnel
certification: Fourth provision,’’ to
prohibit the use of Appendix VII and
Subarticle VIII–2200 of the 2011
Addenda and 2013 Edition of Section XI
of the ASME BPV Code. Licensees are
required to implement Appendix VII
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and Subarticle VIII–2200 of the 2010
Edition of Section XI. This condition
does not constitute a change in NRC
position because the use of the subject
provisions is not currently allowed by
§ 50.55a. Therefore, the addition of this
new condition is not a backfit.
7. Revise § 50.55a(b)(2)(xxi)(A),
‘‘Table IWB–2500–1 examination
requirements: First provision,’’ to
modify the standard for visual
magnification resolution sensitivity and
contrast for visual examinations of
Examination Category B–D components,
making the rule conform with ASME
BPV Code, Section XI requirements for
VT–1 examinations. This revision
removes a condition that was in
addition to the ASME BPV Code
requirements and does not impose a
new requirement. Therefore, this change
is not a backfit.
8. Add a new condition as
§ 50.55a(b)(2)(xxxi), ‘‘Mechanical
clamping devices;’’ to prohibit the use
of mechanical clamping devices in
accordance with IWA–4131.1(c) in the
2010 Edition and IWA–4131.1(d) in the
2011 Addenda through 2013 Edition on
small item Class 1 piping and portions
of a piping system that forms the
containment boundary. This condition
does not constitute a change in NRC
position and does not affect licensees
because the use of the subject provisions
is not currently allowed by § 50.55a.
Therefore, the addition of this new
condition is not a backfit.
9. Add a new condition as
§ 50.55a(b)(2)(xxxii), ‘‘Summary report
submittal,’’ to clarify that licensees
using the 2010 Edition or later editions
and addenda of Section XI must
continue to submit to the NRC the
Preservice and Inservice Summary
Reports required by IWA–6240 of the
2009 Addenda of Section XI. This
condition does not result in a change in
the NRC’s requirements insomuch as
these reports have been required in the
2009 Addenda of Section XI and all
previous editions and addenda.
Therefore, the addition of this new
condition is not a backfit.
10. Add a new condition as
§ 50.55a(b)(2)(xxxiii), ‘‘Risk-Informed
allowable pressure,’’ to prohibit the use
of ASME BPV Code, Section XI,
Appendix G, Paragraph G–2216. The
use of Paragraph G–2216 is not
currently allowed by § 50.55a.
Therefore, the condition does not
constitute a new or changed NRC
position on the lack of acceptability of
Paragraph G–2216. Therefore, the
addition of this new condition is not a
backfit.
11. Add a new condition as
§ 50.55a(b)(2)(xxxiv), ‘‘Nonmandatory
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Appendix U.’’ Paragraph
(b)(2)(xxxiv)(A) requires that repair or
replacement activities temporarily
deferred under the provisions of
Nonmandatory Appendix U shall be
performed during the next scheduled
refueling outage. This condition is
imposed to ensure that repairs/
replacements are performed on
degraded components when a unit is
shutdown for refueling. This change is
consistent with the condition previously
placed on ASME BPV Code Case N–
513–3 and, therefore, does not impose a
new requirement. This change is not a
backfit. Paragraph (b)(2)(xxxiv)(B)
requires that the mandatory appendix in
ASME BPV Code Case N–513–3 be used
in lieu of the appendix referenced in
Paragraph U–S1–4.2.1(c) of Appendix
U. This change is required because the
appendix referenced in Appendix U was
unintentionally omitted. This change is
not a backfit.
12. Add a new condition as
§ 50.55a(b)(2)(xxxv), ‘‘Use of RTT0 in the
KIa and KIc equations,’’ to specify that
when licensees use ASME BPV Code,
Section XI 2013 Edition Nonmandatory
Appendix A, Paragraph A–4200, if T0 is
available, then RTT0 may be used in
place of RTNDT for applications using
the KIc equation and the associated KIc
curve, but not for applications using the
KIa equation and the associated KIa
curve. Conditions on the use of ASME
BPV Code, Section XI, Nonmandatory
Appendices do not constitute
backfitting inasmuch as those
provisions apply to voluntary actions
initiated by the licensee to use the
‘‘nonmandatory compliance’’ provisions
in these Appendices of the rule.
13. Add a new condition as
§ 50.55a(b)(2)(xxxvi), ‘‘Fracture
toughness of irradiated materials,’’ to
require licensees using ASME BPV
Code, Section XI 2013 Edition
Nonmandatory Appendix A, Paragraph
A–4400, to obtain NRC approval before
using irradiated T0 and the associated
RTT0 in establishing fracture toughness
of irradiated materials. Conditions on
the use of ASME BPV Code, Section XI,
Nonmandatory Appendices do not
constitute backfitting inasmuch as those
provisions apply to voluntary actions
initiated by the licensee to use the
‘‘nonmandatory compliance’’ provisions
in these Appendices of the rule.
14. Add a new condition as
§ 50.55a(b)(2)(xxxvii), ‘‘ASME BPV Code
Case N–824,’’ to allow the use of the
code case as conditioned. Conditions on
the use of ASME BPV Code Case N–824
do not constitute backfitting, inasmuch
as the use of this code case is not
required by the NRC but instead is an
alternative which may be voluntarily
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used by the licensee (i.e., a ‘‘voluntary
alternative’’).
OM Code
1. Add a new condition as
§ 50.55a(b)(3)(ii)(A), ‘‘MOV diagnostic
test interval,’’ to require that licensees
evaluate the adequacy of the diagnostic
test intervals established for MOVs
within the scope of OM Code, Appendix
III, not later than 5 years or three
refueling outages (whichever is longer)
from initial implementation of
Appendix III of the OM Code. This
condition represents an exception to a
later OM Code provision but merely
retains the current NRC condition on
ASME OM Code Case OMN–1, and is
therefore not a backfit because the NRC
is not imposing a new requirement.
2. Add a new condition as
§ 50.55a(b)(3)(ii)(B), ‘‘MOV testing
impact on risk,’’ to require that licensees
ensure that the potential increase in
core damage frequency and large early
release frequency associated with the
extension is acceptably small when
extending exercise test intervals for high
risk MOVs beyond a quarterly
frequency. This condition represents an
exception to a later OM Code provision
but merely retains the current NRC
condition on ASME OM Code Case
OMN–1, and is therefore not a backfit
because the NRC is not imposing a new
requirement.
3. Add a new condition as
§ 50.55a(b)(3)(ii)(C), ‘‘MOV risk
categorization,’’ to require, when
applying Appendix III to the OM Code,
that licensees categorize MOVs
according to their safety significance
using the methodology described in OM
Code Case OMN–3 subject to the
conditions discussed in RG 1.192, or
using an MOV risk ranking methodology
accepted by the NRC on a plant-specific
or industry-wide basis in accordance
with the conditions in the applicable
safety evaluation. This condition
represents an exception to a later OM
Code provision but merely retains the
current NRC condition on ASME OM
Code Case OMN–1, and is therefore not
a backfit because the NRC is not
imposing a new requirement.
4. Add a new condition as
§ 50.55a(b)(3)(ii)(D), ‘‘MOV stroke time,’’
to require that, when applying
Paragraph III–3600, ‘‘MOV Exercising
Requirements,’’ of Appendix III to the
OM Code, licensees shall verify that the
stroke time of the MOVs specified in
plant technical specifications satisfies
the assumptions in the plant’s safety
analyses. This condition retains the
MOV stroke time requirement for a
smaller set of MOVs than was specified
in previous editions and addenda of the
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OM Code. The retention of this
requirement is not a backfit.
5. Add new conditions as
§ 50.55a(b)(3)(iii)(A) through (D), ‘‘New
reactors,’’ to apply specific conditions
for IST programs applicable to licensees
of new nuclear power plants in addition
to the provisions of the OM Code as
incorporated by reference with
conditions in § 50.55a. Licensees of
‘‘new reactors’’ are, as identified in the
paragraph: (1) Holders of operating
licenses for nuclear power reactors that
received construction permits under
this part on or after the date 12 months
after August 17, 2017, and (2) holders of
COLs issued under 10 CFR part 52,
whose initial fuel loading occurs on or
after the date 12 months after August 17,
2017. This implementation schedule for
new reactors is consistent with the NRC
regulations in § 50.55a(f)(4)(i). These
conditions represent an exception to a
later OM Code provision but merely
retain a current NRC requirement, and
are therefore not a backfit because the
NRC is not imposing a new requirement.
6. Revise § 50.55a(b)(3)(iv), ‘‘Check
valves (Appendix II),’’ to specify that
Appendix II, ‘‘Check Valve Condition
32971
Monitoring Program,’’ of the OM Code,
2003 Addenda through the 2012
Edition, is acceptable for use with the
following clarification: Trending and
evaluation shall support the
determination that the valve or group of
valves is capable of performing its
intended function(s) over the entire
interval. At least one of the Appendix II
condition monitoring activities for a
valve group shall be performed on each
valve of the group at approximate equal
intervals not to exceed the maximum
interval shown in the following table:
MAXIMUM INTERVALS FOR USE WHEN APPLYING INTERVAL EXTENSIONS
Maximum
interval between
activities of
member valves
in the groups
(years)
Group size
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≥4 .....................................................................................................................................................................
3 .......................................................................................................................................................................
2 .......................................................................................................................................................................
1 .......................................................................................................................................................................
The regulation is being revised to
extend the applicability of this existing
NRC condition on the OM Code to the
2012 Edition of the OM Code and to
update the clarification for the use of
Appendix II. This does not represent a
change in the NRC’s position that the
condition is needed with respect to the
OM Code. Therefore, this condition is
not a backfit.
7. Add a new condition as
§ 50.55a(b)(3)(vii), ‘‘Subsection ISTB,’’
to prohibit the use of Subsection ISTB
in the 2011 Addenda to the OM Code
because the complete set of planned
Code modifications to support the
changes to the comprehensive pump
test acceptance criteria was not made in
that addenda. This condition represents
an exception to a later OM Code
provision but merely limits the use of
the later Code provision, and is
therefore not a backfit because the NRC
is not imposing a new requirement.
8. Add a new condition as
§ 50.55a(b)(3)(viii), ‘‘Subsection ISTE,’’
to allow licensees to implement
Subsection ISTE, ‘‘Risk-Informed
Inservice Testing of Components in
Light-Water Reactor Nuclear Power
Plants,’’ in the OM Code, 2009 Edition,
2011 Addenda and 2012 Edition, where
the licensee has obtained authorization
to implement Subsection ISTE as an
alternative to the applicable IST
requirements in the OM Code on a caseby-case basis in accordance with
§ 50.55a(z). This condition represents an
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exception to a later OM Code provision
but merely limits the use of the later
Code provision, and is therefore not a
backfit because the NRC is not imposing
a new requirement.
9. Add a new condition as
§ 50.55a(b)(3)(ix), ‘‘Subsection ISTF,’’ to
specify that licensees applying
Subsection ISTF, 2012 Edition, shall
satisfy the requirements of Mandatory
Appendix V, ‘‘Pump Periodic
Verification Test Program,’’ of the OM
Code, 2012 Edition. The condition also
specifies that Subsection ISTF, 2011
Addenda, is not acceptable for use. This
condition represents an exception to a
later OM Code provision but merely
limits the use of the later Code
provision, and is therefore not a backfit
because the NRC is not imposing a new
requirement.
10. Add a new condition as
§ 50.55a(b)(3)(x), ‘‘ASME OM Code Case
OMN–20,’’ to allow licensees to
implement OM Code Case OMN–20,
‘‘Inservice Test Frequency,’’ in the OM
Code, 2012 Edition. This condition
allows voluntary action initiated by the
licensee to use the code case and is,
therefore, not a backfit.
11. Add a new condition as
§ 50.55a(b)(3)(xi), ‘‘Valve Position
Indication,’’ to emphasize, when
implementing OM Code (2012 Edition),
Subsection ISTC–3700, ‘‘Position
Verification Testing,’’ licensees shall
implement the OM Code provisions to
verify that valve operation is accurately
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4.5
4.5
6
Not applicable
Maximum
interval
between
activities of
each valve
in the group
(years)
16
12
12
10
indicated. This condition emphasizes
the OM Code requirements for valve
position indication and is not a change
to those requirements. As such, this
condition is not a backfit.
12. Revise § 50.55a(f), ‘‘Preservice and
inservice testing requirements,’’ to
clarify that the OM Code includes
provisions for preservice testing of
components as part of its overall
provisions for IST programs. No
expansion of IST program scope is
intended by this clarification. This
condition does not result in a change in
requirements previously approved in
the Code and is, therefore, not a backfit.
13. Revise § 50.55a(f)(4), ‘‘Inservice
testing standards for operating plants,’’
to state that the paragraph is applicable
to pumps and valves that are within the
scope of the OM Code. Also, revise
§ 50.55a(f)(4) to state that the IST
requirements for pumps and valves that
are within the scope of the OM Code but
are not classified as ASME BPV Code
Class 1, Class 2, or Class 3 may be
satisfied as an augmented IST program
in accordance with § 50.55a(f)(6)(ii)
without requesting relief under
§ 50.55a(f)(5) or alternatives under
§ 50.55a(z). This use of an augmented
IST program may be acceptable
provided the basis for deviations from
the OM Code as incorporated by
reference in this section demonstrates
an acceptable level of quality and safety,
or that implementing the Code
provisions would result in hardship or
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unusual difficulty without a
compensating increase in the level of
quality and safety, where documented
and available for NRC review. These
changes align the scope of pumps and
valves for inservice testing with the
scope defined in the OM Code. These
changes do not result in a change in
requirements previously approved in
the Code, and is therefore not a backfit.
ASME BPV Code Case N–729–4
Revise § 50.55a(g)(6)(ii)(D), ‘‘Reactor
vessel head inspections.’’
On June 22, 2012, the ASME
approved the fourth revision of ASME
BPV Code Case N–729 (N–729–4). The
NRC proposed to update the
requirements of § 50.55a(g)(6)(ii)(D) to
require licensees to implement ASME
BPV Code Case N–729–4, with
conditions. The ASME BPV Code Case
N–729–4 contains similar requirements
as N–729–1; however, N–729–4 also
contains new requirements to address
previous NRC conditions, including
changes to inspection frequency and
qualifications. The new NRC conditions
on the use of ASME BPV Code Case N–
729–4 address operational experience,
clarification of implementation, and the
use of alternatives to the code case.
The current regulatory requirements
for the examination of pressurized water
reactor upper RPV heads that use
nickel-alloy materials are provided in
§ 50.55a(g)(6)(ii)(D). This section was
first created by rulemaking, dated
September 10, 2008 (73 FR 52730), to
require licensees to implement ASME
BPV Code Case N–729–1, with
conditions, instead of the inspections
previously required by the ASME BPV
Code, Section XI. The action did
constitute a backfit; however, the NRC
concluded that imposition of ASME
BPV Code Case N–729–1, as
conditioned, constituted an adequate
protection backfit.
The General Design Criteria (GDC) for
nuclear power plants (appendix A to 10
CFR part 50) or, as appropriate, similar
requirements in the licensing basis for a
reactor facility, provide bases and
requirements for NRC assessment of the
potential for, and consequences of,
degradation of the reactor coolant
pressure boundary (RCPB). The
applicable GDC include GDC 14
(Reactor Coolant Pressure Boundary),
GDC 31 (Fracture Prevention of Reactor
Coolant Pressure Boundary), and GDC
32 (Inspection of Reactor Coolant
Pressure Boundary). General Design
Criterion 14 specifies that the RCPB be
designed, fabricated, erected, and tested
so as to have an extremely low
probability of abnormal leakage, of
rapidly propagating failure, and of gross
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rupture. General Design Criterion 31
specifies that the probability of rapidly
propagating fracture of the RCPB be
minimized. General Design Criterion 32
specifies that components that are part
of the RCPB have the capability of being
periodically inspected to assess their
structural and leak-tight integrity.
The NRC concludes that ASME BPV
Code Case N–729–4, as conditioned,
shall be mandatory in order to ensure
that the requirements of the GDC are
satisfied. Imposition of ASME BPV Code
Case N–729–4, with conditions, ensures
that the ASME BPV Code-allowable
limits will not be exceeded, leakage will
likely not occur, and potential flaws
will be detected before they challenge
the structural or leak-tight integrity of
the RPV upper head within current
nondestructive examination limitations.
The NRC concludes that the regulatory
framework for providing adequate
protection of public health and safety is
accomplished by the incorporation of
ASME BPV Code Case N–729–4 into
§ 50.55a, as conditioned. All current
licensees of U.S. pressurized water
reactors will be required to implement
ASME BPV Code Case N–729–4, as
conditioned. The Code Case provisions
on examination requirements for RPV
upper heads are essentially the same as
those established under ASME BPV
Code Case N–729–1, as conditioned.
One exception is the condition in
§ 50.55a(g)(6)(ii)(D)(3), which will
require, for upper heads with Alloy 600
penetration nozzles, that bare metal
visual examinations be performed each
outage in accordance with Table 1 of
ASME BPV Code Case N–729–4.
Accordingly, the NRC imposition of the
ASME BPV Code Case N–729–4, as
conditioned, may be deemed to be a
modification of the procedures to
operate a facility resulting from the
imposition of the new regulation, and as
such, this rulemaking provision may be
considered backfitting under
§ 50.109(a)(1).
The NRC continues to find that
inspections of RPV upper heads, their
penetration nozzles, and associated
partial penetration welds are necessary
for adequate protection of public health
and safety and that the requirements of
ASME BPV Code Case N–729–4, as
conditioned, represent an acceptable
approach, developed, in part, by a
voluntary consensus standards body for
performing future inspections. The NRC
concludes that approval of ASME BPV
Code Case N–729–4, as conditioned, by
incorporation by reference of the Code
Case into § 50.55a, is necessary to
ensure that the facility provides
adequate protection to the health and
safety of the public and constitutes a
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redefinition of the requirements
necessary to provide reasonable
assurance of adequate protection of
public health and safety. Therefore, a
backfit analysis need not be prepared for
this portion of the rule in accordance
with § 50.109(a)(4)(ii) and (iii).
ASME BPV Code Case N–770–2
Revise § 50.55a(g)(6)(ii)(F),
‘‘Examination requirements for Class 1
piping and nozzle dissimilar metal butt
welds.’’
On June 9, 2011, the ASME approved
the second revision of ASME BPV Code
Case N–770 (N–770–2). The NRC is
updating the requirements of
§ 50.55a(g)(6)(ii)(F) to require licensees
to implement ASME BPV Code Case N–
770–2, with conditions. The ASME BPV
Code Case N–770–2 contains similar
baseline and ISI requirements for
unmitigated nickel-alloy butt welds, and
preservice and ISI requirements for
mitigated butt welds as N–770–1.
However, N–770–2 also contains new
requirements for optimized weld
overlays, a specific mitigation technique
and volumetric inspection coverage.
Further, the NRC conditions on the use
of ASME BPV Code Case N–770–2 have
been modified to address the changes in
the code case, clarify inspection
coverage requirements and require the
development of inspection
qualifications to allow complete weld
inspection coverage in the future.
The current regulatory requirements
for the examination of ASME Class 1
piping and nozzle dissimilar metal butt
welds that use nickel-alloy materials is
provided in § 50.55a(g)(6)(ii)(F). This
section was first created by rulemaking,
dated June 21, 2011 (76 FR 36232), to
require licensees to implement ASME
BPV Code Case N–770–1, with
conditions. The NRC added
§ 50.55a(g)(6)(ii)(F) to require licensees
to implement ASME BPV Code Case N–
770–1, with conditions, instead of the
inspections previously required by the
ASME BPV Code, Section XI. The action
did constitute a backfit; however, the
NRC concluded that imposition of
ASME BPV Code Case N–770–1, as
conditioned, constituted an adequate
protection backfit.
The GDC for nuclear power plants
(appendix A to 10 CFR part 50) or, as
appropriate, similar requirements in the
licensing basis for a reactor facility,
provide bases and requirements for NRC
assessment of the potential for, and
consequences of, degradation of the
RCPB. The applicable GDC include GDC
14 (Reactor Coolant Pressure Boundary),
GDC 31 (Fracture Prevention of Reactor
Coolant Pressure Boundary) and GDC 32
(Inspection of Reactor Coolant Pressure
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Boundary). General Design Criterion 14
specifies that the RCPB be designed,
fabricated, erected, and tested so as to
have an extremely low probability of
abnormal leakage, of rapidly
propagating failure, and of gross
rupture. General Design Criterion 31
specifies that the probability of rapidly
propagating fracture of the RCPB be
minimized. General Design Criterion 32
specifies that components that are part
of the RCPB have the capability of being
periodically inspected to assess their
structural and leak-tight integrity.
The NRC concludes that ASME BPV
Code Case N–770–2, as conditioned,
must be imposed in order to ensure that
the requirements of the GDC are
satisfied. Imposition of ASME BPV Code
Case N–770–2, with conditions, ensures
that the requirements of the GDC are
met for all mitigation techniques
currently in use for Alloy 82/182 butt
welds because ASME BPV Codeallowable limits will not be exceeded,
leakage would likely not occur and
potential flaws will be detected before
they challenge the structural or leaktight integrity of piping welds. All
current licensees of U.S. pressurized
water reactors will be required to
implement ASME BPV Code Case N–
770–2, as conditioned. The Code Case
provisions on examination requirements
for ASME Class 1 piping and nozzle
nickel-alloy dissimilar metal butt welds
are somewhat different from those
established under ASME BPV Code Case
N–770–1, as conditioned, and will
require a licensee to modify its
procedures for inspection of ASME
Class 1 nickel-alloy welds to meet these
requirements. Accordingly, the NRC
imposition of the ASME BPV Code Case
N–770–2, as conditioned, may be
deemed to be a modification of the
procedures to operate a facility resulting
from the imposition of the new
regulation, and as such, this rulemaking
provision may be considered backfitting
under § 50.109(a)(1).
The NRC continues to find that ASME
Class 1 nickel-alloy dissimilar metal
weld inspections are necessary for
adequate protection of public health and
safety, and that the requirements of
ASME BPV Code Case N–770–2, as
conditioned, represent an acceptable
approach developed by a voluntary
consensus standards body for
performing future ASME Class 1 nickelalloy dissimilar metal weld inspections.
The NRC concludes that approval of
ASME BPV Code Case N–770–2, as
conditioned, by incorporation by
reference of the Code Case into § 50.55a,
is necessary to ensure that the facility
provides adequate protection to the
health and safety of the public and
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constitutes a redefinition of the
requirements necessary to provide
reasonable assurance of adequate
protection of public health and safety.
Therefore, a backfit analysis need not be
prepared for this portion of the rule in
accordance with § 50.109(a)(4)(ii) and
(iii).
Conclusion
The NRC finds that incorporation by
reference into § 50.55a of the 2009
Addenda through 2013 Edition of
Section III, Division 1, of the ASME BPV
Code, subject to the identified
conditions; the 2009 Addenda through
2013 Edition of Section XI, Division 1,
of the ASME BPV Code, subject to the
identified conditions; and the 2009
Edition through the 2012 Edition of the
OM Code, subject to the identified
conditions, does not constitute
backfitting or represent an inconsistency
with any issue finality provisions in 10
CFR part 52.
The NRC finds that the incorporation
by reference of Code Cases N–824 and
OMN–20 does not constitute backfitting
or represent an inconsistency with any
issue finality provisions in 10 CFR part
52.
The NRC finds that the inclusion of a
new condition on Code Case N–729–4
and a new condition on Code Case N–
770–2 constitutes backfitting necessary
for adequate protection.
X. 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).
XI. Finding of No Significant
Environmental Impact: Environmental
Assessment
This final rule is in accordance with
the NRC’s policy to incorporate by
reference in § 50.55a new editions and
addenda of the ASME BPV and OM
Codes to provide updated rules for
constructing and inspecting components
and testing pumps, valves, and dynamic
restraints (snubbers) in light-water
nuclear power plants. The ASME Codes
are national voluntary consensus
standards and are required by the
NTTAA to be used by government
agencies unless the use of such a
standard is inconsistent with applicable
law or otherwise impractical. The
National Environmental Policy Act
(NEPA) requires Federal agencies to
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32973
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 proposed
action and alternatives to the proposed
action (42 U.S.C. 4332(C); NEPA Sec.
102(C)).
The NRC has determined under
NEPA, as amended, and the NRC’s
regulations in subpart A of 10 CFR part
51, that this rule is not a major Federal
action significantly affecting the quality
of the human environment and,
therefore, an environmental impact
statement is not required. The
rulemaking does not significantly
increase the probability or consequences
of accidents, no changes are being made
in the types of effluents that may be
released off-site, and there is no
significant increase in public radiation
exposure. The NRC estimates the
radiological dose to plant personnel
performing the inspections required by
ASME BPV Code Case N–770–2 would
be about 3 rem per plant over a 10-year
interval, and a one-time exposure for
mitigating welds of about 30 rem per
plant. The NRC estimates the
radiological dose to plant personnel
performing the inspections required by
ASME BPV Code Case N–729–4 would
be about 3 rem per plant over a 10-year
interval and a one-time exposure for
mitigating welds of about 30 rem per
plant. As required by 10 CFR part 20,
and in accordance with current plant
procedures and radiation protection
programs, plant radiation protection
staff will continue monitoring dose rates
and would make adjustments in
shielding, access requirements,
decontamination methods, and
procedures as necessary to minimize the
dose to workers. The increased
occupational dose to individual workers
stemming from the ASME BPV Code
Case N–770–2 and N–729–4 inspections
must be maintained within the limits of
10 CFR part 20 and as low as reasonably
achievable. Therefore, the NRC
concludes that the increase in
occupational exposure would not be
significant. This final rule does not
involve non-radiological plant effluents
and has no other environmental
impacts. Therefore, no significant nonradiological impacts are associated with
this action. The determination of this
environmental assessment is that there
will be no significant off-site impact to
the public from this action.
XII. Paperwork Reduction Act
Statement
This final rule amends collections of
information subject to the Paperwork
Reduction Act of 1995 (44 U.S.C. 3501
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et seq.). The collections of information
were approved by the Office of
Management and Budget (OMB),
approval number 3150–0011.
Because the rule will reduce the
burden for existing information
collections, the public burden for the
information collections is expected to be
decreased by 58.5 hours per response.
This reduction includes the time for
reviewing instructions, searching
existing data sources, gathering and
maintaining the data needed, and
completing and reviewing the
information collection.
The information collection is being
conducted to document the plans for
and the results of ISI and IST programs.
The records are generally historical in
nature and provide data on which future
activities can be based. The practical
utility of the information collection for
the NRC is that appropriate records are
available for auditing by NRC personnel
to determine if ASME BPV and OM
Code provisions for construction,
inservice inspection, repairs, and
inservice testing are being properly
implemented in accordance with
§ 50.55a, or whether specific
enforcement actions are necessary.
Responses to this collection of
information are generally mandatory
under 10 CFR 50.55a.
You may submit comments on any
aspect of the information collection(s),
including suggestions for reducing the
burden, by the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2011–0088.
• Mail comments to: Information
Services Branch, Office of the Chief
Information Officer, Mail Stop: T–2F43,
U.S. Nuclear Regulatory Commission,
Washington, DC 20555–0001 or to
Aaron Szabo, Desk Officer, Office of
Information and Regulatory Affairs
(3150–0011), NEOB–10202, Office of
Management and Budget, Washington,
DC 20503; telephone: 202–395–3621,
email: oira_submission@omb.eop.gov.
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Public Protection Notification
The NRC may not conduct or sponsor,
and a person is not required to respond
to, a collection of information unless the
document requesting the collection
displays a currently valid OMB control
number.
XIII. Congressional Review Act
This final rule is a rule as defined in
the Congressional Review Act (5 U.S.C.
801–808). However, OMB has not found
it to be a major rule as defined in the
Congressional Review Act.
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XIV. Voluntary Consensus Standards
Section 12(d)(3) of the National
Technology Transfer and Advancement
Act of 1995, Public Law 104–113
(NTTAA), and implementing guidance
in OMB Circular A–119 (February 10,
1998), requires 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 which are not
acceptable to the agency.
In this final rule, the NRC is
continuing its existing practice of
establishing requirements for the design,
construction, operation, ISI
(examination), and IST of nuclear power
plants by approving the use of the latest
editions and addenda of the ASME
Codes in § 50.55a. The ASME Codes are
voluntary consensus standards,
developed by participants with broad
and varied interests, in which all
interested parties (including the NRC
and licensees of nuclear power plants)
participate. Therefore, the NRC’s
incorporation by reference of the ASME
Codes is consistent with the overall
objectives of the NTTAA and OMB
Circular A–119.
In this final rule, the NRC is also
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 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
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licensees of nuclear power plants)
participate. Therefore, the NRC’s
approval of the use of the ASME Code
Cases in this final rule is consistent with
the overall objectives of the NTTAA and
OMB Circular A–119.
As discussed in Section II of this
document, ‘‘Discussion,’’ the NRC is
conditioning the use of certain
provisions of the 2009 Addenda, 2010
Edition, 2011 Addenda, and 2013
Edition of the ASME BPV Code Section
III, Division 1 and Section XI, Division
1. The NRC is also conditioning the use
of certain provisions of the 2009
Edition, the 2011 Addenda, and the
2012 Edition of the OM Code, Division
1. This final rule also includes various
versions of quality assurance standard
NQA–1 and Code Cases N–729–4, N–
770–2, N–824, OMN–20, N–513–3
Mandatory Appendix I, and N–852. In
addition, this final rule does not adopt
(‘‘excludes’’) certain provisions of the
ASME Codes, as discussed in this
statement of considerations and in the
regulatory analysis for this rulemaking.
The NRC staff’s position is that this final
rule complies with the NTTAA and
OMB Circular A–119 despite these
conditions and ‘‘exclusions.’’
If the NRC did not conditionally
accept ASME editions, addenda, and
code cases, the NRC would disapprove
these entirely. The effect would be that
licensees and applicants would submit
a larger number of requests for use of
alternatives under § 50.55a(z), requests
for relief under § 50.55a(f) and (g), or
requests for exemptions under § 50.12
and/or § 52.7. These requests would
likely include broad scope requests for
approval to issue the full scope of the
ASME Code editions and addenda
which would otherwise be approved in
this final rule (i.e., the request would
not be simply for approval of a specific
ASME Code provision with conditions).
These requests would be an unnecessary
additional burden for both the licensee
and the NRC, inasmuch as the NRC has
already determined that the ASME
Codes and Code Cases which are the
subject of this final rule are acceptable
for use (in some cases with conditions).
For these reasons, the NRC concludes
that this final rule’s treatment of ASME
Code editions and addenda, and code
cases and any conditions placed on
them 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 U.S. voluntary consensus
standards bodies for use within the
United States, which the NRC could
incorporate by reference instead of the
ASME Codes. The NRC also did not
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sradovich on DSK3GMQ082PROD with RULES2
identify any voluntary consensus
standards, developed by multinational
voluntary consensus standards bodies
for use on a multinational basis, which
the NRC could incorporate by reference
instead of the ASME Codes. The NRC
identified codes addressing the same
subject as the ASME Codes for use in
individual countries. At least one
country, Korea, directly translated the
ASME Code for use in that country. In
other countries (e.g., Japan), ASME
Codes were the basis for development of
the country’s codes, but the ASME
Codes were substantially modified to
accommodate that country’s regulatory
system and reactor designs. Finally,
there are countries (e.g., the Russian
Federation) where that country’s code
was developed without regard to the
ASME Code. However, some of these
codes may not meet the definition of a
voluntary consensus standard because
they were developed by the state rather
than a voluntary consensus standards
body. The NRC’s evaluation of other
countries’ codes to determine whether
each code provides a comparable or
enhanced level of safety, when
compared against the level of safety
provided under the ASME Codes, would
require a significant expenditure of
agency resources. This expenditure does
not seem justified, given that
substituting another country’s code for
the U.S. voluntary consensus standard
does not appear to substantially further
the apparent underlying objectives of
the NTTAA.
In summary, this final rule satisfies
the requirements of Section 12(d)(3) of
the NTTAA and OMB Circular A–119.
XV. Incorporation by Reference—
Reasonable Availability to Interested
Parties
The NRC is incorporating by reference
recent editions and addenda to the
ASME Codes for nuclear power plants
and a standard for quality assurance.
The NRC is also incorporating by
reference six ASME Code Cases. As
described in the ‘‘Background’’ and
‘‘Discussion’’ sections of this document,
these materials provide rules for safety
governing the design, fabrication, and
inspection of nuclear power plant
components.
The NRC is required by law to obtain
approval for incorporation by reference
from the Office of the Federal Register
(OFR). The OFR’s requirements for
incorporation by reference are set forth
in 1 CFR part 51. On November 7, 2014,
the OFR adopted changes to its
regulations governing incorporation by
reference (79 FR 66267). The OFR
regulations require an agency to include
in a final rule a discussion of the ways
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that the materials the agency
incorporates by reference are reasonably
available to interested parties and how
interested parties can obtain the
materials. The discussion in this section
complies with the requirement for final
rules as set forth in § 51.5(b).
The NRC considers ‘‘interested
parties’’ to include all potential NRC
stakeholders, not only the individuals
and entities regulated or otherwise
subject to the NRC’s regulatory
oversight. These NRC stakeholders are
not a homogenous group, so the
considerations for determining
‘‘reasonable availability’’ vary by class
of interested parties. The NRC identifies
six classes of interested parties with
regard to the material to be incorporated
by reference in an NRC rule:
• Individuals and small entities
regulated or otherwise subject to the
NRC’s regulatory oversight who are
subject to the material to be
incorporated by reference by
rulemaking. This class also includes
applicants and potential applicants for
licenses and other NRC regulatory
approvals. In this context, ‘‘small
entities’’ has the same meaning as a
‘‘small entity’’ under § 2.810.
• Large entities otherwise subject to
the NRC’s regulatory oversight who are
subject to the material to be
incorporated by reference by
rulemaking. This class also includes
applicants and potential applicants for
licenses and other NRC regulatory
approvals. In this context, ‘‘large
entities’’ are those which do not qualify
as a ‘‘small entity’’ under § 2.810.
• Non-governmental organizations
with institutional interests in the
matters regulated by the NRC.
• Other Federal agencies, states, local
governmental bodies (within the
meaning of § 2.315(c)).
• Federally-recognized and Staterecognized 3 Indian tribes.
• Members of the general public (i.e.,
individual, unaffiliated members of the
public who are not regulated or
otherwise subject to the NRC’s
regulatory oversight) who may wish to
gain access to the materials that the NRC
proposes to incorporate by reference in
order to participate in the rulemaking.
The NRC makes the materials to be
incorporated by reference available for
inspection to all interested parties, by
appointment, at the NRC Technical
Library, which is located at Two White
Flint North, 11545 Rockville Pike,
Rockville, Maryland 20852; telephone:
3 State-recognized Indian tribes are not within the
scope of § 2.315(c). However, for purposes of the
NRC’s compliance with 1 CFR 51.5, ‘‘interested
parties’’ includes a broad set of stakeholders
including State-recognized Indian tribes.
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301–415–7000; email:
Library.Resource@nrc.gov.
Interested parties may purchase a
copy of the materials from ASME at
Three Park Avenue, New York, NY
10016, or at the ASME Web site https://
www.asme.org/shop/standards. The
materials are also accessible through
third-party subscription services such as
IHS (15 Inverness Way East, Englewood,
CO 80112; https://global.ihs.com) and
Thomson Reuters Techstreet (3916
Ranchero Dr., Ann Arbor, MI 48108;
https://www.techstreet.com). The
purchase prices for individual
documents range from $225 to $720 and
the cost to purchase all documents is
approximately $9,000.
For the class of interested parties
constituting members of the general
public who wish to gain access to the
materials to be incorporated by
reference in order to participate in the
rulemaking, the NRC recognizes that the
$9,000 cost may be so high that the
materials could be regarded as not
reasonably available for purposes of
commenting on this rulemaking, despite
the NRC’s actions to make the materials
available at the NRC’s PDR.
Accordingly, the NRC sent a letter to
the ASME on April 9, 2015, requesting
that they consider enhancing public
access to these materials during the
public comment period. In an April 21,
2015, letter to the NRC, the ASME
agreed to make the materials available
online in a read-only electronic access
format during the public comment
period.
During the public comment period,
the ASME made publicly-available the
editions and addenda to the ASME
Codes for nuclear power plants, the
ASME standard for quality assurance,
and the ASME Code Cases which the
NRC proposed to incorporate by
reference. The ASME made the
materials publicly-available in read-only
format at the ASME Web site https://
go.asme.org/NRC.
The materials are available to all
interested parties in multiple ways and
in a manner consistent with their
interest in this rulemaking. Therefore,
the NRC concludes that the materials
the NRC is incorporating by reference in
this rulemaking are reasonably available
to all interested parties.
XVI. Availability of Guidance
The NRC will not be issuing guidance
for this rulemaking. The ASME BPV
Code and OM Code provide direction
for the performance of activities to
satisfy the Code requirements for
design, inservice inspection, and
inservice testing of nuclear power plant
SSCs. In addition, the NRC provides
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guidance in this Federal Register notice
for the implementation of the new
conditions on the ASME BPV Code and
OM Code, as necessary. The NRC has a
number of standard review plans
(SRPs), which provide guidance to NRC
reviewers and make communication and
understanding of NRC review processes
available to members of the public and
the nuclear power industry. NUREG–
0800, ‘‘Review of Safety Analysis
Reports for Nuclear Power Plants,’’ has
numerous sections which discuss
implementation of various aspects of the
ASME BPV Code and OM Code (e.g.,
Sections 3.2.2, 3.8.1, 3.8.2, 3.9.3, 3.9.6,
3.9.7, 3.9.8, 3.13, 5.2.1.1, 5.2.1.2, 5.2.4,
and 6.6). The NRC also publishes
Regulatory Guides and Generic
Communications (i.e., Regulatory Issue
Summaries, Information Notices) to
communicate and clarify NRC technical
or policy positions on regulatory matters
which may contain guidance relative to
this rulemaking.
Revision 2 of NUREG–1482,
‘‘Guidelines for Inservice Testing at
Nuclear Power Plants,’’ provides
guidance for the development and
implementation of IST programs at
nuclear power plants. With direction
provided in the ASME BPV and OM
Codes, and guidance in this Federal
Register notice, the NRC has determined
that preparation of a separate guidance
document is not necessary for this
update to § 50.55a. However, the NRC
will consider preparation of a revision
to NUREG–1482 in the future to address
the latest edition of the ASME OM Code
incorporated by reference in § 50.55a.
XVII. Availability of Documents
The NRC is making the documents
identified in Table 2 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 2—AVAILABILITY OF DOCUMENTS
ADAMS Accession No./
FEDERAL REGISTER citation/Web link
Document
sradovich on DSK3GMQ082PROD with RULES2
Proposed Rule Documents:
Proposed Rule—Federal Register Notice .............................................................................................
Draft Regulatory Analysis ......................................................................................................................
Final Rule Documents:
Final Regulatory Analysis ......................................................................................................................
2017 Final Rule (10 CFR 50.55a) American Society of Mechanical Engineers Codes and Code
Cases: Analysis of Public Comments.
Related Documents:
Fatigue and Fracture Mechanics: 33rd Volume, ASTM STP 1417, W.G. Reuter and R.S. Piascik,
Eds., ASTM International, West Conshohocken, PA, 2002.
Final Results from the CARINA Project on Crack Initiation and Arrest of Irradiated German RPV
Steels for Neutron Fluences in the Upper Bound, H. Hein et al., ASTM International, West
Conshohocken, PA, June 2014.
Letter from Brian Thomas, NRC, to Michael Merker, ASME, ‘‘Public Access to Material the NRC
Seeks to Incorporate by Reference into its Regulations,’’ April 9, 2015.
Letter from Mark Maxin, NRC, to Rick Libra, BWRVIP Chairman, ‘‘Safety Evaluation for Electric
Power Research Institute (EPRI) Boiling Water Reactor (BWR) Vessel and Internals Project
(BWRVIP) Report 1003020 (BWRVIP–97), ‘BWR Vessel and Internals Project, Guidelines for
Performing Weld Repairs to Irradiated BWR Internals’ (TAC No. MC3948),’’ June 30, 2008.
Letter from Michael Merker, ASME, to Brian Thomas, NRC; April 21, 2015 .......................................
Licensee Event Report 50–338/2012–001–00 ......................................................................................
NUREG–0800, ‘‘Standard Review Plan for the Review of Safety Analysis Reports for Nuclear
Power Plants, LWR Edition’’.
NUREG–0800, Section 3.9.6, Revision 3, ‘‘Functional Design, Qualification, and Inservice Testing
Programs for Pumps, Valves, and Dynamic Restraints,’’ March 2007.
NUREG–1482, Revision 2, ‘‘Guidelines for Inservice Testing at Nuclear Power Plants: Inservice
Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints
(Snubbers) at Nuclear Power Plants,’’ October 2013.
NUREG–1800, Revision 2, ‘‘Standard Review Plan for Review of License Renewal Applications for
Nuclear Power Plants,’’ December 2010.
NUREG–1801, Revision 2, ‘‘Generic Aging Lessons Learned (GALL) Report,’’ December 2010 .......
NUREG–1950, ‘‘Disposition of Public Comments and Technical Bases for Changes in the License
Renewal Guidance Documents NUREG–1801 and NUREG–1800,’’ April 2011.
NUREG–2124, ‘‘Final Safety Evaluation Report Related to the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and 4,’’ Section 3.9.6, ‘‘Inservice Testing of Pumps and Valves
(Related to RG 1.206, Section C.III.1, Chapter 3, C.I.3.9.6, ‘Functional Design, Qualification, and
Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints’)’’.
NUREG/CR–6860, ‘‘An Assessment of Visual Testing,’’ November 2004 ...........................................
NUREG/CR–6933, ‘‘Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping
Welds Using Advanced Low-Frequency Ultrasonic Methods,’’ March 2007.
NUREG/CR–7122, ‘‘An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless
Steel Pressurizer Surge Line Piping Welds,’’ March 2012.
NRC Generic Letter 89–10, ‘‘Safety-Related Motor-Operated Valve Testing and Surveillance,’’ June
1989.
NRC Generic Letter 90–05, ‘‘Guidance for Performing Temporary Non-Code Repair of ASME Code
Class 1, 2, and 3 Piping (Generic Letter 90–05),’’ June 1990.
NRC Meeting Summary of June 5–7, 2013, Annual Materials Programs Technical Information Exchange Public Meeting.
NRC Meeting Summary of January 19, 2016, Category 2 Public Meeting with Industry Representatives to Discuss Welding on Neutron Irradiated Ferritic and Austenitic Materials.
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80 FR 56820 (September 18, 2015).
ML14170B104.
ML16130A522.
ML16130A531.
https://www.astm.org/DIGITAL_LIBRARY/STP/SOURCE_PAGES/
STP1417.htm.
https://www.astm.org/DIGITAL_LIBRARY/STP/PAGES/
STP157220130113.htm.
ML15085A206.
ML081680730.
ML15112A064.
ML12151A441.
ML070660036.
ML070720041.
ML13295A020.
ML103490036.
ML103490041.
ML11116A062.
ML12271A045.
ML043630040.
ML071020410 and ML071020414.
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32977
TABLE 2—AVAILABILITY OF DOCUMENTS—Continued
ADAMS Accession No./
FEDERAL REGISTER citation/Web link
Document
NRC Meeting Summary of March 2, 2016, Public Meeting on Stakeholder Comments on the Proposed Rule.
NRC Staff Memorandum, ‘‘Consolidation of SECY–94–084 and SECY–95–132,’’ July 24, 1995 .......
NRC Memorandum, ‘‘Staff Requirements—Affirmation Session, 11:30 a.m., Friday, September 10,
1999, Commissioners’ Conference Room, One White Flint North, Rockville, Maryland (Open to
Public Attendance),’’ September 10, 1999.
NRC Regulatory Guide 1.28, Revision 4, ‘‘Quality Assurance Program Criteria (Design and Construction),’’ June 2010.
NRC Regulatory Guide 1.147, Revision 17, ‘‘Inservice Inspection Code Case Acceptability, ASME
Section XI, Division 1,’’ August 2014.
NRC Regulatory Guide 1.174, Revision 2, ‘‘An Approach for Using Probabilistic Risk Assessment in
Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,’’ May 2011.
NRC Regulatory Guide 1.175, ‘‘An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice Testing,’’ August 1998.
NRC Regulatory Guide 1.192, Revision 1, ‘‘Operation and Maintenance Code Case Acceptability,
ASME OM Code,’’ August 2014.
NRC Regulatory Guide 1.200, Revision 2, ‘‘An Approach for Determining the Technical Adequacy
of Probabilistic Risk Assessment Results for Risk-Informed Activities,’’ March 2009.
NRC Regulatory Guide 1.201, Revision 1, ‘‘Guidelines for Categorizing Structures, Systems, and
Components in Nuclear Power Plants According to Their Safety Significance,’’ May 2006.
NRC Regulatory Information Conference, Recent Operating Reactors Materials Issues, Presentation Materials, 2013.
NRC Regulatory Issue Summary 2013–07, ‘‘NRC Staff Position on the Use of American Society of
Mechanical Engineers Certification Mark,’’ May 28, 2013.
Relief Request REP–1 U2, Revision 2 ..................................................................................................
SECY–90–016, ‘‘Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements’’.
SECY–93–087, ‘‘Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced
Light-Water Reactor (ALWR) Designs’’.
SECY–94–084, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of NonSafety Systems in Passive Plant Designs’’.
SECY–95–132, ‘‘Policy and Technical Issues Associated with the Regulatory Treatment of NonSafety Systems (RTNSS) in Passive Plant Designs (SECY–94–084)’’.
Vogtle Electric Generating Plant, Units 3 and 4, Updated Final Safety Analysis Report, Revision 3,
Chapter 3, Section 3.9, Mechanical Systems and Components.
sradovich on DSK3GMQ082PROD with RULES2
List of Subjects in 10 CFR Part 50
Administrative practice and
procedure, Antitrust, Classified
information, Criminal penalties,
Education, Fire prevention, Fire
protection, Incorporation by reference,
Intergovernmental relations, Nuclear
power plants and reactors, Penalties,
Radiation protection, Reactor siting
criteria, Reporting and recordkeeping
requirements, Whistleblowing.
For the reasons set forth in the
preamble, and under the authority of the
Atomic Energy Act of 1954, as amended;
the Energy Reorganization Act of 1974,
as amended; and 5 U.S.C. 552 and 553,
the NRC is adopting the following
amendments to 10 CFR part 50.
PART 50—DOMESTIC LICENSING OF
PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50
continues to read as follows:
■
Authority: Atomic Energy Act of 1954,
secs. 11, 101, 102, 103, 104, 105, 108, 122,
147, 149, 161, 181, 182, 183, 184, 185, 186,
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187, 189, 223, 234 (42 U.S.C. 2014, 2131,
2132, 2133, 2134, 2135, 2138, 2152, 2167,
2169, 2201, 2231, 2232, 2233, 2234, 2235,
2236, 2237, 2239, 2273, 2282); Energy
Reorganization Act of 1974, secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);
Nuclear Waste Policy Act of 1982, sec. 306
(42 U.S.C. 10226); National Environmental
Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C.
3504 note; Sec. 109, Pub. L. 96–295, 94 Stat.
783.
2. In § 50.55a:
a. Revise paragraphs (a) introductory
text, (a)(1)(i) introductory text and
(a)(1)(i)(E)(12) and (13) and add
paragraphs (a)(1)(i)(E)(14) through (17);
■ b. Revise paragraphs (a)(1)(ii)
introductory text and (a)(1)(ii)(C)(48)
and (49) and add paragraphs
(a)(1)(ii)(C)(50) through (53);
■ c. Revise paragraphs (a)(1)(iii)(A)
through (C) and add paragraphs
(a)(1)(iii)(D) through (G);
■ d. Revise paragraph (a)(1)(iv)
introductory text and add paragraphs
(a)(1)(iv)(B) and (C);
■ e. Add paragraph (a)(1)(v);
■ f. Revise paragraphs (b) introductory
text, (b)(1) introductory text and
■
■
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ML090410014.
ML061090627.
https://www.nrc.gov/public-involve/conference-symposia/ric/past/2013/
docs/abstracts/sessionabstract19.html.
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(b)(1)(ii), (iv), and (vii) and add
paragraphs (b)(1)(viii) and (ix);
■ g. Revise paragraphs (b)(2)
introductory text, (b)(2)(vi), and
(b)(2)(viii) introductory text, add
paragraphs (b)(2)(viii)(H) and (I), revise
paragraphs (b)(2)(ix) introductory text,
(b)(2)(ix)(D), and (b)(2)(x) and (xii), add
paragraph (b)(2)(xviii)(D), revise
paragraphs (b)(2)(xxi)(A) and
(b)(2)(xxiii), add and reserve paragraph
(b)(2)(xxx), and add paragraphs
(b)(2)(xxxi) through (xxxvii);
■ h. Revise paragraphs (b)(3)
introductory text and (b)(3)(i) and (ii),
add paragraph (b)(3)(iii), revise
paragraph (b)(3)(iv) introductory text,
and add paragraphs (b)(3)(vii) through
(xi);
■ i. Revise paragraphs (b)(4)
introductory text and (b)(5) and (6);
■ j. Revise paragraphs (f) heading and
introductory text, (f)(2), (f)(3)(iii)(A) and
(B), (f)(3)(iv)(A) and (B), (f)(4)
introductory text, and (f)(4)(i) and (ii);
and
■ k. Revise paragraphs (g) heading and
introductory text, (g)(2), and (g)(3)
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heading, remove paragraph (g)(3)
introductory text, revise paragraphs
(g)(3)(i), (ii), and (v), (g)(4)(i) and (ii),
and (g)(6)(ii)(D)(1) through (4), remove
paragraphs (g)(6)(ii)(D)(5) and (6), revise
paragraphs (g)(6)(ii)(F)(1) through (10),
and add paragraphs (g)(6)(ii)(F)(11)
through (13).
The revisions and additions read as
follows:
sradovich on DSK3GMQ082PROD with RULES2
§ 50.55a
Codes and standards.
(a) Documents approved for
incorporation by reference. The
standards listed in this paragraph (a)
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, by
appointment, at the NRC Technical
Library, which is located at Two White
Flint North, 11545 Rockville Pike,
Rockville, Maryland 20852; telephone:
301–415–7000; email:
Library.Resource@nrc.gov; 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/federalregister/cfr/ibr-locations.html.
(1) * * *
(i) ASME Boiler and Pressure Vessel
Code, Section III. The editions and
addenda for Section III of the ASME
Boiler and Pressure Vessel Code
(excluding Nonmandatory Appendices)
(referred to herein as ASME BPV Code)
are listed in this paragraph (a)(1)(i), but
limited by those provisions identified in
paragraph (b)(1) of this section.
*
*
*
*
*
(E) * * *
(12) 2007 Edition,
(13) 2008 Addenda,
(14) 2009b Addenda (including
Subsection NCA; and Division 1
subsections NB through NH and
Appendices),
(15) 2010 Edition (including
Subsection NCA; and Division 1
subsections NB through NH and
Appendices),
(16) 2011a Addenda (including
Subsection NCA; and Division 1
subsections NB through NH and
Appendices), and
(17) 2013 Edition (including
Subsection NCA; and Division 1
subsections NB through NH and
Appendices).
(ii) ASME Boiler and Pressure Vessel
Code, Section XI. The editions and
addenda for Section XI of the ASME
BPV Code are listed in this paragraph
(a)(1)(ii), but limited by those provisions
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identified in paragraph (b)(2) of this
section.
*
*
*
*
*
(C) * * *
(48) 2007 Edition,
(49) 2008 Addenda,
(50) 2009b Addenda,
(51) 2010 Edition,
(52) 2011a Addenda (Excluding
Article IWB–2000: IWB–2500
‘‘Examination and Inspection:
Examination and Pressure Test
Requirements,’’ Table IWB–2500–1
‘‘Examination Categories,’’ Item
numbers B5.11 and B5.71), and
(53) 2013 Edition (Excluding Article
IWB–2000: IWB–2500 ‘‘Examination
and Inspection: Examination and
Pressure Test Requirements,’’ Table
IWB–2500–1 (B–F) ‘‘Examination
Category B–F, Pressure Retaining
Dissimilar Metal Welds in Vessel
Nozzles,’’ Item numbers B5.11 and
B5.71; Article IWB–3000 ‘‘Acceptance
Standards,’’ IWB–3100 ‘‘Evaluation of
Examination Results,’’ IWB–3110
‘‘Preservice Volumetric and Surface
Examinations,’’ IWB–3112
‘‘Acceptance,’’ paragraph (a)(3); and
Article IWC–3000 ‘‘Acceptance
Standards,’’ IWC–3100 ‘‘Evaluation of
Examination Results,’’ IWC–3110
‘‘Preservice Volumetric and Surface
Examinations,’’ IWC–3112
‘‘Acceptance,’’ paragraph (a)(3)).
(iii) * * *
(A) ASME BPV Code Case N–513–3
Mandatory Appendix I. ASME BPV
Code Case N–513–3, ‘‘Evaluation
Criteria for Temporary Acceptance of
Flaws in Moderate Energy Class 2 or 3
Piping Section XI, Division 1,’’
Mandatory Appendix I, ‘‘Relations for
Fm, Fb, and F for Through-Wall Flaws’’
(Approval Date: January 26, 2009).
ASME BPV Code Case N–513–3
Mandatory Appendix I is referenced in
paragraph (b)(2)(xxxiv)(B) of this
section.
(B) ASME BPV Code Case N–722–1.
ASME BPV 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.
(C) ASME BPV Code Case N–729–4.
ASME BPV Code Case N–729–4,
‘‘Alternative Examination Requirements
for PWR Reactor Vessel Upper Heads
With Nozzles Having Pressure-Retaining
Partial-Penetration Welds Section XI,
Division 1’’ (Approval Date: June 22,
2012), with the conditions in paragraph
(g)(6)(ii)(D) of this section.
(D) ASME BPV Code Case N–770–2.
ASME BPV Code Case N–770–2,
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‘‘Alternative Examination Requirements
and Acceptance Standards for Class 1
PWR Piping and Vessel Nozzle Butt
Welds Fabricated with UNS N06082 or
UNS W86182 Weld Filler Material With
or Without Application of Listed
Mitigation Activities Section XI,
Division 1’’ (Approval Date: June 9,
2011), with the conditions in paragraph
(g)(6)(ii)(F) of this section.
(E) ASME BPV Code Case N–824.
ASME BPV Code Case N–824,
‘‘Ultrasonic Examination of Cast
Austenitic Piping Welds From the
Outside Surface Section XI, Division 1’’
(Approval Date: October 16, 2012), with
the conditions in paragraphs
(b)(2)(xxxvii)(A) through (D) of this
section.
(F) ASME BPV Code Case N–852.
ASME BPV Code Case N–852,
‘‘Application of the ASME NPT Stamp,
Section III, Division 1; Section III,
Division 2; Section III, Division 3;
Section III, Division 5’’ (Approval Date:
February 9, 2015). ASME BPV Code
Case N–852 is referenced in paragraph
(b)(1)(ix) of this section.
(G) ASME OM Code Case OMN–20.
ASME OM Code Case OMN–20,
‘‘Inservice Test Frequency,’’ in the 2012
Edition of the ASME OM Code. OMN–
20 is referenced in paragraph (b)(3)(x) of
this section.
(iv) ASME Operation and
Maintenance Code. The editions and
addenda for the ASME Operation and
Maintenance of Nuclear Power Plants
(various edition titles referred to herein
as ASME OM Code) are listed in this
paragraph (a)(1)(iv), but limited by those
provisions identified in paragraph (b)(3)
of this section.
*
*
*
*
*
(B) ‘‘Operation and Maintenance of
Nuclear Power Plants, Division 1:
Section IST Rules for Inservice Testing
of Light-Water Reactor Power Plants:’’
(1) 2009 Edition; and
(2) 2011 Addenda.
(C) ‘‘Operation and Maintenance of
Nuclear Power Plants, Division 1: OM
Code: Section IST:’’
(1) 2012 Edition.
(2) [Reserved]
(v) ASME Quality Assurance
Requirements. (A) ASME NQA–1,
‘‘Quality Assurance Program
Requirements for Nuclear Facilities:’’
(1) NQA–1—1983 Edition;
(2) NQA–1a—1983 Addenda;
(3) NQA–1b—1984 Addenda;
(4) NQA–1c—1985 Addenda;
(5) NQA–1—1986 Edition;
(6) NQA–1a—1986 Addenda;
(7) NQA–1b—1987 Addenda;
(8) NQA–1c—1988 Addenda;
(9) NQA–1—1989 Edition;
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(10) NQA–1a—1989 Addenda;
(11) NQA–1b—1991 Addenda; and
(12) NQA–1c—1992 Addenda.
(B) ASME NQA–1, ‘‘Quality
Assurance Requirements for Nuclear
Facility Applications:’’
(1) NQA–1—1994 Edition;
(2) NQA–1—2008 Edition; and
(3) NQA–1a—2009 Addenda.
*
*
*
*
*
(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 BPV Code
and the ASME OM Code as specified in
this paragraph (b). 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 10 CFR part 52 is
subject to the following conditions. As
used in this section, references to
Section III refer to Section III of the
ASME BPV Code and include the 1963
Edition through 1973 Winter Addenda
32979
and the 1974 Edition (Division 1)
through the 2013 Edition (Division 1),
subject to the following conditions:
*
*
*
*
*
(ii) Section III condition: Weld leg
dimensions. When applying the 1989
Addenda through the latest edition and
addenda incorporated by reference in
paragraph (a)(1) of this section,
applicants and licensees may not apply
the Section III provisions identified in
Table I of this section for welds with leg
size less than 1.09 tn:
TABLE I—PROHIBITED CODE PROVISIONS
Editions and addenda
1989
1989
2004
2011
Addenda through 2013 Edition ..................................................
Addenda through 2003 Addenda ..............................................
Edition through 2010 Edition .....................................................
Addenda through 2013 Edition ..................................................
*
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Code provision
*
*
*
*
(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 Facility Applications,’’ 1994
Edition, 2008 Edition, and the 2009–1a
Addenda specified in either NCA–4000
or NCA–7000 of that edition and
addenda of Section III may be used by
an applicant or licensee, provided that
the administrative, quality, and
technical provisions contained in that
edition and addenda of Section III are
used in conjunction with the applicant’s
or licensee’s appendix B to this part
quality assurance program; and that the
applicant’s or licensee’s Section III
activities comply with those
commitments contained in the
applicant’s or licensee’s quality
assurance program description. Where
NQA–1 and Section III do not address
the commitments contained in the
applicant’s or licensee’s appendix B
quality assurance program description,
those licensee commitments must be
applied to Section III activities.
*
*
*
*
*
(vii) Section III condition: Capacity
certification and demonstration of
function of incompressible-fluid
pressure-relief valves. When applying
the 2006 Addenda through the 2013
Edition, applicants and licensees may
use paragraph NB–7742, except that
paragraph NB–7742(a)(2) may not be
used. For a valve design of a single size
to be certified over a range of set
pressures, the demonstration of function
tests under paragraph NB–7742 must be
conducted as prescribed in NB–7732.2
on two valves covering the minimum set
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Subparagraph NB–3683.4(c)(1); Subparagraph NB–3683.4(c)(2).
Note 11 to Figure NC–3673.2(b)–1; Note 11 to Figure ND–3673.2(b)–1.
Note 13 to Figure NC–3673.2(b)–1; Note 13 to Figure ND–3673.2(b)–1.
Note 11 to Table NC–3673.2(b)–1; Note 11 to Table ND–3673.2(b)–1.
pressure for the design and the
maximum set pressure that can be
accommodated at the demonstration
facility selected for the test.
(viii) Section III condition: Use of
ASME certification marks. When
applying editions and addenda earlier
than the 2011 Addenda to the 2010
Edition, licensees may use either the
ASME BPV Code Symbol Stamps or the
ASME Certification Marks with the
appropriate certification designators and
class designators as specified in the
2013 Edition through the latest edition
and addenda incorporated by reference
in paragraph (a)(1) of this section.
(ix) Section III Condition: NPT Code
Symbol Stamps. Licensees may use the
NPT Code Symbol Stamp with the
letters arranged horizontally as specified
in ASME BPV Code Case N–852 for the
service life of a component that had the
NPT Code Symbol Stamp applied
during the time period from January 1,
2005, through December 31, 2015.
(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 BPV Code,
and include the 1970 Edition through
the 1976 Winter Addenda and the 1977
Edition through the 2013 Edition,
subject to the following conditions:
*
*
*
*
*
(vi) Section XI condition: Effective
edition and addenda of Subsection IWE
and Subsection IWL. Licensees that
implemented the expedited examination
of containment, in accordance with
Subsection IWE and Subsection IWL,
during the period from September 9,
1996, to September 9, 2001, may use
either the 1992 Edition with the 1992
Addenda or the 1995 Edition with the
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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.
*
*
*
*
*
(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 up to and including the
2008 Addenda must apply paragraph
(b)(2)(viii)(E) of this section. Applicants
or licensees applying Subsection IWL,
2007 Edition with the 2009 Addenda
through the latest edition and addenda
incorporated by reference in paragraph
(a)(1)(ii) of this section, must apply
paragraphs (b)(2)(viii)(H) and (I) of this
section.
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*
*
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(H) Concrete containment
examinations: Eighth provision. For
each inaccessible area of concrete
identified for evaluation under IWL–
2512(a), or identified as susceptible to
deterioration under IWL–2512(b), the
licensee must provide the applicable
information specified in paragraphs
(b)(2)(viii)(E)(1), (2), and (3) of this
section in the ISI Summary Report
required by IWA–6000.
(I) Concrete containment
examinations: Ninth provision. During
the period of extended operation of a
renewed license under part 54 of this
chapter, the licensee must perform the
technical evaluation under IWL–2512(b)
of inaccessible below-grade concrete
surfaces exposed to foundation soil,
backfill, or groundwater at periodic
intervals not to exceed 5 years. In
addition, the licensee must examine
representative samples of the exposed
portions of the below-grade concrete,
when such below-grade concrete is
excavated for any reason.
(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
(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 (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 or licensees
applying Subsection IWE, 2007 Edition
through the latest edition and 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.
*
*
*
*
*
(D) Metal containment examinations:
Fourth provision. This paragraph
(b)(2)(ix)(D) may be used as an
alternative to the requirements of IWE–
2430. 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
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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:
(1) A description of each flaw or area,
including the extent of degradation, and
the conditions that led to the
degradation;
(2) The acceptability of each flaw or
area and the need for additional
examinations to verify that similar
degradation does not exist in similar
components;
(3) A description of necessary
corrective actions; and
(4) The number and type of additional
examinations to ensure detection of
similar degradation in similar
components.
*
*
*
*
*
(x) Section XI condition: Quality
assurance. When applying the editions
and addenda later than the 1989 Edition
of ASME BPV Code, Section XI, the
edition and addenda of NQA–1,
‘‘Quality Assurance Requirements for
Nuclear Facility Applications,’’ 1994
Edition, the 2008 Edition, and the 2009–
1a Addenda specified in either IWA–
1400 or Table IWA 1600–1 of that
edition and addenda of Section XI, may
be used by a licensee provided that the
licensee uses its appendix B to this part
quality assurance program in
conjunction with Section XI
requirements and the commitments
contained in the licensee’s quality
assurance program description. Where
NQA–1 and Section XI do not address
the commitments contained in the
licensee’s appendix B quality assurance
program description, those licensee
commitments must be applied to
Section XI activities.
*
*
*
*
*
(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
approved for use on irradiated material
with the following conditions:
(A) Underwater welding: First
provision. Licensees must obtain NRC
approval in accordance with paragraph
(z) of this section regarding the welding
technique to be used prior to performing
welding on ferritic material exposed to
fast neutron fluence greater than 1 ×
1017 n/cm2 (E > 1 MeV).
(B) Underwater welding: Second
provision. Licensees must obtain NRC
approval in accordance with paragraph
(z) of this section regarding the welding
technique to be used prior to performing
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welding on austenitic material other
than P-No. 8 material exposed to
thermal neutron fluence greater than 1
× 1017 n/cm2 (E < 0.5 eV). Licensees
must obtain NRC approval in
accordance with paragraph (z) regarding
the welding technique to be used prior
to performing welding on P-No. 8
austenitic material exposed to thermal
neutron fluence greater than 1 × 1017
n/cm2 (E < 0.5 eV) and measured or
calculated helium concentration of the
material greater than 0.1 atomic parts
per million.
*
*
*
*
*
(xviii) * * *
(D) NDE personnel certification:
Fourth provision. The use of Appendix
VII and Subarticle VIII–2200 of the 2011
Addenda and 2013 Edition of Section XI
of the ASME BPV Code is prohibited.
When using ASME BPV Code, Section
XI editions and addenda later than the
2010 Edition, licensees and applicants
must use the prerequisites for ultrasonic
examination personnel certifications in
Table VII–4110–1 and Subarticle VIII–
2200, Appendix VIII in the 2010
Edition.
*
*
*
*
*
(xxi) * * *
(A) Table IWB–2500–1 examination
requirements: First provision. The
provisions of Table IWB 2500–1,
Examination Category B–D, Full
Penetration Welded Nozzles in Vessels,
Items B3.40 and B3.60 (Inspection
Program A) and Items B3.120 and
B3.140 (Inspection Program B) of the
1998 Edition must be applied when
using the 1999 Addenda through the
latest edition and addenda incorporated
by reference in paragraph (a)(1)(ii) of
this section. A visual examination with
magnification that has a resolution
sensitivity to resolve 0.044 inch (1.1
mm) lower case characters without an
ascender or descender (e.g., a, e, n, v),
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.
*
*
*
*
*
(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
2009 Addenda, is prohibited.
*
*
*
*
*
(xxx) [Reserved]
(xxxi) Section XI condition:
Mechanical clamping devices. When
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installing a mechanical clamping device
on an ASME BPV Code class piping
system, Appendix W of Section XI shall
be treated as a mandatory appendix and
all of the provisions of Appendix W
shall be met for the mechanical
clamping device being installed.
Additionally, use of IWA–4131.1(c) of
the 2010 Edition of Section XI and
IWA–4131.1(d) of the 2011 Addenda of
the 2010 Edition and later versions of
Section XI is prohibited on small item
Class 1 piping and portions of a piping
system that form the containment
boundary.
(xxxii) Section XI condition:
Summary report submittal. When using
ASME BPV Code, Section XI, 2010
Edition through the latest edition and
addenda incorporated by reference in
paragraph (a)(1)(ii) of this section,
Summary Reports described in IWA–
6000 must be submitted to the NRC as
described in IWA–6240(a) and IWA–
6240(b). Preservice inspection summary
reports shall be submitted prior to the
date of placement of the unit into
commercial service and inservice
inspection summary reports shall be
submitted within 90 calendar days of
the completion of each refueling outage.
(xxxiii) Section XI condition: RiskInformed allowable pressure. The use of
Paragraph G–2216 in Appendix G in the
2011 Addenda and later editions and
addenda of the ASME BPV Code,
Section XI is prohibited.
(xxxiv) Section XI condition:
Nonmandatory Appendix U. When
using Nonmandatory Appendix U of the
2013 Edition of the ASME BPV Code,
Section XI the following conditions
apply:
(A) The repair or replacement
activities temporarily deferred under the
provisions of Nonmandatory Appendix
U must be performed during the next
scheduled refueling outage.
(B) In lieu of the appendix referenced
in paragraph U–S1–4.2.1(c) of Appendix
U the mandatory appendix in ASME
BPV Code Case N–513–3 must be used.
(xxxv) Section XI condition: Use of
RTT0 in the KIa and KIc equations. When
using the 2013 Edition of the ASME
BPV Code, Section XI, Appendix A,
paragraph A–4200, if T0 is available,
then RTT0 may be used in place of
RTNDT for applications using the KIc
equation and the associated KIc curve,
but not for applications using the KIa
equation and the associated KIa curve.
(xxxvi) Section XI condition: Fracture
toughness of irradiated materials. When
using the 2013 Edition of the ASME
BPV Code, Section XI, Appendix A
paragraph A–4400, the licensee shall
obtain NRC approval under paragraph
(z) of this section before using irradiated
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T0 and the associated RTT0 in
establishing fracture toughness of
irradiated materials.
(xxxvii) Section XI condition: ASME
BPV Code Case N–824. Licensees may
use the provisions of ASME BPV Code
Case N–824, ‘‘Ultrasonic Examination of
Cast Austenitic Piping Welds From the
Outside Surface Section XI, Division 1,’’
subject to the following conditions.
(A) Ultrasonic examinations must be
spatially encoded.
(B) Instead of Paragraph 1(c)(1)(–a),
licensees shall use dual, transmitreceive, refracted longitudinal wave,
multi-element phased array search
units.
(C) Instead of Paragraph 1(c)(1)(–c)(–
2), licensees shall use a phased array
search unit with a center frequency of
500 kHz with a tolerance of ± 20
percent.
(D) Instead of Paragraph 1(c)(1)(–d),
the phased array search unit must
produce angles including, but not
limited to, 30 to 55 degrees with a
maximum increment of 5 degrees.
(3) Conditions on ASME OM Code. As
used in this section, references to the
ASME OM Code are to the ASME OM
Code, Subsections ISTA, ISTB, ISTC,
ISTD, ISTE, and ISTF; Mandatory
Appendices I, II, III, and V; and
Nonmandatory Appendices A through H
and J through M, in the 1995 Edition
through the 2012 Edition, as specified in
paragraph (a)(1)(iv) of this section.
Mandatory appendices must be used if
required by the OM Code;
nonmandatory appendices are approved
for use by the NRC but need not be
used. The following conditions are
applicable when implementing the
ASME OM Code:
(i) OM condition: Quality assurance.
When applying editions and addenda of
the ASME OM Code, the requirements
of ASME Standard NQA–1, ‘‘Quality
Assurance Requirements for Nuclear
Facility Applications,’’ 1994 Edition,
2008 Edition, and 2009–1a Addenda,
are acceptable as permitted by either
ISTA 1.4 of the 1995 Edition through
1997 Addenda or ISTA–1500 of the
1998 Edition through the latest edition
and addenda of the ASME OM Code
incorporated by reference in paragraph
(a)(1)(iv) of this section, provided the
licensee uses its appendix B to this part
quality assurance program in
conjunction with the ASME OM Code
requirements and the commitments
contained in the licensee’s quality
assurance program description. Where
NQA–1 and the ASME OM Code do not
address the commitments contained in
the licensee’s appendix B quality
assurance program description, the
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32981
commitments must be applied to ASME
OM Code activities.
(ii) OM condition: Motor-Operated
Valve (MOV) testing. Licensees must
comply with the provisions for testing
MOVs in ASME 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 MOVs
continue to be capable of performing
their design basis safety functions.
Licensees implementing ASME OM
Code, Mandatory Appendix III,
‘‘Preservice and Inservice Testing of
Active Electric Motor Operated Valve
Assemblies in Light-Water Reactor
Power Plants,’’ of the 2009 Edition, 2011
Addenda, and 2012 Edition shall
comply with the following conditions:
(A) MOV diagnostic test interval.
Licensees shall evaluate the adequacy of
the diagnostic test intervals established
for MOVs within the scope of ASME
OM Code, Appendix III, not later than
5 years or three refueling outages
(whichever is longer) from initial
implementation of ASME OM Code,
Appendix III.
(B) MOV testing impact on risk.
Licensees shall ensure that the potential
increase in core damage frequency and
large early release frequency associated
with the extension is acceptably small
when extending exercise test intervals
for high risk MOVs beyond a quarterly
frequency.
(C) MOV risk categorization. When
applying Appendix III to the ASME OM
Code, licensees shall categorize MOVs
according to their safety significance
using the methodology described in
ASME OM Code Case OMN–3,
‘‘Requirements for Safety Significance
Categorization of Components Using
Risk Insights for Inservice Testing of
LWR Power Plants,’’ subject to the
conditions applicable to OMN–3 which
are set forth in Regulatory Guide 1.192,
or using an MOV risk ranking
methodology accepted by the NRC on a
plant-specific or industry-wide basis in
accordance with the conditions in the
applicable safety evaluation.
(D) MOV stroke time. When applying
Paragraph III–3600, ‘‘MOV Exercising
Requirements,’’ of Appendix III to the
ASME OM Code, licensees shall verify
that the stroke time of MOVs specified
in plant technical specifications satisfies
the assumptions in the plant’s safety
analyses.
(iii) OM condition: New reactors. In
addition to complying with the
provisions in the ASME OM Code with
the conditions specified in paragraph
(b)(3) of this section, holders of
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operating licenses for nuclear power
reactors that received construction
permits under this part on or after the
date 12 months after August 17, 2017,
and holders of combined licenses issued
under 10 CFR part 52, whose initial fuel
loading occurs on or after the date 12
months after August 17, 2017, shall also
comply with the following conditions,
as applicable:
(A) Power-operated valves. Licensees
shall periodically verify the capability
of power-operated valves to perform
their design-basis safety functions.
(B) Check valves. Licensees must
perform bi-directional testing of check
valves within the IST program where
practicable.
(C) Flow-induced vibration. Licensees
shall monitor flow-induced vibration
from hydrodynamic loads and acoustic
resonance during preservice testing or
inservice testing to identify potential
adverse flow effects on components
within the scope of the IST program.
(D) High risk non-safety systems.
Licensees shall assess the operational
readiness of pumps, valves, and
dynamic restraints within the scope of
the Regulatory Treatment of Non-Safety
Systems for applicable reactor designs.
(iv) OM condition: Check valves
(Appendix II). Licensees applying
Appendix II, ‘‘Check Valve Condition
Monitoring Program,’’ of the ASME OM
Code, 1995 Edition with the 1996 and
1997 Addenda, shall satisfy the
requirements of paragraphs (b)(3)(iv)(A)
through (C) of this section. Licensees
applying Appendix II, 1998 Edition
through the 2012 Edition, shall satisfy
the requirements of paragraphs
(b)(3)(iv)(A), (B), and (D) of this section.
Appendix II of the ASME OM Code,
2003 Addenda through the 2012
Edition, is acceptable for use with the
following requirements. Trending and
evaluation shall support the
determination that the valve or group of
valves is capable of performing its
intended function(s) over the entire
interval. At least one of the Appendix II
condition monitoring activities for a
valve group shall be performed on each
valve of the group at approximate equal
intervals not to exceed the maximum
interval shown in the following table:
TABLE II—MAXIMUM INTERVALS FOR USE WHEN APPLYING INTERVAL EXTENSIONS
Maximum
interval between
activities of
member valves
in the groups
(years)
Group size
≥4 .....................................................................................................................................................................
3 .......................................................................................................................................................................
2 .......................................................................................................................................................................
1 .......................................................................................................................................................................
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(vii) OM condition: Subsection ISTB.
Subsection ISTB, 2011 Addenda, is
prohibited for use.
(viii) OM condition: Subsection ISTE.
Licensees may not implement the riskinformed approach for inservice testing
(IST) of pumps and valves specified in
Subsection ISTE, ‘‘Risk-Informed
Inservice Testing of Components in
Light-Water Reactor Nuclear Power
Plants,’’ in the ASME OM Code, 2009
Edition, 2011 Addenda, or 2012 Edition,
without first obtaining NRC
authorization to use Subsection ISTE as
an alternative to the applicable IST
requirements in the ASME OM Code,
pursuant to paragraph (z) of this section.
(ix) OM condition: Subsection ISTF.
Licensees applying Subsection ISTF,
2012 Edition, shall satisfy the
requirements of Mandatory Appendix V,
‘‘Pump Periodic Verification Test
Program,’’ of the ASME OM Code, 2012
Edition. Subsection ISTF, 2011
Addenda, is prohibited for use.
(x) OM condition: ASME OM Code
Case OMN–20. Licensees may
implement ASME OM Code Case OMN–
20, ‘‘Inservice Test Frequency,’’ which
is incorporated by reference in
paragraph (a)(1)(iii)(G) of this section,
for editions and addenda of the ASME
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OM Code listed in paragraph (a)(1)(iv) of
this section.
(xi) OM condition: Valve Position
Indication. When implementing ASME
OM Code, 2012 Edition, Subsection
ISTC–3700, ‘‘Position Verification
Testing,’’ licensees shall verify that
valve operation is accurately indicated
by supplementing valve position
indicating lights with other indications,
such as flow meters or other suitable
instrumentation, to provide assurance of
proper obturator position.
(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, as incorporated by reference
in paragraph (a)(3)(i) of this section,
without prior NRC approval, subject to
the following conditions:
*
*
*
*
*
(5) Conditions on inservice inspection
Code Cases. Licensees may apply the
ASME BPV Code Cases listed in NRC
Regulatory Guide 1.147, as incorporated
by reference in paragraph (a)(3)(ii) of
this section, without prior NRC
approval, subject to the following:
PO 00000
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Maximum
interval between
activities of
each valve
in the group
(years)
4.5
4.5
6
Not applicable
16
12
12
10
(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 this
section as listed in Tables 1 and 2 of
NRC Regulatory Guide 1.147, as
incorporated by reference in paragraph
(a)(3)(ii) of this section.
(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 NRC
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Regulatory Guide 1.147. If a licensee has
applied a listed Code Case that is later
listed as annulled in NRC 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 ASME OM Code
Cases. Licensees may apply the ASME
OM Code Cases listed in NRC
Regulatory Guide 1.192, as incorporated
by reference in paragraph (a)(3)(iii) of
this section, 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 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 this section as listed in
Tables 1 and 2 of NRC Regulatory Guide
1.192, as incorporated by reference in
paragraph (a)(3)(iii) of this section.
(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 NRC
Regulatory Guide 1.192. If a licensee has
applied a listed Code Case that is later
listed as annulled in NRC Regulatory
Guide 1.192, the licensee may continue
to apply the Code Case to the end of the
current 120-month interval.
*
*
*
*
*
(f) Preservice and inservice testing
requirements. Systems and components
of boiling and pressurized water-cooled
nuclear power reactors must meet the
requirements for preservice and
inservice testing (referred to in this
paragraph (f) collectively as inservice
testing) of the ASME BPV Code and
ASME OM Code as specified in this
paragraph (f). 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
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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 paragraph (g) of
this section.
*
*
*
*
*
(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 BPV 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 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 or NRC Regulatory Guide 1.192, as
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 or NRC Regulatory Guide 1.192, as
incorporated by reference in paragraphs
(a)(3)(ii) and (iii) of this section,
respectively), subject to the applicable
conditions listed therein.
*
*
*
*
*
(3) * * *
(iii) * * *
(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 BPV 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 or NRC
Regulatory Guide 1.192, as incorporated
by reference in paragraphs (a)(3)(ii) and
(iii) of this section, respectively) applied
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32983
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 BPV 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 OM
Code Cases listed in NRC Regulatory
Guide 1.192, as 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) * * *
(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 BPV 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 BPV
Code Cases listed in NRC Regulatory
Guide 1.147, as 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 BPV 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, as incorporated by
reference in paragraph (a)(3)(iii) of this
section), incorporated by reference in
paragraph (a)(1)(iv) of this section at the
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time the construction permit, combined
license, or design certification is issued.
*
*
*
*
*
(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 within the scope of the ASME OM
Code 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. The inservice test
requirements for pumps and valves that
are within the scope of the ASME OM
Code but are not classified as ASME
BPV Code Class 1, Class 2, or Class 3
may be satisfied as an augmented IST
program in accordance with paragraph
(f)(6)(ii) of this section without
requesting relief under paragraph (f)(5)
of this section or alternatives under
paragraph (z) of this section. This use of
an augmented IST program may be
acceptable provided the basis for
deviations from the ASME OM Code, as
incorporated by reference in this
section, demonstrates an acceptable
level of quality and safety, or that
implementing the Code provisions
would result in hardship or unusual
difficulty without a compensating
increase in the level of quality and
safety, where documented and available
for NRC review.
(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 ASME 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 OM Code Cases
listed in NRC Regulatory Guide 1.192,
as incorporated by reference in
paragraph (a)(3)(iii) of this section,
subject to the conditions listed in
paragraph (b) of this section).
(ii) Applicable IST Code: Successive
120-month intervals. Inservice tests to
verify operational readiness of pumps
and valves, whose function is required
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for safety, conducted during successive
120-month intervals must comply with
the requirements of the latest edition
and addenda of the ASME 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 or NRC Regulatory Guide 1.192 as
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.
*
*
*
*
*
(g) Preservice and 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 paragraph (f) of
this section.
*
*
*
*
*
(2) Accessibility requirements—(i)
Accessibility requirements for plants
with CPs issued between 1971 and 1974.
For a boiling or pressurized watercooled nuclear power facility whose
construction permit was issued on or
after January 1, 1971, but before July 1,
1974, components that are classified as
ASME BPV Code Class 1 and Class 2
and supports for components that are
classified as ASME BPV Code Class 1
and Class 2 must be designed and be
provided with the access necessary to
perform the required preservice and
inservice examinations 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 BPV Code Cases listed in NRC
Regulatory Guide 1.147, as incorporated
by reference in paragraph (a)(3)(ii) of
this section) in effect 6 months before
the date of issuance of the construction
permit.
(ii) Accessibility requirements for
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,
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Fmt 4701
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combined license, or manufacturing
license under part 52 of this chapter,
was issued on or after July 1, 1974,
components that are classified as ASME
BPV Code Class 1, Class 2, and Class 3
and supports for components that are
classified as ASME BPV Code Class 1,
Class 2, and Class 3 must be designed
and provided with the access necessary
to perform the required preservice and
inservice examinations 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 BPV Code Cases listed in NRC
Regulatory Guide 1.147, as incorporated
by reference in paragraph (a)(3)(ii) of
this section) applied to the construction
of the particular component.
(iii) Accessibility requirements:
Meeting later Code 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.
(3) Preservice examination
requirements—(i) Preservice
examination requirements for plants
with CPs issued between 1971 and 1974.
For a boiling or pressurized watercooled nuclear power facility whose
construction permit was issued on or
after January 1, 1971, but before July 1,
1974, components that are classified as
ASME BPV Code Class 1 and Class 2
and supports for components that are
classified as ASME BPV Code Class 1
and Class 2 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 BPV Code Cases listed in NRC
Regulatory Guide 1.147, as incorporated
by reference in paragraph (a)(3)(ii) of
this section) in effect 6 months before
the date of issuance of the construction
permit.
(ii) Preservice examination
requirements for 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, components that are classified
as ASME BPV Code Class 1, Class 2, and
Class 3 and supports for components
that are classified as ASME BPV Code
Class 1, Class 2, and Class 3 must meet
the preservice examination
requirements set forth in the editions
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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 BPV
Code Cases listed in NRC Regulatory
Guide 1.147, as incorporated by
reference in paragraph (a)(3)(ii) of this
section) applied to the construction of
the particular component.
*
*
*
*
*
(v) Preservice examination
requirements: Meeting later Code
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) * * *
(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 ASME 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, when
using ASME BPV Code, Section XI, or
NRC Regulatory Guide 1.192, when
using the ASME OM Code, as
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.
Licensees may, at any time in their 120month ISI interval, elect to use the
Appendix VIII in the latest edition and
addenda of the ASME BPV Code
incorporated by reference in paragraph
(a) of this section, subject to any
applicable conditions listed in
paragraph (b) of this section. Licensees
using this option must also use the same
edition and addenda of Appendix I as
Appendix VIII, including any applicable
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 ASME Code
incorporated by reference in paragraph
(a) of this section 12 months before the
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start of the 120-month inspection
interval (or the optional ASME Code
Cases listed in NRC Regulatory Guide
1.147, when using ASME BPV Code,
Section XI, or NRC Regulatory Guide
1.192, when using the ASME OM Code,
as 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
August 17, 2017, may delay the update
of their Appendix VIII program by up to
18 months after August 17, 2017.
Alternatively, licensees may, at any time
in their 120-month ISI interval, elect to
use the Appendix VIII in the latest
edition and addenda of the ASME BPV
Code incorporated by reference in
paragraph (a) of this section, subject to
any applicable conditions listed in
paragraph (b) of this section. Licensees
using this option must also use the same
Edition and Addenda of Appendix I as
Appendix VIII, including any applicable
conditions listed in paragraph (b) of this
section.
*
*
*
*
*
(6) * * *
(ii) * * *
(D) * * *
(1) Implementation. Holders of
operating licenses or combined licenses
for pressurized-water reactors as of or
after August 17, 2017 shall implement
the requirements of ASME BPV Code
Case N–729–4 instead of ASME BPV
Code Case N–729–1, subject to the
conditions specified in paragraphs
(g)(6)(ii)(D)(2) through (4) of this
section, by the first refueling outage
starting after August 17, 2017.
(2) Appendix I use. Appendix I of
ASME BPV Code Case N–729–4 shall
not be implemented without prior NRC
approval.
(3) Bare metal visual frequency.
Instead of Note 4 of ASME BPV Code
Case N–729–4, the following shall be
implemented. If effective degradation
years (EDY) < 8 and if no flaws are
found that are attributed to primary
water stress corrosion cracking:
(i) A bare metal visual examination is
not required during refueling outages
when a volumetric or surface
examination is performed; and
(ii) If a wetted surface examination
has been performed of all of the partial
penetration welds during the previous
non-visual examination, the
reexamination frequency may be
extended to every third refueling outage
or 5 calendar years, whichever is less,
provided an IWA–2212 VT–2 visual
examination of the head is performed
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32985
under the insulation through multiple
access points in outages that the VE is
not completed. This IWA–2212 VT–2
visual examination may be performed
with the reactor vessel depressurized.
(4) Surface exam acceptance criteria.
In addition to the requirements of
Paragraph –3132.1(b) of ASME BPV
Code Case N–729–4, a component
whose surface examination detects
rounded indications greater than
allowed in Paragraph NB–5352 in size
on the partial-penetration or associated
fillet weld shall be classified as having
an unacceptable indication and
corrected in accordance with the
provisions of paragraph–3132.2 of
ASME BPV Code Case N–729–4.
*
*
*
*
*
(F) * * *
(1) Implementation. Holders of
operating licenses or combined licenses
for pressurized-water reactors as of or
after August 17, 2017, shall implement
the requirements of ASME BPV Code
Case N–770–2 instead of ASME BPV
Code Case N–770–1, subject to the
conditions specified in paragraphs
(g)(6)(ii)(F)(2) through (13) of this
section, by the first refueling outage
starting after August 17, 2017.
(2) Categorization. Full structural
weld overlays, authorized by the NRC
staff in accordance with the alternatives
approval process of this section, may be
categorized as Inspection Items C–1 or
F–1, 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 the purpose of determining ISI
frequencies, all other butt welds that
rely on Alloy 82/182 for structural
integrity shall 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 an alternative code
case Inspection Item is used based on
conformance with an ASME mitigation
code case endorsed in NRC Regulatory
Guide 1.147 with any applying
conditions specified in NRC Regulatory
Guide 1.147, as incorporated by
reference in paragraph (a)(3)(ii) of this
section. Paragraph –1100(e) of ASME
BPV Code Case N–770–2 shall not be
used to exempt welds that rely on Alloy
82/182 for structural integrity from any
requirement of paragraph (g)(6)(ii)(F) of
this section.
(3) Baseline examinations. Baseline
examinations for welds in Table 1 of
ASME BPV Code Case N–770–2,
Inspection Items A–1, A–2, and B, if not
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previously performed or currently
scheduled to be performed in an
ongoing refueling outage as of August
17, 2017, in accordance with paragraph
(g)(6)(ii)(F) of this section, shall be
completed by the end of the next
refueling outage. Previous examinations
of these welds can be credited for
baseline examinations only if they were
performed within the re-inspection
period for the weld item in Table 1 of
ASME BPV Code Case N–770–2 and the
examination of each weld meets the
examination requirements of paragraphs
–2500(a) or –2500(b) of ASME BPV
Code Case N–770–2 as conditioned in
this section. Other previous
examinations that do not meet these
requirements can be used to meet the
baseline examination requirement,
provided NRC approval in accordance
with paragraph (z)(1) or (2) of this
section, is granted prior to the end of the
next refueling outage.
(4) Examination coverage. When
implementing Paragraph –2500(a) of
ASME BPV Code Case N–770–2,
essentially 100 percent of the required
volumetric examination coverage shall
be obtained, including greater than 90
percent of the volumetric examination
coverage for circumferential flaws.
Licensees are prohibited from using
Paragraphs –2500(c) and –2500(d) of
ASME BPV Code Case N–770–2 to meet
examination requirements.
(5) Inlay/onlay inspection frequency.
All hot-leg operating temperature welds
in Inspection Items G, H, J, and K shall
be inspected each inspection interval. A
25 percent sample of Inspection Items
G, H, J, and K cold-leg operating
temperature welds shall be inspected
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whenever the core barrel is removed
(unless it has already been inspected
within the past 10 years) or within 20
years, whichever is less.
(6) Reporting requirements. 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) Defining ‘‘t’’. 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) Optimized weld overlay
examination. Initial inservice
examination of Inspection Item C–2
welds shall be performed between the
third refueling outage and no later than
10 years after application of the overlay.
(9) Deferral. Note (11)(b)(1) in ASME
BPV Code Case N–770–2 shall not be
used to defer the initial inservice
examination of optimized weld overlays
(i.e., Inspection Item C–2 of ASME BPV
Code Case N–770–2).
(10) Examination technique. Note
14(b) of Table 1 and Note (b) of Figure
5(a) of ASME BPV Code Case N–770–2
may only be implemented if the
PO 00000
Frm 00054
Fmt 4701
Sfmt 9990
requirements of Note 14(a) of Table 1 of
ASME BPV Code Case N–770–2 cannot
be met.
(11) Cast stainless steel. Examination
of ASME BPV Code Class 1 piping and
vessel nozzle butt welds involving cast
stainless steel materials, shall be
performed with Appendix VIII,
Supplement 9 qualifications, or
qualifications similar to Appendix VIII,
Supplement 2 or 10 using cast stainless
steel mockups no later than the next
scheduled weld examination after
January 1, 2022, in accordance with the
requirements of Paragraph –2500(a).
(12) Stress improvement inspection
coverage. Under Paragraph I.5.1, for cast
stainless steel items, the required
examination volume shall be examined
by Appendix VIII procedures to the
maximum extent practical including
100 percent of the susceptible material
volume.
(13) Encoded ultrasonic examination.
Ultrasonic examinations of nonmitigated or cracked mitigated
dissimilar metal butt welds in the
reactor coolant pressure boundary must
be performed in accordance with the
requirements of Table 1 for Inspection
Item A–1, A–2, B, E, F–2, J, and K for
100 percent of the required inspection
volume using an encoded method.
*
*
*
*
*
Dated at Rockville, Maryland, this 30th day
of June 2017.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Acting Director, Office of Nuclear Reactor
Regulation.
[FR Doc. 2017–14166 Filed 7–17–17; 8:45 am]
BILLING CODE 7590–01–P
E:\FR\FM\18JYR2.SGM
18JYR2
Agencies
[Federal Register Volume 82, Number 136 (Tuesday, July 18, 2017)]
[Rules and Regulations]
[Pages 32934-32986]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-14166]
[[Page 32933]]
Vol. 82
Tuesday,
No. 136
July 18, 2017
Part II
Nuclear Regulatory Commission
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10 CFR Part 50
Incorporation by Reference of American Society of Mechanical Engineers
Codes and Code Cases; Final Rule
Federal Register / Vol. 82 , No. 136 / Tuesday, July 18, 2017 / Rules
and Regulations
[[Page 32934]]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2011-0088]
RIN 3150-AI97
Incorporation by Reference of American Society of Mechanical
Engineers Codes and 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 recent editions and addenda to
the American Society of Mechanical Engineers (ASME) Codes for nuclear
power plants and a standard for quality assurance. The NRC is also
incorporating by reference six ASME Code Cases. This action is in
accordance with the NRC's policy to periodically update the regulations
to incorporate by reference new editions and addenda of the ASME Codes
and is intended to maintain the safety of nuclear power plants and to
make NRC activities more effective and efficient.
DATES: This final rule is effective on August 17, 2017. The
incorporation by reference of certain publications listed in the
regulation is approved by the Director of the Federal Register as of
August 17, 2017.
ADDRESSES: Please refer to Docket ID NRC-2011-0088 when contacting the
NRC about the availability of information for this action. You may
obtain publicly-available information related to this action by any of
the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2011-0088. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. For
the convenience of the reader, instructions about obtaining materials
referenced in this document are provided in the ``Availability of
Documents'' section.
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: Daniel I. Doyle, Office of Nuclear
Reactor Regulation, telephone: 301-415-3748, email:
Daniel.Doyle@nrc.gov; or Keith Hoffman, Office of Nuclear Reactor
Regulation, telephone: 301-415-1294, email: Keith.Hoffman@nrc.gov. Both
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
The NRC is amending its regulations to incorporate by reference
recent editions and addenda to the ASME Codes for nuclear power plants
and an ASME standard for quality assurance. The NRC is also
incorporating by reference six ASME Code Cases.
This final rule is the latest in a series of rulemakings to amend
the NRC's regulations to incorporate by reference revised and updated
ASME Codes for nuclear power plants. The ASME is a voluntary consensus
standards body, and the ASME Codes are voluntary consensus standards.
The ASME periodically revises and updates its codes for nuclear power
plants by issuing new editions and addenda. The NRC's use of the ASME
Codes is consistent with applicable requirements of the National
Technology Transfer and Advancement Act (NTTAA). This rulemaking is in
accordance with the NRC's policy to update the regulations to
incorporate by reference those new editions and addenda. The
incorporation by reference of the new editions and addenda will
maintain the safety of nuclear power plants, make NRC activities more
effective and efficient, and allow nuclear power plant licensees and
applicants to take advantage of the latest ASME Codes. Additional
discussion of voluntary consensus standards and the NRC's compliance
with the NTTAA is set forth in Section XIV of this document,
``Voluntary Consensus Standards.''
B. Major Provisions
Major provisions of this final rule include:
Incorporation by reference of ASME Codes into the NRC's
regulations and delineation of the NRC's requirements for the use of
these codes, including conditions.
Incorporation by reference of various versions of quality
assurance standard NQA-1 into NRC regulations and approval for their
use.
Incorporation by reference of six ASME Code Cases.
C. Costs and Benefits
The NRC prepared a regulatory analysis (ADAMS Accession No.
ML16130A522) to identify the costs and benefits associated with this
final rule. The regulatory analysis prepared for this rulemaking was
used to determine if the rule is cost-effective, overall, and to help
the NRC evaluate potentially costly conditions placed on specific
provisions of the ASME Codes and Code Cases which are the subject of
this rulemaking. Therefore, the regulatory analysis focuses on the
marginal difference in benefits and costs for each provision of this
final rule relative to the ``no action'' baseline alternative. The
regulatory analysis identified costs and benefits in a quantitative
fashion as well as in a qualitative fashion. An uncertainty analysis
was performed to evaluate the effects of uncertainties in the
quantitative estimation of both costs and benefits, and this analysis
showed the rule alternative is cost effective with over 99 percent
certainty. The standard deviation of the cost estimate net benefit is
$4.1 million.
Table 1--Cost-Benefit Summary
------------------------------------------------------------------------
Alternative 2--
the rule
alternative
net benefits
Objective (costs)
(million) (Net
present value,
7% discount
rate)
------------------------------------------------------------------------
Industry................................................ $11.5
NRC..................................................... 3.28
Net benefit............................................. 14.7
------------------------------------------------------------------------
Table 1 summarizes the costs and benefits for the alternative of
proceeding with the final rule (Alternative 2) and shows that the final
rule is quantitatively cost-beneficial with a net benefit of $14.7
million to both the industry and the NRC when compared to the
regulatory baseline (Alternative 1). The regulatory analysis shows that
implementing the final rule is quantitatively cost-effective and an
efficient use of NRC and Industry resources. Uncertainty analysis shows
a standard deviation of $4.08 million, resulting in a net benefit range
of $8.19 million to $21.6 million. Because the
[[Page 32935]]
rulemaking alternative is cost-effective, the rulemaking approach is
recommended.
There are several benefits associated with this final rule. The new
motor-operated valve (MOV) provisions in this final rule result in over
$25 million in averted costs (7-percent net present value) due to the
removal of quarterly testing requirements and replacing those
requirements with less frequent diagnostic and biannual testing
requirements. Additionally, the provisions in this final rule will
result in averted costs to the NRC and the industry from relief
requests for the code cases in this final rule, in particular the ASME
OMN-20 Code Case Time Period Extension provision, in excess of $5.1
million (7-percent net present value).
Qualitative factors which were considered include regulatory
stability and predictability, regulatory efficiency, and consistency
with the NTTAA. Table 50 in the regulatory analysis includes a
discussion of the costs and benefits that were considered
qualitatively. Considering non-quantified costs and benefits, the
regulatory analysis shows that the rulemaking is justified because the
number and significance of the non-quantified benefits outweigh the
non-quantified costs. Certainly, if the qualitative benefits (including
the safety benefit, regulatory efficiency, and other nonquantified
benefits) are considered together with the quantified benefits, then
the benefits would outweigh the identified quantitative and qualitative
impacts. Therefore, integrating both quantified and non-quantified
costs and benefits, the benefits of the final rule outweigh the
identified quantitative and qualitative impacts attributable to the
final rule.
Table of Contents
I. Background
II. Discussion
A. ASME BPV Code, Section III
B. ASME BPV Code, Section XI
C. OM Code
D. ASME Code Cases
III. Opportunities for Public Participation
IV. NRC Responses to Public Comments
V. Section-by-Section Analysis
VI. Generic Aging Lessons Learned Report
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Plain Writing
XI. Finding of No Significant Impact: Environmental Assessment
XII. Paperwork Reduction Act Statement
XIII. Congressional Review Act
XIV. Voluntary Consensus Standards
XV. Incorporation by Reference--Reasonable Availability to
Interested Parties
XVI. Availability of Guidance
XVII. Availability of Documents
I. Background
The ASME develops and publishes the ASME Boiler and Pressure Vessel
Code (BPV Code), which contains requirements for the design,
construction, and inservice inspection (ISI) of nuclear power plant
components; and the OM Code,\1\ which contains requirements for
inservice testing (IST) of nuclear power plant components. Until 2012,
the ASME issued new editions of the ASME BPV Code every 3 years and
addenda to the editions annually, except in years when a new edition
was issued. Similarly, the ASME periodically published new editions and
addenda of the OM Code. Starting in 2012, the ASME decided to issue
editions of its BPV and OM Codes (no addenda) every 2 years. The new
editions and addenda typically revise provisions of the ASME BPV and OM
Codes (ASME Codes) to broaden their applicability, add specific
elements to current provisions, delete specific provisions, and/or
clarify them to narrow the applicability of the provision. The
revisions to the editions and addenda of the ASME Codes do not
significantly change philosophy or approach.
---------------------------------------------------------------------------
\1\ The editions and addenda of the ASME Code for Operation and
Maintenance of Nuclear Power Plants have had different titles from
2005 to 2012 and are referred to collectively in this rule as the
``OM Code.''
---------------------------------------------------------------------------
It has been the NRC's practice to establish requirements for the
design, construction, operation, ISI examination, and IST of nuclear
power plants by approving the use of editions and addenda of the ASME
Codes in Sec. 50.55a of title 10 of the Code of Federal Regulations
(10 CFR), ``Codes and standards.'' The NRC approves and/or mandates the
use of certain parts of editions and addenda of these ASME Codes in
Sec. 50.55a through the rulemaking process of ``incorporation by
reference.'' Upon incorporation by reference of the ASME Codes into
Sec. 50.55a, the provisions of the ASME Codes are legally-binding NRC
requirements as delineated in Sec. 50.55a and subject to the
conditions on certain specific ASME Code provisions that are set forth
in Sec. 50.55a. The editions and addenda of the ASME BPV and OM Codes
were last incorporated by reference into the regulations in a final
rule dated June 21, 2011 (76 FR 36232), subject to the NRC's
conditions.
The ASME Codes are consensus standards developed by participants
with broad and varied interests, including the NRC and licensees of
nuclear power plants. The ASME's adoption of new editions of, and
addenda to, the ASME Codes does not mean that there is unanimity on
every provision in the ASME Codes. There may be disagreement among the
technical experts, including NRC representatives, on the ASME Code
committees and subcommittees, regarding the acceptability or
desirability of a particular Code provision included in an ASME-
approved Code edition or addenda. If the NRC believes that there is a
significant technical or regulatory concern with a provision in an
ASME-approved Code edition or addenda being considered for
incorporation by reference, then the NRC will condition the use of that
provision when it incorporates by reference that ASME Code edition or
addenda. In some cases, the condition increases the level of safety
afforded by the ASME Code provision or addresses a regulatory issue not
considered by the ASME. In other instances, where research data or
experience has shown that certain Code provisions are unnecessarily
conservative, the condition may provide that the Code provision need
not be complied with in some or all respects. The NRC's conditions are
included in Sec. 50.55a, typically in paragraph (b) of that
regulation. In a Staff Requirements Memorandum (SRM) dated September
10, 1999, the Commission indicated that NRC rulemakings adopting
(incorporating by reference) a voluntary consensus standard must
identify and justify each part of the standard that is not adopted. For
this rulemaking, the provisions of the 2009 Addenda, 2010 Edition, 2011
Addenda, and 2013 Edition of Section III, Division 1; and the 2009
Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition of Section XI,
Division 1, of the ASME BPV Code; and the 2009 Edition, 2011 Addenda,
and 2012 Edition of the OM Code that the NRC is not adopting, or
partially adopting, are identified in the Discussion, Regulatory
Analysis, and Backfitting and Issue Finality sections of this document.
The provisions of those specific editions and addenda and Code Cases
that are the subject of this rulemaking that the NRC finds to be
conditionally acceptable, together with the applicable conditions, are
also identified in the Discussion, Regulatory Analysis, and Backfitting
and Issue Finality sections of this document.
The ASME Codes are voluntary consensus standards, and the NRC's
incorporation by reference of these Codes is consistent with applicable
requirements of the NTTAA. Additional discussion on NRC's compliance
with the NTTAA is set forth in Section XIV
[[Page 32936]]
of this document, ``Voluntary Consensus Standards.''
This final rule reflects the NRC's redesignation of paragraphs
within Sec. 50.55a set forth in a final rule dated November 5, 2014
(79 FR 65776), as corrected on December 11, 2014 (79 FR 73461). The re-
designation of paragraphs was needed to address the Office of the
Federal Register's requirements in 1 CFR part 51 for incorporation by
reference. For additional information on the November 2014 final rule,
please consult the statement of considerations (preamble) for that
final rule.
II. Discussion
The NRC regulations incorporate by reference ASME Codes for nuclear
power plants. The ASME periodically revises and updates its codes for
nuclear power plants. This final rule is the latest in a series of
rulemakings to amend the NRC's regulations to incorporate by reference
revised and updated ASME Codes for nuclear power plants. The proposed
rule which led to this final rule was published on September 18, 2015
(80 FR 56820). This rulemaking is intended to maintain the safety of
nuclear power plants and make NRC activities more effective and
efficient.
The NRC follows a three-step process to determine acceptability of
new provisions in new editions and addenda to the Codes and the need
for conditions on the uses of these Codes. This process was employed in
the review of the Codes that are the subject of this rule. First, the
NRC staff actively participates with other ASME committee members with
full involvement in discussions and technical debates in the
development of new and revised Codes. This includes a technical
justification of each new or revised Code. Second, the NRC committee
representatives discuss the Codes and technical justifications with
other cognizant NRC staff to ensure an adequate technical review.
Third, the NRC position on each Code is reviewed and approved by NRC
management as part of the rule amending Sec. 50.55a to incorporate by
reference new editions and addenda of the ASME Codes and conditions on
their use. This regulatory process, when considered together with the
ASME's own process for developing and approving the ASME Codes,
provides reasonable assurance that the NRC approves for use only those
new and revised Code edition and addenda, with conditions as necessary,
that provide reasonable assurance of adequate protection to public
health and safety, and that do not have significant adverse impacts on
the environment.
The NRC is amending its regulations to incorporate by reference:
The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013
Edition of the ASME BPV Code, Section III, Division 1 and Section XI,
Division 1, with conditions on their use.
The 2009 Edition, the 2011 Addenda, and the 2012 Edition
of Division 1 of the OM Code, with conditions on their use.
ASME Standard NQA-1, ``Quality Assurance Requirements for
Nuclear Facility Applications,'' including several editions and addenda
to NQA-1 from previous years with slightly varying titles as identified
in Sec. 50.55a(a)(1)(v). More specifically, the NRC is incorporating
by reference the 1983 Edition through the 1994 Edition, the 2008
Edition, and the 2009-1a Addenda to the 2008 Edition of ASME NQA-1,
with conditions on their use.
ASME BPV Code Case N-513-3, ``Evaluation Criteria for
Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping
Section XI, Division 1,'' Mandatory Appendix I, ``Relations for Fm, Fb,
and F for Through-Wall Flaws,'' Approval Date: January 26, 2009. This
Code Case has already been approved for use by the NRC in Regulatory
Guide (RG) 1.147 (75 FR 61321; October 5, 2010), but is now being
incorporated by reference in order to adopt a condition on Nonmandatory
Appendix U, which requires the use of this Code Case appendix.
ASME BPV Code Case N-729-4, ``Alternative Examination
Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having
Pressure-Retaining Partial-Penetration Welds Section XI, Division 1,''
ASME approval date: June 22, 2012, with conditions on its use.
ASME BPV Code Case N-770-2, ``Alternative Examination
Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel
Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler
Material With or Without Application of Listed Mitigation Activities,
Section XI, Division 1,'' ASME approval date: June 9, 2011, with
conditions on its use.
ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast
Austenitic Piping Welds From the Outside Surface Section XI, Division
1,'' ASME approval date: October 16, 2012.
ASME BPV Code Case N-852, ``Application of the ASME NPT
Stamp, Section III, Division 1; Section III, Division 2; Section III,
Division 3; Section III, Division 5,'' Approval Date: February 9, 2015.
OM Code Case OMN-20, ``Inservice Test Frequency.''
The current regulations in Sec. 50.55a(a)(1)(ii) incorporate by
reference ASME BPV Code, Section XI, 1970 Edition through the 1976
Winter Addenda; and the 1977 Edition (Division 1) through the 2008
Addenda (Division 1), subject to the existing conditions in Sec.
50.55a(b)(2)(i) through (xxix). This amendment revises Sec.
50.55a(a)(1)(ii) to incorporate by reference the 2009 Addenda (Division
1) through the 2013 Edition (Division 1) of the ASME BPV Code, Section
XI. It also clarifies the wording and adds, removes, or revises some of
the conditions as explained in this document.
The NRC is revising Sec. 50.55a(a)(1)(iv) to incorporate by
reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division
1 of the OM Code. Based on this revision, the NRC regulations will
incorporate by reference in Sec. 50.55a the 1995 Edition through the
2012 Edition of the OM Code.
The NRC reviewed changes to the Codes in the editions and addenda
of the Codes identified in this rulemaking, and published a proposed
rule in the Federal Register setting forth the NRC's proposal to
incorporate by reference the ASME Codes, together with proposed
conditions on their use (80 FR 56820; September 18, 2015). After
consideration of the public comments received on the proposed rule
(public comments are discussed in Section IV of this document, ``NRC
Responses to Public Comments''), the NRC concludes, in accordance with
the process for review of changes to the Codes, that each of the
editions and addenda of the Codes, and the 2008 Edition and the 2009-1a
Addenda of NQA-1, are technically adequate, consistent with current NRC
regulations, and approved for use with specified conditions set forth
in this final rule. Each of the NRC conditions and the reasons for each
condition are discussed in the following sections. The discussions are
organized under the applicable ASME Code and Section.
There is not a separate heading for ASME quality assurance standard
NQA-1 because there are three separate discussions of NQA-1--one under
the heading for ASME BPV Code, Section III, one under the heading for
ASME BPV Code, Section XI, and one under the heading for OM Code--
because there are three conditions related to NQA-1, one in each of
those areas (Sec. 50.55a(b)(1)(iv) for Section III, Sec.
50.55a(b)(2)(x) for Section XI, and Sec. 50.55a(b)(3)(i) for the OM
Code). In addition, administrative and editorial changes to various
paragraphs of Sec. 50.55a are being adopted for accuracy, clarity,
consistency, and general
[[Page 32937]]
administrative convenience. These editorial changes are not further
discussed in this heading, but are described in Section V of this
document, ``Section-by-Section Analysis.''
Four of the six ASME Code Cases being incorporated by reference in
this rulemaking (N-729-4, N-770-2, N-824, and OMN-20) are discussed in
Section II.D of this document, ``ASME Code Cases.'' A fifth ASME Code
Case, N-852, is discussed in Section II.A, ``ASME BPV Code, Section
III,'' because the NRC's approval of that Code Case relates to a
provision of Section III, which is addressed in Sec. 50.55a(b)(1)(ix).
The sixth ASME Code Case, N-513-3, is discussed in Section II.B, ``ASME
BPV Code, Section XI,'' because the NRC's approval of that Code Case
relates to a provision of Section XI, which is addressed in Sec.
50.55a(b)(2)(xxxiv).
A. ASME BPV Code, Section III
10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section
III
The NRC is clarifying that Section III Nonmandatory Appendices are
not incorporated by reference. This language was originally added in a
final rule published on June 21, 2011 (76 FR 36232); however, it was
omitted from the final rule published on November 5, 2014 (79 FR
65776). The NRC is correcting the omission by inserting the
parenthetical clause ``(excluding Nonmandatory Appendices)'' in Sec.
50.55a(a)(1)(i).
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
The NRC is identifying prohibited subparagraphs and notes for each
ASME BPV Code edition and addenda in tabular form as opposed to the
narrative form of the existing regulation. No substantive change to the
requirements is intended by this revision. The NRC believes that
presenting the information in tabular form will increase the clarity
and understandability of the regulation.
The existing condition in Sec. 50.55a(b)(1)(ii) prohibits, for
welds with leg sizes less than 1.09 tn, the use of certain
Code provisions in ASME BPV Code, Section III, Division 1. The Code
provisions provide stress indices for welded joints used in the design
of Class 2 and Class 3 piping. The use of these indices is prohibited
for welds with leg sizes less than 1.09 tn, where
tn is the nominal pipe thickness because this would result
in a weld that would be weaker than the pipe to which it is adjoined
under these dimensions. The location of the prohibited provisions vary
in the Code editions and addenda from the 1989 Addenda through the 2013
Edition, so in this final rule the NRC clearly identifies the
prohibited code provisions in the editions and addenda in a tabular
format.
As an editorial matter, this final rule identifies the prohibited
ASME BPV Code provisions as ``notes,'' which is the term used by the
ASME, rather than ``footnotes.'' The NRC is using the terminology used
by the ASME for clarity.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
The NRC is approving for use the version of NQA-1 referenced in the
2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code,
Section III, Subsection NCA, Article 7000, which this rule is also
incorporating by reference. This allows applicants and licensees to use
the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2010
and later editions and addenda of Section III.
In the 2010 Edition of ASME BPV Code, Section III, Subsection NCA,
Article NCA-4000, ``Quality Assurance,'' was updated to require N-Type
Certificate Holders to comply with the requirements of Part 1 of the
2008 Edition and the 2009-1a Addenda of ASME Standard NQA-1, ``Quality
Assurance Requirements for Nuclear Facility Applications,'' as modified
and supplemented in NCA-4120(b) and NCA-4134. In addition, NCA-4110(b)
was revised to remove the reference to a specific edition and addenda
of ASME NQA-1, and Table NCA-7100-2, ``Standards and Specifications
Referenced in Division 1,'' was updated to require the 2008 Edition and
2009-1a Addenda of NQA-1 when using the 2010 Edition of Section III. In
light of these changes, the NRC reviewed the 2008 Edition and the 2009-
1a Addenda of NQA-1 and compared it to previously approved versions of
NQA-1 and found that there were no significant differences. In
addition, the NRC reviewed the changes to Subsection NCA that reference
the 2008 Edition and 2009-1a Addenda of NQA-1, compared them to
previously approved versions of Subsection NCA, and found that there
were no significant differences. Therefore, the NRC has concluded that
these editions and addenda of NQA-1 are acceptable for use.
The NRC is revising Sec. 50.55a(b)(1)(iv) to clarify that an
applicant's or licensee's commitments addressing those areas where NQA-
1 either does not address a requirement in appendix B to 10 CFR part
50, ``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' or is less stringent than the comparable
appendix B requirement govern the applicant's or licensee's Section III
activities. The clarification is consistent with Sec. 50.55a(b)(2)(x)
and (b)(3)(i). The NQA-1 provides the ASME's method for establishing
and implementing a quality assurance (QA) program for the design and
construction of nuclear power plants and fuel reprocessing plants.
However, NQA-1, as modified and supplemented in NCA-4120(b) and NCA-
4134, does not address some of the requirements of appendix B to 10 CFR
part 50. In some cases, the provisions of NQA-1 are less stringent than
the comparable appendix B requirements. Therefore, in order to meet the
requirements of appendix B, an applicant's or licensee's QA program
description must contain commitments addressing those provisions of
appendix B which are not covered by NQA-1, as well as provisions that
supplement or replace the NQA-1 provisions where the appendix B
requirement is more stringent.
Finally, the NRC is removing the reference in Sec.
50.55a(b)(1)(iv) to versions of NQA-1 older than the 1994 Edition
because the NRC did not receive any adverse comments from any applicant
or licensee about removing versions of NQA-1 older than the 1994
Edition from the regulation. The NRC received only one comment
regarding NQA-1. The comment expressed support for incorporation by
reference of NQA-1 and did not respond to the NRC's request for comment
regarding the removal of references to older versions of NQA-1.
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification
and Demonstration of Function of Incompressible-Fluid Pressure-Relief
Valves
The NRC is revising Sec. 50.55a(b)(1)(vii) so that the existing
condition prohibiting the use of paragraph NB-7742(a)(2) of the 2006
Addenda through the 2007 Edition, up to and including the 2008 Addenda,
is extended to include the editions and addenda up to the 2013 Edition,
which are the subject of this rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME
Certification Marks
The NRC is adding Sec. 50.55a(b)(1)(viii) to allow licensees to
use either the ASME BPV Code Symbol Stamps of editions and addenda
earlier than the 2011 Addenda to the 2010 Edition of the ASME BPV Code
or the ASME Certification Marks with the appropriate
[[Page 32938]]
certification designators and class designators as specified in the
2013 Edition through the latest edition and addenda incorporated by
reference in Sec. 50.55a.
The ASME BPV Code requires, in certain instances, that components
be stamped. The stamp signifies that the component has been designed,
fabricated, examined and tested, as specified in the ASME BPV Code. The
stamp also signifies that the required ASME BPV Code data report forms
have been completed, and the authorized inspector has inspected the
item and authorized the application of the ASME BPV Code Symbol Stamp.
The ASME has instituted changes in the BPV Code to consolidate the
different ASME BPV Code Symbol Stamps into a common ASME Certification
Mark. This action was implemented in the 2011 Addenda to the 2010
Edition of the ASME BPV Code. As of the end of 2012, ASME no longer
utilizes the ASME BPV Code Symbol Stamp. Licensees, however, may not
have updated to the edition or addenda that identifies the use of the
ASME Certification Mark. Nevertheless, licensees are legally required
to implement the ASME BPV Code Edition and Addenda identified as their
current code of record. As ASME components are procured, these
components may be received with the ASME Certification Mark, while the
licensee's current code of record may require the component to have the
ASME BPV Code Symbol Stamp. Installation of a component under such
circumstances would not be in compliance with the regulations that the
licensees are required to meet.
Both the ASME Certification Mark and the ASME BPV Code Symbol Stamp
are official ASME methods of certifying compliance with the Code.
Although these ASME Certification Marks differ slightly in appearance,
they serve the same purpose of certifying code compliance by the ASME
Certificate Holder and continue to provide for the same level of
quality assurance for the application of the ASME Certification Mark as
was required for the application of the ASME BPV Code Symbol Stamp. The
new ASME Certification Mark represents a small, non-safety significant
modification of ASME's trademark. As such, it does not change the
technical requirements of the Code. The ASME has confirmed that the
Certification Mark with designator is equivalent to the corresponding
BPV Code Symbol Stamp. Based on statements made by ASME in a letter
dated August 17, 2012, the NRC has concluded that the ASME BPV Code
Symbol Stamps and ASME Certification Mark with code-specific
designators are equivalent with respect to their certification of
compliance with the BPV Code. The NRC discussed this issue in
Regulatory Issue Summary 2013-07, ``NRC Staff Position on the Use of
American Society of Mechanical Engineers Certification Mark,'' dated
May 28, 2013.
10 CFR 50.55a(b)(1)(ix) Section III Condition: NPT Code Symbol Stamps
The NRC is adding Sec. 50.55a(b)(1)(ix) to allow licensees to use
the NPT Code Symbol Stamp with the letters arranged horizontally as
specified in ASME BPV Code Case N-852 for the service life of a
component that had the NPT Code Symbol Stamp applied during the time
period from January 1, 2005, through December 31, 2015.
Public comments on the use of the NPT Code Symbol requested that
the NRC accept the NPT Code Symbol Stamp having the NPT letters
arranged horizontally as an acceptable NPT Stamp to certify Code
compliance for fabricated items that have already been stamped prior to
receiving a replacement NPT Code Symbol Stamp from the ASME. The
comments requested that the NRC include acceptance of Code Case N-852
in this final rule for this purpose. Within the context of its Code
rules, ASME asserts that the NPT Code Symbol Stamp having the NPT
letters arranged horizontally, although differing slightly in
appearance from the NPT Code Symbol Stamp as illustrated in Section
III, Table NCA-8100-1 of the ASME BPV Code, 2010 Edition and earlier
editions and addenda, serves the same purpose of certifying Code
compliance by the ASME NPT Certificate Holder with confirmation by the
Authorized Nuclear Inspector and provides the same level of quality
assurance. In addition, ASME indicated that on or after January 1,
2016, the ASME will no longer authorize use of the NPT Code Symbol
Stamp having the NPT letters arranged horizontally. Accordingly, on or
after January 1, 2016, fabricated items will only be stamped with the
NPT Code Symbol Stamp as illustrated in Section III, Table NCA-8100-1
of the ASME BPV Code, 2010 Edition and earlier editions and addenda.
The NRC agrees in general with this comment, in which the ASME
asserts that the ASME NPT Code Symbol Stamp with the letters arranged
horizontally to be equivalent to the ``N over PT'' ASME NPT Code Symbol
Stamp. Therefore, using either Code Symbol Stamp serves the same
purpose of certifying code compliance by the ASME Certificate Holder
with confirmation by the Authorized Nuclear Inspector and provides the
same level of quality assurance. The NRC also notes that the same
administrative and technical requirements in the ASME Code still apply
whether an ASME NPT Code Symbol Stamp with the letters arranged
horizontally or an ``N over PT'' ASME NPT Code Symbol Stamp is applied.
However, since this NPT Code Symbol Stamp having the NPT letters
arranged horizontally will only be applied onto fabricated components
from the time period of January 1, 2005, through December 31, 2015, the
time period for when this NPT Code Symbol Stamp was applied to the
component should be limited to these dates to prevent inadvertent
fraudulent material. Therefore, the NRC agrees that the ASME BPV Code
Case N-852 is acceptable for the service life of the component that had
the NPT Code Symbol stamp applied from the time period of January 1,
2005, through December 31, 2015. In response to this comment, the NRC
added Sec. 50.55a(b)(1)(ix) to include a statement that licensees may
use the NPT Code Symbol Stamp with the letters arranged horizontally as
specified in ASME BPV Code Case N-852 for the service life of a
component that had the NPT Code Symbol Stamp applied during the time
period from January 1, 2005, through December 31, 2015. The NRC is
incorporating by reference ASME BPV Code Case N-852 in Sec.
50.55a(a)(1)(iii)(F) because it is referenced in Sec.
50.55a(b)(1)(ix).
Although the proposed rule did not include this Code Case, the NRC
has determined that the incorporation by reference of this Code Case at
the final rule stage is a logical outgrowth of the proposed rule. The
NRC's intent to ensure that Sec. 50.55a identify all ASME-approved
methods for labelling Code components is apparent from the statement of
considerations for the proposed rule. See 80 FR 56820 (September 18,
2015) at 56823-56824. The NRC did not entirely achieve that purpose,
and this resulted in public comments seeking approval of this Code
Case, which supports the proposition that the public had a reasonable
opportunity to either propose the correction, with conditions as the
commenter believes are necessary or desirable, or to indicate why the
(anticipated) correction should not be made. Therefore, the NRC
concludes that it may incorporate by reference ASME BPV Code Case N-
852.
[[Page 32939]]
B. ASME BPV Code, Section XI
10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section
XI
In the proposed rule, the NRC proposed a revision to Sec.
50.55a(a)(1)(ii) that would have clarified that Section XI Nonmandatory
Appendix U of the 2013 Edition of ASME BPV Code, Section XI was not
incorporated by reference and therefore not approved for use. After
considering public comments, the NRC has determined that it will not
exclude Appendix U from the incorporation by reference because it is
the integration of ASME BPV Code Cases N-513-3, ``Evaluation Criteria
for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3
Piping Section XI, Division 1,'' and N-705, ``Evaluation Criteria for
Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3
Vessels and Tanks Section XI, Division 1,'' into Section XI. The NRC
has approved Code Cases N-513-3 and N-705 in RG 1.147. However, as
described in the discussion for Sec. 50.55a(b)(2)(xxxiv) in Section
II.B, ``ASME BPV Code Section XI,'' the NRC has found it necessary to
adopt two new conditions to the use of Nonmandatory Appendix U.
The NRC is adopting two conditions in the language of Sec.
50.55a(a)(1)(ii)(C)(52) and (53) to address two inconsistencies that
were identified between the NRC's position in a proposed rule regarding
the acceptability of ASME Code Cases (81 FR 10780; March 2, 2016) (2016
Code Case proposed rule) and the proposed rule for this rulemaking (80
FR 56820; September 18, 2015). The first inconsistency is that the
NRC's proposed conditions on ASME BPV Code Case N-799, ``Dissimilar
Metal Welds Joining Vessel Nozzles to Components,'' in the 2016 Code
Case proposed rule were not reflected in the 2015 proposed rule for
this rulemaking, even though the technical content of ASME BPV Code
Case N-799 has been incorporated into the 2011 Addenda and 2013 Edition
of ASME BPV Code, Section XI. The second inconsistency is that the
NRC's proposed disapproval of ASME BPV Code Case N-813, ``Alternative
Requirements for Preservice Volumetric and Surface Examination,'' in
the 2016 Code Case proposed rule was not reflected in the 2015 proposed
rule for this rulemaking, even though the technical content of ASME BPV
Code Case N-813 has been incorporated into the 2013 Edition of the ASME
BPV Code, Section XI as IWB-3112(a)(3) and IWC-3112(a)(3). To address
these two inconsistencies, the NRC is excluding these ASME BPV Code,
Section XI items from incorporation by reference, as reflected in Sec.
50.55a(a)(1)(ii)(C)(52) and (53) of the final rule. The NRC plans to
complete the development of the regulatory approaches for examination
of component-to-component welds for new construction plants and the
acceptance of preservice flaws by analytical evaluation for operating
plants and include them in a future rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and
Addenda of Subsection IWE and Subsection IWL
The NRC is revising Sec. 50.55a(b)(2)(vi) to expressly state that
licensees that implemented the expedited examination of containment
during the 5-year period from September 9, 1996, to September 9, 2001,
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),
when implementing the initial 120-month inspection interval for the
containment ISI requirements of this section.
The expedited examination involved the completion of the first set
of examinations of the first or initial 120-month containment
inspection interval. It is noted that all of the operating reactors in
the previously stated class would have gone past their initial 120-
month inspection interval by 2011. The change removes the possibility
of misinterpretation of the provision as requiring plants that do not
fall in the previously stated class, such as reactors licensed after
September 9, 2001, to use the 1992 Edition with 1992 Addenda or the
1995 Edition with 1996 Addenda of Subsection IWE and Subsection IWL,
Section XI for implementing the initial 120-month inspection interval
of the containment ISI program. Applicants and licensees that do not
fall in the previously stated class must use Code editions and addenda
in accordance with Sec. 50.55a(g)(4)(i) and (ii), respectively, for
the initial and successive 120-month containment ISI intervals.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment
Examinations
The NRC is revising Sec. 50.55a(b)(2)(viii) by removing the
condition for using the 2007 Edition with 2009 Addenda through the 2013
Edition of Subsection IWL requiring compliance with Sec.
50.55a(b)(2)(viii)(E). To support the removal of the condition, the NRC
is adding new requirements governing the performance and documentation
of concrete containment examinations in Sec. 50.55a(b)(2)(viii)(H) and
(I), which are discussed separately in the next two headings.
Section 50.55a(b)(2)(viii)(E) is one of several conditions that
apply to the inservice examination of concrete containments using
Subsection IWL of various editions and addenda of the ASME BPV Code,
Section XI, incorporated by reference in Sec. 50.55a(a)(1)(ii). The
NRC is removing the condition in Sec. 50.55a(b)(2)(viii)(E) when
applying the 2007 Edition with 2009 Addenda through the 2013 Edition of
Subsection IWL because its intent has been incorporated into the Code
in the new provision IWL-2512, ``Inaccessible Areas.''
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth
Provision
The NRC is adding Sec. 50.55a(b)(2)(viii)(H) to specify the
information that must be provided in the ISI Summary Report required by
IWA-6000, when inaccessible concrete surfaces are evaluated under the
new Code provision IWL-2512. This new condition replaces the existing
condition in Sec. 50.55a(b)(2)(viii)(E), when using the 2007 Edition
with the 2009 Addenda through the 2013 Edition of Subsection IWL.
The existing condition in Sec. 50.55a(b)(2)(viii)(E) of the
current rule requires that, for Class CC applications, the licensee
shall evaluate the acceptability of inaccessible areas when conditions
exist in accessible areas that could indicate the presence of or result
in degradation to such inaccessible areas, and provide the evaluation
information required by Sec. 50.55a(b)(2)(viii)(E)(1), (2), and (3) in
the IWA-6000 ISI Summary Report.
In the 2009 Addenda Subsection IWL, the ASME revised existing
provisions IWL-1220 and IWL-2510 and added the new provision IWL-2512
intended to incorporate the condition in Sec. 50.55a(b)(2)(viii)(E)
into Subsection IWL. The IWL-2510, ``Surface Examination,'' was
restructured into new paragraphs in IWL-2511, ``Accessible Areas,''
with almost the same provisions as the previous IWL-2510 and IWL-2512,
``Inaccessible Areas,'' to be specific to examinations required for
accessible areas, and differentiate between those and the new
requirements for inaccessible areas. The inaccessible areas addressed
by the new IWL-2512 are: (1) Concrete surfaces obstructed by adjacent
structures, parts or appurtenances (e.g., generally above-grade
inaccessible areas); and (2)
[[Page 32940]]
concrete surfaces made inaccessible by foundation material or backfill
(e.g., below-grade inaccessible areas).
The revised IWL-2511(a) has a new requirement that states that,
``If the Responsible Engineer determines that observed suspect
conditions indicate the presence of, or could result in, degradation of
inaccessible areas, the requirements of IWL-2512(a) shall be met.'' The
new IWL-2512(a) requires the ``Responsible Engineer'' to evaluate
suspect conditions and specify the type and extent of examinations, if
any, required to be performed on inaccessible surface areas described
in the previous paragraph. The acceptability of the evaluated
inaccessible area would be determined either based on the evaluation or
based on the additional examinations, if determined to be required. The
new IWL-2512(b) further requires a periodic technical evaluation of
below-grade inaccessible areas of concrete to be performed to determine
and manage its susceptibility to degradation regardless of whether
suspect conditions exist in accessible areas that would warrant an
evaluation of inaccessible areas based on the condition in Sec.
50.55a(b)(2)(viii)(E). Therefore, the revised IWL-2511(a) and new IWL-
2512 code provisions address the evaluation and acceptability of
inaccessible areas consistent with the existing condition in Sec.
50.55a(b)(2)(viii)(E), with one exception. The exception is that the
new IWL-2512 provision does not explicitly require the information
specified in Sec. 50.55a(b)(2)(viii)(E)(1), (2), and (3) of the
existing condition to be provided in the IWA-6000 ISI Summary Report.
For these reasons, the NRC is identifying the information that must
be provided in the ISI Summary Report required by IWA-6000 when
inaccessible concrete surfaces are evaluated under the new code
provision IWL-2512. This new condition replaces the existing condition
in Sec. 50.55a(b)(2)(viii)(E) when using the 2007 Edition with the
2009 Addenda through the 2013 Edition of Subsection IWL. The
information required by the new condition must be provided when
inaccessible concrete areas are evaluated per IWL-2512(a) for
degradation based on suspect conditions found in accessible areas, as
well as when periodic technical evaluations of inaccessible below-grade
concrete areas required by IWL-2512(b) are performed.
10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth
Provision
The NRC is adding Sec. 50.55a(b)(2)(viii)(I) to place a condition
on the periodic technical evaluation requirements in the new IWL-
2512(b), for consistency with NUREG-1801, Revision 2, ``Generic Aging
Lessons Learned (GALL) Report,'' with regard to aging management of
below-grade containment concrete surfaces. The new IWL-2512(b)
provision is applicable to inaccessible below-grade concrete surfaces
exposed to foundation soil, backfill, or groundwater. This condition
would apply only during the period of extended operation of a renewed
license under 10 CFR part 54, when using IWL-2512(b) of the 2007
Edition with 2009 Addenda through the 2013 Edition of Subsection IWL.
In the 2009 Addenda of Subsection IWL, the ASME added new Code
provisions, IWL-2512(b) and (c) as well as a new line item L1.13 in
Table IWL-2500-1, intended to specifically address aging management
concerns with potentially unidentified degradation of inaccessible
below-grade containment concrete areas and to be responsive to actions
outlined in the GALL Report related to aging management of inaccessible
below-grade concrete surfaces. It is noted that these new Code
provisions are an enhancement to the requirement of the existing
condition in Sec. 50.55a(b)(2)(viii)(E) to specifically address aging
management of inaccessible below-grade containment concrete areas and
is generally acceptable to the NRC.
The new IWL-2512(b) provides requirements for systematically
performing a periodic technical evaluation of concrete surfaces exposed
to foundation soil, backfill, or groundwater to determine
susceptibility of the concrete to deterioration that could affect its
ability to perform its intended design function under conditions
anticipated through the service life of the structure. It requires the
technical evaluation to be performed and documented at periodic
intervals not to exceed 10 years regardless of whether conditions exist
in accessible areas that would warrant an evaluation of inaccessible
areas by the existing condition in Sec. 50.55a(b)(2)(viii)(E), which
the NRC finds reasonable for the initial 40-year operating license
period. The new IWL-2512(b) further provides the specific elements,
including aging mechanisms considered, that the technical evaluation
should include, as well as the definition of an aggressive below-grade
environment. The new IWL-2512(c) requires that the evaluation results
of IWL-2512(b) be used to define and document the condition monitoring
program, if determined to be required, including required examinations
and frequencies, to be implemented for the management of degradation
and aging effects of the below-grade concrete surface areas. If it is
determined that additional examinations are required, these
examinations of inaccessible below-grade areas will be implemented in
accordance with new line item L1.13 in Table IWL-2500-1 under
Examination Category L-A, Concrete, with acceptance criteria based on
IWL-3210. It should be noted that a technical evaluation approach, such
as in IWL-2512(b), could be used, and is generally used, to determine
acceptability of a below-grade inaccessible area to satisfy the
condition in Sec. 50.55a(b)(2)(viii)(E).
The technical evaluation requirements in IWL-2512(b) assist in
determining the susceptibility to degradation and manage aging effects
of inaccessible below-grade concrete surfaces, before the loss of
intended function. The requirements are based on, and are generally
consistent with, the guidance in the GALL Report, with the following
two exceptions. The first exception is that IWL-2512(b) requires the
technical evaluation to determine the susceptibility of the concrete to
degradation and the ability to perform the intended design function
through its service life at periodic intervals not to exceed 10 years.
The aging management programs (AMPs) for safety-related structures
(e.g., Structures Monitoring) in the GALL Report require such
evaluation to be performed at intervals not to exceed 5 years, which is
also consistent with applicant commitments during review of license
renewal applications. The second exception is that IWL-2512(b) requires
that examination of representative samples of below-grade concrete be
performed if excavated for any reason when an aggressive below-grade
environment is present. However, the NRC notes that the AMPs (X1.S6
Structures Monitoring and X1.S7 Water Control Structures) in the GALL
Report require the same examination even for a non-aggressive below-
grade environment.
Based on these reasons, the NRC is adding Sec.
50.55a(b)(2)(viii)(I) to place a condition on the periodic technical
evaluation requirements in IWL-2512(b) for consistency with the GALL
Report, when addressing the two exceptions previously described with
respect to aging management of inaccessible below-grade concrete
components of the
[[Page 32941]]
containment. The new condition requires that, during the period of
extended operation of a renewed license, the technical evaluation under
IWL-2512(b) of inaccessible below-grade concrete surfaces exposed to
foundation soil, backfill, or groundwater be performed at periodic
intervals not to exceed 5 years, as opposed to the 10-year interval in
IWL-2512. In addition, the condition requires the examination of
representative samples of the exposed portions of the below-grade
concrete be performed when excavated for any reason as opposed to IWL-
2512, which limits the examination to excavations in aggressive, below-
grade environments. Since the GALL Report is the technical basis
document for license renewal, this new condition applies only during
the period of extended operation of a renewed license under 10 CFR part
54, when using IWL-2512(b) of the 2007 Edition with 2009 Addenda
through the 2013 Edition of Subsection IWL, Section XI.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment
Examinations
The NRC is extending the applicability of the existing conditions
in Sec. 50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J), governing
examinations of metal containments and the liners of concrete
containments under Subsection IWE, to the ASME BPV Code editions and
addenda which are the subject of this rulemaking (i.e., the 2007
Edition with 2009 Addenda through the 2013 Edition). The last sentence
of Sec. 50.55a(b)(2)(ix) prior to this final rule stated that the
referenced conditions were applicable only to addenda, but not to
editions, approved by the NRC after the 2007 Edition of the ASME BPV
Code. To rectify this, the NRC is revising the last sentence of Sec.
50.55a(b)(2)(ix) to refer to the latest ``edition and'' addenda after
the 2007 Edition which are incorporated by reference into Sec. 50.55a.
The NRC reviewed the Code changes in Subsection IWE of the 2009
Addenda through the 2013 Edition of ASME BPV Code, Section XI, and
noted that all of the changes were editorial or administrative with the
intent to improve the clarity of the existing requirements or correct
errors by errata. There were no changes to Subsection IWE in the Code
editions and addenda that are the subject of this rulemaking that the
NRC believes would require new regulatory conditions to ensure safety,
nor do the changes to Subsection IWE address the NRC's reasons for
adopting the conditions on the use of Subsection IWE. Accordingly, the
NRC is extending the applicability of the existing conditions (by
adding the words ``edition and'' to Sec. 50.55a(b)(2)(ix) as
discussed) without any change to the provisions of the conditions.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
The NRC is approving for use the version of NQA-1 referenced in the
2009 Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition of the
ASME BPV Code, Section XI, Table IWA 1600-1, ``Referenced Standards and
Specifications,'' which this rule is also incorporating by reference.
This allows, but does not require, licensees to use the 1994 Edition or
the 2008 Edition and the 2009-1a Addenda of NQA-1 when using the 2009
Addenda and later editions and addenda of Section XI.
In the 2013 Edition of ASME BPV Code, Section XI, Table IWA 1600-1
was updated to allow licensees to use the 1994 Edition or the 2008
Edition with the 2009-1a Addenda of NQA-1 when using the 2013 Edition
of Section XI. In the 2010 Edition of ASME BPV Code, Section XI, IWA-
1400, ``Owner's Responsibilities,'' Subparagraph (n)(2) was updated to
reference the NQA-1 Part I, Basic Requirements and Supplementary
Requirements for Nuclear Facilities. In the 2009 Addenda of the 2007
Edition of ASME BPV Code, Section XI, Table IWA-1600-1, ``Referenced
Standards and Specifications,'' was updated to allow licensees to use
the 1994 Edition of NQA-1. The NRC reviewed the 2008 Edition and the
2009-1a Addenda of NQA-1 and compared it to previously approved
versions of NQA-1 and found that there were no significant differences.
Therefore, the NRC has concluded that these editions and addenda of
NQA-1 are acceptable for use.
The NRC is amending Sec. 50.55a(b)(2)(x) to clarify that a
licensee's commitments addressing those areas where NQA-1 either does
not address a requirements in appendix B to 10 CFR part 50 or is less
stringent than the comparable appendix B requirement govern the
licensee's Section XI activities. The clarification is consistent with
Sec. 50.55a(b)(1)(iv) and (b)(3)(i). The ASME's method for
establishing and implementing a QA program for the design and
construction of nuclear power plants and fuel reprocessing plants is
described in NQA-1. However, NQA-1 does not address some of the
requirements of appendix B to 10 CFR part 50. In some cases, the
provisions of NQA-1 are less stringent than the comparable appendix B
requirements. Therefore, in order to meet the requirements of appendix
B, a licensee's QA program description must contain commitments
addressing those provisions of appendix B which are not covered by NQA-
1, as well as provisions that supplement or replace the NQA-1
provisions where the appendix B requirement is more stringent.
Finally, the NRC is removing the reference in Sec. 50.55a(b)(2)(x)
to versions of NQA-1 older than the 1994 Edition because the NRC did
not receive any adverse comments from any applicant or licensee
regarding concerns about removing versions of NQA-1 older than the 1994
Edition from the regulation. The NRC received only one comment
regarding NQA-1. The comment expressed support for incorporation by
reference of NQA-1 and did not respond to the NRC's request for comment
regarding the removal of references to older versions of NQA-1.
10 CFR 50.55a(b)(2)(xii) Section XI Condition: Underwater Welding
The NRC is revising Sec. 50.55a(b)(2)(xii) to allow underwater
welding on irradiated materials in accordance with IWA-4660,
``Underwater Welding,'' of Section XI, 1997 Addenda through the latest
edition and addenda incorporated by reference in Sec.
50.55a(a)(1)(ii). The conditions for which underwater welding would be
permitted without prior NRC approval are based on technical factors,
such as neutron fluence and, for certain material classes, helium
concentration.
The existing condition in Sec. 50.55a(b)(2)(xii) does not allow
underwater welding on irradiated materials by prohibiting the use of
IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through
the latest edition and addenda incorporated by reference in Sec.
50.55a(a)(1)(ii) on materials that are irradiated; however, there are
two problems with the restriction in Sec. 50.55a(b)(2)(xii). First,
the neutron fluence threshold above which a material is considered to
be irradiated is not defined in Sec. 50.55a(b)(2)(xii). Second,
studies such as those documented in Boiling Water Reactor Vessel and
Internals Project (BWRVIP) Report 1003020 (BWRVIP-97) have shown that
reactor internals can tolerate some neutron irradiation without
suffering damage to weldability, as long as the helium concentration in
the material does not exceed a certain threshold. The NRC completed its
Safety Evaluation of BWRVIP-97 in May 2008 and concluded that
implementation of the guidelines in the BWRVIP-97 report, with some
modifications as documented in the
[[Page 32942]]
NRC Safety Evaluation dated June 30, 2008, will provide an acceptable
technical basis for the design of weld repairs based on the helium
content of irradiated reactor vessel internals. The current version of
Sec. 50.55a(b)(2)(xii) does not define a threshold of helium
concentration below which the material is considered to be weldable.
The most recent editions of the ASME BPV Code state in Article IWA-
4660 that underwater welding may not be performed on irradiated
materials other than P-No. 8 materials containing less than 0.1 atomic
parts per million (appm) measured or calculated helium content
generated through irradiation. Some editions and addenda of the ASME
BPV Code prior to 2010 state in Article IWA-4660 that underwater
welding may only be performed in applications not predicted to exceed a
thermal neutron fluence of 1 x 10\17\ n/cm\2\. Other editions and
addenda of the ASME BPV Code prior to 2010 do not restrict the
underwater welding of irradiated materials. Therefore, there is
inconsistent treatment among the various editions and addenda of the
ASME BPV Code on the underwater welding of irradiated materials.
Current ASME BPV Code and Code Case requirements for welding on
irradiated materials, other than the underwater welding requirements
specified in IWA-4660, are inconsistent. Thresholds for weldability may
be stated in terms of fast neutron fluence, thermal neutron fluence, or
helium concentration. In some cases, thresholds are not defined and the
Code or Code Case simply states that consideration must be given to
irradiation effects when welding. The NRC believes that thresholds for
welding on irradiated materials should be based on the current
understanding of irradiation damage, as supported by technical studies
(such as BWRVIP-97) which have been evaluated by the NRC. In addition,
the NRC believes that these thresholds should be consistently applied
for all Code and Code Case applications.
During the public comment period for this rulemaking, a
representative of ASME recommended that Sec. 50.55a(b)(2)(xii) be
revised such that it applies only to those editions and addenda earlier
than the 2010 Edition. The effect of such a revision would be to allow
welding on P-No. 8 materials containing less than 0.1 appm measured or
calculated helium content generated through irradiation. However, this
proposed revision would not be consistent with other ASME BPV Code or
Code Case requirements for welding on irradiated materials, and this
proposed revision does not address standards for welding on material
classes other than P-No. 8. Instead the NRC is adopting conditions that
would apply to all materials and which can be consistently applied for
all Code and Code Case applications. The first condition, Sec.
50.55a(b)(2)(xii)(A), is based on fast neutron fluence and applies to
ferritic materials. The second condition, Sec. 50.55a(b)(2)(xii)(B),
is based on helium content and/or thermal fluence and applies to
austenitic materials. For P-No. 8 austenitic materials, the evaluation
of BWRVIP-97 supports a weldability threshold based on helium content
and thermal fluence. For austenitic materials other than P-No. 8, there
are insufficient data to support a weldability threshold based on
helium content, and, therefore, the NRC is adopting a weldability
threshold based on thermal fluence only.
The conditions for which underwater welding are permitted, as
stated in the revision of Sec. 50.55a(b)(2)(xii), were determined, in
part, based on technical discussions in a Category 2 public meeting
with industry representatives held on January 19, 2016. The NRC later
presented the new conditions at a public meeting held on March 2, 2016.
There were no comments on this change from the attendees at the March
2, 2016, public meeting. Summaries of the January 19 and March 2, 2016,
public meetings are available in ADAMS under Accession Nos. ML16050A383
and ML16069A408, respectively.
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth
Provision
The NRC is adding Sec. 50.55a(b)(2)(xviii)(D) to prohibit
applicants and licensees from using the ultrasonic examination
nondestructive examination (NDE) personnel certification requirements
in Section XI, Appendix VII and Subarticle VIII-2200 of the 2011
Addenda and 2013 Edition of the ASME BPV Code. Paragraph (b)(2)(xviii)
currently includes conditions on the certification of NDE personnel. In
addition, the new paragraph will require applicants and licensees to
use the 2010 Edition, Table VII-4110-1 training hour requirements for
Levels I, II, and III ultrasonic examination personnel, and the 2010
Edition, Subarticle VIII-2200 of Appendix VIII prerequisites for
personnel requirements. In the 2011 Addenda and 2013 Edition, the ASME
BPV Code added an accelerated Appendix VII training process for
certification of ultrasonic examination personnel based on training and
prior experience, and separated the Appendix VII training requirements
from the Appendix VIII qualification requirements. These new ASME BPV
Code provisions will provide personnel in training with less experience
and exposure to representative flaws in representative materials and
configurations common to operating nuclear power plants, and they would
permit personnel with prior non-nuclear ultrasonic examination
experience to qualify for examinations in nuclear power plants without
exposure to the variety of defects, examination conditions, components,
and regulations common to operating nuclear power plants.
The impact of reduced training and nuclear power plant
familiarization is unknown. The ASME BPV Code supplants training hours
and field experience without a technical basis, minimum defined
training criteria, process details, or standardization. For these
reasons, the NRC is prohibiting the use of Appendix VII and Subarticle
VIII-2200 of the 2011 Addenda and 2013 Edition. The NRC is requiring
applicants and licensees using the 2011 Addenda and 2013 Edition to use
the prerequisites for ultrasonic examination personnel certifications
in Table VII-4110-1 and Subarticle VIII-2200, Appendix VIII in the 2010
Edition.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements:
First Provision
The NRC is revising Sec. 50.55a(b)(2)(xxi)(A) to modify the
standard for visual magnification resolution sensitivity and contrast
for visual examinations performed on Examination Category B-D
components instead of ultrasonic examinations, making the rule conform
with ASME BPV Code, Section XI requirements for VT-1 examinations. The
character recognition rules are used in ASME BPV Code, Section XI,
Table IWA-2211-1 for VT-1 tests, and are the standard tests used for
resolution and contrast checks of the VT-1 equipment. This revision
essentially removed a requirement that was an addition to ASME BPV Code
that required 1-mil wires to be used in licensees' Sensitivity,
Resolution, and Contrast Standard targets. In 2004, the NRC published
NUREG/CR-6860, ``An Assessment of Visual Testing,'' showing that a
linear target, such as a wire, is not an effective method for testing
the resolution of a video camera system. In addition, Boiling Water
Reactor Vessel and Internals Project Report 105696 (BWRVIP-03) was
changed to eliminate a \1/2\ mil wire from the Sensitivity, Resolution,
and Contrast Standards due to similar concerns.
[[Page 32943]]
Simple line detection can be a poor performance standard, allowing
detection of a highly blurred image. This does not emulate sharpness
quality recognition for evaluation of weld discontinuities. The 750
[mu]m (30 mil) and the even smaller 25 [mu]m (1 mil) widths should not
be used as performance standards because they do not determine image
sharpness. This technique only measures the ``visible minimum'' for
long linear indications, and does not measure a system's resolution or
recognition limits. If the wire, or printed line, has a strong enough
contrast against the background, then a linear feature well below the
resolution of a system can be detected.
10 CFR 50.55a(b)(2)(xxiii) Section XI Condition: Evaluation of
Thermally Cut Surfaces
The NRC is revising Sec. 50.55a(b)(2)(xxiii) to clarify that this
condition, prohibiting the ASME BPV Code provisions allowing
elimination of mechanical processing of thermally cut surfaces under
certain circumstances, only applies to the 2001 Edition through the
2009 Addenda.
10 CFR 50.55a(b)(2)(xxx) Section XI Condition: Steam Generator
Preservice Examinations
In the proposed rule, the NRC proposed adding Sec.
50.55a(b)(2)(xxx), with a condition regarding steam generator
preservice examinations. The NRC received requests for clarification of
the proposed condition, including elaboration on the kind of preservice
examination that should be performed. The NRC agrees with the need for
this clarification; however, during the development of the final rule,
the NRC determined that additional time was needed to evaluate this
proposed condition. Therefore, to ensure that this rulemaking is
concluded as timely as possible, the NRC is not including this
condition in this final rule and will address the need for a condition
in a future rulemaking. The NRC has concluded that omitting this
condition does not present a health or safety concern because licensees
are currently performing appropriate steam generator preservice
inspections under existing programs.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping
Devices
The NRC is adding Sec. 50.55a(b)(2)(xxxi) to require the use of
Nonmandatory Appendix W when using a mechanical clamping device on an
ASME BPV Code Class piping system. This condition, in part, clearly
prohibits the use of mechanical clamping devices on small item Class 1
piping and portions of piping systems that form the containment
boundary. This condition also maintains the previously required design
and testing requirements for the implementation of mechanical clamping
devices on ASME BPV Code Class piping systems.
In the 2010 Edition of the ASME BPV Code, a change was made to
include mechanical clamping devices under the small items exclusion
rules of IWA-4131. Currently in the 2007 Edition/2008 Addenda of
Section XI under IWA-4133, ``Mechanical Clamping Devices Used as Piping
Pressure Boundary,'' mechanical clamping devices may be used only if
they meet the requirements of Mandatory Appendix IX of Section XI of
the ASME BPV Code. Article IX-1000 (c) of Appendix IX prohibits the use
of mechanical clamping devices on (1) Class 1 piping and (2) portions
of a piping system that form the containment boundary.
In the 2010 Edition, IWA-4133 was modified to allow use of IWA-
4131.1(c) for the installation of mechanical clamping devices. This
change allowed the use of small items exclusion rules in the
installation of mechanical clamping devices. Subparagraph IWA-4131.1(c)
was added such that mechanical clamping devices installed on items
classified as ``small items'' under IWA-4131, including Class 1 piping
and portions of a piping system that form the containment boundary,
would be allowed without a repair/replacement plan, pressure testing,
services of an Authorized Inspection Agency, and completion of the NIS-
2 form. The NRC, in accordance with the previously approved IWA-4133 of
the 2007 Edition/2008 Addenda of the ASME BPV Code, does not believe
that the ASME has provided a sufficient technical basis to support the
use of mechanical clamping devices on Class 1 piping or portions of a
piping system that form the containment boundary as a permanent repair.
Furthermore, the NRC finds that the ASME has not provided any basis for
the small item exemption allowing the installation of mechanical clamps
on these components. In the 2011 Addenda of the ASME BPV Code, IWA-
4131.1(c) was relocated to IWA-4131.1(d). To add clarity to the
condition, the NRC has included statements such that the implementation
of these paragraphs is now prohibited.
In the 2013 Edition, Mandatory Appendix IX of Section XI of the
ASME BPV Code was changed to Nonmandatory Appendix W of Section XI of
the ASME BPV Code. The NRC found insufficient basis to make this
change, removing the mandatory requirements for the use of mechanical
clamping devices on ASME BPV Code Class piping systems. By taking this
action, the ASME BPV Code now allows mechanical clamping devices to be
installed in various methods through interpretations of the ASME BPV
Code that do not maintain the requirements for design and testing of
the formerly mandatory Appendix IX. Therefore, to clarify the
requirement for the implementation of mechanical clamps in ASME BPV
Code Class systems, the NRC requires the use of Appendix W of Section
XI when using mechanical clamping devices, and prohibits the use of
mechanical clamping devices on small item Class 1 piping and portions
of a piping system that form the containment boundary, as would
otherwise be permitted under IWA-4131.1(c) in the 2010 Edition and IWA-
4131.1(d) in the 2011 Addenda through 2013 Edition.
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report
Submittal
The NRC is adding Sec. 50.55a(b)(2)(xxxii) to require licensees
using the 2010 Edition and later editions and addenda of Section XI to
continue to submit Summary Reports as required in IWA-6240 of the 2009
Addenda.
Prior to the 2010 Edition, Section XI required the preservice
summary report to be submitted prior to the date of placement of the
unit into commercial service, and the inservice summary report to be
submitted within 90 calendar days of the completion of each refueling
outage. In the 2010 Edition, IWA-6240 was revised to state, ``Summary
reports shall be submitted to the enforcement and regulatory
authorities having jurisdiction at the plant site, if required by these
authorities.'' This change in the 2010 Edition could lead to confusion
as to whether or not the summary reports need to be submitted to the
NRC, as well as the time for submitting the reports, if they were
required. The NRC concludes that summary reports must continue to be
submitted to the NRC in a timely manner because they provide valuable
information regarding examinations performed, conditions noted,
corrective actions taken, and the implementation status of preservice
inspection and ISI programs. Therefore, the NRC is adding Sec.
50.55a(b)(2)(xxxii) to ensure that preservice and inservice summary
reports will continue to be submitted within the timeframes currently
[[Page 32944]]
established in Section XI editions and addenda prior to the 2010
Edition.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed
Allowable Pressure
The NRC is adding Sec. 50.55a(b)(2)(xxxiii) to prohibit the use of
Appendix G, Paragraph G-2216, in the 2011 Addenda and later editions
and addenda of the ASME BPV Code, Section XI. The 2011 Addenda of the
ASME BPV Code included, for the first time, a risk-informed methodology
to compute allowable pressure as a function of inlet temperature for
reactor heat-up and cool-down at rates not to exceed 100 degrees F/hr
(56 degrees C/hr). This methodology was developed based upon
probabilistic fracture mechanics (PFM) evaluations that investigated
the likelihood of reactor pressure vessel (RPV) failure based on
specific heat-up and cool-down scenarios.
During the ASME's consideration of this change, the NRC staff noted
that additional requirements would need to be placed on the use of this
alternative. For example, the NRC staff indicated that it would be
important for a licensee who wishes to utilize such a risk-informed
methodology for determining plant-specific pressure-temperature limits
to ensure that the material condition of its facility is consistent
with assumptions made in the PFM evaluations that supported the
development of the methodology. One aspect of this would be evaluating
plant-specific ISI data to determine whether the facility's RPV flaw
distribution was consistent with the flaw distribution assumed in the
supporting PFM evaluations. This consideration is consistent with a
similar requirement established by the NRC in Sec. 50.61a,
``Alternative Fracture Toughness Requirements for Protection against
Pressurized Thermal Shock Events.'' The PFM methodology that supports
Sec. 50.61a is very similar to that which was used to support ASME BPV
Code, Section XI, Appendix G, Paragraph G-2216. These concerns with the
Paragraph G-2216 methodology for computing allowable pressure as a
function of inlet temperature for reactor heat-up and cooldown were not
addressed by the ASME. Accordingly, the NRC is prohibiting the use of
Paragraph G-2216 in Appendix G of the 2010 Edition. The continued use
of the deterministic methodology of Section XI, Appendix G to generate
Pressure-Temperature (P-T) limits remains acceptable.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Nonmandatory Appendix
U
The NRC is adding Sec. 50.55a(b)(2)(xxxiv) to require that two
conditions, (A) and (B), be satisfied when using Nonmandatory Appendix
U of the 2013 Edition of the ASME BPV Code, Section XI. In the proposed
rule, the NRC had proposed to exclude Nonmandatory Appendix U from the
incorporation by reference and therefore not approve it for use. After
considering public comments, the NRC has incorporated by reference
Appendix U in this final rule because it integrates ASME BPV Code Cases
N-513-3, ``Evaluation Criteria for Temporary Acceptance of Flaws in
Moderate Energy Class 2 or 3 Piping Section XI, Division 1,'' and N-
705, ``Evaluation Criteria for Temporary Acceptance of Degradation in
Moderate Energy Class 2 or 3 Vessels and Tanks Section XI, Division
1,'' into Section XI. The NRC has approved the use of ASME BPV Code
Cases N-513-3 and N-705 in RG 1.147, which allows licensees to use
these code cases without prior permission from the NRC.
The first condition on the use of Appendix U is set forth in Sec.
50.55a(b)(2)(xxxiv)(A) of this final rule and requires that an ASME BPV
Code repair or replacement activity temporarily deferred under the
provisions of Nonmandatory Appendix U to the 2013 Edition of the ASME
BPV Code, Section XI, must be performed during the next scheduled
outage. This condition is consistent with the NRC's condition on the
use of ASME BPV Code Case N-513-3 in RG 1.147, Revision 17. Appendix U
defines that the evaluation period is the operational time for which
the temporary acceptance criteria are satisfied but not exceeding 26
months from the initial discovery of the condition. Original versions
of ASME BPV Code Case N-513 stated, in part, that certain flaws may be
acceptable without performing a repair/replacement activity for a
limited time, not to exceed the time to the next scheduled outage. The
NRC staff found that the acceptance of ASME BPV Code Case N-513 was
based on allowing continued plant operation with a monitored and
evaluated low safety significant degraded condition for a limited time
until plant shutdown. By allowing use of this Appendix, this option is
allowed rather than requiring an unnecessary plant shutdown to repair
the degradation. However, the NRC believes once the plant is shut down,
the degraded piping must be repaired.
The second condition on the use of Appendix U is set forth in Sec.
50.55a(b)(2)(xxxiv)(B) of this final rule. This paragraph requires the
use of the mandatory appendix in ASME BPV Code Case N-513-3 in lieu of
the appendix referenced in paragraph U-S1-4.2.1(c) of Appendix U (which
was inadvertently omitted from Appendix U). The NRC is incorporating by
reference the mandatory appendix in ASME BPV Code Case N-513-3 in Sec.
50.55a(a)(1)(iii)(A) because it is referenced in Sec.
50.55a(b)(2)(xxxiv)(B).
A proposed condition on Disposition of Flaws in Class 3 Components,
which was located in Sec. 50.55a(b)(2)(xxxiv) of the proposed rule, is
not included in this final rule based on public comments that the error
has been corrected by ASME in published erratum.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0
in the KIa and KIc Equations
The NRC is adding Sec. 50.55a(b)(2)(xxxv) to specify that when
licensees use the 2013 Edition of the ASME BPV Code, Section XI,
Appendix A, Paragraph A-4200, if T0 is available, then
RTT0 may be used in place of RTNDT for
applications using the KIc equation and the associated
KIc curve, but not for applications using the KIa
equation and the associated KIa curve.
Nonmandatory Appendix A provides a procedure based on linear
elastic fracture mechanics (LEFM) for determining the acceptability of
flaws that have been detected during inservice inspections that exceed
the allowable flaw indication standards of IWB-3500. Sub-article A-4200
provides a procedure for determining fracture toughness of the material
used in the LEFM analysis. The NRC staff's concern is related to the
proposed insertion regarding an alternative based on the use of the
Master Curve methodology to determine the nil-ductility transition
reference temperature RTNDT, which is an important parameter
in determining the fracture toughness of the material. Specifically,
the insertion proposed to use the Master Curve reference temperature
RTT0, which is defined as RTT0 = T0 +
35 [deg]F, where T0 is a material-specific temperature value
determined in accordance with ASTM E1921, ``Standard Test Method for
Determination of Reference Temperature, T0, for Ferritic
Steels in the Transition Range,'' to index (shift) the fracture
toughness KIc curve, based on the lower bound of static
initiation critical stress intensity factor, as well as the
KIa curve, based on the lower bound
[[Page 32945]]
of crack arrest critical stress intensity factor.
While use of RTT0 to index the KIc curve is
acceptable, using RTT0 to index the KIa curve is
questionable. This concern is based on the data analysis in ``A
Physics-Based Model for the Crack Arrest Toughness of Ferritic
Steels,'' written by NRC staff member Mark Kirk and published in
``Fatigue and Fracture Mechanics, 33rd Volume, ASTM STP 1417'' which
indicated that the crack arrest data does not support using
RTT0 as RTNDT to index the KIa curve.
This is also confirmed by industry data disclosed in a presentation,
``Final Results from the CARINA Project on Crack Initiation and Arrest
of Irradiated German RPV Steels for Neutron Fluences in the Upper
Bound,'' by AREVA at the 26th Symposium on Effects of Radiation on
Nuclear Materials (June 12-13, 2013, Indianapolis, Indiana, USA). The
NRC staff recognized that the proposed insertion is consistent with
ASME BPV Code Case N-629, ``Use of Fracture Toughness Test Data to
Establish Reference Temperature for Pressure Retaining Materials,''
which was accepted by the NRC without conditions. In addition to the
current NRC effort, the appropriate ASME BPV Code committee is in the
process of correcting this issue in a future revision of Appendix A of
Section XI.
With this condition, users of Appendix A can avoid using an
erroneous fracture toughness KIa value in their LEFM
analysis for determining the acceptability of a detected flaw in
applicable components. Therefore, the NRC is adding a condition which
permits the use of RTT0 in place of RTNDT in
applications using the KIc equation and the associated
KIc curve, but does not permit the use of RTT0 in
place of RTNDT in applications using the KIa
equation and the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of
Irradiated Materials
The NRC is adding Sec. 50.55a(b)(2)(xxxvi) to require licensees
using ASME BPV Code, Section XI, 2013 Edition, Appendix A, Paragraph A-
4400, to obtain NRC approval under Sec. 50.55a(z) before using
irradiated T0 and the associated RTT0 in
establishing fracture toughness of irradiated materials.
Sub-article A-4400 provides guidance for considering irradiation
effects on materials. The NRC staff's concern is related to use of
RTT0 based on measured T0 of the irradiated
materials. Specifically, the NRC staff has concerns over this sentence
in the proposed insertion: ``Measurement of RTT0 of
unirradiated or irradiated materials as defined in A-4200(b) is
permitted, including use of the procedures given in ASTM E1921 to
obtain direct measurement of irradiated T0.''
Permission of measurement of RTT0 of irradiated
materials, without providing guidelines regarding how to use the
measured parameter in determining the fracture toughness of the
irradiated materials, may mislead the users of Appendix A into adopting
methodology that has not been accepted by the NRC. With this condition,
users of Appendix A can avoid inappropriately using a fracture
toughness KIc value based on the irradiated T0
and the associated RTT0 in their LEFM analysis for
determining the acceptability of a detected flaw in applicable
components.
10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements
The NRC is adding new paragraphs (g)(2)(i), (ii), and (iii) and
revising current paragraphs (g) introductory text, (g)(2), (g)(3)
introductory text, and (g)(3)(i), (ii), and (v) to distinguish the
requirements for accessibility and preservice examination from those
for inservice inspection in Sec. 50.55a(g). In addition, consistent
with other paragraphs of this section, headings are added to the
subordinate paragraphs of (g) in order to enhance readability of the
regulation. No substantive change to the requirements are intended by
these revisions.
C. OM Code
10 CFR 50.55a(b)(3) Conditions on ASME OM Code
The NRC is revising Sec. 50.55a(b)(3) to clarify that Subsections
ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II,
III, and V; and Nonmandatory Appendices A through H and J through M of
the OM Code are each incorporated by reference into Sec. 50.55a. The
NRC is also clarifying that the OM Code Nonmandatory Appendices
incorporated by reference into Sec. 50.55a are approved for use, but
are not mandated. The Nonmandatory Appendices may be used by applicants
and licensees of nuclear power plants, subject to the conditions in
Sec. 50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
The NRC is revising Sec. 50.55a(b)(3)(i) to allow use of the 1994
Edition, 2008 Edition, and the 2009-1a Addenda of NQA-1, ``Quality
Assurance Requirements for Nuclear Facility Applications.'' The NRC
reviewed these editions and addenda, compared them to the previously
approved versions of NQA-1, and found that there were no significant
differences.
The NRC is removing the reference in Sec. 50.55a(b)(3)(i) to
versions of NQA-1 older than the 1994 Edition, inasmuch as these
versions do not appear to be in use at any nuclear power plant. The NRC
did not receive any adverse comments from any applicant or licensee
regarding concerns about removing versions of NQA-1 older than the 1994
Edition from the regulation. The NRC received one comment regarding
NQA-1, supporting incorporation by reference of NQA-1 but not
responding to the NRC's request for comment regarding the removal of
references to older versions of NQA-1. Accordingly, the NRC concludes
that removal of NQA-1 versions older than the 1994 Edition will not
have any adverse effect on licensees, and the final rule removes these
older versions from Sec. 50.55a(b)(3)(i).
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV)
Testing
The NRC is revising Sec. 50.55a(b)(3)(ii) to reflect the new
Appendix III, ``Preservice and Inservice Testing of Active Electric
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,''
of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition. Appendix
III of the OM Code establishes provisions for periodic verification of
the design-basis capability of MOVs within the scope of the IST
program. Appendix III of the OM Code reflects the incorporation of OM
Code Cases OMN-1, ``Alternative Rules for Preservice and Inservice
Testing of Active Electric Motor-Operated Valve Assemblies in Light-
Water Reactor Power Plants,'' and OMN-11, ``Risk-Informed Testing for
Motor-Operated Valves.'' The NRC is adding four new conditions on the
use of Mandatory Appendix III in new Sec. 50.55a(b)(3)(ii)(A), (B),
(C), and (D) to address periodic verification of MOV design-basis
capability. These new conditions are discussed in the next four
sections.
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval (First
Condition on Use of Mandatory Appendix III)
In the proposed rule, the NRC specified in Sec.
50.55a(b)(3)(ii)(A) that licensees evaluate the adequacy of the
diagnostic test interval for each MOV and adjust the interval as
necessary, but not later than 5 years or three refueling outages
(whichever is longer) from initial implementation of OM Code,
[[Page 32946]]
Appendix III. Paragraph III-3310(b) in Appendix III includes a
provision stating that if insufficient data exist to determine the IST
interval, then MOV inservice testing shall be conducted every two
refueling outages or 3 years (whichever is longer) until sufficient
data exist, from an applicable MOV or MOV group, to justify a longer
IST interval. As discussed in a final rule published September 22, 1999
(64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC
considers it appropriate to include a modification requiring licensees
to evaluate the information obtained for each MOV, during the first 5
years or three refueling outages (whichever is longer) of the use of
Appendix III to validate assumptions made in justifying a longer test
interval.
In response to public comments, the NRC revised Sec.
50.55a(b)(3)(ii)(A) to clarify its intent for licensees to evaluate the
test interval within 5 years or three refueling outages (whichever is
longer) following implementation of Appendix III to the OM Code, rather
than implying that every MOV must be tested within 5 years or three
refueling outages of the initial implementation of Appendix III. For
example, the condition allows grouping of MOVs to share test
information in the evaluation of the MOV periodic verification
intervals within 5 years or three refueling outages (whichever is
longer) of the implementation of OM Code, Appendix III. Therefore,
Sec. 50.55a(b)(3)(ii)(A) of this final rule states that licensees
shall evaluate the adequacy of the diagnostic test intervals
established for MOVs within the scope of OM Code, Mandatory Appendix
III, not later than 5 years or three refueling outages (whichever is
longer) from initial implementation of OM Code, Appendix III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk (Second Condition
on Use of Mandatory Appendix III)
The NRC is adding Sec. 50.55a(b)(3)(ii)(B) to require that when
using Mandatory Appendix III, licensees ensure that the potential
increase in core damage frequency (CDF) and large early release
frequency (LERF) associated with the extension is acceptably small when
extending exercise test intervals for high risk MOVs beyond a quarterly
frequency. As discussed in a final rule published September 22, 1999
(64 FR 51386), with respect to the use of OM Code Case OMN-1, the NRC
considers it important for licensees to have sufficient information
from the specific MOV, or similar MOVs, to demonstrate that exercising
on a refueling outage frequency does not significantly affect component
performance. The information may be obtained by grouping similar MOVs
and establishing periodic exercising intervals of MOVs in the group
over the refueling interval.
Section 50.55a(b)(3)(ii)(B) requires that the increase in the
overall plant CDF and LERF resulting from the extension be acceptably
small. As presented in RG 1.174, ``An Approach for Using Probabilistic
Risk Assessment [PRA] in Risk-Informed Decisions on Plant-Specific
Changes to the Licensing Basis,'' the NRC considers acceptably small
changes to be relative and to depend on the current plant CDF and LERF.
For plants with total baseline CDF of 10-\4\ per year or
less, acceptably small means CDF increases of up to 10-\5\
per year; and for plants with total baseline CDF greater than
10-\4\ per year, acceptably small means CDF increases of up
to 10-\6\ per year. For plants with total baseline LERF of
10-\5\ per year or less, acceptably small LERF increases are
considered to be up to 10-\6\ per year; and for plants with
total baseline LERF greater than 10-\5\ per year, acceptably
small LERF increases are considered to be up to 10-\7\ per
year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization (Third Condition on
Use of Mandatory Appendix III)
The NRC is adding Sec. 50.55a(b)(3)(ii)(C) to require, when
applying Mandatory Appendix III, that licensees categorize MOVs
according to their safety significance using the methodology described
in OM Code Case OMN-3, ``Requirements for Safety Significance
Categorization of Components Using Risk Insights for Inservice Testing
of LWR Power Plants,'' subject to the conditions discussed in RG 1.192,
or using an MOV risk ranking methodology accepted by the NRC on a
plant-specific or industry-wide basis in accordance with the conditions
in the applicable safety evaluation. Paragraph III-3720 in Appendix III
to the OM Code states that when applying risk insights, each MOV shall
be evaluated and categorized using a documented risk ranking
methodology. Further, Appendix III only addresses risk ranking
methodologies that include two risk categories. In light of the
potential extension of quarterly test intervals for high risk MOVs and
the relaxation of IST activities for low risk MOVs based on risk
insights, the NRC has determined that the rule should specify that
plant-specific or industry-wide risk ranking methodologies must have
been accepted by the NRC through RG 1.192 (which accepts OM Code Case
OMN-3 with the specified conditions) or the issuance of safety
evaluations. As noted in the response to public comments, the intent of
this condition is to indicate that when applying Appendix III to the OM
Code, licensees may use either a two-risk category approach (high or
low) or a three-risk category approach (high, medium, and low),
provided the risk ranking method has been accepted by the NRC.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time (Fourth Condition on Use of
Mandatory Appendix III)
The NRC is adding Sec. 50.55a(b)(3)(ii)(D) to require that when a
licensee applies Paragraph III-3600, ``MOV Exercising Requirements,''
of Appendix III to the OM Code, the licensee verify that the stroke
time of the MOV satisfies the assumptions in the plant's safety
analyses. Previous editions and addenda of the OM Code specified that
the licensee must perform quarterly MOV stroke time measurements that
could be used to verify that the MOV stroke time satisfies the
assumptions in the safety analyses consistent with plant TS. The need
for verification of the MOV stroke time during periodic exercising is
consistent with the NRC's lessons learned from the implementation of OM
Code Case OMN-1. However, Paragraph III-3600 of Appendix III of the
versions of the OM Code that will be incorporated by reference in this
rulemaking no longer require the verification of MOV stroke time during
periodic exercising. For this reason, the NRC is adopting this new
condition, which will effectively retain the need to verify that the
MOV stroke time during periodic exercising satisfies the assumptions in
the plant's safety analyses.
Based on the discussion during the public webinar on March 2, 2016,
the NRC revised the condition to clarify that it applies to MOVs
referenced in the plant TS. In particular, the NRC revised the
condition to indicate that when a licensee applies Paragraph III-3600
of Appendix III to the OM Code, the licensee shall verify that the
stroke time of MOVs specified in plant technical specifications
satisfies the assumptions in the plant's safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors
The NRC is adding Sec. 50.55a(b)(3)(iii) to apply specific
conditions for IST programs applicable to licensees of new nuclear
power plants in addition to the provisions of the OM Code as
incorporated by reference with conditions in Sec. 50.55a. Licensees of
``new reactors'' are, as identified in the paragraph: (1) Holders of
operating
[[Page 32947]]
licenses for nuclear power reactors that received construction permits
under this part on or after the date 12 months after August 17, 2017,
and (2) holders of combined licenses (COLs) issued under 10 CFR part
52, whose initial fuel loading occurs on or after the date 12 months
after August 17, 2017. This implementation schedule for new reactors is
consistent with the NRC regulations governing inservice testing in
Sec. 50.55a(f)(4)(i).
Commission Papers SECY-90-016, ``Evolutionary Light Water Reactor
(LWR) Certification Issues and Their Relationship to Current Regulatory
Requirements;'' SECY-93-087, ``Policy, Technical, and Licensing Issues
Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR)
Designs;'' SECY-94-084, ``Policy and Technical Issues Associated with
the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant
Designs;'' and SECY-95-132, ``Policy and Technical Issues Associated
with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive
Plant Designs (SECY-94-084),'' discuss IST programs for new reactors
licensed under 10 CFR part 52.
In recognition of new reactor designs and lessons learned from
nuclear power plant operating experience, the ASME is updating the OM
Code to incorporate improved IST provisions for components used in
nuclear power plants that were issued (or will be issued) construction
permits, or COLs, on or following January 1, 2000 (defined in the OM
Code as post-2000 plants). The first phase of the ASME effort
incorporated IST provisions that specify full flow pump testing and
other clarifications for post-2000 plants in the OM Code beginning with
the 2011 Addenda. The second phase of the ASME effort incorporated
preservice and inservice inspection and surveillance provisions for
pyrotechnic-actuated (squib) valves in the 2012 Edition of the OM Code.
The ASME is considering further modifications to the OM Code to address
additional lessons learned from valve operating experience and new
reactor issues. As described in the following paragraphs, Sec.
50.55a(b)(3)(iii) will include four specific conditions which are
discussed in the following paragraphs.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves
The NRC is adding Sec. 50.55a(b)(3)(iii)(A) to require that
licensees within the scope of Sec. 50.55a(b)(3)(iii) periodically
verify the capability of power-operated valves (POVs) to perform their
design-basis safety functions. While Appendix III to the OM Code
addresses this requirement for MOVs with the conditions specified in
Sec. 50.55a, applicable applicants and licensees will need to develop
programs to periodically verify the design-basis capability of other
POVs. The NRC's Regulatory Issue Summary 2000-03, ``Resolution of
Generic Issue 158: Performance of Safety-Related Power-Operated Valves
Under Design Basis Conditions,'' provides attributes for a successful
long-term periodic verification program for POVs by incorporating
lessons learned from MOV performance at operating nuclear power plants
and research programs. Implementation of Appendix III to the OM Code as
accepted in Sec. 50.55a(b)(3)(ii) satisfies Sec. 50.55a(b)(3)(iii)(A)
for MOVs.
Section 50.55a(b)(3)(iii)(A) is consistent with the Commission
policy for new reactors summarized in an NRC Staff Memorandum,
``Consolidation of SECY-94-084 and SECY-95-132,'' dated July 24, 1995,
that (a) the design capability of safety-related POVs should be
demonstrated by a qualification test prior to installation; (b) prior
to initial startup, POV capability under design-basis differential
pressure and flow should be verified by a pre-operational test; and (c)
during the operational phase, POV capability under design-basis
differential pressure and flow should be verified periodically through
a program similar to that developed for MOVs in Generic Letter 89-10,
``Safety-Related Motor-Operated Valve Testing and Surveillance,'' dated
June 28, 1989.\2\
---------------------------------------------------------------------------
\2\ The NRC issued seven supplements to provide guidance for the
implementation of the MOV testing program requested in Generic
Letter 89-10. The supplements to Generic Letter 89-10 did not modify
the substance of the MOV testing program requested in Generic Letter
89-10 to provide reasonable assurance in the capability of safety-
related MOVs to perform their design-basis functions.
---------------------------------------------------------------------------
The condition in Sec. 50.55a(b)(3)(iii)(A) specifies with the same
level of detail as the condition in Sec. 50.55a(b)(3)(ii) that nuclear
power plant licensees must establish a program to ensure the continued
capability of MOVs in performing their design-basis safety functions.
When establishing the MOV periodic verification condition, the NRC
provided guidance in the final rule published September 22, 1999 (64 FR
51370), for licensees to develop acceptable programs that would satisfy
the MOV periodic verification condition. Similarly, the NRC staff is
providing guidance herein for new reactor applicants and licensees to
develop acceptable programs to periodically verify the capability of
POVs to perform their design-basis safety functions.
In NUREG-2124, ``Final Safety Evaluation Report [FSER] Related to
the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and
4,'' the NRC staff found the provisions established by the COL
applicant for Vogtle Units 3 and 4 in its Final Safety Analysis Report
(FSAR), Revision 5, Section 3.9.6.2.2, ``Valve Testing,'' to
periodically verify the capability of POVs (such as air-operated valves
(AOVs), solenoid-operated valves (SOVs), and hydraulic-operated valves
(HOVs)) to perform their design-basis safety functions to be
acceptable. In particular, the Vogtle Units 3 and 4 FSAR specifies
that:
Power-operated valves other than active MOVs are exercised
quarterly in accordance with OM ISTC, unless justification is
provided in the inservice testing program for testing these valves
at other than Code mandated frequencies. Although the design basis
capability of power-operated valves is verified as part of the
design and qualification process, power-operated valves that perform
an active safety function are tested again after installation in the
plant, as required, to ensure valve setup is acceptable to perform
their required functions, consistent with valve qualification. These
tests, which are typically performed under static (no flow or
pressure) conditions, also document the ``baseline'' performance of
the valves to support maintenance and trending programs. During the
testing, critical parameters needed to ensure proper valve setup are
measured. Depending on the valve and actuator type, these parameters
may include seat load, running torque or thrust, valve travel,
actuator spring rate, bench set and regulator supply pressure.
Uncertainties associated with performance of these tests and use of
the test results (including those associated with measurement
equipment and potential degradation mechanisms) are addressed
appropriately. Uncertainties may be considered in the specification
of acceptable valve setup parameters or in the interpretation of the
test results (or a combination of both). Uncertainties affecting
both valve function and structural limits are addressed. Additional
testing is performed as part of the air-operated valve (AOV)
program, which includes the key elements for an AOV Program as
identified in the JOG AOV program document, Joint Owners Group Air
Operated Valve Program Document, Revision 1, December 13, 2000
(References 203 and 204) [JOG AOV Program Document, Revision 1,
December 13, 2000 (ADAMS Accession No. ML010950310), and NRC comment
letter dated October 8, 1999, to Nuclear Energy Institute (ADAMS
Accession No. ML020360077)]. The AOV program incorporates the
attributes for a successful power-operated valve long-term periodic
verification program, as discussed in Regulatory Issue Summary 2000-
03, Resolution of Generic Safety Issue 158: Performance of Safety-
Related Power-Operated Valves Under Design Basis
[[Page 32948]]
Conditions, by incorporating lessons learned from previous nuclear
power plant operations and research programs as they apply to the
periodic testing of air- and other power-operated valves included in
the IST program.
For example, key lessons learned addressed in the AOV program
include:
Valves are categorized according to their safety
significance and risk ranking.
Setpoints for AOVs are defined based on current vendor
information or valve qualification diagnostic testing, such that the
valve is capable of performing its design-basis function(s).
Periodic static testing is performed, at a minimum on
high risk (high safety significance) valves, to identify potential
degradation, unless those valves are periodically cycled during
normal plant operation, under conditions that meet or exceed the
worst case operating conditions within the licensing basis of the
plant for the valve, which would provide adequate periodic
demonstration of AOV capability. If required based on valve
qualification or operating experience, periodic dynamic testing is
performed to re-verify the capability of the valve to perform its
required functions.
Sufficient diagnostics are used to collect relevant
data (e.g., valve stem thrust and torque, fluid pressure and
temperature, stroke time, operating and/or control air pressure,
etc.) to verify the valve meets the functional requirements of the
qualification specification.
Test frequency is specified, and is evaluated each
refueling outage based on data trends as a result of testing.
Frequency for periodic testing is in accordance with References 203
and 204, with a minimum of 5 years (or 3 refueling cycles) of data
collected and evaluated before extending test intervals.
Post-maintenance procedures include appropriate
instructions and criteria to ensure baseline testing is re-performed
as necessary when maintenance on the valve, repair or replacement,
have the potential to affect valve functional performance.
Guidance is included to address lessons learned from
other valve programs specific to the AOV program.
Documentation from AOV testing, including maintenance
records and records from the corrective action program are retained
and periodically evaluated as a part of the AOV program.
* * * * *
The attributes of the AOV testing program described above, to
the extent that they apply to and can be implemented on other
safety-related power-operated valves, such as electro-hydraulic
operated valves, are applied to those other power-operated valves.''
(Vogtle Electric Generating Plant, Units 3 and 4, Updated Final
Safety Analysis Report (UFSAR), Section 3.9.6, ``Inservice Testing
of Pumps and Valves'')
Applicable applicants and licensees may follow the method described
in the Vogtle Units 3 and 4 FSAR in satisfying Sec.
50.55a(b)(3)(iii)(A), or may establish a different method, subject to
evaluation by the NRC during the licensing process or inspections.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
The NRC is adding Sec. 50.55a(b)(3)(iii)(B) to require that
licensees within the scope of Sec. 50.55a(b)(3)(iii) perform bi-
directional testing of check valves within the IST program where
practicable. Nuclear power plant operating experience has revealed that
testing check valves in only the flow direction can result in
significant degradation, such as a missing valve disc, not being
identified by the test. Nonmandatory Appendix M, ``Design Guidance for
Nuclear Power Plant Systems and Component Testing,'' to OM Code, 2011
Addenda and 2012 Edition, includes guidance for the design of new
reactors to enable bi-directional testing of check valves. New reactor
designs will provide the capability for licensees of new nuclear power
plants to perform bi-directional testing of check valves within the IST
program. Bi-directional testing of check valves in new reactors, as
required by Sec. 50.55a(b)(3)(iii)(B), could be accomplished by valve-
specific testing or condition monitoring activities in accordance with
Appendix II to the OM Code as accepted in Sec. 50.55a. The NRC is
specifying this provision for bi-directional testing of check valves
for new reactors in Sec. 50.55a(b)(3)(iii)(B) to emphasize that new
reactors should include the capability for bi-directional testing of
check valves as part of their initial design.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration
In the proposed rule, the NRC proposed adding Sec.
50.55a(b)(3)(iii)(C) to require that licensees subject to Sec.
50.55a(b)(3)(iii) monitor flow-induced vibration (FIV) from
hydrodynamic loads and acoustic resonance during preservice testing and
inservice testing to identify potential adverse flow effects that might
impact components within the scope of the IST program.
Nuclear power plant operating experience has revealed the potential
for adverse flow effects from vibration caused by hydrodynamic loads
and acoustic resonance on components in the reactor coolant, steam, and
feedwater systems. Therefore, the licensee will be required to address
potential adverse flow effects on safety-related pumps, valves, and
dynamic restraints within the IST program in the reactor coolant,
steam, and feedwater systems from hydraulic loading and acoustic
resonance during plant operation. In response to public comments, the
NRC revised Sec. 50.55a(b)(3)(iii)(C) to clarify its intent that FIV
monitoring of components may be conducted during preservice testing or
inservice testing. This requirement will confirm that piping,
components, restraints, and supports have been designed and installed
to withstand the dynamic effects of steady-state FIV and anticipated
operational transient conditions. As part of preservice testing
activities, the initial test program may be used to verify that safety-
related piping and components are properly installed and supported such
that vibrations caused by steady-state or dynamic effects do not result
in excessive stress or fatigue in safety-related plant systems.
In the Vogtle Units 3 and 4 FSER, the NRC staff found the
provisions established by the COL applicant for Vogtle Units 3 and 4 in
its FSAR, Revision 5, Section 3.9, ``Mechanical Systems and
Components,'' Section 14.2.9, ``Preoperational Test Descriptions,'' and
Section 14.2.10, ``Startup Test Procedures,'' with incorporation by
reference of corresponding sections of the AP1000 Design Control
Document (DCD), to monitor FIV from hydrodynamic loads and acoustic
resonance during preservice testing or inservice testing to be
acceptable. In particular, the NRC staff stated in the Vogtle Units 3
and 4 FSER:
AP1000 DCD Tier 2, Section 3.9.2, ``Dynamic Testing and
Analysis,'' describes tests to confirm that piping, components,
restraints, and supports have been designed to withstand the dynamic
effects of steady-state FIV and anticipated operational transient
conditions. Section 14.2.9.1.7, ``Expansion, Vibration and Dynamic
Effects Testing,'' in AP1000 DCD Tier 2, Chapter 14, ``Initial Test
Program,'' states that the purpose of the expansion, vibration and
dynamic effects testing is to verify that safety-related, high
energy piping and components are properly installed and supported
such that, in addition to other factors, vibrations caused by
steady-state or dynamic effects do not result in excessive stress or
fatigue to safety-related plant systems. Nuclear power plant
operating experience has revealed the potential for adverse flow
effects from vibration caused by hydrodynamic loads and acoustic
resonance on reactor coolant, steam, and feedwater systems. . . . In
its response, SNC [Vogtle Units 3 and 4 COL applicant] stated that
it intended to use the overall Initial Test Program to demonstrate
that the plant has been constructed as designed and the systems
perform consistent with design requirements. SNC referenced the
provisions in the AP1000 DCD for vibration monitoring and testing to
be implemented at VEGP. For example, the applicant notes that AP1000
DCD Tier 2, Section 3.9.2.1, ``Piping Vibration, Thermal Expansion
and Dynamic Effects,'' specifies that the preoperational test
[[Page 32949]]
program for ASME BPV Code, Section III, Class 1, 2, and 3 piping
systems simulates actual operating modes to demonstrate that
components comprising these systems meet functional design
requirements and that piping vibrations are within acceptable
levels. SNC indicates that the planned vibration testing program
described in AP1000 DCD Tier 2, Sections 14.2.9 and 14.2.10, with
the preservice and IST programs described in AP1000 DCD Tier 2,
Sections 3.9.3.4.4 and 3.9.6, will confirm component installation in
accordance with design requirements, and address the effects of
steady-state (flow-induced) and transient vibration to ensure the
operability of valves and dynamic restraints in the IST Program. The
NRC staff considers the response by SNC clarifies its application of
the provisions in the AP1000 DCD to ensure that potential adverse
flow effects will be addressed at VEGP. Therefore, the NRC staff
considers Standard Content Open Item 3.9-5 to be resolved for the
VEGP COL application.'' (NUREG-2124, ``Final Safety Evaluation
Report Related to the Combined Licenses for Vogtle Electric
Generating Plant, Units 3 and 4,'' Section 3.9.6, ``Inservice
Testing of Pumps and Valves (Related to RG 1.206, Section C.III.1,
Chapter 3, C.I.3.9.6, `Functional Design, Qualification, and
Inservice Testing Programs for Pumps, Valves, and Dynamic
Restraints')'').
As clarified in the final rule in response to public comments, a
licensee may monitor components for adverse FIV effects during
preservice testing or IST activities.
Applicable applicants and licensees may either apply the methods
described in the Vogtle Units 3 and 4 FSAR in satisfying Sec.
50.55a(b)(3)(iii)(C) or develop their own plant-specific methods to
satisfy Sec. 50.55a(b)(3)(iii)(C) for NRC review during the licensing
process.
10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems
The NRC is adding Sec. 50.55a(b)(3)(iii)(D) to require that
licensees within the scope of Sec. 50.55a(b)(3)(iii) establish a
program to assess the operational readiness of pumps, valves, and
dynamic restraints within the scope of the Regulatory Treatment of Non-
Safety Systems (RTNSS) for applicable reactor designs. As of the time
of this final rule, these are designs which have been certified in a
design certification rule under 10 CFR part 52. In SECY-94-084 and
SECY-95-132, the Commission discusses RTNSS policy and technical issues
associated with passive plant designs. Some new nuclear power plants
have advanced light-water reactor (ALWR) designs that use passive
safety systems that rely on natural forces, such as density
differences, gravity, and stored energy to supply safety injection
water and to provide reactor core and containment cooling. Active
systems in passive ALWR designs are categorized as non-safety systems
with limited exceptions. Active systems in passive ALWR designs provide
the first line of defense to reduce challenges to the passive systems
in the event of a transient at the nuclear power plant. Active systems
that provide a defense-in-depth function in passive ALWR designs need
not meet all of the acceptance criteria for safety-related systems.
However, there should be a high level of confidence that these active
systems will be available and reliable when needed. The combined
activities to provide confidence in the capability of these active
systems in passive ALWR designs to perform their functions important to
safety are referred to as the RTNSS program. In the NRC Staff
Memorandum, ``Consolidation of SECY-94-084 and SECY-95-132,'' dated
July 24, 1995, the NRC staff provided a consolidated list of the
approved policy and technical positions associated with RTNSS equipment
in passive plant designs discussed in SECY-94-084 and SECY-95-132. This
new paragraph specifies the need for licensees to assess the
operational readiness of RTNSS pumps, valves, and dynamic restraints.
The July 24, 1995, staff memorandum summarizes the Commission
policy positions related to inservice testing of RTNSS pumps and valves
as follows:
The staff also concluded that additional inservice testing
requirements may be necessary for certain pumps and valves in
passive plant designs. The unique passive plant design relies
significantly on passive safety systems, but also depends on non-
safety systems (which are traditionally safety-related systems in
current light-water reactors) to prevent challenges to passive
systems. Therefore, the reliable performance of individual
components is a very significant factor in enhancing the safety of
passive plant design. The staff recommends that the following
provisions be applied to passive ALWR plants to ensure reliable
component performance.
1. Important non-safety-related components are not required to
meet criteria similar to safety-grade criteria. However, the non-
safety-related piping systems with functions that have been
identified as being important by the RTNSS process should be
designed to accommodate testing of pumps and valves to assure that
the components meet their intended functions. Specific positions on
the inservice testing requirements for those components will be
determined as a part of the staff's review of plant-specific
implementation of the regulatory treatment of non-safety systems for
passive reactor designs.
2. . . . The vendors for advanced passive reactors, for which
the final designs are not complete, have sufficient time to include
provisions in their piping system designs to allow testing at power.
Quarterly testing is the base testing frequency in the Code and the
original intent of the Code. Furthermore, the COL holder may need to
test more frequently than during cold shutdowns or at every
refueling outage to ensure that the reliable performance of
components is commensurate with the importance of the safety
functions to be performed and with system reliability goals.
Therefore, to the extent practicable, the passive ALWR piping
systems should be designed to accommodate the applicable Code
requirements for the quarterly testing of valves. However, design
configuration changes to accommodate Code-required quarterly testing
should be done only if the benefits of the test outweigh the
potential risk.
3. The passive system designs should incorporate provisions (1)
to permit all critical check valves to be tested for performance, to
the extent practicable, in both forward- and reverse-flow
directions, although the demonstration of a non-safety direction
test need not be as rigorous as the corresponding safety direction
test, and (2) to verify the movement of each check valve's obturator
during inservice testing by observing a direct instrumentation
indication of the valve position such as a position indicator or by
using nonintrusive test methods.
4. . . . Similarly, to the extent practicable, the design of
non-safety-related piping systems with functions under design-basis
condition that have been identified as being important by the RTNSS
process should incorporate provisions to periodically test power-
operated valves in the system during operations to assure that the
valves meet their intended functions under design-basis conditions.
5. . . . Mispositioning may occur through actions taken locally
(manual or electrical), at a motor control center, or in the control
room, and includes deliberate changes of valve position to perform
surveillance testing. The staff will determine if and the extent to
which this concept should be applied to MOVs in important non-
safety-related systems when the staff reviews the implementation of
the regulatory treatment of non-safety systems.'' (NRC Staff
Memorandum, ``Consolidation of SECY-94-084 and SECY-95-132,'' July
24, 1995, pages 26-28).
Consistent with the Commission policy for RTNSS equipment, Sec.
50.55a(b)(3)(iii)(D) specifies that new reactor licensees shall assess
the operational readiness of pumps, valves, and dynamic restraints
within the RTNSS scope. This regulatory requirement will allow
licensees flexibility in developing programs to assess operational
readiness of RTNSS components that satisfy the Commission policy.
Guidance on the implementation of the Commission policy for RTNSS
equipment is set forth in NRC Inspection Procedure 73758, ``Part 52,
Functional Design and Qualification, and Preservice and Inservice
Testing Programs for Pumps, Valves and
[[Page 32950]]
Dynamic Restraints,'' dated April 19, 2013.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
The NRC is revising Sec. 50.55a(b)(3)(iv) to address Appendix II,
``Check Valve Condition Monitoring Program,'' provided in the 2003
Addenda through the 2012 Edition of the OM Code. In the proposed rule,
the NRC proposed a condition in Sec. 50.55a(b)(3)(iv) to provide
assurance that the valve or group of valves is capable of performing
its intended function(s) over the entire interval. Public comments
indicated that the proposed condition could be misinterpreted.
Therefore, the NRC revised the proposed condition to clarify that the
implementation of Appendix II must include periodic sampling of the
check valves over the maximum interval allowed by Appendix II for the
check valve condition monitoring program. A new table was added to the
paragraph to specify the maximum intervals between check valve
condition monitoring activities when applying interval extensions.
The conditions currently specified for the use of Appendix II, 1995
Edition with the 1996 and 1997 Addenda, and 1998 Edition through the
2002 Addenda, of the OM Code remain unchanged by this final rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
The NRC is adding a new condition, Sec. 50.55a(b)(3)(vii), to
prohibit the use of Subsection ISTB, ``Inservice Testing of Pumps in
Light-Water Reactor Nuclear Power Plants,'' in the 2011 Addenda of the
OM Code. In the 2011 Addenda to the OM Code, the upper end of the
``Acceptable Range'' and the ``Required Action Range'' for flow and
differential or discharge pressure for comprehensive pump testing in
Subsection ISTB was raised to higher values. The NRC staff on the OM
Code committee accepted the proposed increase of the upper end of the
``Acceptable Range'' and ``Required Action Range'' with the planned
addition of a requirement for a pump periodic verification test program
in the OM Code. However, the 2011 Addenda to the OM Code did not
include the requirement for a pump periodic verification test program.
Since then, the 2012 Edition of the OM Code has incorporated Mandatory
Appendix V, ``Pump Periodic Verification Test Program,'' which supports
the changes to the acceptable and required action ranges for
comprehensive pump testing in Subsection ISTB. Therefore, the new Sec.
50.55a(b)(3)(vii) prohibits the use of Subsection ISTB in the 2011
Addenda of the OM Code. Licensees will be allowed to apply Subsection
ISTB with the revised acceptable and required action ranges in the 2012
Edition of the OM Code as incorporated by reference in Sec. 50.55a.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
The NRC is adding Sec. 50.55a(b)(3)(viii) to specify that
licensees who wish to implement Subsection ISTE, ``Risk-Informed
Inservice Testing of Components in Light-Water Reactor Nuclear Power
Plants,'' of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition,
must request and obtain NRC approval in accordance with Sec. 50.55a(z)
to apply Subsection ISTE on a plant-specific basis as a risk-informed
alternative to the applicable IST requirements in the OM Code.
In the 2009 Edition of the OM Code, the ASME included new
Subsection ISTE that describes a voluntary risk-informed approach in
developing an IST program for pumps and valves at nuclear power plants.
If a licensee chooses to implement this risk-informed IST approach,
Subsection ISTE indicates that all requirements in Subsection ISTA,
``General Requirements,'' Subsection ISTB, and Subsection ISTC,
``Inservice Testing of Valves in Light-Water Reactor Nuclear Power
Plants,'' of the OM Code continue to apply, except those identified in
Subsection ISTE. The ASME selected risk-informed guidance from OM Code
Cases OMN-1, OMN-3, OMN-4, ``Requirements for Risk Insights for
Inservice Testing of Check Valves at LWR Power Plants,'' OMN-7,
``Alternative Requirements for Pump Testing,'' OMN-11, and OMN-12,
``Alternative Requirements for Inservice Testing Using Risk Insights
for Pneumatically and Hydraulically Operated Valve Assemblies in Light-
Water Reactor Power Plants,'' in preparing Subsection ISTE of the OM
Code.
During development of Subsection ISTE, the NRC staff participating
on the OM Code committees indicated that the conditions specified in RG
1.192 for the use of the applicable OM Code Cases need to be considered
when evaluating the acceptability of the implementation of Subsection
ISTE. In addition, the NRC staff noted that several aspects of
Subsection ISTE will need to be addressed on a case-by-case basis when
determining the acceptability of its implementation. Therefore, the new
condition in Sec. 50.55a(b)(3)(viii) requires that licensees who wish
to implement Subsection ISTE of the OM Code must request approval from
the NRC to apply Subsection ISTE on a plant-specific basis as a risk-
informed alternative to the applicable IST requirements in the OM Code.
Nuclear power plant applicants for construction permits under 10
CFR part 50, or combined licenses for construction and operation under
10 CFR part 52, may describe their proposed implementation of the risk-
informed IST approach specified in Subsection ISTE of the OM Code for
NRC review in their applications.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
The NRC is adding a condition on the use of Subsection ISTF in
Sec. 50.55a(b)(3)(ix). First, the condition states that Subsection
ISTF, 2011 Addenda, is prohibited for use. Second, the condition
specifies that licensees applying Subsection ISTF, ``Inservice Testing
of Pumps in Light-Water Reactor Nuclear Power Plants--Post-2000
Plants,'' in the 2012 Edition of the OM Code shall satisfy the
requirements of Mandatory Appendix V, ``Pump Periodic Verification Test
Program,'' of the OM Code, 2012 Edition.
As previously discussed regarding the new condition in Sec.
50.55a(b)(3)(vii), the upper end of the ``Acceptable Range'' and the
``Required Action Range'' for flow and differential or discharge
pressure for comprehensive pump testing in Subsection ISTB in the OM
Code was raised to higher values in combination with the incorporation
of Mandatory Appendix V, ``Pump Periodic Verification Test Program.''
However, the 2011 Addenda of the OM Code does not include Appendix V.
In addition, Subsection ISTF in the 2011 Addenda and 2012 Edition of
the OM Code does not include a requirement for a pump periodic
verification test program. Therefore, the new condition in Sec.
50.55a(b)(3)(ix) requires that the provisions of Appendix V be applied
when implementing Subsection ISTF of the 2012 Edition of the OM Code to
support the application of the upper end of the Acceptable Range and
the Required Action Range for flow and differential or discharge
pressure for inservice pump testing in Subsection ISTF.
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
The NRC is adding Sec. 50.55a(b)(3)(xi) to emphasize the
provisions in OM Code, 2012 Edition, Subsection ISTC-3700, ``Position
Verification Testing,'' to verify that valve obturator position is
accurately indicated. Subsection ISTC-3700 of the OM Code requires that
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valves with remote position indicators shall be observed locally at
least once every 2 years to verify that valve operation is accurately
indicated. Subsection ISTC-3700 states that where practicable, this
local observation should be supplemented by other indications, such as
the use of flow meters or other suitable instrumentation to verify
obturator position. Subsection ISTC-3700 also states that where local
observation is not possible, other indications shall be used for
verification of valve operation. Nuclear power plant operating
experience has revealed that reliance on indicating lights and stem
travel are not sufficient to satisfy the requirement in ISTC-3700 to
verify that valve operation is accurately indicated. Appendix A,
``General Design Criteria for Nuclear Power Plants,'' to 10 CFR part 50
requires that where generally recognized codes and standards are used,
they shall be identified and evaluated to determine their
applicability, adequacy, and sufficiency, and shall be supplemented or
modified as necessary to assure a quality product in keeping with the
required safety function. This new condition specifies that when
implementing OM Code, Subsection ISTC-3700, licensees shall verify that
valve operation is accurately indicated by supplementing valve position
indicating lights with other indications, such as flow meters or other
suitable instrumentation, to provide assurance of proper obturator
position. The OM Code specifies obturator movement verification in
order to detect certain internal valve failure modes consistent with
the definition of `exercising' found in ISTA-2000, ``Definitions,''
(i.e., demonstration that the moving parts of a component function).
Verification of the ability of an obturator to change or maintain
position is an essential element of valve operational readiness
determination, which is a fundamental aspect of the OM Code.
The NRC initially emphasized the ASME OM Code requirement for valve
position indication in 1995 in the original issuance of NUREG-1482,
``Guidelines for Inservice Testing at Nuclear Power Plants,'' paragraph
4.2.5. The NRC's position is further elaborated in NUREG-1482 (Revision
2), ``Guidelines for Inservice Testing at Nuclear Power Plants:
Inservice Testing of Pumps and Valves and Inservice Examination and
Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants,''
paragraph 4.2.7. As discussed in NUREG-1482 (Revision 2), ISTC-3700
allows flexibility to licensees in verifying that operation of valves
with remote position indicators is accurately indicated. For example,
NUREG-1482 refers to various methods to verify valve operation, such as
nonintrusive techniques, flow initiation or absence of flow, leak
testing, and pressure testing. The extent of verification necessary for
valve operation to satisfy ISTC-3700 will depend on the type of valve,
the sophistication of the diagnostic equipment used in testing the
valve, possible failure modes of the valve, and the operating history
of the valve and similar valve types. To satisfy ISTC-3700, the
licensee is responsible for developing and implementing a method to
provide reasonable assurance that valve operation is accurately
indicated.
The NRC is requiring this condition for the implementation of the
2012 Edition of the OM Code for the 120-month IST interval in order to
allow additional time for licensees to comply with this condition.
10 CFR 50.55a(f): Preservice and Inservice Testing Requirements
The NRC is revising the introductory text of Sec. 50.55a(f) to
indicate that systems and components must meet the requirements for
``preservice and inservice testing'' in the applicable ASME Codes and
that both activities are referred to as ``inservice testing'' in the
remainder of paragraph (f). The change clarifies that the OM Code
includes provisions for preservice testing of components as part of its
overall provisions for IST programs. No expansion of IST program scope
was intended by this clarification.
In the proposed rule, the NRC included references to the OM Code in
Sec. 50.55a(f)(3)(iii)(A), Class 1 Pumps and Valves: First Provision;
Sec. 50.55a(f)(3)(iii)(B), Class 1 Pumps and Valves: Second Provision;
Sec. 50.55a(f)(3)(iv)(A), Class 2 and 3 Pumps and Valves: First
Provision; and Sec. 50.55a(f)(3)(iv)(B): Second Provision; to align
the regulatory language with the current ASME OM Code used for IST
programs. Because Sec. 50.55a(f)(3)(iii) and (iv) specifically
reference Class 1, 2, or 3 pumps and valves, the proposed changes to
these paragraphs referencing the OM Code are unnecessary and have not
been adopted in this final rule.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for
Operating Plants
The NRC is revising Sec. 50.55a(f)(4) to clarify that this
paragraph is applicable to pumps and valves that are within the scope
of the OM Code. This revision aligns the scope of pumps and valves for
inservice testing with the scope defined in the OM Code.
Public comments on the alignment of the IST program scope in Sec.
50.55a(f)(4) indicated that the nuclear industry is addressing the
requirements in 10 CFR part 50, appendices A and B, to establish an IST
program for safety-related pumps and valves that are not classified as
ASME BPV Code Class 1, 2, or 3 components through either the OM Code
provisions or augmented IST programs. For example, one public commenter
indicated that generally, augmented IST programs are designed to meet
the OM Code where practicable, but relief requests are not required
when alternate testing is necessary. The NRC regulations in Sec.
50.55a address the concept of augmented IST programs for pumps and
valves at nuclear power plants. For example, Sec. 50.55a(f)(6)(ii),
``Augmented IST requirements,'' indicates that the licensee may follow
an augmented IST program for pumps and valves for which the NRC deems
that added assurance of operational readiness is necessary. The NRC
finds that an augmented IST program as addressed in Sec.
50.55a(f)(6)(ii) is acceptable for safety-related pumps and valves that
are not classified as ASME BPV Code Class 1, 2, or 3 components.
Public commenters were concerned that the alignment of the scope of
the OM Code and Sec. 50.55a would cause a potential paperwork burden
for the submittal of relief or alternative requests for safety-related
pumps and valves that are not classified as ASME BPV Code Class 1, 2,
or 3 components. In response to these comments, the NRC included a
provision in Sec. 50.55a(f)(4) that the IST requirements for pumps and
valves that are within the scope of the OM Code but are not classified
as ASME BPV Code Class 1, Class 2, or Class 3 may be satisfied as an
augmented IST program in accordance with Sec. 50.55a(f)(6)(ii) without
requesting relief under Sec. 50.55a(f)(5) or alternatives under Sec.
50.55a(z). This use of an augmented IST program may be acceptable
provided the basis for deviations from the OM Code, as incorporated by
reference in this section, demonstrates an acceptable level of quality
and safety, or that implementing the Code provisions would result in
hardship or unusual difficulty without a compensating increase in the
level of quality and safety, where documented and available for NRC
review. This additional provision avoids the potential paperwork burden
for the submittal of relief or alternative requests by allowing the
licensee to maintain the documentation demonstrating an acceptable
level of quality and safety on site for NRC review, as appropriate. The
[[Page 32952]]
documentation and availability of the basis for deviations from the OM
Code for NRC review are acceptable for pumps and valves within the
scope of the OM Code but not classified as ASME BPV Code Class 1, 2, or
3, based on their lower safety significance in comparison to ASME BPV
Code Class 1, 2, and 3 pumps and valves.
10 CFR 50.55a(g)(4) Inservice Inspection Standards Requirement for
Operating Plants
The NRC recognizes that updating an Appendix VIII program is a
complex and time-consuming process. The NRC also recognizes that
licensees would face the possibility of needing to maintain multiple
Appendix VIII programs if units were to update their ISI programs on
different dates. Maintaining certifications to multiple Appendix VIII
programs would be very complicated, while not improving the
effectiveness of the programs. Based on public comments, and to assist
licensees in updating and coordinating their ISI programs, the NRC is
adding two options to the regulations. First, the NRC is revising Sec.
50.55a(g)(4)(i) and (ii) to clarify that a licensee whose ISI interval
commences during the 12- to 18-month period after the approval date of
this final rule, may delay the update of their Appendix VIII program by
up to 18 months after the approval date of this final rule. This will
provide licensees with enough time to incorporate the changes for the
new Appendix VIII program. Second, the NRC is adding the option for
licensees to update their ISI program to use the latest edition and
addenda of Appendix VIII incorporated by reference in Sec.
50.55a(a)(1) at any time in the licensee's ten-year interval. Licensees
can normally update their ISI programs using all or portions of newer
versions of ASME BPV Code Section XI under Sec. 50.55a(g)(4)(iv),
subject to NRC review and approval. While some requests to use portions
of ASME BPV Code Section XI require a detailed review by the NRC, a
licensee asking to use the entire latest incorporated-by-reference
version of Appendix VIII would certainly be approved by the NRC staff
in this process. This provision will, therefore, allow licensees to use
the latest incorporated version of Appendix VIII, as long as it is
coupled with the same edition and addenda of Appendix I, without the
NRC review and approval process. This will allow licensees to
coordinate their ISI programs and use the latest approved version of
Appendix VIII without the delay imposed by submitting a relief request
under Sec. 50.55a (g)(4)(iv).
D. ASME Code Cases
Administrative Changes to References in Sec. 50.55a to NRC Regulatory
Guides Identifying ASME Code Cases Approved for Use by the NRC
The NRC is removing the revision number of the three RGs currently
approved by the Office of the Federal Register for incorporation by
reference throughout the substantive provisions of Sec. 50.55a
addressing the ASME Code Cases, i.e., paragraphs (b) through (g). The
revision numbers for the RGs approved for incorporation by reference
(currently, RG 1.84, RG 1.147, and RG 1.192) will be retained in Sec.
50.55a(a)(3)(i) through (iii), where the RGs are listed by full title,
including revision number. These changes simplify the regulatory
language containing cross-references to these RGs and reduce the
possibility of NRC error in preparing future amendments to Sec. 50.55a
with respect to these RGs. These changes are administrative in nature
and do not change substantive requirements with respect to the RGs and
the Code Cases listed in the RGs.
Administrative Changes To Comply With Requirements for Incorporation by
Reference
The NRC is revising Sec. 50.55a(a)(1)(iii) to maintain the ASME
Code Cases in alphanumeric order.
Organization of NRC's Discussion of the Six ASME Code Cases
Incorporated by Reference in This Final Rule
The discussions under the following headings address four of the
six ASME Code Cases being incorporated by reference in this rulemaking
(N-729-4, N-770-2, N-824, and OMN-20). A fifth ASME Code Case, N-852,
is discussed in Section II.A, ``ASME BPV Code, Section III,'' because
the NRC's approval of that Code Case relates to a provision of Section
III, which is addressed in Sec. 50.55a(b)(1)(ix). The sixth ASME Code
Case, N-513-3, is discussed in Section II.B, ``ASME BPV Code, Section
XI,'' because the NRC's approval of that Code Case relates to a
provision of Section XI, which is addressed in Sec.
50.55a(b)(2)(xxxiv).
ASME BPV Code Case N-729-4
On September 10, 2008, the NRC issued a final rule to update Sec.
50.55a to the 2004 Edition of the ASME BPV Code (73 FR 52730). As part
of the final rule, Sec. 50.55a(g)(6)(ii)(D) implemented an augmented
ISI program for the examination of pressurized water reactor RPV upper
head penetration nozzles and associated partial penetration welds. The
program required the implementation of ASME BPV Code Case N-729-1, with
certain conditions.
The application of ASME BPV Code Case N-729-1 was necessary because
the inspections required by the 2004 Edition of the ASME BPV Code,
Section XI were not written to address degradation of the RPV upper
head penetration nozzles and associated welds by primary water stress
corrosion cracking (PWSCC). The safety consequences of inadequate
inspections can be significant. The NRC's determination that the ASME
BPV Code required inspections are inadequate is based upon operating
experience and analysis. The absence of an effective inspection regime
could, over time, result in unacceptable circumferential cracking, or
the degradation of the RPV upper head or other reactor coolant system
components by leakage assisted corrosion. These degradation mechanisms
increase the probability of a loss-of-coolant accident.
Examination frequencies and methods for RPV upper head penetration
nozzles and welds are provided in ASME BPV Code Case N-729-1. The use
of code cases is voluntary, so these provisions were developed, in
part, with the expectation that the NRC would incorporate the code case
by reference into the CFR. Therefore, the NRC adopted rule language in
Sec. 50.55a(g)(6)(ii)(D) requiring implementation of ASME BPV Code
Case N-729-1, with conditions, in order to enhance the examination
requirements in the ASME BPV Code, Section XI for RPV upper head
penetration nozzles and welds. The examinations conducted in accordance
with ASME BPV Code Case N-729-1 provide reasonable assurance that ASME
BPV Code allowable limits will not be exceeded and that PWSCC will not
lead to failure of the RPV upper head penetration nozzles or welds.
However, the NRC concluded that certain conditions were needed in
implementing the examinations in ASME BPV Code Case N-729-1. These
conditions are set forth in Sec. 50.55a(g)(6)(ii)(D).
On June 22, 2012, the ASME approved the fourth revision of ASME BPV
Code Case N-729 (N-729-4). This revision changed certain requirements
based on a consensus review of inspection techniques and frequencies.
These changes were deemed necessary by the ASME to supersede the
previous requirements under N-729-1 to establish an effective long-term
inspection program for the RPV upper head penetration nozzles and
associated welds in pressurized water reactors. The
[[Page 32953]]
major changes included incorporation of previous NRC conditions in the
CFR. Minor changes were also made to address editorial issues, to
correct figures or to add clarity.
The NRC is updating the requirements of Sec. 50.55a(g)(6)(ii)(D)
to require licensees to implement ASME BPV Code Case N-729-4, with
conditions. One existing condition on ASME BPV Code Case N-729-1 has
been modified, four existing conditions are being deleted in this final
rule, one existing condition is being redesignated without substantive
change, and two new conditions--in Sec. 50.55a(g)(6)(ii)(D)(3) and
(4)--are adopted in this final rule in order to address the changes in
ASME BPV Code Case N-729-4. The NRC's revisions to the conditions are
discussed under the next three headings. As discussed earlier, this
final rule incorporates by reference ASME BPV Code Case N-729-4 into
Sec. 50.55a(a)(1)(iii)(C).
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
The NRC is revising Sec. 50.55a(g)(6)(ii)(D)(1) to change the
version of ASME BPV Code Case N-729 from N-729-1 to N-729-4 for the
reasons previously set forth. Due to the incorporation of N-729-4, the
date to establish applicability for licensed pressurized water reactors
will be changed to the effective date of this final rule.
10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6) (Removed)
The NRC is removing the existing conditions in Sec.
50.55a(g)(6)(ii)(D)(2) through (5) and redesignating the condition
currently in Sec. 50.55a(g)(6)(ii)(D)(6) as Sec.
50.55a(g)(6)(ii)(D)(2) without any substantive change. The existing
conditions in Sec. 50.55a(g)(6)(ii)(D)(2) through (5) have all been
incorporated either verbatim or more conservatively in the revisions to
ASME BPV Code Case N-729, up to version N-729-4. Therefore, there is no
reason to retain these conditions in Sec. 50.55a.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency (New
Condition)
The NRC is adopting a new condition in Sec. 50.55a(g)(6)(ii)(D)(3)
to modify the option in ASME BPV Code Case N-729-4 to extend bare metal
visual inspections of the RPV upper head surface beyond the frequency
listed in Table 1 of the Code Case. Previously, upper heads aged with
less than eight effective degradation years were considered to have a
low probability of initiating PWSCC, the cracking mechanism of concern.
This ranking of effective degradation years was based on a simple time
at temperature correlation. All of the upper heads within this
category, with the exception of new heads using Alloy 600 penetration
nozzles, were considered to have lower susceptibility to cracking due
to the upper heads being at or near the cold leg operating temperature
of the reactor coolant system. Therefore, these plants were referred to
as having ``cold heads.'' All of the upper heads that had experienced
cracking prior to 2006 were near the hot leg operating temperature of
the reactor coolant system, which validated the time at temperature
model.
In 2006, one of the 21 ``cold head'' plants identified two
indications within a penetration nozzle and the associated partial
penetration weld. Then, between 2006 and 2013, five of the 21 ``cold
head'' plants identified multiple indications within fifteen different
penetration nozzles and the associated partial penetration welds. None
of these indications caused leakage, and volumetric examination of the
penetration nozzles showed that no flaws in the nozzle material had
grown through-wall; however, this increasing trend creates a reasonable
safety concern.
Recent operational experience has shown that the volumetric
inspection of penetration nozzles, at the current inspection frequency,
is adequate to identify indications in the nozzle material prior to
leakage; however, volumetric examinations cannot be performed on the
partial penetration welds. Therefore, given the additional cracking
identified at cold leg temperatures, the NRC staff has concerns about
the adequacy of the partial penetration weld examinations.
Leakage from a partial penetration weld into the annulus between
the nozzle and head material can cause corrosion of the low alloy steel
head. While initially limited in leak rate, due to limited surface area
of the weld being in contact with the annulus region, corrosion of the
vessel head material can expose more of the weld surface to the
annulus, allowing a greater leak rate. Since an indication in the weld
cannot be identified by a volumetric inspection, a postulated crack
through the weld, just about to cause leakage, could exist as a plant
performed its last volumetric and/or bare metal visual examination of
the upper head material. This gives the crack years to breach the
surface and leak prior to the next scheduled visual examination.
Only a surface examination of the wetted surface of the partial
penetration weld can reliably detect flaws in the weld. Unfortunately,
this examination cannot size the flaws in the weld, and, if performed
manually, requires significant radiological dose to examine all of the
partial penetration welds on the upper head. As such, the available
techniques are only able to detect a flaw after it has caused leakage.
These techniques are a bare metal visual examination or a volumetric
leak path assessment performed on the frequency of the volumetric
examination.
Volumetric leak path examinations are only done during outages when
a volumetric examination of the nozzle is performed. Therefore, under
the current requirements allowed by Note 4 of ASME BPV Code Case N-729-
4, leakage from a crack in the weld of a ``cold head'' plant could
start and continue to grow for the 5 years between the required bare
metal visual examinations to detect leakage through the partial
penetration weld.
Given the additional cracking identified at cold leg temperatures
of upper head penetration nozzles and associated welds, the NRC finds
limited basis to continue to categorize these ``cold head'' plants as
having a low susceptibility to crack initiation. The NRC is increasing
the frequency of the bare metal visual examinations of ``cold heads''
to identify potential leakage as soon as reasonably possible due to the
volumetric examination limitations. Therefore, the NRC is conditioning
Note 4 of ASME BPV Code Case N-729-4 to require a bare metal visual
exam during each outage in which a volumetric exam is not performed.
The NRC also will allow ``cold head'' plants to extend their bare metal
visual inspection frequency from once each refueling outage, as stated
in Table 1 of N-729-1, to once every 5 years, but only if the licensee
performed a wetted surface examination of all of the partial
penetration welds during the previous volumetric examination. Applying
the conditioned bare metal visual inspection frequency or a volumetric
examination each outage will allow licensees to identify any potential
leakage through the partial penetration welds prior to significant
degradation of the low alloy steel head material, thereby providing
reasonable assurance of the structural integrity of the reactor coolant
pressure boundary.
These issues, including the operational experience, the fact that
volumetric examination is not available to interrogate the partial
penetration welds, and potential regulatory options, were discussed
publicly at multiple ASME BPV Code meetings, at the annual Materials
Programs Technical Information Exchange public meeting
[[Page 32954]]
held at the NRC Headquarters in June 2013, and at the 2013 NRC
Regulatory Information Conference.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria (New
Condition)
The NRC is adopting a new condition in Sec. 50.55a(g)(6)(ii)(D)(4)
to define surface examination acceptance criteria. Paragraph -3132(b)
of ASME BPV Code Case N-729-4 sets forth the acceptance criteria for
surface examinations. In general, throughout Section XI of the ASME BPV
Code, the acceptance criteria for surface examinations default to
Section III, Paragraph NB-5352, ``Acceptance Standards.'' Typically,
for rounded indications, the indication was only unacceptable if it was
greater than \3/16\-inch in size. The NRC requested that the code case
authors include a requirement that any size rounded indication causing
nozzle leakage is unacceptable due to operating experience identifying
PWSCC under rounded indications less than \3/16\-inch in size.
Recently, the ASME BPV Code Committee approved an interpretation of
the language in Paragraph -3132(b), which implied that any size rounded
indication is acceptable unless there is relevant indication of nozzle
leakage, even those greater than \3/16\-inch. The NRC does not agree
with the interpretation and maintains its original position on rounded
indications that any size rounded indication is unacceptable if there
is an indication of leakage. Since the adoption of ASME BPV Code Case
N-729-1 into Sec. 50.55a(g)(6)(ii)(D), all licensees have used the
NRC's position in implementing Paragraph -3132(b), even after the
recent ASME BPV Code Committee interpretation approval over NRC
objection.
Therefore, in order to ensure compliance with the previous and
ongoing requirement, the NRC is revising condition Sec.
50.55a(g)(6)(ii)(D)(4) to include clarity within the acceptance
criteria for surface examinations. The current edition requirements of
NB-5352 of ASME BPV Code, Section III for the licensee's ongoing 10-
year inservice inspection interval shall be met.
ASME BPV Code Case N-770-2
On June 21, 2011 (76 FR 36232), the NRC issued a final rule, which
included Sec. 50.55a(g)(6)(ii)(F) that requires the implementation of
ASME BPV Code Case N-770-1, ``Alternative Examination Requirements and
Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt
Welds Fabricated with UNS N06082 or UNS N86182 Weld Filler Material
With or Without Application of Listed Mitigation Activities,'' with
certain conditions.
On June 9, 2011, the ASME approved the second revision of ASME BPV
Code Case N-770 (N-770-2). The major changes from N-770-1 to N-770-2
included establishing new ASME BPV Code Case, Table 1, inspection item
classifications for optimized weld overlays and allowing alternatives
when complete inspection coverage cannot be met. Minor changes were
also made to address editorial issues, to correct figures, or to add
clarity. The NRC found that the updates and improvements in N-770-2 are
sufficient to update Sec. 50.55a(g)(6)(ii)(F).
The NRC, therefore, is updating the requirements of Sec.
50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV Code
Case N-770-2, with conditions. The NRC conditions have been modified to
address the changes in ASME BPV Code Case N-770-2 and to ensure that
this regulatory framework will provide adequate protection of public
health and safety. The following sections discuss each of the NRC's
changes to the conditions on ASME BPV Code Case N-770-2. As discussed
earlier, this final rule incorporates by reference ASME BPV Code Case
N-770-2 into Sec. 50.55a(a)(1)(iii)(D).
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(1) to change the
version of ASME BPV Code Case N-770 from N-770-1 to N-770-2 and to
require its implementation, with conditions, to incorporate the updates
and improvements contained in N-770-2. The NRC will allow licensees to
begin using N-770-2 on the effective date of this rule.
10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(2) to provide
clarification regarding categorization of each Alloy 82/182 butt weld,
mitigated or not, under N-770-2. This paragraph also clarifies the
NRC's position that Paragraph -1100(e) shall not be used to exempt
welds that rely on Alloy 82/182 for structural integrity from more
frequent ISI schedules until the NRC has reviewed and authorized an
alternative categorization for the weld. Additionally, the NRC will
change the inspection item categories for full structural weld overlays
from C to C-1 and F to F-1 due to reclassification under ASME BPV Code
Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(3) to clarify the
baseline examination requirements by stating that previously-conducted
examinations, in order to count as baseline examinations, must meet the
requirements of ASME BPV Code Case N-770-2, as conditioned in this
section. The 2011 rule required the use of ASME BPV Code Section XI
Appendix VIII qualifications for baseline examinations, which is
stricter than N-770-2 and does not provide requirements for optimized
weld overlays. The revision also updates the deadline for baseline
examination requirements, since the January 20, 2012, deadline from the
previous rule has passed. Finally, upon implementation of this rule, if
a licensee is currently in an outage, then the baseline inspection
requirement can be met by performing the inspections in accordance with
the previous regulatory requirements of Sec. 50.55a(g)(6)(ii)(F), in
lieu of the examination requirements of Paragraphs -2500(a) or -2500(b)
of ASME BPV Code Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(4) to define
examination coverage for circumferential flaws and to prohibit the use
of Paragraph -2500(d) of ASME BPV Code Case N-770-2 which, in some
circumstances, allows unacceptably low examination coverage. Paragraph
-2500(d) of N-770-2 would allow the reduction of circumferential
volumetric examination coverage with analytical evaluation. Paragraph -
2500(c) was previously prohibited from use, and it continues to be
prohibited. The NRC is establishing an essentially 100 percent
volumetric examination coverage requirement, including greater than 90
percent of the required volumetric examination coverage, for
circumferential flaws to provide reasonable assurance of structural
integrity of all ASME BPV Code Class 1 butt welds susceptible to PWSCC.
Therefore, the NRC is adopting a condition prohibiting the use of
Paragraphs -2500(c) and -2500(d). A licensee may request approval for
use of these paragraphs under 10 CFR 50.55a(z).
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(5) to add the
explanatory heading, ``Inlay/onlay
[[Page 32955]]
inspection frequency,'' and to make minor editorial corrections.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(6) to add the
explanatory heading, ``Reporting requirements.''
10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(7) to add the
explanatory heading, ``Defining `t'.''
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(8) to add the
explanatory heading, ``Optimized weld overlay examination,'' and to
maintain the requirement for the timing of the initial inservice
examination of optimized weld overlays.
Uncracked welds mitigated with optimized weld overlays were re-
categorized by ASME BPV Code Case N-770-2 from Inspection Item D to
Inspection Item C-2; however, the initial inspection requirement was
not incorporated into the Code Case for Inspection Item C-2. The NRC
has determined that uncracked welds mitigated with an optimized weld
overlay must have an initial inservice examination no sooner than the
third refueling outage and no later than 10 years following the
application of the weld overlay to identify unacceptable crack growth.
Optimized weld overlays establish compressive stress on the inner half
thickness of the weld, but the outer half thickness may also be under
tensile stress. The requirement for an initial inservice examination no
sooner than the third refueling outage and no later than 10 years
following the application of the weld overlay is based on the design of
optimized weld overlays, which require the outer quarter thickness of
the susceptible material to provide structural integrity for the weld.
Therefore, the NRC is continuing adoption of the condition, which
requires the initial inservice examination of uncracked welds mitigated
by optimized weld overlay (i.e., the welds which are subject to
Inspection Item C-2 of ASME BPV Code Case N-770-2) within the specified
timeframe.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(9) to add the
explanatory heading, ``Deferral,'' and to address changes in ASME BPV
Code Case N-770-2 which allow the deferral of the first inservice
examination of uncracked welds mitigated with optimized weld overlays,
Inspection Item C-2.
Previously, under N-770-1, the initial inservice examination of
these welds was not allowed to be deferred. Allowing deferral of the
initial inservice examination in accordance with N-770-2 could, in
certain circumstances, allow the initial inservice examination to be
performed up to 20 years after installation. Therefore, the NRC is
adopting a condition which would preclude the deferral of the initial
inservice examination of uncracked welds mitigated by optimized weld
overlays.
10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(10) to add the
explanatory heading, ``Examination technique,'' and to address changes
in ASME BPV Code Case N-770-2. Note 14(a) of Table 1 of ASME BPV Code
Case N-770-2 provides the previously required full examination
requirement for optimized weld overlays. The language of ASME BPV Code
Case N-770-2, however, does not require the implementation of the full
examination requirements of Note 14(a) of Table 1, if possible, before
implementing the reduced examination coverage requirements of Note
14(b) of Table 1 or Note (b) of Figure 5(a). The NRC agrees that
reduced examination coverage is the best alternative if the full
examination cannot be met; however, the full examination requirement
should be implemented, if possible, before the option of reduced
examination coverage is allowed. Therefore, the NRC is modifying the
current condition in Sec. 50.55a(g)(6)(ii)(F)(10) to allow the use of
Note 14(b) of Table 1 and Note (b) of Figure 5(a) of ASME BPV Code Case
N-770-2 only after the determination that the requirements of Note
14(a) of Table 1 of ASME BPV Code Case N-770-2 cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(11) to address
examination requirements through cast stainless steel materials by
requiring the use of Appendix VIII qualifications to meet the
inspection requirements of Paragraph -2500(a) of ASME BPV Code Case N-
770-2. The requirements for volumetric examination of butt welds
through cast stainless steel materials are currently being developed as
Supplement 9 to the ASME BPV Code, Section XI, Appendix VIII. In
accordance with Appendix VIII for supplements that have not been
developed, the requirements of Appendix III apply. Appendix III
requirements are not equivalent to Appendix VIII requirements. For the
volumetric examination of ASME BPV Code Class 1 welds, the NRC has
established the requirement for examination qualification under the
Appendix VIII. Therefore, the NRC is adopting a condition requiring the
use of Appendix VIII qualifications to meet the inspection requirements
of Paragraph -2500(a) of ASME BPV Code Case N-770-2 by January 1, 2022.
The development of a sufficient number of mockups would be required
to establish an Appendix VIII program for examination of ASME BPV Code
Class 1 piping and vessel nozzle butt welds through cast stainless
steel materials. The NRC recognizes that significant time and resources
are required to create mockups and to allow for qualification of
equipment, procedures and personnel. Therefore, the NRC is requiring
licensees to use these Appendix VIII qualifications no later than their
first scheduled weld examinations involving cast stainless steel
materials occurring after January 1, 2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(12) to clarify the
examination coverage requirements allowed under Appendix I of ASME BPV
Code Case N-770-2 for butt welds joining cast stainless steel material.
Under current ASME BPV Code, Section XI, Appendix VIII requirements,
the volumetric examination of butt welds through cast stainless steel
materials is under Supplement 9. Supplement 9 rules are still being
developed by the ASME BPV Code. Therefore, it is currently impossible
to meet the requirement of Paragraph I.5.1 for butt welds joining cast
stainless steel material.
The material of concern is the weld material susceptible to PWSCC
adjoining the cast stainless steel material. Appendix VIII qualified
procedures are available to perform the inspection of the susceptible
weld material, but they are not qualified to inspect the cast stainless
steel materials. Therefore, the NRC is adopting a condition changing
the inspection volume for stress-improved dissimilar metal welds with
cast stainless steel from the ASME BPV Code Section XI requirements to
``the maximum extent practical including 100 percent of the susceptible
material volume.'' This will
[[Page 32956]]
remain applicable until an Appendix VIII qualified procedure for the
inspection through cast stainless steel materials is available in
accordance with the new condition in Sec. 50.55a(g)(6)(ii)(F)(11).
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(13) to require the
encoding of ultrasonic volumetric examinations of Inspection Items A-1,
A-2, B, E, F-2, J, and K in Table 1 of N-770-2. A human performance gap
has been found between some ultrasonic testing procedures, as
demonstrated during ASME BPV Code, Section XI, Appendix VIII
qualification versus as applied in the field.
The human factors that contributed to the licensee-performed
examinations which failed to identify significant flaws at North Anna
Power Station, Unit 1 in 2012 (Licensee Event Report 50-338/2012-001-
00) and at Diablo Canyon Nuclear Power Plant in 2013 (Relief Request
REP-1 U2, Revision 2) can be avoided by the use of encoded ultrasonic
examinations. Encoded ultrasonic examinations electronically store both
the positional and ultrasonic information from the inspections. Encoded
examinations allow for the inspector to evaluate the data and search
for indications outside of a time limited environment to assure that
the inspection was conducted properly and to allow for sufficient time
to analyze the data. Additionally, the encoded examination would allow
for an independent review of the data by other inspectors or an
independent third party. Finally, the encoded examination could be
compared to previous and/or future encoded examinations to determine if
flaws are present and flaw growth rates. Therefore, the NRC is adopting
a condition requiring the use of encoding for ultrasonic volumetric
examinations of non-mitigated or cracked mitigated dissimilar metal
butt welds in the reactor coolant pressure boundary which are within
the scope of ASME BPV Code Case N-770-2.
ASME BPV Code Case N-824
10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii) to allow licensees to
use the provisions of ASME BPV Code Case N-824, ``Ultrasonic
Examination of Cast Austenitic Piping Welds From the Outside Surface
Section XI, Division 1,'' subject to four conditions in Sec.
50.55a(b)(2)(xxxvii)(A) through (D), when implementing inservice
examinations in accordance with the ASME BPV Code, Section XI
requirements.
During the construction of nuclear power plants, it was recognized
that the grain structure of cast austenitic stainless steel (CASS)
could prevent effective ultrasonic inspections of piping welds where
one or both sides of the welds were constructed of CASS. The high
strength and toughness of CASS (prior to thermal embrittlement) made it
desirable as a building material despite this known inspection issue.
This choice of construction materials has rendered many pressure
boundary components without a means to reliably inspect them
volumetrically. While there is no operational experience of a CASS
component failing, as part of the reactor pressure boundary, inservice
volumetric inspection of these components is necessary to provide
reasonable assurance of their structural integrity.
The current regulatory requirements for the examination of CASS,
provided in Sec. 50.55a, do not provide sufficient guidance to assure
that the CASS components are being inspected adequately. To illustrate
that ASME BPV Code does not provide adequate guidance, ASME BPV Code,
Section XI, Appendix III, Supplement 1 states, ``Cast materials may
preclude meaningful examinations because of geometry and attenuation
variables.'' For this reason, over the past several decades, licensees
have been unable to perform effective inspections of welds joining CASS
components. To allow for continued operation of their plants, licensees
submitted hundreds of requests for relief from the ASME BPV Code
requirements for inservice inspection of CASS components to the NRC,
resulting in a significant regulatory burden.
The recent advances in inspection technology are driving renewed
work at ASME BPV Code meetings to produce Section XI, Appendix VIII,
Supplement 9 to resolve the CASS inspection issue, but it will be years
before these code updates will be published, as well as additional time
to qualify and approve procedures for use in the field. Until then,
licensees would still use the requirements of ASME BPV Code Section XI,
Appendix III, Supplement 1, which states that inspection of CASS
materials meeting the ASME BPV Code requirements may not be meaningful.
Consequently, less effective examinations would continue to be used in
the field, and more relief requests would be generated between now and
the implementation of Supplement 9.
The NRC commissioned a research program to determine the
effectiveness of the new technologies for inspections of CASS
components in an effort to resolve some of the known inspection issues.
The result of this work is published in NUREG/CR-6933, ``Assessment of
Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using
Advanced Low-Frequency Ultrasonic Methods'', March 2007, and NUREG/CR-
7122, ``An Evaluation of Ultrasonic Phased Array Testing for Cast
Austenitic Stainless Steel Pressurizer Surge Line Piping Welds,'' March
2012. Based on the improvements in ultrasonic inspection technology and
techniques for CASS components, the ASME approved BPV Code Case N-824
(N-824) on October 16, 2012, which describes how to develop a procedure
capable of meaningfully inspecting welds in CASS components.
Effective examinations of CASS components require the use of lower
frequencies and larger transducers than are typically used for
ultrasonic inspections of piping welds and would require licensees to
modify their inspection procedures. The NRC recognizes that requiring
the use of spatial encoding will limit the full implementation of ASME
BPV Code Case N-824, as spatial encoding is not practical for many weld
configurations.
At this time, the use of ASME BPV Code Case N-824, as conditioned,
is the most effective known method for adequately examining welds with
one or more CASS components. With the use of ASME BPV Code Case N-824,
as conditioned, licensees will be able to take full credit for
completion of the Sec. 50.55a required inservice volumetric inspection
of welds involving CASS components. The implementation of ASME BPV Code
Case N-824, as conditioned, will have the dual effect of improving the
rigor of required volumetric inspections and reducing the number of
uninspectable Class 1 and Class 2 pressure retaining welds.
The NRC concludes that incorporation of ASME BPV Code Case N-824,
subject to the four conditions in Sec. 50.55a(b)(2)(xxxvii)(A) through
(D), will significantly improve the flaw detection capability of
ultrasonic inspection of CASS components until Supplement 9 is
implemented, thereby providing reasonable assurance of leak tightness
and structural integrity. Additionally, it will reduce the regulatory
burden on licensees and allow licensees to submit fewer relief requests
for welds in CASS materials. The four conditions on the use of ASME BPV
Code Case N-824,
[[Page 32957]]
Sec. 50.55a(b)(2)(xxxvii)(A) through (D), are discussed in the next
four headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) (First Condition on Use of ASME BPV Code
Case N-824)
The NRC, based upon NUREG/CR-6933 and NUREG/CR-7122, has determined
that inspections of CASS materials are very challenging, and sufficient
technical basis exists to condition the code case to bring the code
case into agreement with the NUREG/CR reports. The NUREG/CR reports
also show that CASS materials produce high levels of coherent noise.
The noise signals can be confusing and mask flaw indications. Use of
encoded inspection data allows the inspector to mitigate this problem
through the ability to electronically manipulate the data, which allows
for discrimination between coherent noise and flaw indications. The NRC
found that encoding CASS inspection data provides significant detection
benefits. Therefore, the NRC is adding a condition in Sec.
50.55a(b)(2)(xxxvii)(A) to require the use of encoded data when
utilizing N-824 for the examination of CASS components.
10 CFR 50.55a(b)(2)(xxxvii)(B) (Second Condition on Use of ASME BPV
Code Case N-824)
The use of dual element phased-array search units showed the most
promise in obtaining meaningful responses from flaws. For this reason,
the NRC is adding a condition in Sec. 50.55a(b)(2)(xxxvii)(B) to
require the use of dual, transmit-receive, refracted longitudinal wave,
multi-element phased array search units when utilizing N-824 for the
examination of CASS components.
10 CFR 50.55a(b)(2)(xxxvii)(C) (Third Condition on Use of ASME BPV Code
Case N-824)
The optimum inspection frequencies for examining CASS components of
various thicknesses are described in NUREG/CR-6933 and NUREG/CR-7122.
For this reason, the NRC is adding a condition in Sec.
50.55a(b)(2)(xxxvii)(C) to require that ultrasonic examinations
performed to implement ASME BPV Code Case N-824 on piping greater than
1.6 inches (41 mm) thick shall use a phased array search unit with a
center frequency of 500 kHz with a tolerance of + /- 20 percent.
10 CFR 50.55a(b)(2)(xxxvii)(D) (Fourth Condition on Use of ASME BPV
Code Case N-824)
NUREG/CR-6933 shows that the grain structure of CASS can reduce the
effectiveness of some inspection angles. For this reason, the NRC is
adding a condition in Sec. 50.55a(b)(2)(xxxvii)(D) to require that
ultrasonic examinations performed to implement ASME BPV Code Case N-824
shall use a phased array search unit which produces angles including,
but not limited to, 30 to 55 degrees with a maximum increment of 5
degrees.
OM Code Case OMN-20
10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20
The NRC is adding Sec. 50.55a(b)(3)(x) to allow licensees to
implement OM Code Case OMN-20, ``Inservice Test Frequency,'' in the OM
Code, 2012 Edition, for the editions and addenda of the OM Code that
are listed in Sec. 50.55a(a)(1)(iv) as being approved for
incorporation by reference. As a conforming change, Sec.
50.55a(a)(1)(iii)(G) is being added to incorporate by reference OM Code
Case OMN-20 into Sec. 50.55a.
Surveillance Requirement (SR) 3.0.3 from TS 5.5.6, ``Inservice
Testing Program,'' allows licensees to apply a delay period before
declaring the SR for TS equipment ``not met'' when the licensee
inadvertently exceeds or misses the time limit for performing TS
surveillance. Licensees have been applying SR 3.0.3 to inservice tests.
The NRC has determined that licensees cannot use TS 5.5.6 to apply SR
3.0.3 to inservice tests under Sec. 50.55a(f) that are not associated
with a TS surveillance. To invoke SR 3.0.3, the licensee shall first
discover that a TS surveillance was not performed at its specified
frequency. Therefore, the delay period that SR 3.0.3 provides does not
apply to non-TS support components tested under Sec. 50.55a(f). The OM
Code does not provide for any inservice test frequency reductions or
extensions. In order to provide inservice test frequency reductions or
extensions that can no longer be provided by SR 3.0.3 from TS 5.5.6,
the ASME has developed OM Code Case OMN-20. The NRC has reviewed OM
Code Case OMN-20 and has found it acceptable for use. The NRC
determined that OM Code Case OMN-20 may be safely used for all
licensees using editions and addenda of the OM Code that are listed in
Sec. 50.55a(a)(1)(iv). The NRC will include OM Code Case OMN-20 in the
next revision of RG 1.192, at which time a conforming change will be
made to delete both this paragraph and Sec. 50.55a(a)(1)(iii)(G).
III. Opportunities for Public Participation
The proposed rule was published on September 18, 2015, for a 75-day
comment period (80 FR 56820). The public comment period closed on
December 2, 2015.
After the close of the public comment period, the NRC held a public
meeting on March 2, 2016, to discuss the proposed rule, to answer
questions on specific provisions of the proposed rule, and to discuss
public comments received on the proposed rule in order to enhance the
NRC's understanding of the comments. The public meeting summary is
available in ADAMS under Accession No. ML16069A408.
IV. NRC Responses to Public Comments
The NRC received 27 letters and emails in response to the
opportunity for public comment on the proposed rule. These comment
submissions were submitted by the following commenters (listed in order
of receipt):
1. Private citizen, Edward Cavey
2. Private citizen, Dale Matthews
3. Private citizen, Ron Clow
4. ASME
5. Iddeal Solutions, LLC
6. Electric Power Research Institute (EPRI)
7. Private citizen, William Taylor
8. ASME
9. Private citizen, Dan Nowakowski
10. Wolf Creek Nuclear Operating Corporation
11. Northern States Power Company--Minnesota
12. FirstEnergy Nuclear Operating Company
13. PSEG Nuclear
14. Dominion Resources Services, Inc.
15. Private citizen, Terence Chan
16. Nuclear Energy Institute
17. EPRI
18. Duke Energy
19. Private Citizen, William Taylor
20. Dominion Engineering, Inc.
21. Tennessee Valley Authority
22. Southern Nuclear Operating Company
23. Prairie Island Nuclear Plant
24. Inservice Test Owners Group
25. Exelon Generation Company
26. EPRI
27. EPRI
In general, the comments:
Suggested revising or rewording conditions to make them
clearer.
Supported incorporation of Code Cases N-729-4, N-770-2, N-
824, or OMN-20 into Sec. 50.55a.
Supported the proposed changes to add or remove
conditions.
Opposed proposed conditions.
Supplied additional information for NRC consideration.
Proposed rewriting or renumbering of paragraphs.
[[Page 32958]]
Asked questions or requested information from the NRC.
Due to the large number of comments received and the length of the
NRC's responses, this document summarizes the NRC's response to
comments in areas of particular interest to stakeholders that prompted
the NRC to make changes in this final rule from what was proposed. A
discussion of all comments and complete NRC responses are presented in
a separate document, ``2017 Final Rule (10 CFR 50.55a) American Society
of Mechanical Engineers Codes and Code Cases: Analysis of Public
Comments,'' (ADAMS Accession No. ML16130A531).
10 CFR 50.55a(a)(1)(ii), (b)(2); Nonmandatory Appendix U
Public commenters were concerned that the NRC was proposing to
exclude incorporating by reference Nonmandatory Appendix U because
Nonmandatory Appendix U is the incorporation of the provisions of ASME
BPV Code Cases N-513-3 and N-705, without any technical changes, into
the Section XI Code. The NRC agrees with this comment, in that ASME BPV
Code Cases N-513-3 and N-705 have been approved in RG 1.147. Based on
these comments, the NRC has removed the proposed exclusion of
Nonmandatory Appendix U from this final rule. However, the NRC has
found it necessary to apply two new conditions in Sec.
50.55a(b)(2)(xxxiv)(A) and (B) to Nonmandatory Appendix U. The first
condition provides regulatory consistency with the approval of the code
cases in RG 1.147. The second condition requires the use of an Appendix
from ASME BPV Code Case N-513-3 that was unintentionally omitted from
Appendix U. The NRC discussed these changes at the March 2, 2016,
public meeting, and the NRC considered the public feedback from that
meeting when developing this final rule.
10 CFR 50.55a(b)(2)(xii), Underwater Welding
Public commenters were concerned that the proposed rule continued
to prohibit the use of underwater welding in Sec. 50.55a(b)(2)(xii),
when changes were made to address this condition in the 2010 Edition of
Section XI. The NRC agrees that the condition should be modified to
address the changes in the Code. After consideration of the public
comments, the NRC noted other inconsistencies for addressing welding on
irradiated materials that appear in the Code and in some Code Cases.
Section 50.55a(b)(2)(xii) of this final rule reflects a change to
include two conditions that provide consistency for welding of
irradiated materials. The NRC discussed these changes at the March 2,
2016, public meeting, and the NRC considered the public feedback from
that meeting when developing this final rule.
10 CFR 50.55a(b)(2)(xxxi), Mechanical Clamping Devices
Public commenters were concerned that the wording of the proposed
condition in Sec. 50.55a(b)(2)(xxxi) was unclear and that citing the
specific paragraphs of Section XI to which the NRC is taking exception
would be clearer. The NRC agrees. To clarify the requirement for the
implementation of mechanical clamps, the condition was changed to
require the use of Appendix W of Section XI when using mechanical
clamps. Additionally, use of IWA-4131.1(c) of the 2010 Edition of
Section XI and IWA-4131.1(d) of the 2011 Addenda of the 2010 Edition
and later versions of Section XI is prohibited. Identifying these
specific subparagraphs was deemed necessary, as they may have caused
confusion with the intended purpose of the original proposed condition
in maintaining the previous regulatory requirements for mechanical
clamping devices. Section 50.55a(b)(2)(xxxi) of this final rule
reflects this change.
10 CFR 50.55a(b)(2)(xxxvii), ASME BPV Code Case N-824
Public commenters had concerns with conditions proposed on ASME BPV
Code Case N-824, ``Ultrasonic Examination of Cast Austenitic Piping
Welds From the Outside Surface Section XI, Division 1,'' in Sec.
50.55a(b)(2)(xxxvii)(A) through (E). There were concerns that the
conditions would limit the use of Code Case N-824 and that some
conditions did not have a sufficient technical basis. The NRC partially
agreed with the comments requesting the removal and modification of
some conditions in Sec. 50.55a(b)(2)(xxxvii) restricting the
frequencies and angles usable on some cast austenitic welds. Based on
the public comments, one condition was removed entirely and two others
were modified. Section 50.55a(b)(2)(xxxvii)(A) through (D) of this
final rule contain the modified and reduced conditions on the use of
ASME BPV Code Case N-824. The NRC discussed these changes at the March
2, 2016, public meeting, and the NRC considered the public feedback
from that meeting when developing this final rule.
10 CFR 50.55a(b)(3)(xi), OM Condition: Valve Position Indication
Public commenters raised concerns regarding the proposed condition
in Sec. 50.55a(b)(3)(xi) to emphasize the OM Code provisions in
Subsection ISTC-3700, ``Position Verification Testing,'' to verify that
valve operation is accurately indicated. Public commenters indicated
that because of the significance of implementing the condition, some
licensees might need time to revise or create procedures to govern the
implementation of this condition. Public commenters also suggested that
the condition be limited to active valves. The NRC partially agrees and
partially disagrees with these comments. The NRC agrees that additional
time to implement the condition regarding valve position verification
is appropriate. Therefore, the NRC has revised the condition to
indicate that it will be effective with implementation of the 2012
Edition of the OM Code. The NRC staff does not agree with the
suggestion to limit the condition to active valves because the OM Code
requires that passive valves undergo periodic verification of position
indication.
V. Section-by-Section Analysis
Administrative Changes
The NRC is removing the revision number of the three RGs currently
approved by the Office of the Federal Register for incorporation by
reference throughout the substantive provisions of Sec. 50.55a
addressing the ASME Code Cases, i.e., paragraphs (b) through (g). The
revision numbers for the RGs approved for incorporation by reference
(currently, RG 1.84, RG 1.147, and RG 1.192) will be retained in Sec.
50.55a(a)(3)(i) through (iii), where the RGs are listed by full title,
including revision number. That paragraph identifies the specific
materials which the Office of the Federal Register has approved for
incorporation by reference, as required by Office of the Federal
Register requirements in 1 CFR 51.9. Readers would need to refer to
Sec. 50.55a(a) to determine the specific revision of the relevant RG
that is approved for incorporation by reference by the Office of the
Federal Register. These changes are administrative in nature and do not
change substantive requirements with respect to the RGs and the Code
Cases listed in the RGs.
10 CFR 50.55a(a) Documents Approved for Incorporation by Reference
The NRC is revising the incorporation by reference language to
update the
[[Page 32959]]
contact information for the NRC Technical Library.
10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section
III
The NRC is revising Sec. 50.55a(a)(1)(i) to clarify that Section
III Nonmandatory Appendices of the listed editions and addenda are
excluded from the incorporation by reference. The exclusion was
originally added in a final rule published on June 21, 2011 (76 FR
36232); however, it was erroneously omitted from the final rule
published on November 5, 2014 (79 FR 65776). The NRC is correcting the
omission in this final rule by inserting ``(excluding Nonmandatory
Appendices)'' in Sec. 50.55a(a)(1)(i). The NRC is relocating the
definition of the term ``BPV Code,'' which is used throughout the
section, from Sec. 50.55a(b) to Sec. 50.55a(a)(1)(i).
10 CFR 50.55a(a)(1)(i)(E) ``Rules for Construction of Nuclear Facility
Components--Division 1''
The NRC is revising Sec. 50.55a(a)(1)(i)(E) to add ASME BPV Code,
Section III 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013 Edition.
10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section
XI
The NRC is revising Sec. 50.55a(a)(1)(ii) to include two minor
editorial changes: to replace ``Boiler and Pressure Vessel Code'' with
``BPV Code'' and to replace ``limited to'' with ``limited by.''
10 CFR 50.55a(a)(1)(ii)(C)(52) and (53) ``Rules for Inservice
Inspection of Nuclear Power Plant Components--Division 1''
The NRC is revising Sec. 50.55a(a)(1)(ii)(C)(52) and (53) to add
ASME BPV Code, Section XI 2009 Addenda, 2010 Edition, 2011 Addenda, and
2013 Edition. The examination requirements for Examination Category B-
F, Item Numbers B5.11 and B5.71, Nozzle-to-Component Butt Welds in the
2011 Addenda and the 2013 Edition of ASME BPV Code, Section XI are
expressly excluded from the incorporation by reference in Sec.
50.55a(a)(1)(ii)(C)(52) and, therefore, not approved for use.
Similarly, the requirements of IWB-3112(a)(3) and IWC-3112(a)(3) in the
2013 Edition of ASME BPV Code, Section XI are expressly excluded from
the incorporation by reference in Sec. 50.55a(a)(1)(ii)(C)(53) and are
not approved for use.
10 CFR 50.55a(a)(1)(iii)(A) ASME BPV Code Case N-513-3 Mandatory
Appendix I
The NRC is revising Sec. 50.55a(a)(1)(iii)(A) to include
information for a new standard that is being incorporated by reference,
entitled, ``ASME BPV Code Case N-513-3 Mandatory Appendix I.''
10 CFR 50.55a(a)(1)(iii)(B) ASME BPV Code Case N-722-1
The NRC is revising Sec. 50.55a(a)(1)(iii)(B) to maintain
alphanumeric order for the ASME Code Cases listed in Sec.
50.55a(a)(1)(iii). ASME BPV Code Case N-722-1 was previously approved
for incorporation by reference.
10 CFR 50.55a(a)(1)(iii)(C) ASME BPV Code Case N-729-4
The NRC is revising Sec. 50.55a(a)(1)(iii)(C) to add the title
``ASME BPV Code Case N-729-4,'' and include information for the
standard that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(D) ASME BPV Code Case N-770-2
The NRC is adding Sec. 50.55a(a)(1)(iii)(D) to add the title
``ASME BPV Code Case N-770-2,'' and include information for the
standard that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(E) ASME BPV Code Case N-824
The NRC is adding Sec. 50.55a(a)(1)(iii)(E) to include information
for a new standard that is being incorporated by reference, entitled,
``ASME BPV Code Case N-824.''
10 CFR 50.55a(a)(1)(iii)(F) ASME BPV Code Case N-852
The NRC is adding Sec. 50.55a(a)(1)(iii)(F) to include information
for a new standard that is being incorporated by reference, entitled,
``ASME BPV Code Case N-852.''
10 CFR 50.55a(a)(1)(iii)(G) ASME OM Code Case OMN-20
The NRC is adding Sec. 50.55a(a)(1)(iii)(G) to include information
for a new standard that is being incorporated by reference, entitled,
``ASME OM Code Case OMN-20.''
10 CFR 50.55a(a)(1)(iv) ASME Operation and Maintenance Code
The NRC is revising Sec. 50.55a(a)(1)(iv) to correct the title of
the OM Code and to relocate the definition of the term ``OM Code,''
which is used throughout the section, from Sec. 50.55a(b) to Sec.
50.55a(a)(1)(iv).
10 CFR 50.55a(a)(1)(iv)(B) ``Operation and Maintenance of Nuclear Power
Plants, Division 1: Section IST Rules for Inservice Testing of Light-
Water Reactor Power Plants''
The NRC is adding new Sec. 50.55a(a)(1)(iv)(B) to include ASME OM
Code 2009 Edition and 2011 Addenda.
10 CFR 50.55a(a)(1)(iv)(C) ``Operation and Maintenance of Nuclear Power
Plants, Division 1: OM Code: Section IST''
The NRC is adding new Sec. 50.55a(a)(1)(iv)(C) to include ASME OM
Code 2012 Edition.
10 CFR 50.55a(a)(1)(v) ASME Quality Assurance Requirements
The NRC is adding new Sec. 50.55a(a)(1)(v) to include information
regarding NQA-1 standards and add the title ``ASME Quality Assurance
Requirements'' for ASME NQA-1 Code as part of NRC titling convention.
10 CFR 50.55a(b) Use and Conditions on the Use of Standards
The NRC is revising Sec. 50.55a(b) to correct the title of the
ASME OM Code.
10 CFR 50.55a(b)(1) Conditions on ASME BPV Code Section III
The NRC is revising Sec. 50.55a(b)(1) to reflect the latest
edition incorporated by reference, the 2013 Edition.
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
The NRC is revising Sec. 50.55a(b)(1)(ii) to clarify rule language
and add Table I, which clarifies prohibited Section III provisions for
welds with leg size less than 1.09 tn in tabular form.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
The NRC is revising Sec. 50.55a(b)(1)(iv) to clarify that it
allows, but does not require, applicants and licensees to use the 2008
Edition through the 2009-1a Addenda of NQA-1 when applying the 2010
Edition and later editions of the ASME BPV Code, Section III, up to the
2013 Edition. Applicants and licensees are required to meet appendix B
of 10 CFR part 50, and NQA-1 is one way of meeting portions of appendix
B. An applicant or licensee may select any version of NQA-1 that has
been approved for use in Sec. 50.55a, but they must also use the
administrative, quality, and technical provisions contained in the
version of NCA-4000 referencing that Edition or Addenda of
[[Page 32960]]
NQA-1 selected by the applicant or licensee.
NQA-1 provides a method for establishing and implementing a QA
program for the design and construction of nuclear power plants and
fuel reprocessing plants; however, NQA-1, as modified and supplemented
by NCA-4000, does not meet all of the requirements of appendix B to 10
CFR part 50.
Section 50.55a(b)(1)(iv) clarifies that applicants and licensees
using NQA-1 are also required to meet appendix B to 10 CFR part 50 and
the commitments contained in their QA program descriptions. To meet the
requirements of appendix B, when using NQA-1 during the design and
construction phase, applicants and licensees must address, in their
quality program description, those areas where NQA-1 is insufficient to
meet appendix B. Additional guidance and regulatory positions on how to
meet appendix B when using NQA-1 are provided in RG 1.28, ``Quality
Assurance Program Criteria (Design and Construction).''
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification
and Demonstration of Function of Incompressible-Fluid Pressure-Relief
Valves
The NRC is revising Sec. 50.55a(b)(1)(vii) to reflect the editions
and addenda of the ASME BPV Code incorporated by reference in this
rulemaking.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME
Certification Marks
The NRC is adding Sec. 50.55a(b)(1)(viii) to allow licensees to
use either the ASME BPV Code Symbol Stamp or ASME Certification Mark
with the appropriate certification designator and class designator as
specified in the 2013 Edition through the latest edition and addenda
incorporated by reference in Sec. 50.55a.
10 CFR 50.55a(b)(1)(ix) Section III Condition: NPT Code Symbol Stamps
The NRC is adding Sec. 50.55a(b)(1)(ix) to allow licensees to use
the NPT Code Symbol Stamp with the letters arranged horizontally as
specified in ASME BPV Code Case N-852 for the service life of a
component that had the NPT Code Symbol Stamp applied during the time
period from January 1, 2005, through December 31, 2015.
10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI
The NRC is revising Sec. 50.55a(b)(2) to reflect the editions and
addenda of the ASME BPV Code incorporated by reference in this
rulemaking.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and
Addenda of Subsection IWE and Subsection IWL
The NRC is revising Sec. 50.55a(b)(2)(vi) to clarify that the
provision applies only to the class of licensees of operating reactors
that were required by previous versions of Sec. 50.55a to develop and
implement a containment ISI program in accordance with Subsection IWE
and Subsection IWL, and complete an expedited examination of
containment during the 5-year period from September 9, 1996 to
September 9, 2001.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment
Examinations
The NRC is revising Sec. 50.55a(b)(2)(viii) by removing the
condition for using the 2007 Edition with 2009 Addenda through the 2013
Edition of Subsection IWL requiring compliance with Sec.
50.55a(b)(2)(viii)(E). To support the removal of the condition, the NRC
is adding new requirements governing the performance and documentation
of concrete containment examinations in Sec. 50.55a(b)(2)(viii)(H) and
(I), which are discussed separately in the next two headings.
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth
Provision
The NRC is adding Sec. 50.55a(b)(2)(viii)(H) to require licensees
to provide the applicable information specified in paragraphs
(b)(2)(viii)(E)(1), (2), and (3) of this section in the ISI Summary
Report required by IWA-6000 for each inaccessible concrete surface area
evaluated under the new code provision IWL-2512 of the 2009 Addenda up
to and including the 2013 Edition.
10 CFR 50.55a(b)(2)(viii)(I) Concrete Containment Examinations: Ninth
Provision
The NRC is adding Sec. 50.55a(b)(2)(viii)(I) to provide a new
condition requiring the technical evaluation required by IWL-2512(b) of
the 2009 Addenda up to and including the 2013 Edition of inaccessible
below-grade concrete surfaces exposed to foundation soil, backfill, or
groundwater be performed at periodic intervals not to exceed 5 years.
In addition, the licensee must examine representative samples of the
exposed portions of the below-grade concrete, when such below-grade
concrete is excavated for any reason. The condition applies only to
holders of renewed licenses under 10 CFR part 54 during the period of
extended operation (i.e., beyond the expiration date of the original
40-year license) of a renewed license when using IWL-2512(b) of the
2007 Edition with 2009 Addenda through the latest edition and addenda
in Sec. 50.55a(a)(1)(ii)--the 2013 Edition under this final rule.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment
Examinations
The NRC is revising Sec. 50.55a(b)(2)(ix) to continue to apply the
existing conditions in Sec. 50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B)
and (J) with respect to the metal containment examination requirements
in Subsection IWE up to and including the 2013 Edition (and all future
editions and addenda of the ASME BPV Code which the NRC incorporates by
reference into Sec. 50.55a). The NRC is accomplishing this by adding
the words ``edition and'' to the last sentence in Sec.
50.55a(b)(2)(ix).
10 CFR 50.55a(b)(2)(ix)(D) Metal Containment Examinations: Fourth
Provision
The NRC is revising the rule text in Sec. 50.55a(b)(2)(ix)(D) to
improve clarity. Section 50.55a(b)(2)(ix)(D) introductory text and
(b)(2)(ix)(D)(1) are combined. The information required to be included
in the ISI Summary report is now all on the same paragraph level. No
substantive change to the requirements is intended by this revision.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
The NRC is revising Sec. 50.55a(b)(2)(x) to clarify that it
allows, but does not require, licensees to use the 1994 Edition or the
2008 Edition through the 2009-1a Addenda of NQA-1 when applying the
2009 Addenda and later editions and addenda of the ASME BPV Code,
Section XI, up to the 2013 Edition. Licensees are required to meet
appendix B of 10 CFR part 50, and NQA-1 is one way of meeting portions
of appendix B. A licensee may select any version of NQA-1 that has been
approved for use in Sec. 50.55a.
NQA-1 provides a method for establishing and implementing a QA
program for the design and construction of nuclear power plants and
fuel reprocessing plants; however, NQA-1 does not meet all of the
requirements of appendix B to 10 CFR part 50. Section 50.55a(b)(2)(x)
clarifies that licensees using NQA-1 are also required to meet appendix
B to 10 CFR part 50 and the commitments contained in their QA program
descriptions. To meet the
[[Page 32961]]
requirements of appendix B, when using NQA-1 during ISI phase,
licensees must address, in their quality program description, those
areas where NQA-1 is insufficient to meet appendix B. Additional
guidance and regulatory positions on how to meet appendix B when using
NQA-1 are provided in RG 1.28.
10 CFR 50.55a(b)(2)(xii) Section XI Condition: Underwater Welding
The NRC is revising Sec. 50.55a(b)(2)(xii) to allow underwater
welding on irradiated materials in accordance with IWA-4660 under
certain conditions. Licensees are allowed to perform welding on
irradiated materials if certain neutron fluence criteria and, for
certain material classes, helium concentration criteria are not
exceeded. If these criteria are exceeded, the licensee is prohibited
from performing welding on irradiated materials unless the licensee
obtains NRC approval in accordance with Sec. 50.55a(z).
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth
Provision
The NRC is adding Sec. 50.55a(b)(2)(xviii)(D) to provide a new
condition prohibiting the use of Appendix VII and Subarticle VIII-2200
of the 2011 Addenda and 2013 Edition of Section XI of the ASME BPV
Code. Licensees are required to implement Appendix VII and Subarticle
VIII-2200 of the 2010 Edition of Section XI.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements:
First Provision
The NRC is revising Sec. 50.55a(b)(2)(xxi)(A) to modify the
standard for visual magnification resolution sensitivity and contrast
for visual examinations performed on Examination Category B-D
components instead of ultrasonic examinations. A visual examination
with magnification that has a resolution sensitivity to resolve 0.044
inch (1.1 mm) lower case characters without an ascender or descender
(e.g., a, e, n, v), utilizing the allowable flaw length criteria in
Table IWB-3512-1, 1997 Addenda through the latest edition and addenda
incorporated by reference in Sec. 50.55a(a)(1)(ii), with a limiting
assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed
instead of an ultrasonic examination. This revision removes a
requirement that was in addition to the ASME BPV Code that required 1-
mil wires to be used in licensees' Sensitivity, Resolution, and
Contrast Standard targets.
10 CFR 50.55a(b)(2)(xxiii) Section XI Condition: Evaluation of
Thermally Cut Surfaces
The NRC is revising Sec. 50.55a(b)(2)(xxiii) to modify the
applicability of the condition. The condition will only apply to the
2001 Edition through the 2009 Addenda IWA-4461.4, which was revised in
the 2010 Edition to remove paragraph IWA-4461.4.2, which permitted an
application specific evaluation of thermally cut surfaces in lieu of a
thermal metal removal process qualification.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping
Devices
The NRC is adding Sec. 50.55a(b)(2)(xxxi) to provide a new
condition maintaining the requirement to use Appendix IX, now
renumbered as Appendix W, when installing a mechanical clamping device
on an ASME BPV Code Class piping system. Additionally, the condition
prohibits the use of mechanical clamping devices in accordance with the
changes made to IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d) in
the 2011 Addenda through 2013 Edition on small item Class 1 piping and
portions of a piping system that form the containment boundary.
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report
Submittal
The NRC is adding Sec. 50.55a(b)(2)(xxxii) to provide a new
condition requiring licensees using the 2010 Edition or later editions
and addenda of Section XI to follow the requirements of IWA-6240 of the
2009 Addenda of Section XI for the submittal of Preservice and
Inservice Summary Reports. The condition also describes the timing of
the submission of the Summary Reports by referencing the specific
Section XI paragraph IWA-6240(b) in the regulation.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed
Allowable Pressure
The NRC is adding Sec. 50.55a(b)(2)(xxxiii) to provide a new
condition to prohibit the use of Appendix G, Paragraph G-2216, in the
2011 Addenda and later editions and addenda of the ASME BPV Code,
Section XI.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Nonmandatory Appendix
U
The NRC is adding Sec. 50.55a(b)(2)(xxxiv)(A) and (B) to require
that two conditions be satisfied when using Nonmandatory Appendix U of
the 2013 Edition of the ASME BPV Code, Section XI. Paragraph
(b)(2)(xxxiv)(A) requires that an ASME BPV Code repair or replacement
activity temporarily deferred under the provisions of Nonmandatory
Appendix U to the 2013 Edition of the ASME BPV Code, Section XI, shall
be performed during the next scheduled refueling outage. Paragraph
(b)(2)(xxxiv)(B) requires the use of the mandatory appendix in ASME BPV
Code Case N-513-3, in lieu of the appendix referenced in paragraph U-
S1-4.2.1(c) of Appendix U, which was inadvertently omitted from
Appendix U.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0
in the KIa and KIc Equations
The NRC is adding Sec. 50.55a(b)(2)(xxxv) to provide a new
condition to specify that when licensees use ASME BPV Code, Section XI,
2013 Edition, Appendix A, paragraph A-4200, if T0 is
available, then RTT0 may be used in place of
RTNDT for applications using the KIc equation and
the associated KIc curve, but not for applications using the
KIa equation and the associated KIa curve.
10 CFR 50.55a(b)(2)(xxxvi) Section XI Condition: Fracture Toughness of
Irradiated Materials
The NRC is adding Sec. 50.55a(b)(2)(xxxvi) to provide a new
condition requiring licensees using ASME BPV Code, Section XI, 2013
Edition, Appendix A, paragraph A-4400, to obtain NRC approval under
Sec. 50.55a(z) before using irradiated T0 and the
associated RTT0 in establishing fracture toughness of
irradiated materials.
10 CFR 50.55a(b)(2)(xxxvii) Section XI Condition: ASME BPV Code Case N-
824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii) to provide a new
provision that allows licensees to implement ASME BPV Code Case N-824,
``Ultrasonic Examination of Cast Austenitic Piping Welds From the
Outside Surface Section XI, Division 1,'' subject to four conditions in
paragraphs (b)(2)(xxxvii)(A) through (D). Each of these paragraphs are
discussed in the following headings.
10 CFR 50.55a(b)(2)(xxxvii)(A) Section XI Condition: ASME BPV Code Case
N-824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii)(A) to add a new
condition that requires ultrasonic examinations performed to implement
ASME BPV Code Case N-824 to be spatially encoded.
[[Page 32962]]
10 CFR 50.55a(b)(2)(xxxvii)(B) Section XI Condition: ASME BPV Code Case
N-824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii)(B) to add a new
condition that requires that ultrasonic examinations performed to
implement ASME BPV Code Case N-824 shall use dual, transmit-receive,
refracted longitudinal wave, multi-element phased array search units
instead of the requirements of Paragraph 1(c)(1)(-a) of N-824.
10 CFR 50.55a(b)(2)(xxxvii)(C) Section XI Condition: ASME BPV Code Case
N-824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii)(C) to add a new
condition that requires that ultrasonic examinations performed to
implement ASME BPV Code Case N-824 on piping greater than 1.6 inches
(41 mm) thick shall use a phased array search unit with a center
frequency of 500 kHz with a tolerance of + /- 20 percent instead of the
requirements of Paragraph 1(c)(1)(-c)(-2).
10 CFR 50.55a(b)(2)(xxxvii)(D) Section XI Condition: ASME BPV Code Case
N-824
The NRC is adding Sec. 50.55a(b)(2)(xxxvii)(D) to add a new
condition that requires that ultrasonic examinations performed to
implement ASME BPV Code Case N-824 shall use a phased array search unit
which produces angles including, but not limited to, 30 to 55 degrees
with a maximum increment of 5 degrees instead of the requirements of
Paragraph 1(c)(1)(-d).
10 CFR 50.55a(b)(3) Conditions on ASME OM Code
The NRC is revising Sec. 50.55a(b)(3) to clarify that Subsections
ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II,
III, and V; and Nonmandatory Appendices A through H and J through M of
the OM Code are each incorporated by reference into Sec. 50.55a. The
NRC is also clarifying that the OM Code Nonmandatory Appendices
incorporated by reference into Sec. 50.55a are approved for use, but
are not mandated. The Nonmandatory Appendices may be used by applicants
and licensees of nuclear power plants, subject to the conditions in
Sec. 50.55a(b)(3).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
The NRC is revising Sec. 50.55a(b)(3)(i) to allow licensees to use
the 1994 Edition, 2008 Edition, and 2009-1a Addenda of NQA-1 when using
the 1995 Edition through the 2012 Edition of the OM Code. Licensees are
required to meet appendix B to 10 CFR part 50, and NQA-1 is one way of
meeting portions of appendix B.
NQA-1 provides a method for establishing and implementing a QA
program for the design and construction of nuclear power plants and
fuel reprocessing plants; however, NQA-1 does not meet all of the
requirements of appendix B to 10 CFR part 50. Section 50.55a(b)(3)(i)
clarifies that licensees using NQA-1 are also required to meet appendix
B to 10 CFR part 50 and the commitments contained in their QA program
descriptions. To meet the requirements of appendix B, licensees must
address, in their quality program description, those areas where NQA-1
is insufficient to meet appendix B. Additional guidance and regulatory
positions on how to meet appendix B when using NQA-1 are provided in RG
1.28.
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV)
Testing
The NRC is revising Sec. 50.55a(b)(3)(ii) to set forth four
conditions on the use of mandatory Appendix III, ``Preservice and
Inservice Testing of Active Electric Motor Operated Valve Assemblies in
Light-Water Reactor Power Plants,'' in the OM Code, 2009 Edition, 2011
Addenda, and 2012 Edition. The four conditions, which are set forth in
paragraphs (b)(3)(ii)(A) through (D), are discussed in the next four
headings.
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval
The NRC is adding Sec. 50.55a(b)(3)(ii)(A) to require that
licensees evaluate the adequacy of the diagnostic test intervals
established for MOVs within the scope of OM Code, Appendix III, not
later than 5 years or three refueling outages (whichever is longer)
from initial implementation of OM Code, Appendix III.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk
The NRC is adding Sec. 50.55a(b)(3)(ii)(B) to require that
licensees ensure that the potential increase in CDF and LERF associated
with the extension is acceptably small when extending exercise test
intervals for high risk MOVs beyond a quarterly frequency. As specified
in RG 1.192, when extending exercise test intervals for high risk MOVs
beyond a quarterly frequency, licensees must ensure that the potential
increase in CDF and risk associated with the extension is small and
consistent with the intent of the Commission's Safety Goal Policy
Statement. As discussed earlier in Section II, the NRC provides
guidance in RG 1.174 that acceptably small changes are relative and
depend on the current plant CDF and LERF. For plants with total
baseline CDF of 10-4 per year or less, acceptably small
means CDF increases of up to 10-5 per year; and for plants
with total baseline CDF greater than 10-4 per year,
acceptably small means CDF increases of up to 10-6 per year.
For plants with total baseline LERF of 10-5 per year or
less, acceptably small LERF increases are considered to be up to
10-6 per year; and for plants with total baseline LERF
greater than 10-5 per year, acceptably small LERF increases
are considered to be up to 10-7 per year.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization
The NRC is adding Sec. 50.55a(b)(3)(ii)(C) to require, when
applying Appendix III to the OM Code, that licensees categorize MOVs
according to their safety significance using the methodology described
in OM Code Case OMN-3 subject to the conditions discussed in RG 1.192,
or using an MOV risk ranking methodology accepted by the NRC on a
plant-specific or industry-wide basis in accordance with the conditions
in the applicable safety evaluation.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time
The NRC is adding Sec. 50.55a(b)(3)(ii)(D) to require, when
applying Paragraph III-3600, ``MOV Exercising Requirements,'' of
Appendix III to the OM Code, licensees shall verify that the stroke
time of MOVs specified in plant technical specifications satisfies the
assumptions in the plant's safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors
The NRC is adding Sec. 50.55a(b)(3)(iii) to specify that, in
addition to complying with the provisions in the OM Code as required
with the conditions specified in Sec. 50.55a(b)(3), holders of
operating licenses for nuclear power reactors that received
construction permits under this part on or after the date 12 months
after August 17, 2017, and holders of COLs issued under 10 CFR part 52,
whose initial fuel loading occurs on or after the date 12 months after
August 17, 2017, shall also comply with four condition on power-
operated valves, check valves, flow-induced vibration, and operational
readiness of high-risk non-safety systems, to the extent applicable.
These four conditions, which are set forth in
[[Page 32963]]
Sec. 50.55a(b)(3)(iii)(A), (B), (C), and (D), are discussed in the
next four headings.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves (First Condition on
New Reactors)
The NRC is adding Sec. 50.55a(b)(3)(iii)(A) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) periodically verify the
capability of power-operated valves (POVs) to perform their design-
basis safety functions.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves (Second Condition on New
Reactors)
The NRC is adding Sec. 50.55a(b)(3)(iii)(B) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) perform bi-directional
testing of check valves within the IST program where practicable.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration (Third Condition on
New Reactors)
The NRC is adding Sec. 50.55a(b)(3)(iii)(C) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) monitor flow-induced
vibration (FIV) from hydrodynamic loads and acoustic resonance during
preservice testing or inservice testing to identify potential adverse
flow effects that might impact components within the scope of the IST
program.
10 CFR 50.55a(b)(3)(iii)(D) High Risk Non-Safety Systems (Fourth
Condition on New Reactors)
The NRC is adding Sec. 50.55a(b)(3)(iii)(D) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) establish a program to
assess the operational readiness of pumps, valves, and dynamic
restraints within the scope of the Regulatory Treatment of Non-Safety
Systems for applicable reactor designs. As of the time of this final
rule, these are designs which have been certified in a design
certification rule under 10 CFR part 52. This final rule refers to
these RTNSS components using the term, ``high risk non-safety
systems.''
As noted by the public commenters, ASME is preparing guidance for
new reactor licensees to use in developing programs for the treatment
of RTNSS equipment. The NRC staff is participating on the OM Code
committees to assist in developing guidance for the treatment of RTNSS
equipment that is consistent with Commission policy. Guidance on the
implementation of the Commission policy for RTNSS equipment is set
forth in NRC Inspection Procedure 73758, ``Part 52, Functional Design
and Qualification, and Preservice and Inservice Testing Programs for
Pumps, Valves and Dynamic Restraints,'' dated April 19, 2013.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
The NRC is revising Sec. 50.55a(b)(3)(iv) to extend the existing
conditions on the use of Appendix II to the new Editions and Addenda
which are the subject of this rulemaking. These conditions are that:
(i) Trending and evaluation shall support the determination that the
valve or group of valves is capable of performing its intended
function(s) over the entire interval; and (ii) at least one of the
Appendix II condition monitoring activities for a valve group shall be
performed on each valve of the group at approximate equal intervals not
to exceed the maximum interval shown in the following table:
Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
Maximum Maximum interval
interval between between
activities of activities of
Group size member valves in each valve in
the groups the group
(years) (years)
------------------------------------------------------------------------
>=4................................. 4.5 16
3................................... 4.5 12
2................................... 6 12
1................................... Not applicable 10
------------------------------------------------------------------------
The conditions currently specified for the use of Appendix II, 1995
Edition with the 1996 and 1997 Addenda, and 1998 Edition through the
2002 Addenda, of the OM Code remain the same in this final rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
The NRC is adding Sec. 50.55a(b)(3)(vii) to prohibit the use of
Subsection ISTB in the 2011 Addenda to the OM Code.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
The NRC is adding Sec. 50.55a(b)(3)(viii) to specify that
licensees who wish to implement Subsection ISTE, ``Risk-Informed
Inservice Testing of Components in Light-Water Reactor Nuclear Power
Plants,'' of the OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition,
must first request and obtain NRC approval in accordance with Sec.
50.55a(z) to apply Subsection ISTE on a plant-specific basis as a risk-
informed alternative to the applicable IST requirements in the OM Code.
The NRC will evaluate Sec. 50.55a(z) requests for approval to
implement Subsection ISTE in accordance with the following
considerations. These considerations are consistent with the guidance
provided in RG 1.174.
1. Scope of Risk-Informed IST Program
Subsection ISTE-1100, ``Applicability,'' establishes the component
safety categorization methodology and process for dividing the
population of pumps and valves, as identified in the IST Program Plan,
into high safety significant component (HSSC) and low safety
significant component (LSSC) categories. When establishing a risk-
informed IST program, the licensee should address a wide range of
components important to safety at the nuclear power plant that includes
both safety-related and nonsafety-related components. These components
might extend beyond the scope of the OM Code.
2. Risk-Ranking Methodology
The licensee should specify, in its request for authorization to
implement a risk-informed IST program, the methodology to be applied in
risk ranking its components. ISTE-4000, ``Specific Component
Categorization Requirements,'' incorporates OM Code Case OMN-3 for the
categorization of pumps and valves in developing a risk-informed IST
program. The OMN-3 Code Case methodology for risk ranking uses two
categories of safety
[[Page 32964]]
significance. The NRC staff has also accepted other methodologies for
risk ranking that use three categories of safety significance.
3. Safety Significance Categorization
The licensee should categorize components according to their safety
significance based on the methodology described in Subsection ISTE with
the applicable conditions on the use of OM Code Case OMN-3 specified in
RG 1.192, or use other risk ranking methodologies accepted by the NRC
on a plant-specific or industry-wide basis with applicable conditions
specified by the NRC for their acceptance. The licensee should address
the seven conditions in RG 1.192 for the use of OM Code Case OMN-3, as
appropriate, in developing the risk-informed IST program described in
Subsection ISTE. With respect to the provisions in Subsection ISTE,
these conditions are:
(a) The implementation of ISTE-1100 should include within the scope
of a licensee's risk-informed IST program non-ASME OM Code pumps and
valves categorized as HSSCs that might not currently be included in the
IST program at the nuclear power plant.
(b) The decision criteria discussed in ISTE-4410, ``Decision
Criteria,'' and Nonmandatory Appendix L, ``Acceptance Guidelines,'' of
the OM Code for evaluating the acceptability of aggregate risk effects
(i.e., for CDF and LERF) should be consistent with the guidance
provided in RG 1.174, ``An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis.''
(c) The implementation of ISTE-4440, ``Defense in Depth,'' should
be consistent with the guidance contained in Section 2.2.1, ``Defense-
in-Depth Evaluation,'' and Section 2.2.2, ``Safety Margin Evaluation,''
of RG 1.175, ``An Approach for Plant-Specific, Risk-Informed
Decisionmaking: Inservice Testing.''
(d) The implementation of ISTE-4500, ``Inservice Testing Program,''
and ISTE-6100, ``Performance Monitoring,'' should be consistent with
the guidance contained in Section 3.2, ``Program Implementation,'' and
Section 3.3, ``Performance Monitoring,'' of RG 1.175.
(e) The implementation of ISTE-3210, ``Plant-Specific PRA,'' should
be consistent with the guidance that the Owner is responsible for
demonstrating and justifying the technical adequacy of the PRA analyses
used as the basis to perform component risk ranking and for estimating
the aggregate risk impact. For example, RG 1.200, ``An Approach for
Determining the Technical Adequacy of Probabilistic Risk Assessment
Results for Risk-Informed Activities,'' and RG 1.201, ``Guidelines for
Categorizing Structures, Systems, and Components in Nuclear Power
Plants According to their Safety Significance,'' provide guidance for
PRA technical adequacy and component risk ranking.
(f) The implementation of ISTE-4240, ``Reconciliation,'' should
specify that 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 LSSC.
(g) The implementation of ISTE-3220, ``Living PRA,'' should be
consistent with the following: (i) 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; (ii)
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 (iii) the review of the
plant changes must be performed in a timely manner and must be
performed once every two refueling outages, or as required by Sec.
50.71(h)(2) for COL holders.
4. Pump Testing
Subsection ISTE-5100, ``Pumps,'' incorporates OM Code Case OMN-7
for risk-informed testing of pumps categorized as LSSCs. Subsection
ISTE-5100 allows the interval for Group A and Group B testing of LSSC
pumps specified in Subsection ISTB of the OM Code to be extended from
the current 3-month interval to intervals of 6 months or 2 years.
Subsection ISTE-5100 eliminates the requirement in Subsection ISTB to
perform comprehensive pump testing for LSSC pumps. Table ISTE-5121-1,
``LSSC Pump Testing,'' specifies that pump operation may be required
more frequently than the specified test frequency (6 months) to meet
vendor recommendations. Subsection ISTE-4500, ``Inservice Testing
Program,'' specifies in ISTE-4510, ``Maximum Testing Interval,'' that
the maximum testing interval shall be based on the more limiting of (a)
the results of the aggregate risk, or (b) the performance history of
the component. ISTE-5130, ``Maximum Test Interval--Pre-2000 Plants,''
specifies that the most limiting interval for LSSC pump testing shall
be determined from ISTE-4510 and ISTE-5120, ``Low Safety Significant
Pump Testing.'' The ASME developed the comprehensive pump test
requirements in the OM Code to address weaknesses in the Code
requirements to assess the operational readiness of pumps to perform
their design-basis safety function. Therefore, the licensee should
ensure that testing under Subsection ISTE will provide assurance of the
operational readiness of pumps in each safety significant
categorization to perform their design-basis safety function as
described in RGs 1.174 and 1.175.
5. Motor-Operated Valve Testing
Subsection ISTE-5300, ``Motor Operated Valve Assemblies,'' provides
a risk-informed IST approach instead of the IST requirements for MOVs
in Mandatory Appendix III to the OM Code. The ASME prepared Appendix
III to the OM Code to replace the requirement for quarterly stroke-time
testing of MOVs with a program of periodic exercising and diagnostic
testing to address lessons learned from nuclear power plant operating
experience and industry and regulatory research programs for MOV
performance. Subsection ISTC of the OM Code specifies the
implementation of Appendix III for periodic exercising and diagnostic
testing of MOVs to replace quarterly stroke-time testing previously
required for MOVs. Appendix III incorporates provisions that allow a
risk-informed IST approach for MOVs as described in OM Code Cases OMN-1
and OMN-11. Subsection ISTE-5300 is not consistent with the provisions
for the risk-informed IST program for MOVs specified in Appendix III to
the OM Code (and Code Cases OMN-1 and 11). Therefore, licensees who
wish to implement Subsection ISTE should address the provisions in
paragraph III-3700, ``Risk-Informed MOV Inservice Testing,'' of
Appendix III to the OM Code as incorporated by reference in Sec.
50.55a, with the applicable conditions, instead of ISTE-5300.
6. Pneumatically and Hydraulically Operated Valve Testing
Subsection ISTE-5400, ``Pneumatically and Hydraulically Operated
Valves,'' specifies that licensees test their AOVs and HOVs in
accordance with Appendix IV to the OM Code. Subsection ISTE-5400
indicates that Appendix IV is in the course of preparation. The NRC
staff will need to review Appendix IV prior to accepting its use as
part of Subsection ISTE. Therefore, licensees who wish to implement
Subsection ISTE should describe the planned IST provisions for AOVs and
HOVs in its request for approval to implement Subsection ISTE.
[[Page 32965]]
7. Pump Periodic Verification Test
Subsection ISTE does not include a requirement to implement the
pump periodic verification test program specified in Mandatory Appendix
V to the OM Code, 2012 Edition. Therefore, licensee should address the
consideration of a pump periodic verification test program in its risk-
informed IST program, proposed as part of the authorization request to
implement Subsection ISTE.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
The NRC is adding Sec. 50.55a(b)(3)(ix) to specify that licensees
applying Subsection ISTF, ``Inservice Testing of Pumps in Light-Water
Reactor Nuclear Power Plants--Post-2000 Plants,'' in the 2012 Edition
of the OM Code shall satisfy the requirements of Mandatory Appendix V,
``Pump Periodic Verification Test Program,'' of the OM Code, 2012
Edition. The paragraph also states that Subsection ISTF, 2011 Addenda,
is not acceptable for use.
10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20
The NRC is adding Sec. 50.55a(b)(3)(x) to allow licensees to
implement OM Code Case OMN-20, ``Inservice Test Frequency,'' in the OM
Code, 2012 Edition, for the editions and addenda of the OM Code that
are listed in Sec. 50.55a(a)(1)(iv).
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
The NRC is adding Sec. 50.55a(b)(3)(xi) to emphasize the
provisions in the OM Code, 2012 Edition, Subsection ISTC-3700,
``Position Verification Testing,'' to verify that valve obturator
position is accurately indicated. The OM Code, Subsection ISTC-3700
requires valves with remote position indicators shall be observed
locally at least once every 2 years to verify that valve operation is
accurately indicated. Licensees will be required to implement the
condition when adopting the 2012 Edition of the OM Code as their Code
of Record for the applicable 120-month IST interval.
10 CFR 50.55a(f) Preservice and Inservice Testing Requirements
The NRC is revising the heading for Sec. 50.55a(f) and clarifying
that the OM Code includes provisions for preservice testing of
components as part of its overall provisions for IST programs.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for
Operating Plants
The NRC is revising Sec. 50.55a(f)(4) to ensure that the paragraph
is applicable to pumps and valves that are within the scope of the OM
Code. The NRC is also including an additional provision in Sec.
50.55a(f)(4) stating that the IST requirements for pumps and valves
that are within the scope of the OM Code but are not classified as ASME
BPV Code Class 1, Class 2, or Class 3 may be satisfied as an augmented
IST program, in accordance with Sec. 50.55a(f)(6)(ii), without
requesting relief under Sec. 50.55a(f)(5) or alternatives under Sec.
50.55a(z). This use of an augmented IST program may be acceptable
provided the basis for deviations from the OM Code, as incorporated by
reference in this section, demonstrates an acceptable level of quality
and safety, or that implementing the Code provisions would result in
hardship or unusual difficulty without a compensating increase in the
level of quality and safety, where documented and available for NRC
review. These changes align the scope of pumps and valves for inservice
testing with the scope defined in the OM Code without imposing an
unnecessary paperwork burden on nuclear power plant licensees for the
submittal of relief and alternative requests for pumps and valves
within the scope of the OM Code but not classified as ASME BPV Code
Class 1, Class 2, or Class 3 components.
10 CFR 50.55a(g) Preservice and Inservice Inspection Requirements
The NRC is revising the heading in Sec. 50.55a(g), adding new
paragraphs (g)(2)(i), (ii), and (iii), and revising current paragraphs
(g) introductory text, (g)(2), (g)(3) introductory text, and (g)(3)(i),
(ii), and (v) to distinguish the requirements for accessibility,
preservice examination, and inservice inspection. No substantive change
to the requirements is intended by these revisions.
10 CFR 50.55a(g)(4) Inservice Inspection Standards Requirement for
Operating Plants
The NRC is revising Sec. 50.55a(g)(4)(ii) to add an implementation
period of 18-months for licensees whose ISI interval commences during
the 12 through 18-month period after the publication of this final
rule. The NRC is also revising Sec. 50.55a(g)(4)(i) and (ii) to add a
provision allowing licensees to adopt the latest version of Appendix
VIII of the ASME BPV Code edition or addenda listed in Sec.
50.55a(a)(1) at any time in the licensee's 120-month ISI interval.
10 CFR 50.55a(g)(6)(ii)(D) Augmented ISI Requirements: Reactor Vessel
Head Inspections
The NRC is revising Sec. 50.55a(g)(6)(ii)(D) to reflect the NRC's
approval of ASME BPV Code Case N-729-4, which supersedes the NRC's
earlier approval of ASME BPV Code Case N-729-1. The revisions include
changes to the conditions governing the use of the Code Case to reflect
the change from N-729-1 to N-729-4. The effect of these changes is to
require licensees to implement an augmented ISI program for the
examination of the pressurized water reactor RPV upper head
penetrations. The following discussions provide a more detailed
discussion of the revisions to Sec. 50.55a(g)(6)(ii)(D).
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
The NRC is revising Sec. 50.55a(g)(6)(ii)(D)(1) to require
licensees to implement an augmented ISI program for the examination of
the pressurized water reactor RPV upper head penetrations meeting ASME
BPV Code Case N-729-4 instead of the previously approved requirements
to use ASME BPV Code Case N-729-1, as conditioned by the NRC.
Removal of Existing Conditions in 10 CFR 50.55a(g)(6)(ii)(D)(2) Through
(5)
The NRC is removing the existing conditions in Sec.
50.55a(g)(6)(ii)(D)(2) through (5) and redesignating the existing
condition in Sec. 50.55a(g)(6)(ii)(D)(6) as Sec.
50.55a(g)(6)(ii)(D)(2).
10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I Use
The NRC is revising the existing condition in Sec.
50.55a(g)(6)(ii)(D)(6), which is redesignated as Sec.
50.55a(g)(6)(ii)(D)(2) in this final rule, to require NRC approval
prior to implementing Appendix I of ASME BPV Code Case N-729-4.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency
The NRC is adding a new condition in Sec. 50.55a(g)(6)(ii)(D)(3)
which requires cold head plants with less than eight effective
degradation years (EDY<8) without PWSCC flaws to perform a bare metal
visual examination (VE) each outage a volumetric exam is not performed
and allows these plants to extend the bare metal visual inspection
frequency from once each refueling outage, as stated in Table 1 of N-
729-4, to once every 5 years, only if the licensee performed a wetted
surface examination of all of the partial
[[Page 32966]]
penetration welds during the previous volumetric examination. In
addition, this new condition clarifies that a bare metal visual
examination is not required during refueling outages when a volumetric
or surface examination is performed of the partial penetration welds.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria
The NRC is adding a new condition in Sec. 50.55a(g)(6)(ii)(D)(4)
clarifying that rounded indications found by surface examinations of
the partial-penetration or associated fillet welds in accordance with
N-729-4 must meet the acceptance criteria for surface examinations of
paragraph NB-5352 of ASME 2013 Edition of Section III for the
licensee's ongoing 10-year ISI interval.
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(1) to require
licensees to implement an augmented ISI program for the examination of
ASME Class 1 piping and nozzle butt welds meeting ASME BPV Code Case N-
770-2 instead of the previously approved ASME BPV Code Case N-770-1.
Furthermore, the NRC is revising Sec. 50.55a(g)(6)(ii)(F)(1) to
update the date of applicability for pressurized water reactors, to
note the change to implement ASME BPV Code Case N-770-2 instead of N-
770-1, and to reflect the number of conditions which must be applied.
10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(2) to clarify the
requirements for licensees to establish the initial categorization of
each weld and modify the wording to reflect the ASME BPV Code Case N-
770-2 change in the inspection item category for full structural weld
overlays (C to C-1 and F to F-1). Additionally, the NRC is adding a
sentence which clarifies the NRC position that Paragraph -1100(e) of
ASME BPV Code Case N-770-2 shall not be used to exempt welds that rely
on Alloy 82/182 for structural integrity from any requirement of Sec.
50.55a(g)(6)(ii)(F).
10 CFR 50.55a(g)(6)(ii)(F)(3) Baseline Examinations
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(3) to clarify the
current requirement in this paragraph to complete baseline examinations
by stating that previously-conducted examinations, in order to count as
baseline examinations, must meet the requirements of ASME BPV Code Case
N-770-2, as conditioned in this section. Additionally, this condition
clarifies that the examination coverage requirements, for a licensee to
count previous inspections as baseline examinations, must meet the
examination coverage requirements described in Paragraphs -2500(a) or -
2500(b) of ASME BPV Code Case N-770-2, as conditioned by the NRC in
this section. Upon implementation of this rule, if a licensee is
currently in an outage, then the baseline inspection requirement can be
met by performing the inspections in accordance with the previous
regulatory requirements of Sec. 50.55a(g)(6)(ii)(F), in lieu of the
examination requirements of Paragraphs -2500(a) or -2500(b) of ASME BPV
Code Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(4) Examination Coverage
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(4) to clarify that
when licensees are implementing paragraph -2500(a) of ASME BPV Code
Case N-770-2, essentially 100 percent of the required volumetric
examination coverage shall be obtained, including greater than 90
percent volumetric examination coverage is obtained for circumferential
flaws, to continue the restriction on the licensee's use of Paragraph -
2500(c) and to continue the restriction that the use of new Paragraph -
2500(d) of ASME BPV Code Case N-770-2 is not allowed without prior NRC
review and approval in accordance with Sec. 50.55a(z), as it would
permit a reduction in volumetric examination coverage for
circumferential flaws. However, a licensee may request approval for use
of these paragraphs under Sec. 50.55a(z), and the NRC may approve the
request if technically justified.
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(5) to add an
explanatory heading, ``Inlay/onlay inspection frequency,'' and to make
minor editorial corrections without substantive changes in the
requirement.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(6) to add an
explanatory heading, ``Reporting requirements.''
10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(7) to add an
explanatory heading, ``Defining `t'.''
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(8) to add an
explanatory heading, ``Optimized weld overlay examination,'' and to
continue the current condition located in Sec. 50.55a(g)(6)(ii)(F)(9)
which requires that the initial examination of optimized weld overlays
(i.e., Inspection Item C-2 of ASME BPV Code Case N-770-2) be performed
between the third refueling outage and no later than 10 years after
application of the overlay and delete the other current examination
requirements for optimized weld overlay examination frequency, as these
requirements were included in the revision from N-770-1 to N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(9) Deferral
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(9) to add an
explanatory heading, ``Deferral,'' and to modify the current condition
to continue denial of the deferral of the initial inservice examination
of uncracked welds mitigated by optimized weld overlays. These welds
shall continue to have their initial inservice examinations as
prescribed in N-770-1 within 10 years of the application of the
optimized weld overlay and not allow deferral of this initial
examination. Subsequent inservice examinations may be deferred as
allowed by N-770-2. Additionally, the modified condition will delete
the current condition on examination requirements for the deferral of
welds mitigated by inlay, onlay, stress improvement and optimized weld
overlay, as these requirements were, with one exception (i.e.,
optimized weld overlay), included in the revision from N-770-1 to N-
770-2.
10 CFR 50.55a(g)(6)(ii)(F)(10) Examination Technique
The NRC is revising Sec. 50.55a(g)(6)(ii)(F)(10) to add an
explanatory heading, ``Examination technique,'' and to modify the
current condition to allow the previously prohibited alternate
examination requirements of Note (b) of Figure 5(a) of ASME BPV Code
Cases N-770-1 and N-770-2 and the same requirements in Note 14(b) of
Table 1 of ASME BPV Code Case N-770-2 for optimized weld overlays only
if the full examination requirements of Note 14(a) of Table 1 of
[[Page 32967]]
ASME BPV Code Case N-770-2 cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(11) to provide a new
condition requiring licensees to establish a Section XI, Appendix VIII,
qualification requirement for ultrasonic inspection of cast stainless
steel and through cast stainless steel to meet the examination
requirements of Paragraph -2500(a) of ASME BPV Code Case N-770-2 by
January 1, 2022.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(12) to provide a new
condition that would allow licensees to implement a stress improvement
mitigation technique for items containing cast stainless steel that
would meet the requirements of Appendix I of ASME BPV Code Case N-770-
2, if the required examination volume can be examined by Appendix VIII
procedures to the maximum extent practical including 100 percent of the
susceptible material volume.
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination
The NRC is adding Sec. 50.55a(g)(6)(ii)(F)(13) to provide a new
condition requiring licensees to perform encoded examinations of 100
percent of the required inspection volume when required to perform
volumetric examinations of all non-mitigated and cracked mitigated butt
welds in the reactor coolant pressure boundary in accordance with ASME
BPV Code Case N-770-2.
VI. Generic Aging Lessons Learned Report
Background
In December 2010, the NRC issued NUREG-1801, Revision 2, for
applicants to use in preparing their license renewal applications. The
GALL Report provides aging management programs (AMPs) that the NRC
staff has concluded are sufficient for aging management in accordance
with the license renewal rule, as required in Sec. 54.21(a)(3). In
addition, NUREG-1800, Revision 2, ``Standard Review Plan for Review of
License Renewal Applications for Nuclear Power Plants,'' was issued in
December 2010 to ensure the quality and uniformity of NRC staff reviews
of license renewal applications and to present a well-defined basis on
which the NRC staff evaluates the applicant's aging management programs
and activities. In April 2011, the NRC issued NUREG-1950, ``Disposition
of Public Comments and Technical Bases for Changes in the License
Renewal Guidance Documents NUREG-1801 and NUREG-1800,'' which describes
the technical bases for the changes in Revision 2 of the GALL Report
and Revision 2 of the SRP for review of license renewal applications.
Revision 2 of the GALL Report, in Sections XI.M1, XI.S1, XI.S2, and
XI.S3, describes the evaluation and technical bases for determining the
sufficiency of ASME BPV Code Subsections IWB, IWC, IWD, IWE, IWF, and
IWL for managing aging during the period of extended operation. In
addition, many other AMPs in the GALL Report rely, in part but to a
lesser degree, on the requirements specified in the ASME BPV Code,
Section XI. Revision 2 of the GALL Report also states that the 1995
Edition through the 2004 Edition of the ASME BPV Code, Section XI,
Subsections IWB, IWC, IWD, IWE, IWF, and IWL, as modified and limited
by Sec. 50.55a, were found to be acceptable editions and addenda for
complying with the requirements of Sec. 54.21(a)(3), unless
specifically noted in certain sections of the GALL Report. The GALL
Report further states that the future Federal Register notices that
amend Sec. 50.55a will discuss the acceptability of editions and
addenda more recent than the 2004 edition for their applicability to
license renewal.
In a final rule issued on June 21, 2011 (76 FR 36232), subsequent
to Revision 2 of the GALL Report, the NRC found that the 2004 Edition
with the 2005 Addenda through the 2007 Edition with the 2008 Addenda of
Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD, IWE, IWF,
and IWL, as subject to the conditions in Sec. 50.55a, are acceptable
for the AMPs in the GALL Report and the conclusions of the GALL Report
remain valid with the augmentations specifically noted in the GALL
Report.
Evaluation With Respect to Aging Management
As part of this rulemaking, the NRC evaluated whether those AMPs in
Revision 2 of the GALL Report which rely upon Subsections IWB, IWC,
IWD, IWE, IWF, and IWL of Section XI in the editions and addenda of the
ASME BPV Code incorporated by reference into Sec. 50.55a, continue to
be acceptable if the AMP relies upon the versions of these Subsections
in the 2007 Edition with the 2009 Addenda through the 2013 Edition. The
NRC finds that the 2007 Edition with the 2009 Addenda through the 2013
Edition of Section XI of the ASME BPV Code, Subsections IWB, IWC, IWD,
IWE, IWF, and IWL, as subject to the conditions of this rule, are
acceptable for the AMPs in the GALL Report and the conclusions of the
GALL Report remain valid with the augmentations specifically noted in
the GALL Report. Accordingly, an applicant for license renewal may use,
in its plant-specific license renewal application, Subsections IWB,
IWC, IWD, IWE, IWF, and IWL of Section XI of the 2007 Edition with the
2009 Addenda through the 2013 Edition of the ASME BPV Code, as subject
to the conditions in this rule, without additional justification.
Similarly, a licensee approved for license renewal that relied on
the GALL AMPs may use Subsections IWB, IWC, IWD, IWE, IWF, and IWL of
Section XI of the 2007 Edition with the 2009 Addenda through the 2013
Edition of the ASME BPV Code. However, a licensee must assess and
follow applicable NRC requirements with regard to changes to its
licensing basis. Some of the AMPs in the GALL Report recommend
augmentation of certain Code requirements in order to ensure adequate
aging management for license renewal. The technical and regulatory
aspects of the AMPs for which augmentations are recommended also apply
if the editions or addenda from the 2007 Edition with the 2009 Addenda
through the 2013 Edition of Section XI of the ASME BPV Code are used to
meet the requirements of Sec. 54.21(a)(3). The NRC staff evaluated the
changes in the 2007 Edition with the 2009 Addenda through the 2013
Edition of Section XI of the ASME BPV Code to determine if the
augmentations described in the GALL Report remain necessary. The NRC
staff's evaluation has concluded that the augmentations described in
the GALL Report are necessary to ensure adequate aging management. For
example, Table IWB-2500-1, in the 2007 Edition with the 2009 Addenda of
ASME BPV Code, Section XI, Subsection IWB, requires surface examination
of ASME BPV Code Class 1 branch pipe connection welds less than nominal
pipe size (NPS) 4 under Examination Category B-J. However, the NRC
staff finds that volumetric or opportunistic destructive examination,
rather than surface examination, is necessary to adequately detect and
manage the aging effect due to stress corrosion cracking or thermal,
mechanical and vibratory loadings in the components for the period of
extended operation. Therefore, GALL Report Section XI.M35, ``One-Time
Inspection of ASME BPV Code Class 1 Small-Bore Piping,'' includes the
augmentation of the requirements in ASME BPV Code, Section XI,
[[Page 32968]]
Subsection IWB to perform a one-time inspection of a sample of ASME BPV
Code Class 1 piping less than NPS 4 and greater than or equal to NPS 1
using volumetric or opportunistic destructive examination. The GALL
Report addresses this augmentation to confirm that there is no need to
manage age-related degradation through periodic volumetric inspections
or that an existing AMP (for example, Water Chemistry AMP) is effective
to manage the aging effect due to stress corrosion cracking or thermal,
mechanical and vibratory loadings for the period of extended operation.
A license renewal applicant may either augment its AMPs as described in
the GALL Report, or propose alternatives for the NRC to review as part
of the applicant's plant-specific justification for its AMPs.
VII. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC
certifies that this rule does not have a significant economic impact on
a substantial number of small entities. This final rule affects only
the licensing and operation of nuclear power plants. The companies that
own these plants do not fall within the scope of the definition of
``small entities'' set forth in the Regulatory Flexibility Act or the
size standards established by the NRC (Sec. 2.810).
VIII. Regulatory Analysis
The NRC has prepared a final regulatory analysis on this
regulation. The analysis examines the costs and benefits of the
alternatives considered by the NRC. The regulatory analysis is
available as indicated in the ``Availability of Documents'' section of
this document.
IX. Backfitting and Issue Finality
Introduction
The NRC's Backfit Rule in Sec. 50.109 states that the NRC shall
require the backfitting of a facility only when it finds the action to
be justified under specific standards stated in the rule. Section
50.109(a)(1) defines backfitting as the modification of or addition to
systems, structures, components, or design of a facility; the design
approval or manufacturing license for a facility; or the procedures or
organization required to design, construct, or operate a facility. Any
of these modifications or additions may result from a new or amended
provision in the NRC's rules or the imposition of a regulatory position
interpreting the NRC's rules that is either new or different from a
previously applicable NRC position after issuance of the construction
permit or the operating license or the design approval.
Section 50.55a requires nuclear power plant licensees to:
Construct ASME BPV Code Class 1, 2, and 3 components in
accordance with the rules provided in Section III, Division 1, of the
ASME BPV Code (``Section III'').
Inspect Class 1, 2, 3, Class MC, and Class CC components
in accordance with the rules provided in Section XI, Division 1, of the
ASME BPV Code (``Section XI'').
Test Class 1, 2, and 3 pumps, valves, and dynamic
restraints (snubbers) in accordance with the rules provided in the OM
Code.
This final rule is incorporating by reference the 2009 Addenda,
2010 Edition, 2011 Addenda, and the 2013 Edition of the ASME BPV Code,
Section III, Division 1 and ASME BPV Code, Section XI, Division 1,
including NQA-1 (with conditions on its use), as well as the 2009
Edition and 2011 Addenda and 2012 Edition of the OM Code and Code Cases
N-770-2 and N-729-4.
The ASME BPV and OM Codes are national consensus standards
developed by participants with broad and varied interests, in which all
interested parties (including the NRC and utilities) participate. A
consensus process involving a wide range of stakeholders is consistent
with the NTTAA, inasmuch as the NRC has determined that there are sound
regulatory reasons for establishing regulatory requirements for design,
maintenance, ISI, and IST by rulemaking. The process also facilitates
early stakeholder consideration of backfitting issues. Therefore, the
NRC believes that the NRC need not address backfitting with respect to
the NRC's general practice of incorporating by reference updated ASME
Codes.
Overall Backfitting Considerations: Section III of the ASME BPV Code
Incorporation by reference of more recent editions and addenda of
Section III of the ASME BPV Code does not affect a plant that has
received a construction permit or an operating license or a design that
has been approved. This is because the edition and addenda to be used
in constructing a plant are, under Sec. 50.55a, determined based on
the date of the construction permit, and are not changed thereafter,
except voluntarily by the licensee. The incorporation by reference of
more recent editions and addenda of Section III ordinarily applies only
to applicants after the effective date of a final rule incorporating
these new editions and addenda. Therefore, incorporation by reference
of a more recent edition and addenda of Section III does not constitute
``backfitting'' as defined in Sec. 50.109(a)(1).
Overall Backfitting Considerations: Section XI of the ASME BPV Code and
the OM Code
Incorporation by reference of more recent editions and addenda of
Section XI of the ASME BPV Code and the OM Code affects the ISI and IST
programs of operating reactors. However, the Backfit Rule generally
does not apply to incorporation by reference of later editions and
addenda of the ASME BPV Code (Section XI) and OM Code. As previously
mentioned, the NRC's longstanding regulatory practice has been to
incorporate later versions of the ASME Codes into Sec. 50.55a. Under
Sec. 50.55a, licensees shall revise their ISI and IST programs every
120 months to the latest edition and addenda of Section XI of the ASME
BPV Code and the OM Code incorporated by reference into Sec. 50.55a 12
months before the start of a new 120-month ISI and IST interval.
Therefore, when the NRC approves and requires the use of a later
version of the Code for ISI and IST, it is implementing this
longstanding regulatory practice and requirement.
Other circumstances where the NRC does not apply the Backfit Rule
to the approval and requirement to use later Code editions and addenda
are as follows:
1. When the NRC takes exception to a later ASME BPV Code or OM Code
provision but merely retains the current existing requirement,
prohibits the use of the later Code provision, limits the use of the
later Code provision, or supplements the provisions in a later Code.
The Backfit Rule does not apply because the NRC is not imposing new
requirements. However, the NRC explains any such exceptions to the Code
in the statement of considerations and regulatory analysis for the
rule.
2. When an NRC exception relaxes an existing ASME BPV Code or OM
Code provision but does not prohibit a licensee from using the existing
Code provision. The Backfit Rule does not apply because the NRC is not
imposing new requirements.
3. The NRC's consideration of backfitting for modifications and
limitations imposed during previous routine updates of Sec. 50.55a
have established a precedent for determining the kinds of modifications
or limitations which should be considered backfitting, or require a
backfit analysis (e.g., final rule dated September 10, 2008 (73 FR
52730), and a correction dated October 2, 2008 (73 FR 57235)). The
consideration of backfitting and issue
[[Page 32969]]
finality with respect to the modifications and limitations in this
rulemaking are consistent with the consideration and application of
backfitting and issue finality requirements to analogous modifications
and limitations in previous Sec. 50.55a rulemakings.
The incorporation by reference and adoption of a requirement
mandating the use of a later ASME BPV Code or OM Code may constitute
backfitting in some circumstances. In these cases, the NRC would
perform a backfit analysis or documented evaluation in accordance with
Sec. 50.109. These include the following:
1. When the NRC endorses a later provision of the ASME BPV Code or
OM Code that takes a substantially different direction from the
existing requirements, the action is treated as a backfit (e.g., 61 FR
41303 (August 8, 1996)).
2. When the NRC requires implementation of a later ASME BPV Code or
OM Code provision on an expedited basis, the action is treated as a
backfit. This applies when implementation is required sooner than it
would be required if the NRC simply endorsed the Code without any
expedited language (e.g., 64 FR 51370 (September 22, 1999)).
3. When the NRC takes an exception to an ASME BPV Code or OM Code
provision and imposes a requirement that is substantially different
from the existing requirement as well as substantially different from
the later Code (e.g., 67 FR 60529 (September 26, 2002)).
Detailed Backfitting Discussion: Changes Beyond Those Necessary To
Incorporate by Reference the New ASME BPV and OM Code Provisions
This section discusses the backfitting considerations for all the
changes to Sec. 50.55a that go beyond the minimum changes necessary
and required to adopt the new ASME Code Addenda into Sec. 50.55a.
ASME BPV Code, Section III
1. Revise Sec. 50.55a(b)(1)(ii), ``Weld leg dimensions,'' to
clarify rule language and add Table I, which clarifies prohibited
Section III provisions for welds with leg sizes less than 1.09
tn in tabular form. This change does not alter the original
intent of this requirement and, therefore, does not impose a new
requirement. Therefore, this change is not a backfit.
2. Revise Sec. 50.55a(b)(1)(iv), ``Quality assurance,'' to require
that when applying editions and addenda later than the 1989 Edition of
Section III, the requirements of NQA-1, 1994 Edition, 2008 Edition, and
the 2009-1a Addenda are acceptable for use, provided that the edition
and addenda of NQA-1 specified in either NCA-4000 or NCA-7000 is used
in conjunction with the administrative, quality, and technical
provisions contained in the edition and addenda of Section III being
used. This revision clarifies the current requirements, and is
considered to be consistent with the meaning and intent of the current
requirements, and therefore is not considered to result in a change in
requirements. Therefore, this change is not a backfit.
3. Add a new condition as Sec. 50.55a(b)(1)(viii), ``Use of ASME
Certification Marks,'' to allow licensees to use either the ASME BPV
Code Symbol Stamp or ASME Certification Mark with the appropriate
certification designator and class designator as specified in the 2013
Edition through the latest edition and addenda incorporated by
reference in Sec. 50.55a. This condition does not result in a change
in requirements previously approved in the Code and, therefore, is not
a backfit.
ASME BPV Code, Section XI
1. Revise Sec. 50.55a(b)(2)(vi), ``Effective edition and addenda
of Subsection IWE and Subsection IWL,'' to clarify that the provision
applies only to the class of licensees of operating reactors that were
required by previous versions of Sec. 50.55a to develop, implement a
containment ISI program in accordance with Subsection IWE and
Subsection IWL, and complete an expedited examination of containment
during the 5-year period from September 9, 1996, to September 9, 2001.
This revision clarifies the current requirements, is considered to be
consistent with the meaning and intent of the current requirements, and
is not considered to result in a change in requirements. Therefore,
this change is not a backfit.
2. Revise Sec. 50.55a(b)(2)(viii), ``Concrete containment
examinations,'' so that when using the 2007 Edition with 2009 Addenda
through the 2013 Edition of Subsection IWL, the conditions in Sec.
50.55a(b)(2)(viii)(E) do not apply, but the new conditions in Sec.
50.55a(b)(2)(viii)(H) and (I) do apply. This revision does not require
Sec. 50.55a(b)(2)(viii)(E) to be used when following the 2007 Edition
with 2009 Addenda through the 2013 Edition of Subsection IWL because
most of its requirements have been included in IWL-2512, ``Inaccessible
Areas.'' Therefore, this change is not a backfit because the
requirements have not changed. The revision to add the condition in
Sec. 50.55a(b)(2)(viii)(H) captures the reporting requirements of the
current Sec. 50.55a(b)(2)(viii)(E) which were not included in IWL-
2512. Therefore, this change is not a backfit because the requirements
have not changed. The revision to add the condition in Sec.
50.55a(b)(2)(viii)(I) addresses a new code provision in IWL-2512(b) for
evaluation of below-grade concrete surfaces during the period of
extended operation of a renewed license. The condition assures
consistency with the GALL Report, Revision 2, and applies to plants
going forward using the 2007 Edition with 2009 Addenda through the 2013
Edition of Subsection IWL. The requirements remain unchanged from the
recommendations in the GALL Report and, therefore, this change is not a
backfit.
3. Revise Sec. 50.55a(b)(2)(ix), ``Metal containment
examinations,'' to extend the applicability of the existing conditions
in Sec. 50.55a(b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) to the 2007
Edition with 2009 Addenda through the 2013 Edition of Subsection IWE.
This condition does not result in a change to current requirements, and
is therefore not a backfit.
4. Revise Sec. 50.55a(b)(2)(x), ``Quality assurance,'' to require
that when applying the editions and addenda later than the 1989 Edition
of ASME BPV Code, Section XI, the requirements of NQA-1, 1994 Edition,
the 2008 Edition, and the 2009-1a Addenda specified in either IWA-1400
or Table IWA 1600-1, ``Referenced Standards and Specifications,'' of
that edition and addenda of Section XI are acceptable for use, provided
the licensee uses its appendix B to 10 CFR part 50 QA program in
conjunction with Section XI requirements. This revision clarifies the
current requirements, which the NRC considers to be consistent with the
meaning and intent of the current requirements. Therefore, the NRC does
not consider the clarification to be a change in requirements.
Therefore, this change is not a backfit.
5. Revise Sec. 50.55a(b)(2)(xii), ``Underwater welding,'' to allow
underwater welding on irradiated materials under certain conditions.
The revision eliminates the prohibition on welding on irradiated
materials. Therefore, this change is not a backfit.
6. Add a new condition as Sec. 50.55a(b)(2)(xviii)(D), ``NDE
personnel certification: Fourth provision,'' to prohibit the use of
Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013
Edition of Section XI of the ASME BPV Code. Licensees are required to
implement Appendix VII
[[Page 32970]]
and Subarticle VIII-2200 of the 2010 Edition of Section XI. This
condition does not constitute a change in NRC position because the use
of the subject provisions is not currently allowed by Sec. 50.55a.
Therefore, the addition of this new condition is not a backfit.
7. Revise Sec. 50.55a(b)(2)(xxi)(A), ``Table IWB-2500-1
examination requirements: First provision,'' to modify the standard for
visual magnification resolution sensitivity and contrast for visual
examinations of Examination Category B-D components, making the rule
conform with ASME BPV Code, Section XI requirements for VT-1
examinations. This revision removes a condition that was in addition to
the ASME BPV Code requirements and does not impose a new requirement.
Therefore, this change is not a backfit.
8. Add a new condition as Sec. 50.55a(b)(2)(xxxi), ``Mechanical
clamping devices;'' to prohibit the use of mechanical clamping devices
in accordance with IWA-4131.1(c) in the 2010 Edition and IWA-4131.1(d)
in the 2011 Addenda through 2013 Edition on small item Class 1 piping
and portions of a piping system that forms the containment boundary.
This condition does not constitute a change in NRC position and does
not affect licensees because the use of the subject provisions is not
currently allowed by Sec. 50.55a. Therefore, the addition of this new
condition is not a backfit.
9. Add a new condition as Sec. 50.55a(b)(2)(xxxii), ``Summary
report submittal,'' to clarify that licensees using the 2010 Edition or
later editions and addenda of Section XI must continue to submit to the
NRC the Preservice and Inservice Summary Reports required by IWA-6240
of the 2009 Addenda of Section XI. This condition does not result in a
change in the NRC's requirements insomuch as these reports have been
required in the 2009 Addenda of Section XI and all previous editions
and addenda. Therefore, the addition of this new condition is not a
backfit.
10. Add a new condition as Sec. 50.55a(b)(2)(xxxiii), ``Risk-
Informed allowable pressure,'' to prohibit the use of ASME BPV Code,
Section XI, Appendix G, Paragraph G-2216. The use of Paragraph G-2216
is not currently allowed by Sec. 50.55a. Therefore, the condition does
not constitute a new or changed NRC position on the lack of
acceptability of Paragraph G-2216. Therefore, the addition of this new
condition is not a backfit.
11. Add a new condition as Sec. 50.55a(b)(2)(xxxiv),
``Nonmandatory Appendix U.'' Paragraph (b)(2)(xxxiv)(A) requires that
repair or replacement activities temporarily deferred under the
provisions of Nonmandatory Appendix U shall be performed during the
next scheduled refueling outage. This condition is imposed to ensure
that repairs/replacements are performed on degraded components when a
unit is shutdown for refueling. This change is consistent with the
condition previously placed on ASME BPV Code Case N-513-3 and,
therefore, does not impose a new requirement. This change is not a
backfit. Paragraph (b)(2)(xxxiv)(B) requires that the mandatory
appendix in ASME BPV Code Case N-513-3 be used in lieu of the appendix
referenced in Paragraph U-S1-4.2.1(c) of Appendix U. This change is
required because the appendix referenced in Appendix U was
unintentionally omitted. This change is not a backfit.
12. Add a new condition as Sec. 50.55a(b)(2)(xxxv), ``Use of
RTT0 in the KIa and KIc equations,''
to specify that when licensees use ASME BPV Code, Section XI 2013
Edition Nonmandatory Appendix A, Paragraph A-4200, if T0 is
available, then RTT0 may be used in place of
RTNDT for applications using the KIc equation and
the associated KIc curve, but not for applications using the
KIa equation and the associated KIa curve.
Conditions on the use of ASME BPV Code, Section XI, Nonmandatory
Appendices do not constitute backfitting inasmuch as those provisions
apply to voluntary actions initiated by the licensee to use the
``nonmandatory compliance'' provisions in these Appendices of the rule.
13. Add a new condition as Sec. 50.55a(b)(2)(xxxvi), ``Fracture
toughness of irradiated materials,'' to require licensees using ASME
BPV Code, Section XI 2013 Edition Nonmandatory Appendix A, Paragraph A-
4400, to obtain NRC approval before using irradiated T0 and
the associated RTT0 in establishing fracture toughness of
irradiated materials. Conditions on the use of ASME BPV Code, Section
XI, Nonmandatory Appendices do not constitute backfitting inasmuch as
those provisions apply to voluntary actions initiated by the licensee
to use the ``nonmandatory compliance'' provisions in these Appendices
of the rule.
14. Add a new condition as Sec. 50.55a(b)(2)(xxxvii), ``ASME BPV
Code Case N-824,'' to allow the use of the code case as conditioned.
Conditions on the use of ASME BPV Code Case N-824 do not constitute
backfitting, inasmuch as the use of this code case is not required by
the NRC but instead is an alternative which may be voluntarily used by
the licensee (i.e., a ``voluntary alternative'').
OM Code
1. Add a new condition as Sec. 50.55a(b)(3)(ii)(A), ``MOV
diagnostic test interval,'' to require that licensees evaluate the
adequacy of the diagnostic test intervals established for MOVs within
the scope of OM Code, Appendix III, not later than 5 years or three
refueling outages (whichever is longer) from initial implementation of
Appendix III of the OM Code. This condition represents an exception to
a later OM Code provision but merely retains the current NRC condition
on ASME OM Code Case OMN-1, and is therefore not a backfit because the
NRC is not imposing a new requirement.
2. Add a new condition as Sec. 50.55a(b)(3)(ii)(B), ``MOV testing
impact on risk,'' to require that licensees ensure that the potential
increase in core damage frequency and large early release frequency
associated with the extension is acceptably small when extending
exercise test intervals for high risk MOVs beyond a quarterly
frequency. This condition represents an exception to a later OM Code
provision but merely retains the current NRC condition on ASME OM Code
Case OMN-1, and is therefore not a backfit because the NRC is not
imposing a new requirement.
3. Add a new condition as Sec. 50.55a(b)(3)(ii)(C), ``MOV risk
categorization,'' to require, when applying Appendix III to the OM
Code, that licensees categorize MOVs according to their safety
significance using the methodology described in OM Code Case OMN-3
subject to the conditions discussed in RG 1.192, or using an MOV risk
ranking methodology accepted by the NRC on a plant-specific or
industry-wide basis in accordance with the conditions in the applicable
safety evaluation. This condition represents an exception to a later OM
Code provision but merely retains the current NRC condition on ASME OM
Code Case OMN-1, and is therefore not a backfit because the NRC is not
imposing a new requirement.
4. Add a new condition as Sec. 50.55a(b)(3)(ii)(D), ``MOV stroke
time,'' to require that, when applying Paragraph III-3600, ``MOV
Exercising Requirements,'' of Appendix III to the OM Code, licensees
shall verify that the stroke time of the MOVs specified in plant
technical specifications satisfies the assumptions in the plant's
safety analyses. This condition retains the MOV stroke time requirement
for a smaller set of MOVs than was specified in previous editions and
addenda of the
[[Page 32971]]
OM Code. The retention of this requirement is not a backfit.
5. Add new conditions as Sec. 50.55a(b)(3)(iii)(A) through (D),
``New reactors,'' to apply specific conditions for IST programs
applicable to licensees of new nuclear power plants in addition to the
provisions of the OM Code as incorporated by reference with conditions
in Sec. 50.55a. Licensees of ``new reactors'' are, as identified in
the paragraph: (1) Holders of operating licenses for nuclear power
reactors that received construction permits under this part on or after
the date 12 months after August 17, 2017, and (2) holders of COLs
issued under 10 CFR part 52, whose initial fuel loading occurs on or
after the date 12 months after August 17, 2017. This implementation
schedule for new reactors is consistent with the NRC regulations in
Sec. 50.55a(f)(4)(i). These conditions represent an exception to a
later OM Code provision but merely retain a current NRC requirement,
and are therefore not a backfit because the NRC is not imposing a new
requirement.
6. Revise Sec. 50.55a(b)(3)(iv), ``Check valves (Appendix II),''
to specify that Appendix II, ``Check Valve Condition Monitoring
Program,'' of the OM Code, 2003 Addenda through the 2012 Edition, is
acceptable for use with the following clarification: Trending and
evaluation shall support the determination that the valve or group of
valves is capable of performing its intended function(s) over the
entire interval. At least one of the Appendix II condition monitoring
activities for a valve group shall be performed on each valve of the
group at approximate equal intervals not to exceed the maximum interval
shown in the following table:
Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
Maximum
Maximum interval
interval between between
Group size activities of activities of
member valves each valve in
in the groups the group
(years) (years)
------------------------------------------------------------------------
>=4................................. 4.5 16
3................................... 4.5 12
2................................... 6 12
1................................... Not applicable 10
------------------------------------------------------------------------
The regulation is being revised to extend the applicability of this
existing NRC condition on the OM Code to the 2012 Edition of the OM
Code and to update the clarification for the use of Appendix II. This
does not represent a change in the NRC's position that the condition is
needed with respect to the OM Code. Therefore, this condition is not a
backfit.
7. Add a new condition as Sec. 50.55a(b)(3)(vii), ``Subsection
ISTB,'' to prohibit the use of Subsection ISTB in the 2011 Addenda to
the OM Code because the complete set of planned Code modifications to
support the changes to the comprehensive pump test acceptance criteria
was not made in that addenda. This condition represents an exception to
a later OM Code provision but merely limits the use of the later Code
provision, and is therefore not a backfit because the NRC is not
imposing a new requirement.
8. Add a new condition as Sec. 50.55a(b)(3)(viii), ``Subsection
ISTE,'' to allow licensees to implement Subsection ISTE, ``Risk-
Informed Inservice Testing of Components in Light-Water Reactor Nuclear
Power Plants,'' in the OM Code, 2009 Edition, 2011 Addenda and 2012
Edition, where the licensee has obtained authorization to implement
Subsection ISTE as an alternative to the applicable IST requirements in
the OM Code on a case-by-case basis in accordance with Sec. 50.55a(z).
This condition represents an exception to a later OM Code provision but
merely limits the use of the later Code provision, and is therefore not
a backfit because the NRC is not imposing a new requirement.
9. Add a new condition as Sec. 50.55a(b)(3)(ix), ``Subsection
ISTF,'' to specify that licensees applying Subsection ISTF, 2012
Edition, shall satisfy the requirements of Mandatory Appendix V, ``Pump
Periodic Verification Test Program,'' of the OM Code, 2012 Edition. The
condition also specifies that Subsection ISTF, 2011 Addenda, is not
acceptable for use. This condition represents an exception to a later
OM Code provision but merely limits the use of the later Code
provision, and is therefore not a backfit because the NRC is not
imposing a new requirement.
10. Add a new condition as Sec. 50.55a(b)(3)(x), ``ASME OM Code
Case OMN-20,'' to allow licensees to implement OM Code Case OMN-20,
``Inservice Test Frequency,'' in the OM Code, 2012 Edition. This
condition allows voluntary action initiated by the licensee to use the
code case and is, therefore, not a backfit.
11. Add a new condition as Sec. 50.55a(b)(3)(xi), ``Valve Position
Indication,'' to emphasize, when implementing OM Code (2012 Edition),
Subsection ISTC-3700, ``Position Verification Testing,'' licensees
shall implement the OM Code provisions to verify that valve operation
is accurately indicated. This condition emphasizes the OM Code
requirements for valve position indication and is not a change to those
requirements. As such, this condition is not a backfit.
12. Revise Sec. 50.55a(f), ``Preservice and inservice testing
requirements,'' to clarify that the OM Code includes provisions for
preservice testing of components as part of its overall provisions for
IST programs. No expansion of IST program scope is intended by this
clarification. This condition does not result in a change in
requirements previously approved in the Code and is, therefore, not a
backfit.
13. Revise Sec. 50.55a(f)(4), ``Inservice testing standards for
operating plants,'' to state that the paragraph is applicable to pumps
and valves that are within the scope of the OM Code. Also, revise Sec.
50.55a(f)(4) to state that the IST requirements for pumps and valves
that are within the scope of the OM Code but are not classified as ASME
BPV Code Class 1, Class 2, or Class 3 may be satisfied as an augmented
IST program in accordance with Sec. 50.55a(f)(6)(ii) without
requesting relief under Sec. 50.55a(f)(5) or alternatives under Sec.
50.55a(z). This use of an augmented IST program may be acceptable
provided the basis for deviations from the OM Code as incorporated by
reference in this section demonstrates an acceptable level of quality
and safety, or that implementing the Code provisions would result in
hardship or
[[Page 32972]]
unusual difficulty without a compensating increase in the level of
quality and safety, where documented and available for NRC review.
These changes align the scope of pumps and valves for inservice testing
with the scope defined in the OM Code. These changes do not result in a
change in requirements previously approved in the Code, and is
therefore not a backfit.
ASME BPV Code Case N-729-4
Revise Sec. 50.55a(g)(6)(ii)(D), ``Reactor vessel head
inspections.''
On June 22, 2012, the ASME approved the fourth revision of ASME BPV
Code Case N-729 (N-729-4). The NRC proposed to update the requirements
of Sec. 50.55a(g)(6)(ii)(D) to require licensees to implement ASME BPV
Code Case N-729-4, with conditions. The ASME BPV Code Case N-729-4
contains similar requirements as N-729-1; however, N-729-4 also
contains new requirements to address previous NRC conditions, including
changes to inspection frequency and qualifications. The new NRC
conditions on the use of ASME BPV Code Case N-729-4 address operational
experience, clarification of implementation, and the use of
alternatives to the code case.
The current regulatory requirements for the examination of
pressurized water reactor upper RPV heads that use nickel-alloy
materials are provided in Sec. 50.55a(g)(6)(ii)(D). This section was
first created by rulemaking, dated September 10, 2008 (73 FR 52730), to
require licensees to implement ASME BPV Code Case N-729-1, with
conditions, instead of the inspections previously required by the ASME
BPV Code, Section XI. The action did constitute a backfit; however, the
NRC concluded that imposition of ASME BPV Code Case N-729-1, as
conditioned, constituted an adequate protection backfit.
The General Design Criteria (GDC) for nuclear power plants
(appendix A to 10 CFR part 50) or, as appropriate, similar requirements
in the licensing basis for a reactor facility, provide bases and
requirements for NRC assessment of the potential for, and consequences
of, degradation of the reactor coolant pressure boundary (RCPB). The
applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC
31 (Fracture Prevention of Reactor Coolant Pressure Boundary), and GDC
32 (Inspection of Reactor Coolant Pressure Boundary). General Design
Criterion 14 specifies that the RCPB be designed, fabricated, erected,
and tested so as to have an extremely low probability of abnormal
leakage, of rapidly propagating failure, and of gross rupture. General
Design Criterion 31 specifies that the probability of rapidly
propagating fracture of the RCPB be minimized. General Design Criterion
32 specifies that components that are part of the RCPB have the
capability of being periodically inspected to assess their structural
and leak-tight integrity.
The NRC concludes that ASME BPV Code Case N-729-4, as conditioned,
shall be mandatory in order to ensure that the requirements of the GDC
are satisfied. Imposition of ASME BPV Code Case N-729-4, with
conditions, ensures that the ASME BPV Code-allowable limits will not be
exceeded, leakage will likely not occur, and potential flaws will be
detected before they challenge the structural or leak-tight integrity
of the RPV upper head within current nondestructive examination
limitations. The NRC concludes that the regulatory framework for
providing adequate protection of public health and safety is
accomplished by the incorporation of ASME BPV Code Case N-729-4 into
Sec. 50.55a, as conditioned. All current licensees of U.S. pressurized
water reactors will be required to implement ASME BPV Code Case N-729-
4, as conditioned. The Code Case provisions on examination requirements
for RPV upper heads are essentially the same as those established under
ASME BPV Code Case N-729-1, as conditioned. One exception is the
condition in Sec. 50.55a(g)(6)(ii)(D)(3), which will require, for
upper heads with Alloy 600 penetration nozzles, that bare metal visual
examinations be performed each outage in accordance with Table 1 of
ASME BPV Code Case N-729-4. Accordingly, the NRC imposition of the ASME
BPV Code Case N-729-4, as conditioned, may be deemed to be a
modification of the procedures to operate a facility resulting from the
imposition of the new regulation, and as such, this rulemaking
provision may be considered backfitting under Sec. 50.109(a)(1).
The NRC continues to find that inspections of RPV upper heads,
their penetration nozzles, and associated partial penetration welds are
necessary for adequate protection of public health and safety and that
the requirements of ASME BPV Code Case N-729-4, as conditioned,
represent an acceptable approach, developed, in part, by a voluntary
consensus standards body for performing future inspections. The NRC
concludes that approval of ASME BPV Code Case N-729-4, as conditioned,
by incorporation by reference of the Code Case into Sec. 50.55a, is
necessary to ensure that the facility provides adequate protection to
the health and safety of the public and constitutes a redefinition of
the requirements necessary to provide reasonable assurance of adequate
protection of public health and safety. Therefore, a backfit analysis
need not be prepared for this portion of the rule in accordance with
Sec. 50.109(a)(4)(ii) and (iii).
ASME BPV Code Case N-770-2
Revise Sec. 50.55a(g)(6)(ii)(F), ``Examination requirements for
Class 1 piping and nozzle dissimilar metal butt welds.''
On June 9, 2011, the ASME approved the second revision of ASME BPV
Code Case N-770 (N-770-2). The NRC is updating the requirements of
Sec. 50.55a(g)(6)(ii)(F) to require licensees to implement ASME BPV
Code Case N-770-2, with conditions. The ASME BPV Code Case N-770-2
contains similar baseline and ISI requirements for unmitigated nickel-
alloy butt welds, and preservice and ISI requirements for mitigated
butt welds as N-770-1. However, N-770-2 also contains new requirements
for optimized weld overlays, a specific mitigation technique and
volumetric inspection coverage. Further, the NRC conditions on the use
of ASME BPV Code Case N-770-2 have been modified to address the changes
in the code case, clarify inspection coverage requirements and require
the development of inspection qualifications to allow complete weld
inspection coverage in the future.
The current regulatory requirements for the examination of ASME
Class 1 piping and nozzle dissimilar metal butt welds that use nickel-
alloy materials is provided in Sec. 50.55a(g)(6)(ii)(F). This section
was first created by rulemaking, dated June 21, 2011 (76 FR 36232), to
require licensees to implement ASME BPV Code Case N-770-1, with
conditions. The NRC added Sec. 50.55a(g)(6)(ii)(F) to require
licensees to implement ASME BPV Code Case N-770-1, with conditions,
instead of the inspections previously required by the ASME BPV Code,
Section XI. The action did constitute a backfit; however, the NRC
concluded that imposition of ASME BPV Code Case N-770-1, as
conditioned, constituted an adequate protection backfit.
The GDC for nuclear power plants (appendix A to 10 CFR part 50) or,
as appropriate, similar requirements in the licensing basis for a
reactor facility, provide bases and requirements for NRC assessment of
the potential for, and consequences of, degradation of the RCPB. The
applicable GDC include GDC 14 (Reactor Coolant Pressure Boundary), GDC
31 (Fracture Prevention of Reactor Coolant Pressure Boundary) and GDC
32 (Inspection of Reactor Coolant Pressure
[[Page 32973]]
Boundary). General Design Criterion 14 specifies that the RCPB be
designed, fabricated, erected, and tested so as to have an extremely
low probability of abnormal leakage, of rapidly propagating failure,
and of gross rupture. General Design Criterion 31 specifies that the
probability of rapidly propagating fracture of the RCPB be minimized.
General Design Criterion 32 specifies that components that are part of
the RCPB have the capability of being periodically inspected to assess
their structural and leak-tight integrity.
The NRC concludes that ASME BPV Code Case N-770-2, as conditioned,
must be imposed in order to ensure that the requirements of the GDC are
satisfied. Imposition of ASME BPV Code Case N-770-2, with conditions,
ensures that the requirements of the GDC are met for all mitigation
techniques currently in use for Alloy 82/182 butt welds because ASME
BPV Code-allowable limits will not be exceeded, leakage would likely
not occur and potential flaws will be detected before they challenge
the structural or leak-tight integrity of piping welds. All current
licensees of U.S. pressurized water reactors will be required to
implement ASME BPV Code Case N-770-2, as conditioned. The Code Case
provisions on examination requirements for ASME Class 1 piping and
nozzle nickel-alloy dissimilar metal butt welds are somewhat different
from those established under ASME BPV Code Case N-770-1, as
conditioned, and will require a licensee to modify its procedures for
inspection of ASME Class 1 nickel-alloy welds to meet these
requirements. Accordingly, the NRC imposition of the ASME BPV Code Case
N-770-2, as conditioned, may be deemed to be a modification of the
procedures to operate a facility resulting from the imposition of the
new regulation, and as such, this rulemaking provision may be
considered backfitting under Sec. 50.109(a)(1).
The NRC continues to find that ASME Class 1 nickel-alloy dissimilar
metal weld inspections are necessary for adequate protection of public
health and safety, and that the requirements of ASME BPV Code Case N-
770-2, as conditioned, represent an acceptable approach developed by a
voluntary consensus standards body for performing future ASME Class 1
nickel-alloy dissimilar metal weld inspections. The NRC concludes that
approval of ASME BPV Code Case N-770-2, as conditioned, by
incorporation by reference of the Code Case into Sec. 50.55a, is
necessary to ensure that the facility provides adequate protection to
the health and safety of the public and constitutes a redefinition of
the requirements necessary to provide reasonable assurance of adequate
protection of public health and safety. Therefore, a backfit analysis
need not be prepared for this portion of the rule in accordance with
Sec. 50.109(a)(4)(ii) and (iii).
Conclusion
The NRC finds that incorporation by reference into Sec. 50.55a of
the 2009 Addenda through 2013 Edition of Section III, Division 1, of
the ASME BPV Code, subject to the identified conditions; the 2009
Addenda through 2013 Edition of Section XI, Division 1, of the ASME BPV
Code, subject to the identified conditions; and the 2009 Edition
through the 2012 Edition of the OM Code, subject to the identified
conditions, does not constitute backfitting or represent an
inconsistency with any issue finality provisions in 10 CFR part 52.
The NRC finds that the incorporation by reference of Code Cases N-
824 and OMN-20 does not constitute backfitting or represent an
inconsistency with any issue finality provisions in 10 CFR part 52.
The NRC finds that the inclusion of a new condition on Code Case N-
729-4 and a new condition on Code Case N-770-2 constitutes backfitting
necessary for adequate protection.
X. 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).
XI. Finding of No Significant Environmental Impact: Environmental
Assessment
This final rule is in accordance with the NRC's policy to
incorporate by reference in Sec. 50.55a new editions and addenda of
the ASME BPV and OM Codes to provide updated rules for constructing and
inspecting components and testing pumps, valves, and dynamic restraints
(snubbers) in light-water nuclear power plants. The ASME Codes are
national voluntary consensus standards and are required by the NTTAA to
be used by government agencies unless the use of such a standard is
inconsistent with applicable law or otherwise impractical. The National
Environmental Policy Act (NEPA) requires Federal 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 proposed action and alternatives to
the proposed action (42 U.S.C. 4332(C); NEPA Sec. 102(C)).
The NRC has determined under NEPA, as amended, and the NRC's
regulations in subpart A of 10 CFR part 51, that this rule is not a
major Federal action significantly affecting the quality of the human
environment and, therefore, an environmental impact statement is not
required. The rulemaking does not significantly increase the
probability or consequences of accidents, no changes are being made in
the types of effluents that may be released off-site, and there is no
significant increase in public radiation exposure. The NRC estimates
the radiological dose to plant personnel performing the inspections
required by ASME BPV Code Case N-770-2 would be about 3 rem per plant
over a 10-year interval, and a one-time exposure for mitigating welds
of about 30 rem per plant. The NRC estimates the radiological dose to
plant personnel performing the inspections required by ASME BPV Code
Case N-729-4 would be about 3 rem per plant over a 10-year interval and
a one-time exposure for mitigating welds of about 30 rem per plant. As
required by 10 CFR part 20, and in accordance with current plant
procedures and radiation protection programs, plant radiation
protection staff will continue monitoring dose rates and would make
adjustments in shielding, access requirements, decontamination methods,
and procedures as necessary to minimize the dose to workers. The
increased occupational dose to individual workers stemming from the
ASME BPV Code Case N-770-2 and N-729-4 inspections must be maintained
within the limits of 10 CFR part 20 and as low as reasonably
achievable. Therefore, the NRC concludes that the increase in
occupational exposure would not be significant. This final rule does
not involve non-radiological plant effluents and has no other
environmental impacts. Therefore, no significant non-radiological
impacts are associated with this action. The determination of this
environmental assessment is that there will be no significant off-site
impact to the public from this action.
XII. Paperwork Reduction Act Statement
This final rule amends collections of information subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501
[[Page 32974]]
et seq.). The collections of information were approved by the Office of
Management and Budget (OMB), approval number 3150-0011.
Because the rule will reduce the burden for existing information
collections, the public burden for the information collections is
expected to be decreased by 58.5 hours per response. This reduction
includes the time for reviewing instructions, searching existing data
sources, gathering and maintaining the data needed, and completing and
reviewing the information collection.
The information collection is being conducted to document the plans
for and the results of ISI and IST programs. The records are generally
historical in nature and provide data on which future activities can be
based. The practical utility of the information collection for the NRC
is that appropriate records are available for auditing by NRC personnel
to determine if ASME BPV and OM Code provisions for construction,
inservice inspection, repairs, and inservice testing are being properly
implemented in accordance with Sec. 50.55a, or whether specific
enforcement actions are necessary. Responses to this collection of
information are generally mandatory under 10 CFR 50.55a.
You may submit comments on any aspect of the information
collection(s), including suggestions for reducing the burden, by the
following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2011-0088.
Mail comments to: Information Services Branch, Office of
the Chief Information Officer, Mail Stop: T-2F43, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001 or to Aaron Szabo,
Desk Officer, Office of Information and Regulatory Affairs (3150-0011),
NEOB-10202, Office of Management and Budget, Washington, DC 20503;
telephone: 202-395-3621, email: oira_submission@omb.eop.gov.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
the collection displays a currently valid OMB control number.
XIII. Congressional Review Act
This final rule is a rule as defined in the Congressional Review
Act (5 U.S.C. 801-808). However, OMB has not found it to be a major
rule as defined in the Congressional Review Act.
XIV. Voluntary Consensus Standards
Section 12(d)(3) of the National Technology Transfer and
Advancement Act of 1995, Public Law 104-113 (NTTAA), and implementing
guidance in OMB Circular A-119 (February 10, 1998), requires 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 which are not
acceptable to the agency.
In this final rule, the NRC is continuing its existing practice of
establishing requirements for the design, construction, operation, ISI
(examination), and IST of nuclear power plants by approving the use of
the latest editions and addenda of the ASME Codes in Sec. 50.55a. The
ASME Codes are voluntary consensus standards, developed by participants
with broad and varied interests, in which all interested parties
(including the NRC and licensees of nuclear power plants) participate.
Therefore, the NRC's incorporation by reference of the ASME Codes is
consistent with the overall objectives of the NTTAA and OMB Circular A-
119.
In this final rule, the NRC is also 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 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 in this final rule is consistent with the overall objectives of
the NTTAA and OMB Circular A-119.
As discussed in Section II of this document, ``Discussion,'' the
NRC is conditioning the use of certain provisions of the 2009 Addenda,
2010 Edition, 2011 Addenda, and 2013 Edition of the ASME BPV Code
Section III, Division 1 and Section XI, Division 1. The NRC is also
conditioning the use of certain provisions of the 2009 Edition, the
2011 Addenda, and the 2012 Edition of the OM Code, Division 1. This
final rule also includes various versions of quality assurance standard
NQA-1 and Code Cases N-729-4, N-770-2, N-824, OMN-20, N-513-3 Mandatory
Appendix I, and N-852. In addition, this final rule does not adopt
(``excludes'') certain provisions of the ASME Codes, as discussed in
this statement of considerations and in the regulatory analysis for
this rulemaking. The NRC staff's position is that this final rule
complies with the NTTAA and OMB Circular A-119 despite these conditions
and ``exclusions.''
If the NRC did not conditionally accept ASME editions, addenda, and
code cases, the NRC would disapprove these entirely. The effect would
be that licensees and applicants would submit a larger number of
requests for use of alternatives under Sec. 50.55a(z), requests for
relief under Sec. 50.55a(f) and (g), or requests for exemptions under
Sec. 50.12 and/or Sec. 52.7. These requests would likely include
broad scope requests for approval to issue the full scope of the ASME
Code editions and addenda which would otherwise be approved in this
final rule (i.e., the request would not be simply for approval of a
specific ASME Code provision with conditions). These requests would be
an unnecessary additional burden for both the licensee and the NRC,
inasmuch as the NRC has already determined that the ASME Codes and Code
Cases which are the subject of this final rule are acceptable for use
(in some cases with conditions). For these reasons, the NRC concludes
that this final rule's treatment of ASME Code editions and addenda, and
code cases and any conditions placed on them 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 U.S. voluntary consensus standards bodies for use within
the United States, which the NRC could incorporate by reference instead
of the ASME Codes. The NRC also did not
[[Page 32975]]
identify any voluntary consensus standards, developed by multinational
voluntary consensus standards bodies for use on a multinational basis,
which the NRC could incorporate by reference instead of the ASME Codes.
The NRC identified codes addressing the same subject as the ASME Codes
for use in individual countries. At least one country, Korea, directly
translated the ASME Code for use in that country. In other countries
(e.g., Japan), ASME Codes were the basis for development of the
country's codes, but the ASME Codes were substantially modified to
accommodate that country's regulatory system and reactor designs.
Finally, there are countries (e.g., the Russian Federation) where that
country's code was developed without regard to the ASME Code. However,
some of these codes may not meet the definition of a voluntary
consensus standard because they were developed by the state rather than
a voluntary consensus standards body. The NRC's evaluation of other
countries' codes to determine whether each code provides a comparable
or enhanced level of safety, when compared against the level of safety
provided under the ASME Codes, would require a significant expenditure
of agency resources. This expenditure does not seem justified, given
that substituting another country's code for the U.S. voluntary
consensus standard does not appear to substantially further the
apparent underlying objectives of the NTTAA.
In summary, this final rule satisfies the requirements of Section
12(d)(3) of the NTTAA and OMB Circular A-119.
XV. Incorporation by Reference--Reasonable Availability to Interested
Parties
The NRC is incorporating by reference recent editions and addenda
to the ASME Codes for nuclear power plants and a standard for quality
assurance. The NRC is also incorporating by reference six ASME Code
Cases. As described in the ``Background'' and ``Discussion'' sections
of this document, these materials provide rules for safety governing
the design, fabrication, and inspection of nuclear power plant
components.
The NRC is required by law to obtain approval for incorporation by
reference from the Office of the Federal Register (OFR). The OFR's
requirements for incorporation by reference are set forth in 1 CFR part
51. On November 7, 2014, the OFR adopted changes to its regulations
governing incorporation by reference (79 FR 66267). The OFR regulations
require an agency to include in a final rule a discussion of the ways
that the materials the agency incorporates by reference are reasonably
available to interested parties and how interested parties can obtain
the materials. The discussion in this section complies with the
requirement for final rules as set forth in Sec. 51.5(b).
The NRC considers ``interested parties'' to include all potential
NRC stakeholders, not only the individuals and entities regulated or
otherwise subject to the NRC's regulatory oversight. These NRC
stakeholders are not a homogenous group, so the considerations for
determining ``reasonable availability'' vary by class of interested
parties. The NRC identifies six classes of interested parties with
regard to the material to be incorporated by reference in an NRC rule:
Individuals and small entities regulated or otherwise
subject to the NRC's regulatory oversight who are subject to the
material to be incorporated by reference by rulemaking. This class also
includes applicants and potential applicants for licenses and other NRC
regulatory approvals. In this context, ``small entities'' has the same
meaning as a ``small entity'' under Sec. 2.810.
Large entities otherwise subject to the NRC's regulatory
oversight who are subject to the material to be incorporated by
reference by rulemaking. This class also includes applicants and
potential applicants for licenses and other NRC regulatory approvals.
In this context, ``large entities'' are those which do not qualify as a
``small entity'' under Sec. 2.810.
Non-governmental organizations with institutional
interests in the matters regulated by the NRC.
Other Federal agencies, states, local governmental bodies
(within the meaning of Sec. 2.315(c)).
Federally-recognized and State-recognized \3\ Indian
tribes.
---------------------------------------------------------------------------
\3\ State-recognized Indian tribes are not within the scope of
Sec. 2.315(c). However, for purposes of the NRC's compliance with 1
CFR 51.5, ``interested parties'' includes a broad set of
stakeholders including State-recognized Indian tribes.
---------------------------------------------------------------------------
Members of the general public (i.e., individual,
unaffiliated members of the public who are not regulated or otherwise
subject to the NRC's regulatory oversight) who may wish to gain access
to the materials that the NRC proposes to incorporate by reference in
order to participate in the rulemaking.
The NRC makes the materials to be incorporated by reference
available for inspection to all interested parties, by appointment, at
the NRC Technical Library, which is located at Two White Flint North,
11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-
7000; email: Library.Resource@nrc.gov.
Interested parties may purchase a copy of the materials from ASME
at Three Park Avenue, New York, NY 10016, or at the ASME Web site
https://www.asme.org/shop/standards. The materials are also accessible
through third-party subscription services such as IHS (15 Inverness Way
East, Englewood, CO 80112; https://global.ihs.com) and Thomson Reuters
Techstreet (3916 Ranchero Dr., Ann Arbor, MI 48108; https://www.techstreet.com). The purchase prices for individual documents range
from $225 to $720 and the cost to purchase all documents is
approximately $9,000.
For the class of interested parties constituting members of the
general public who wish to gain access to the materials to be
incorporated by reference in order to participate in the rulemaking,
the NRC recognizes that the $9,000 cost may be so high that the
materials could be regarded as not reasonably available for purposes of
commenting on this rulemaking, despite the NRC's actions to make the
materials available at the NRC's PDR.
Accordingly, the NRC sent a letter to the ASME on April 9, 2015,
requesting that they consider enhancing public access to these
materials during the public comment period. In an April 21, 2015,
letter to the NRC, the ASME agreed to make the materials available
online in a read-only electronic access format during the public
comment period.
During the public comment period, the ASME made publicly-available
the editions and addenda to the ASME Codes for nuclear power plants,
the ASME standard for quality assurance, and the ASME Code Cases which
the NRC proposed to incorporate by reference. The ASME made the
materials publicly-available in read-only format at the ASME Web site
https://go.asme.org/NRC.
The materials are available to all interested parties in multiple
ways and in a manner consistent with their interest in this rulemaking.
Therefore, the NRC concludes that the materials the NRC is
incorporating by reference in this rulemaking are reasonably available
to all interested parties.
XVI. Availability of Guidance
The NRC will not be issuing guidance for this rulemaking. The ASME
BPV Code and OM Code provide direction for the performance of
activities to satisfy the Code requirements for design, inservice
inspection, and inservice testing of nuclear power plant SSCs. In
addition, the NRC provides
[[Page 32976]]
guidance in this Federal Register notice for the implementation of the
new conditions on the ASME BPV Code and OM Code, as necessary. The NRC
has a number of standard review plans (SRPs), which provide guidance to
NRC reviewers and make communication and understanding of NRC review
processes available to members of the public and the nuclear power
industry. NUREG-0800, ``Review of Safety Analysis Reports for Nuclear
Power Plants,'' has numerous sections which discuss implementation of
various aspects of the ASME BPV Code and OM Code (e.g., Sections 3.2.2,
3.8.1, 3.8.2, 3.9.3, 3.9.6, 3.9.7, 3.9.8, 3.13, 5.2.1.1, 5.2.1.2,
5.2.4, and 6.6). The NRC also publishes Regulatory Guides and Generic
Communications (i.e., Regulatory Issue Summaries, Information Notices)
to communicate and clarify NRC technical or policy positions on
regulatory matters which may contain guidance relative to this
rulemaking.
Revision 2 of NUREG-1482, ``Guidelines for Inservice Testing at
Nuclear Power Plants,'' provides guidance for the development and
implementation of IST programs at nuclear power plants. With direction
provided in the ASME BPV and OM Codes, and guidance in this Federal
Register notice, the NRC has determined that preparation of a separate
guidance document is not necessary for this update to Sec. 50.55a.
However, the NRC will consider preparation of a revision to NUREG-1482
in the future to address the latest edition of the ASME OM Code
incorporated by reference in Sec. 50.55a.
XVII. Availability of Documents
The NRC is making the documents identified in Table 2 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 2--Availability of Documents
------------------------------------------------------------------------
ADAMS Accession No./
Document Federal Register
citation/Web link
------------------------------------------------------------------------
Proposed Rule Documents:
Proposed Rule--Federal Register Notice...... 80 FR 56820 (September
18, 2015).
Draft Regulatory Analysis................... ML14170B104.
Final Rule Documents:
Final Regulatory Analysis................... ML16130A522.
2017 Final Rule (10 CFR 50.55a) American ML16130A531.
Society of Mechanical Engineers Codes and
Code Cases: Analysis of Public Comments.
Related Documents:
Fatigue and Fracture Mechanics: 33rd Volume, https://www.astm.org/
ASTM STP 1417, W.G. Reuter and R.S. DIGITAL_LIBRARY/STP/
Piascik, Eds., ASTM International, West SOURCE_PAGES/
Conshohocken, PA, 2002. STP1417.htm.
Final Results from the CARINA Project on https://www.astm.org/
Crack Initiation and Arrest of Irradiated DIGITAL_LIBRARY/STP/
German RPV Steels for Neutron Fluences in PAGES/
the Upper Bound, H. Hein et al., ASTM STP157220130113.htm.
International, West Conshohocken, PA, June
2014.
Letter from Brian Thomas, NRC, to Michael ML15085A206.
Merker, ASME, ``Public Access to Material
the NRC Seeks to Incorporate by Reference
into its Regulations,'' April 9, 2015.
Letter from Mark Maxin, NRC, to Rick Libra, ML081680730.
BWRVIP Chairman, ``Safety Evaluation for
Electric Power Research Institute (EPRI)
Boiling Water Reactor (BWR) Vessel and
Internals Project (BWRVIP) Report 1003020
(BWRVIP-97), `BWR Vessel and Internals
Project, Guidelines for Performing Weld
Repairs to Irradiated BWR Internals' (TAC
No. MC3948),'' June 30, 2008.
Letter from Michael Merker, ASME, to Brian ML15112A064.
Thomas, NRC; April 21, 2015.
Licensee Event Report 50-338/2012-001-00.... ML12151A441.
NUREG-0800, ``Standard Review Plan for the ML070660036.
Review of Safety Analysis Reports for
Nuclear Power Plants, LWR Edition''.
NUREG-0800, Section 3.9.6, Revision 3, ML070720041.
``Functional Design, Qualification, and
Inservice Testing Programs for Pumps,
Valves, and Dynamic Restraints,'' March
2007.
NUREG-1482, Revision 2, ``Guidelines for ML13295A020.
Inservice Testing at Nuclear Power Plants:
Inservice Testing of Pumps and Valves and
Inservice Examination and Testing of
Dynamic Restraints (Snubbers) at Nuclear
Power Plants,'' October 2013.
NUREG-1800, Revision 2, ``Standard Review ML103490036.
Plan for Review of License Renewal
Applications for Nuclear Power Plants,''
December 2010.
NUREG-1801, Revision 2, ``Generic Aging ML103490041.
Lessons Learned (GALL) Report,'' December
2010.
NUREG-1950, ``Disposition of Public Comments ML11116A062.
and Technical Bases for Changes in the
License Renewal Guidance Documents NUREG-
1801 and NUREG-1800,'' April 2011.
NUREG-2124, ``Final Safety Evaluation Report ML12271A045.
Related to the Combined Licenses for Vogtle
Electric Generating Plant, Units 3 and 4,''
Section 3.9.6, ``Inservice Testing of Pumps
and Valves (Related to RG 1.206, Section
C.III.1, Chapter 3, C.I.3.9.6, `Functional
Design, Qualification, and Inservice
Testing Programs for Pumps, Valves, and
Dynamic Restraints')''.
NUREG/CR-6860, ``An Assessment of Visual ML043630040.
Testing,'' November 2004.
NUREG/CR-6933, ``Assessment of Crack ML071020410 and
Detection in Heavy-Walled Cast Stainless ML071020414.
Steel Piping Welds Using Advanced Low-
Frequency Ultrasonic Methods,'' March 2007.
NUREG/CR-7122, ``An Evaluation of Ultrasonic ML12087A004.
Phased Array Testing for Cast Austenitic
Stainless Steel Pressurizer Surge Line
Piping Welds,'' March 2012.
NRC Generic Letter 89-10, ``Safety-Related ML031150300.
Motor-Operated Valve Testing and
Surveillance,'' June 1989.
NRC Generic Letter 90-05, ``Guidance for ML031140590.
Performing Temporary Non-Code Repair of
ASME Code Class 1, 2, and 3 Piping (Generic
Letter 90-05),'' June 1990.
NRC Meeting Summary of June 5-7, 2013, ML14003A230.
Annual Materials Programs Technical
Information Exchange Public Meeting.
NRC Meeting Summary of January 19, 2016, ML16050A383.
Category 2 Public Meeting with Industry
Representatives to Discuss Welding on
Neutron Irradiated Ferritic and Austenitic
Materials.
[[Page 32977]]
NRC Meeting Summary of March 2, 2016, Public ML16069A408.
Meeting on Stakeholder Comments on the
Proposed Rule.
NRC Staff Memorandum, ``Consolidation of ML003708048.
SECY-94-084 and SECY-95-132,'' July 24,
1995.
NRC Memorandum, ``Staff Requirements-- ML003755050.
Affirmation Session, 11:30 a.m., Friday,
September 10, 1999, Commissioners'
Conference Room, One White Flint North,
Rockville, Maryland (Open to Public
Attendance),'' September 10, 1999.
NRC Regulatory Guide 1.28, Revision 4, ML100160003.
``Quality Assurance Program Criteria
(Design and Construction),'' June 2010.
NRC Regulatory Guide 1.147, Revision 17, ML13339A689.
``Inservice Inspection Code Case
Acceptability, ASME Section XI, Division
1,'' August 2014.
NRC Regulatory Guide 1.174, Revision 2, ``An ML100910006.
Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on
Plant-Specific Changes to the Licensing
Basis,'' May 2011.
NRC Regulatory Guide 1.175, ``An Approach ML003740149.
for Plant-Specific, Risk-Informed
Decisionmaking: Inservice Testing,'' August
1998.
NRC Regulatory Guide 1.192, Revision 1, ML13340A034.
``Operation and Maintenance Code Case
Acceptability, ASME OM Code,'' August 2014.
NRC Regulatory Guide 1.200, Revision 2, ``An ML090410014.
Approach for Determining the Technical
Adequacy of Probabilistic Risk Assessment
Results for Risk-Informed Activities,''
March 2009.
NRC Regulatory Guide 1.201, Revision 1, ML061090627.
``Guidelines for Categorizing Structures,
Systems, and Components in Nuclear Power
Plants According to Their Safety
Significance,'' May 2006.
NRC Regulatory Information Conference, https://www.nrc.gov/
Recent Operating Reactors Materials Issues, public-involve/
Presentation Materials, 2013. conference-symposia/
ric/past/2013/docs/
abstracts/
sessionabstract-
19.html.
NRC Regulatory Issue Summary 2013-07, ``NRC ML13003A207.
Staff Position on the Use of American
Society of Mechanical Engineers
Certification Mark,'' May 28, 2013.
Relief Request REP-1 U2, Revision 2......... ML13232A308.
SECY-90-016, ``Evolutionary Light Water ML003707849.
Reactor (LWR) Certification Issues and
Their Relationship to Current Regulatory
Requirements''.
SECY-93-087, ``Policy, Technical, and ML003708021.
Licensing Issues Pertaining to Evolutionary
and Advanced Light-Water Reactor (ALWR)
Designs''.
SECY-94-084, ``Policy and Technical Issues ML003708068.
Associated with the Regulatory Treatment of
Non-Safety Systems in Passive Plant
Designs''.
SECY-95-132, ``Policy and Technical Issues ML003708005.
Associated with the Regulatory Treatment of
Non-Safety Systems (RTNSS) in Passive Plant
Designs (SECY-94-084)''.
Vogtle Electric Generating Plant, Units 3 ML14183B276.
and 4, Updated Final Safety Analysis
Report, Revision 3, Chapter 3, Section 3.9,
Mechanical Systems and Components.
------------------------------------------------------------------------
List of Subjects in 10 CFR Part 50
Administrative practice and procedure, Antitrust, Classified
information, Criminal penalties, Education, Fire prevention, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Penalties, Radiation protection,
Reactor siting criteria, Reporting and recordkeeping requirements,
Whistleblowing.
For the reasons set forth in the preamble, and under the authority
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. The authority citation for part 50 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103,
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135,
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236,
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs.
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste
Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504
note; Sec. 109, Pub. L. 96-295, 94 Stat. 783.
0
2. In Sec. 50.55a:
0
a. Revise paragraphs (a) introductory text, (a)(1)(i) introductory text
and (a)(1)(i)(E)(12) and (13) and add paragraphs (a)(1)(i)(E)(14)
through (17);
0
b. Revise paragraphs (a)(1)(ii) introductory text and (a)(1)(ii)(C)(48)
and (49) and add paragraphs (a)(1)(ii)(C)(50) through (53);
0
c. Revise paragraphs (a)(1)(iii)(A) through (C) and add paragraphs
(a)(1)(iii)(D) through (G);
0
d. Revise paragraph (a)(1)(iv) introductory text and add paragraphs
(a)(1)(iv)(B) and (C);
0
e. Add paragraph (a)(1)(v);
0
f. Revise paragraphs (b) introductory text, (b)(1) introductory text
and (b)(1)(ii), (iv), and (vii) and add paragraphs (b)(1)(viii) and
(ix);
0
g. Revise paragraphs (b)(2) introductory text, (b)(2)(vi), and
(b)(2)(viii) introductory text, add paragraphs (b)(2)(viii)(H) and (I),
revise paragraphs (b)(2)(ix) introductory text, (b)(2)(ix)(D), and
(b)(2)(x) and (xii), add paragraph (b)(2)(xviii)(D), revise paragraphs
(b)(2)(xxi)(A) and (b)(2)(xxiii), add and reserve paragraph
(b)(2)(xxx), and add paragraphs (b)(2)(xxxi) through (xxxvii);
0
h. Revise paragraphs (b)(3) introductory text and (b)(3)(i) and (ii),
add paragraph (b)(3)(iii), revise paragraph (b)(3)(iv) introductory
text, and add paragraphs (b)(3)(vii) through (xi);
0
i. Revise paragraphs (b)(4) introductory text and (b)(5) and (6);
0
j. Revise paragraphs (f) heading and introductory text, (f)(2),
(f)(3)(iii)(A) and (B), (f)(3)(iv)(A) and (B), (f)(4) introductory
text, and (f)(4)(i) and (ii); and
0
k. Revise paragraphs (g) heading and introductory text, (g)(2), and
(g)(3)
[[Page 32978]]
heading, remove paragraph (g)(3) introductory text, revise paragraphs
(g)(3)(i), (ii), and (v), (g)(4)(i) and (ii), and (g)(6)(ii)(D)(1)
through (4), remove paragraphs (g)(6)(ii)(D)(5) and (6), revise
paragraphs (g)(6)(ii)(F)(1) through (10), and add paragraphs
(g)(6)(ii)(F)(11) through (13).
The revisions and additions read as follows:
Sec. 50.55a Codes and standards.
(a) Documents approved for incorporation by reference. The
standards listed in this paragraph (a) 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, by appointment, at the NRC Technical Library,
which is located at Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland 20852; telephone: 301-415-7000; email:
Library.Resource@nrc.gov; 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) * * *
(i) ASME Boiler and Pressure Vessel Code, Section III. The editions
and addenda for Section III of the ASME Boiler and Pressure Vessel Code
(excluding Nonmandatory Appendices) (referred to herein as ASME BPV
Code) are listed in this paragraph (a)(1)(i), but limited by those
provisions identified in paragraph (b)(1) of this section.
* * * * *
(E) * * *
(12) 2007 Edition,
(13) 2008 Addenda,
(14) 2009b Addenda (including Subsection NCA; and Division 1
subsections NB through NH and Appendices),
(15) 2010 Edition (including Subsection NCA; and Division 1
subsections NB through NH and Appendices),
(16) 2011a Addenda (including Subsection NCA; and Division 1
subsections NB through NH and Appendices), and
(17) 2013 Edition (including Subsection NCA; and Division 1
subsections NB through NH and Appendices).
(ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions
and addenda for Section XI of the ASME BPV Code are listed in this
paragraph (a)(1)(ii), but limited by those provisions identified in
paragraph (b)(2) of this section.
* * * * *
(C) * * *
(48) 2007 Edition,
(49) 2008 Addenda,
(50) 2009b Addenda,
(51) 2010 Edition,
(52) 2011a Addenda (Excluding Article IWB-2000: IWB-2500
``Examination and Inspection: Examination and Pressure Test
Requirements,'' Table IWB-2500-1 ``Examination Categories,'' Item
numbers B5.11 and B5.71), and
(53) 2013 Edition (Excluding Article IWB-2000: IWB-2500
``Examination and Inspection: Examination and Pressure Test
Requirements,'' Table IWB-2500-1 (B-F) ``Examination Category B-F,
Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles,'' Item
numbers B5.11 and B5.71; Article IWB-3000 ``Acceptance Standards,''
IWB-3100 ``Evaluation of Examination Results,'' IWB-3110 ``Preservice
Volumetric and Surface Examinations,'' IWB-3112 ``Acceptance,''
paragraph (a)(3); and Article IWC-3000 ``Acceptance Standards,'' IWC-
3100 ``Evaluation of Examination Results,'' IWC-3110 ``Preservice
Volumetric and Surface Examinations,'' IWC-3112 ``Acceptance,''
paragraph (a)(3)).
(iii) * * *
(A) ASME BPV Code Case N-513-3 Mandatory Appendix I. ASME BPV Code
Case N-513-3, ``Evaluation Criteria for Temporary Acceptance of Flaws
in Moderate Energy Class 2 or 3 Piping Section XI, Division 1,''
Mandatory Appendix I, ``Relations for Fm, Fb, and F for Through-Wall
Flaws'' (Approval Date: January 26, 2009). ASME BPV Code Case N-513-3
Mandatory Appendix I is referenced in paragraph (b)(2)(xxxiv)(B) of
this section.
(B) ASME BPV Code Case N-722-1. ASME BPV 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.
(C) ASME BPV Code Case N-729-4. ASME BPV Code Case N-729-4,
``Alternative Examination Requirements for PWR Reactor Vessel Upper
Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds
Section XI, Division 1'' (Approval Date: June 22, 2012), with the
conditions in paragraph (g)(6)(ii)(D) of this section.
(D) ASME BPV Code Case N-770-2. ASME BPV Code Case N-770-2,
``Alternative Examination Requirements and Acceptance Standards for
Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS
N06082 or UNS W86182 Weld Filler Material With or Without Application
of Listed Mitigation Activities Section XI, Division 1'' (Approval
Date: June 9, 2011), with the conditions in paragraph (g)(6)(ii)(F) of
this section.
(E) ASME BPV Code Case N-824. ASME BPV Code Case N-824,
``Ultrasonic Examination of Cast Austenitic Piping Welds From the
Outside Surface Section XI, Division 1'' (Approval Date: October 16,
2012), with the conditions in paragraphs (b)(2)(xxxvii)(A) through (D)
of this section.
(F) ASME BPV Code Case N-852. ASME BPV Code Case N-852,
``Application of the ASME NPT Stamp, Section III, Division 1; Section
III, Division 2; Section III, Division 3; Section III, Division 5''
(Approval Date: February 9, 2015). ASME BPV Code Case N-852 is
referenced in paragraph (b)(1)(ix) of this section.
(G) ASME OM Code Case OMN-20. ASME OM Code Case OMN-20, ``Inservice
Test Frequency,'' in the 2012 Edition of the ASME OM Code. OMN-20 is
referenced in paragraph (b)(3)(x) of this section.
(iv) ASME Operation and Maintenance Code. The editions and addenda
for the ASME Operation and Maintenance of Nuclear Power Plants (various
edition titles referred to herein as ASME OM Code) are listed in this
paragraph (a)(1)(iv), but limited by those provisions identified in
paragraph (b)(3) of this section.
* * * * *
(B) ``Operation and Maintenance of Nuclear Power Plants, Division
1: Section IST Rules for Inservice Testing of Light-Water Reactor Power
Plants:''
(1) 2009 Edition; and
(2) 2011 Addenda.
(C) ``Operation and Maintenance of Nuclear Power Plants, Division
1: OM Code: Section IST:''
(1) 2012 Edition.
(2) [Reserved]
(v) ASME Quality Assurance Requirements. (A) ASME NQA-1, ``Quality
Assurance Program Requirements for Nuclear Facilities:''
(1) NQA-1--1983 Edition;
(2) NQA-1a--1983 Addenda;
(3) NQA-1b--1984 Addenda;
(4) NQA-1c--1985 Addenda;
(5) NQA-1--1986 Edition;
(6) NQA-1a--1986 Addenda;
(7) NQA-1b--1987 Addenda;
(8) NQA-1c--1988 Addenda;
(9) NQA-1--1989 Edition;
[[Page 32979]]
(10) NQA-1a--1989 Addenda;
(11) NQA-1b--1991 Addenda; and
(12) NQA-1c--1992 Addenda.
(B) ASME NQA-1, ``Quality Assurance Requirements for Nuclear
Facility Applications:''
(1) NQA-1--1994 Edition;
(2) NQA-1--2008 Edition; and
(3) NQA-1a--2009 Addenda.
* * * * *
(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 BPV Code and the ASME
OM Code as specified in this paragraph (b). 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 10
CFR part 52 is subject to the following conditions. As used in this
section, references to Section III refer to Section III of the ASME BPV
Code and include the 1963 Edition through 1973 Winter Addenda and the
1974 Edition (Division 1) through the 2013 Edition (Division 1),
subject to the following conditions:
* * * * *
(ii) Section III condition: Weld leg dimensions. When applying the
1989 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1) of this section, applicants and licensees
may not apply the Section III provisions identified in Table I of this
section for welds with leg size less than 1.09 tn:
Table I--Prohibited Code Provisions
------------------------------------------------------------------------
Editions and addenda Code provision
------------------------------------------------------------------------
1989 Addenda through 2013 Edition..... Subparagraph NB-3683.4(c)(1);
Subparagraph NB-3683.4(c)(2).
1989 Addenda through 2003 Addenda..... Note 11 to Figure NC-3673.2(b)-
1; Note 11 to Figure ND-
3673.2(b)-1.
2004 Edition through 2010 Edition..... Note 13 to Figure NC-3673.2(b)-
1; Note 13 to Figure ND-
3673.2(b)-1.
2011 Addenda through 2013 Edition..... Note 11 to Table NC-3673.2(b)-1;
Note 11 to Table ND-3673.2(b)-
1.
------------------------------------------------------------------------
* * * * *
(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
Facility Applications,'' 1994 Edition, 2008 Edition, and the 2009-1a
Addenda specified in either NCA-4000 or NCA-7000 of that edition and
addenda of Section III may be used by an applicant or licensee,
provided that the administrative, quality, and technical provisions
contained in that edition and addenda of Section III are used in
conjunction with the applicant's or licensee's appendix B to this part
quality assurance program; and that the applicant's or licensee's
Section III activities comply with those commitments contained in the
applicant's or licensee's quality assurance program description. Where
NQA-1 and Section III do not address the commitments contained in the
applicant's or licensee's appendix B quality assurance program
description, those licensee commitments must be applied to Section III
activities.
* * * * *
(vii) Section III condition: Capacity certification and
demonstration of function of incompressible-fluid pressure-relief
valves. When applying the 2006 Addenda through the 2013 Edition,
applicants and licensees may use paragraph NB-7742, except that
paragraph NB-7742(a)(2) may not be used. For a valve design of a single
size to be certified over a range of set pressures, the demonstration
of function tests under paragraph NB-7742 must be conducted as
prescribed in NB-7732.2 on two valves covering the minimum set pressure
for the design and the maximum set pressure that can be accommodated at
the demonstration facility selected for the test.
(viii) Section III condition: Use of ASME certification marks. When
applying editions and addenda earlier than the 2011 Addenda to the 2010
Edition, licensees may use either the ASME BPV Code Symbol Stamps or
the ASME Certification Marks with the appropriate certification
designators and class designators as specified in the 2013 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1) of this section.
(ix) Section III Condition: NPT Code Symbol Stamps. Licensees may
use the NPT Code Symbol Stamp with the letters arranged horizontally as
specified in ASME BPV Code Case N-852 for the service life of a
component that had the NPT Code Symbol Stamp applied during the time
period from January 1, 2005, through December 31, 2015.
(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 BPV Code, and include the 1970 Edition through the 1976 Winter
Addenda and the 1977 Edition through the 2013 Edition, subject to the
following conditions:
* * * * *
(vi) Section XI condition: Effective edition and addenda of
Subsection IWE and Subsection IWL. Licensees that implemented the
expedited examination of containment, in accordance with Subsection IWE
and Subsection IWL, during the period from September 9, 1996, to
September 9, 2001, 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.
* * * * *
(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 up to and including the
2008 Addenda must apply paragraph (b)(2)(viii)(E) of this section.
Applicants or licensees applying Subsection IWL, 2007 Edition with the
2009 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section, must apply
paragraphs (b)(2)(viii)(H) and (I) of this section.
* * * * *
[[Page 32980]]
(H) Concrete containment examinations: Eighth provision. For each
inaccessible area of concrete identified for evaluation under IWL-
2512(a), or identified as susceptible to deterioration under IWL-
2512(b), the licensee must provide the applicable information specified
in paragraphs (b)(2)(viii)(E)(1), (2), and (3) of this section in the
ISI Summary Report required by IWA-6000.
(I) Concrete containment examinations: Ninth provision. During the
period of extended operation of a renewed license under part 54 of this
chapter, the licensee must perform the technical evaluation under IWL-
2512(b) of inaccessible below-grade concrete surfaces exposed to
foundation soil, backfill, or groundwater at periodic intervals not to
exceed 5 years. In addition, the licensee must examine representative
samples of the exposed portions of the below-grade concrete, when such
below-grade concrete is excavated for any reason.
(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 (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 (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 or licensees applying Subsection IWE, 2007 Edition through
the latest edition and 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.
* * * * *
(D) Metal containment examinations: Fourth provision. This
paragraph (b)(2)(ix)(D) may be used as an alternative to the
requirements of IWE-2430. 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:
(1) A description of each flaw or area, including the extent of
degradation, and the conditions that led to the degradation;
(2) The acceptability of each flaw or area and the need for
additional examinations to verify that similar degradation does not
exist in similar components;
(3) A description of necessary corrective actions; and
(4) The number and type of additional examinations to ensure
detection of similar degradation in similar components.
* * * * *
(x) Section XI condition: Quality assurance. When applying the
editions and addenda later than the 1989 Edition of ASME BPV Code,
Section XI, the edition and addenda of NQA-1, ``Quality Assurance
Requirements for Nuclear Facility Applications,'' 1994 Edition, the
2008 Edition, and the 2009-1a Addenda specified in either IWA-1400 or
Table IWA 1600-1 of that edition and addenda of Section XI, may be used
by a licensee provided that the licensee uses its appendix B to this
part quality assurance program in conjunction with Section XI
requirements and the commitments contained in the licensee's quality
assurance program description. Where NQA-1 and Section XI do not
address the commitments contained in the licensee's appendix B quality
assurance program description, those licensee commitments must be
applied to Section XI activities.
* * * * *
(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 approved for use on irradiated material
with the following conditions:
(A) Underwater welding: First provision. Licensees must obtain NRC
approval in accordance with paragraph (z) of this section regarding the
welding technique to be used prior to performing welding on ferritic
material exposed to fast neutron fluence greater than 1 x 10\17\ n/
cm\2\ (E > 1 MeV).
(B) Underwater welding: Second provision. Licensees must obtain NRC
approval in accordance with paragraph (z) of this section regarding the
welding technique to be used prior to performing welding on austenitic
material other than P-No. 8 material exposed to thermal neutron fluence
greater than 1 x 10\17\ n/cm\2\ (E < 0.5 eV). Licensees must obtain NRC
approval in accordance with paragraph (z) regarding the welding
technique to be used prior to performing welding on P-No. 8 austenitic
material exposed to thermal neutron fluence greater than 1 x 10\17\ n/
cm\2\ (E < 0.5 eV) and measured or calculated helium concentration of
the material greater than 0.1 atomic parts per million.
* * * * *
(xviii) * * *
(D) NDE personnel certification: Fourth provision. The use of
Appendix VII and Subarticle VIII-2200 of the 2011 Addenda and 2013
Edition of Section XI of the ASME BPV Code is prohibited. When using
ASME BPV Code, Section XI editions and addenda later than the 2010
Edition, licensees and applicants must use the prerequisites for
ultrasonic examination personnel certifications in Table VII-4110-1 and
Subarticle VIII-2200, Appendix VIII in the 2010 Edition.
* * * * *
(xxi) * * *
(A) Table IWB-2500-1 examination requirements: First provision. The
provisions of Table IWB 2500-1, Examination Category B-D, Full
Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60
(Inspection Program A) and Items B3.120 and B3.140 (Inspection Program
B) of the 1998 Edition must be applied when using the 1999 Addenda
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(ii) of this section. A visual examination with
magnification that has a resolution sensitivity to resolve 0.044 inch
(1.1 mm) lower case characters without an ascender or descender (e.g.,
a, e, n, v), 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.
* * * * *
(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 2009 Addenda, is prohibited.
* * * * *
(xxx) [Reserved]
(xxxi) Section XI condition: Mechanical clamping devices. When
[[Page 32981]]
installing a mechanical clamping device on an ASME BPV Code class
piping system, Appendix W of Section XI shall be treated as a mandatory
appendix and all of the provisions of Appendix W shall be met for the
mechanical clamping device being installed. Additionally, use of IWA-
4131.1(c) of the 2010 Edition of Section XI and IWA-4131.1(d) of the
2011 Addenda of the 2010 Edition and later versions of Section XI is
prohibited on small item Class 1 piping and portions of a piping system
that form the containment boundary.
(xxxii) Section XI condition: Summary report submittal. When using
ASME BPV Code, Section XI, 2010 Edition through the latest edition and
addenda incorporated by reference in paragraph (a)(1)(ii) of this
section, Summary Reports described in IWA-6000 must be submitted to the
NRC as described in IWA-6240(a) and IWA-6240(b). Preservice inspection
summary reports shall be submitted prior to the date of placement of
the unit into commercial service and inservice inspection summary
reports shall be submitted within 90 calendar days of the completion of
each refueling outage.
(xxxiii) Section XI condition: Risk-Informed allowable pressure.
The use of Paragraph G-2216 in Appendix G in the 2011 Addenda and later
editions and addenda of the ASME BPV Code, Section XI is prohibited.
(xxxiv) Section XI condition: Nonmandatory Appendix U. When using
Nonmandatory Appendix U of the 2013 Edition of the ASME BPV Code,
Section XI the following conditions apply:
(A) The repair or replacement activities temporarily deferred under
the provisions of Nonmandatory Appendix U must be performed during the
next scheduled refueling outage.
(B) In lieu of the appendix referenced in paragraph U-S1-4.2.1(c)
of Appendix U the mandatory appendix in ASME BPV Code Case N-513-3 must
be used.
(xxxv) Section XI condition: Use of RTT0 in the KIa and KIc
equations. When using the 2013 Edition of the ASME BPV Code, Section
XI, Appendix A, paragraph A-4200, if T0 is available, then
RTT0 may be used in place of RTNDT for
applications using the KIc equation and the associated
KIc curve, but not for applications using the KIa
equation and the associated KIa curve.
(xxxvi) Section XI condition: Fracture toughness of irradiated
materials. When using the 2013 Edition of the ASME BPV Code, Section
XI, Appendix A paragraph A-4400, the licensee shall obtain NRC approval
under paragraph (z) of this section before using irradiated
T0 and the associated RTT0 in establishing
fracture toughness of irradiated materials.
(xxxvii) Section XI condition: ASME BPV Code Case N-824. Licensees
may use the provisions of ASME BPV Code Case N-824, ``Ultrasonic
Examination of Cast Austenitic Piping Welds From the Outside Surface
Section XI, Division 1,'' subject to the following conditions.
(A) Ultrasonic examinations must be spatially encoded.
(B) Instead of Paragraph 1(c)(1)(-a), licensees shall use dual,
transmit-receive, refracted longitudinal wave, multi-element phased
array search units.
(C) Instead of Paragraph 1(c)(1)(-c)(-2), licensees shall use a
phased array search unit with a center frequency of 500 kHz with a
tolerance of 20 percent.
(D) Instead of Paragraph 1(c)(1)(-d), the phased array search unit
must produce angles including, but not limited to, 30 to 55 degrees
with a maximum increment of 5 degrees.
(3) Conditions on ASME OM Code. As used in this section, references
to the ASME OM Code are to the ASME OM Code, Subsections ISTA, ISTB,
ISTC, ISTD, ISTE, and ISTF; Mandatory Appendices I, II, III, and V; and
Nonmandatory Appendices A through H and J through M, in the 1995
Edition through the 2012 Edition, as specified in paragraph (a)(1)(iv)
of this section. Mandatory appendices must be used if required by the
OM Code; nonmandatory appendices are approved for use by the NRC but
need not be used. The following conditions are applicable when
implementing the ASME OM Code:
(i) OM condition: Quality assurance. When applying editions and
addenda of the ASME OM Code, the requirements of ASME Standard NQA-1,
``Quality Assurance Requirements for Nuclear Facility Applications,''
1994 Edition, 2008 Edition, and 2009-1a Addenda, are acceptable as
permitted by either ISTA 1.4 of the 1995 Edition through 1997 Addenda
or ISTA-1500 of the 1998 Edition through the latest edition and addenda
of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv)
of this section, provided the licensee uses its appendix B to this part
quality assurance program in conjunction with the ASME OM Code
requirements and the commitments contained in the licensee's quality
assurance program description. Where NQA-1 and the ASME OM Code do not
address the commitments contained in the licensee's appendix B quality
assurance program description, the commitments must be applied to ASME
OM Code activities.
(ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees
must comply with the provisions for testing MOVs in ASME 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 MOVs continue to be capable of performing their
design basis safety functions. Licensees implementing ASME OM Code,
Mandatory Appendix III, ``Preservice and Inservice Testing of Active
Electric Motor Operated Valve Assemblies in Light-Water Reactor Power
Plants,'' of the 2009 Edition, 2011 Addenda, and 2012 Edition shall
comply with the following conditions:
(A) MOV diagnostic test interval. Licensees shall evaluate the
adequacy of the diagnostic test intervals established for MOVs within
the scope of ASME OM Code, Appendix III, not later than 5 years or
three refueling outages (whichever is longer) from initial
implementation of ASME OM Code, Appendix III.
(B) MOV testing impact on risk. Licensees shall ensure that the
potential increase in core damage frequency and large early release
frequency associated with the extension is acceptably small when
extending exercise test intervals for high risk MOVs beyond a quarterly
frequency.
(C) MOV risk categorization. When applying Appendix III to the ASME
OM Code, licensees shall categorize MOVs according to their safety
significance using the methodology described in ASME OM Code Case OMN-
3, ``Requirements for Safety Significance Categorization of Components
Using Risk Insights for Inservice Testing of LWR Power Plants,''
subject to the conditions applicable to OMN-3 which are set forth in
Regulatory Guide 1.192, or using an MOV risk ranking methodology
accepted by the NRC on a plant-specific or industry-wide basis in
accordance with the conditions in the applicable safety evaluation.
(D) MOV stroke time. When applying Paragraph III-3600, ``MOV
Exercising Requirements,'' of Appendix III to the ASME OM Code,
licensees shall verify that the stroke time of MOVs specified in plant
technical specifications satisfies the assumptions in the plant's
safety analyses.
(iii) OM condition: New reactors. In addition to complying with the
provisions in the ASME OM Code with the conditions specified in
paragraph (b)(3) of this section, holders of
[[Page 32982]]
operating licenses for nuclear power reactors that received
construction permits under this part on or after the date 12 months
after August 17, 2017, and holders of combined licenses issued under 10
CFR part 52, whose initial fuel loading occurs on or after the date 12
months after August 17, 2017, shall also comply with the following
conditions, as applicable:
(A) Power-operated valves. Licensees shall periodically verify the
capability of power-operated valves to perform their design-basis
safety functions.
(B) Check valves. Licensees must perform bi-directional testing of
check valves within the IST program where practicable.
(C) Flow-induced vibration. Licensees shall monitor flow-induced
vibration from hydrodynamic loads and acoustic resonance during
preservice testing or inservice testing to identify potential adverse
flow effects on components within the scope of the IST program.
(D) High risk non-safety systems. Licensees shall assess the
operational readiness of pumps, valves, and dynamic restraints within
the scope of the Regulatory Treatment of Non-Safety Systems for
applicable reactor designs.
(iv) OM condition: Check valves (Appendix II). Licensees applying
Appendix II, ``Check Valve Condition Monitoring Program,'' of the ASME
OM Code, 1995 Edition with the 1996 and 1997 Addenda, shall satisfy the
requirements of paragraphs (b)(3)(iv)(A) through (C) of this section.
Licensees applying Appendix II, 1998 Edition through the 2012 Edition,
shall satisfy the requirements of paragraphs (b)(3)(iv)(A), (B), and
(D) of this section. Appendix II of the ASME OM Code, 2003 Addenda
through the 2012 Edition, is acceptable for use with the following
requirements. Trending and evaluation shall support the determination
that the valve or group of valves is capable of performing its intended
function(s) over the entire interval. At least one of the Appendix II
condition monitoring activities for a valve group shall be performed on
each valve of the group at approximate equal intervals not to exceed
the maximum interval shown in the following table:
Table II--Maximum Intervals for Use When Applying Interval Extensions
------------------------------------------------------------------------
Maximum Maximum
interval between interval between
activities of activities of
Group size member valves each valve in
in the groups the group
(years) (years)
------------------------------------------------------------------------
>=4................................. 4.5 16
3................................... 4.5 12
2................................... 6 12
1................................... Not applicable 10
------------------------------------------------------------------------
* * * * *
(vii) OM condition: Subsection ISTB. Subsection ISTB, 2011 Addenda,
is prohibited for use.
(viii) OM condition: Subsection ISTE. Licensees may not implement
the risk-informed approach for inservice testing (IST) of pumps and
valves specified in Subsection ISTE, ``Risk-Informed Inservice Testing
of Components in Light-Water Reactor Nuclear Power Plants,'' in the
ASME OM Code, 2009 Edition, 2011 Addenda, or 2012 Edition, without
first obtaining NRC authorization to use Subsection ISTE as an
alternative to the applicable IST requirements in the ASME OM Code,
pursuant to paragraph (z) of this section.
(ix) OM condition: Subsection ISTF. Licensees applying Subsection
ISTF, 2012 Edition, shall satisfy the requirements of Mandatory
Appendix V, ``Pump Periodic Verification Test Program,'' of the ASME OM
Code, 2012 Edition. Subsection ISTF, 2011 Addenda, is prohibited for
use.
(x) OM condition: ASME OM Code Case OMN-20. Licensees may implement
ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' which is
incorporated by reference in paragraph (a)(1)(iii)(G) of this section,
for editions and addenda of the ASME OM Code listed in paragraph
(a)(1)(iv) of this section.
(xi) OM condition: Valve Position Indication. When implementing
ASME OM Code, 2012 Edition, Subsection ISTC-3700, ``Position
Verification Testing,'' licensees shall verify that valve operation is
accurately indicated by supplementing valve position indicating lights
with other indications, such as flow meters or other suitable
instrumentation, to provide assurance of proper obturator position.
(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, as incorporated by reference in
paragraph (a)(3)(i) of this section, without prior NRC approval,
subject to the following conditions:
* * * * *
(5) Conditions on inservice inspection Code Cases. Licensees may
apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.147, as
incorporated by reference in paragraph (a)(3)(ii) of this section,
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 this section as listed in Tables 1 and 2 of NRC
Regulatory Guide 1.147, as incorporated by reference in paragraph
(a)(3)(ii) of this section.
(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 NRC
[[Page 32983]]
Regulatory Guide 1.147. If a licensee has applied a listed Code Case
that is later listed as annulled in NRC 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 ASME OM Code Cases. Licensees may apply the ASME
OM Code Cases listed in NRC Regulatory Guide 1.192, as incorporated by
reference in paragraph (a)(3)(iii) of this section, 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 this section as listed in Tables 1 and 2 of NRC
Regulatory Guide 1.192, as incorporated by reference in paragraph
(a)(3)(iii) of this section.
(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 NRC Regulatory Guide 1.192. If a licensee has applied a
listed Code Case that is later listed as annulled in NRC Regulatory
Guide 1.192, the licensee may continue to apply the Code Case to the
end of the current 120-month interval.
* * * * *
(f) Preservice and inservice testing requirements. Systems and
components of boiling and pressurized water-cooled nuclear power
reactors must meet the requirements for preservice and inservice
testing (referred to in this paragraph (f) collectively as inservice
testing) of the ASME BPV Code and ASME OM Code as specified in this
paragraph (f). 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 paragraph (g) of this section.
* * * * *
(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 BPV 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 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 or NRC Regulatory Guide 1.192, as
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 or NRC Regulatory Guide 1.192, as incorporated by reference
in paragraphs (a)(3)(ii) and (iii) of this section, respectively),
subject to the applicable conditions listed therein.
* * * * *
(3) * * *
(iii) * * *
(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 BPV 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 or
NRC Regulatory Guide 1.192, as 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 BPV 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 OM Code
Cases listed in NRC Regulatory Guide 1.192, as 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) * * *
(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 BPV 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 BPV Code Cases listed in NRC Regulatory
Guide 1.147, as 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 BPV 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, as
incorporated by reference in paragraph (a)(3)(iii) of this section),
incorporated by reference in paragraph (a)(1)(iv) of this section at
the
[[Page 32984]]
time the construction permit, combined license, or design certification
is issued.
* * * * *
(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 within the scope of
the ASME OM Code 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. The inservice test requirements for
pumps and valves that are within the scope of the ASME OM Code but are
not classified as ASME BPV Code Class 1, Class 2, or Class 3 may be
satisfied as an augmented IST program in accordance with paragraph
(f)(6)(ii) of this section without requesting relief under paragraph
(f)(5) of this section or alternatives under paragraph (z) of this
section. This use of an augmented IST program may be acceptable
provided the basis for deviations from the ASME OM Code, as
incorporated by reference in this section, demonstrates an acceptable
level of quality and safety, or that implementing the Code provisions
would result in hardship or unusual difficulty without a compensating
increase in the level of quality and safety, where documented and
available for NRC review.
(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 ASME 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 OM Code Cases
listed in NRC Regulatory Guide 1.192, as incorporated by reference in
paragraph (a)(3)(iii) of this section, subject to the conditions listed
in paragraph (b) of this section).
(ii) Applicable IST Code: Successive 120-month intervals. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during successive 120-month
intervals must comply with the requirements of the latest edition and
addenda of the ASME 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 or NRC Regulatory Guide 1.192 as 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.
* * * * *
(g) Preservice and 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 paragraph (f) of this section.
* * * * *
(2) Accessibility requirements--(i) Accessibility requirements for
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 that are classified as ASME BPV Code Class 1 and Class 2 and
supports for components that are classified as ASME BPV Code Class 1
and Class 2 must be designed and be provided with the access necessary
to perform the required preservice and inservice examinations 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 BPV Code Cases listed in NRC Regulatory Guide 1.147,
as incorporated by reference in paragraph (a)(3)(ii) of this section)
in effect 6 months before the date of issuance of the construction
permit.
(ii) Accessibility requirements for 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, components
that are classified as ASME BPV Code Class 1, Class 2, and Class 3 and
supports for components that are classified as ASME BPV Code Class 1,
Class 2, and Class 3 must be designed and provided with the access
necessary to perform the required preservice and inservice examinations
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 BPV Code Cases listed in NRC Regulatory
Guide 1.147, as incorporated by reference in paragraph (a)(3)(ii) of
this section) applied to the construction of the particular component.
(iii) Accessibility requirements: Meeting later Code 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.
(3) Preservice examination requirements--(i) Preservice examination
requirements for 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 that are classified as ASME BPV Code Class 1
and Class 2 and supports for components that are classified as ASME BPV
Code Class 1 and Class 2 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 BPV Code Cases listed in
NRC Regulatory Guide 1.147, as incorporated by reference in paragraph
(a)(3)(ii) of this section) in effect 6 months before the date of
issuance of the construction permit.
(ii) Preservice examination requirements for 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, components that are classified as ASME BPV Code Class 1, Class 2,
and Class 3 and supports for components that are classified as ASME BPV
Code Class 1, Class 2, and Class 3 must meet the preservice examination
requirements set forth in the editions
[[Page 32985]]
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 BPV Code Cases listed in NRC Regulatory Guide 1.147, as
incorporated by reference in paragraph (a)(3)(ii) of this section)
applied to the construction of the particular component.
* * * * *
(v) Preservice examination requirements: Meeting later Code
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) * * *
(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 ASME 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, when
using ASME BPV Code, Section XI, or NRC Regulatory Guide 1.192, when
using the ASME OM Code, as 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. Licensees may, at
any time in their 120-month ISI interval, elect to use the Appendix
VIII in the latest edition and addenda of the ASME BPV Code
incorporated by reference in paragraph (a) of this section, subject to
any applicable conditions listed in paragraph (b) of this section.
Licensees using this option must also use the same edition and addenda
of Appendix I as Appendix VIII, including any applicable 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 ASME 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, when using ASME
BPV Code, Section XI, or NRC Regulatory Guide 1.192, when using the
ASME OM Code, as 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 August
17, 2017, may delay the update of their Appendix VIII program by up to
18 months after August 17, 2017. Alternatively, licensees may, at any
time in their 120-month ISI interval, elect to use the Appendix VIII in
the latest edition and addenda of the ASME BPV Code incorporated by
reference in paragraph (a) of this section, subject to any applicable
conditions listed in paragraph (b) of this section. Licensees using
this option must also use the same Edition and Addenda of Appendix I as
Appendix VIII, including any applicable conditions listed in paragraph
(b) of this section.
* * * * *
(6) * * *
(ii) * * *
(D) * * *
(1) Implementation. Holders of operating licenses or combined
licenses for pressurized-water reactors as of or after August 17, 2017
shall implement the requirements of ASME BPV Code Case N-729-4 instead
of ASME BPV Code Case N-729-1, subject to the conditions specified in
paragraphs (g)(6)(ii)(D)(2) through (4) of this section, by the first
refueling outage starting after August 17, 2017.
(2) Appendix I use. Appendix I of ASME BPV Code Case N-729-4 shall
not be implemented without prior NRC approval.
(3) Bare metal visual frequency. Instead of Note 4 of ASME BPV Code
Case N-729-4, the following shall be implemented. If effective
degradation years (EDY) < 8 and if no flaws are found that are
attributed to primary water stress corrosion cracking:
(i) A bare metal visual examination is not required during
refueling outages when a volumetric or surface examination is
performed; and
(ii) If a wetted surface examination has been performed of all of
the partial penetration welds during the previous non-visual
examination, the reexamination frequency may be extended to every third
refueling outage or 5 calendar years, whichever is less, provided an
IWA-2212 VT-2 visual examination of the head is performed under the
insulation through multiple access points in outages that the VE is not
completed. This IWA-2212 VT-2 visual examination may be performed with
the reactor vessel depressurized.
(4) Surface exam acceptance criteria. In addition to the
requirements of Paragraph -3132.1(b) of ASME BPV Code Case N-729-4, a
component whose surface examination detects rounded indications greater
than allowed in Paragraph NB-5352 in size on the partial-penetration or
associated fillet weld shall be classified as having an unacceptable
indication and corrected in accordance with the provisions of
paragraph-3132.2 of ASME BPV Code Case N-729-4.
* * * * *
(F) * * *
(1) Implementation. Holders of operating licenses or combined
licenses for pressurized-water reactors as of or after August 17, 2017,
shall implement the requirements of ASME BPV Code Case N-770-2 instead
of ASME BPV Code Case N-770-1, subject to the conditions specified in
paragraphs (g)(6)(ii)(F)(2) through (13) of this section, by the first
refueling outage starting after August 17, 2017.
(2) Categorization. Full structural weld overlays, authorized by
the NRC staff in accordance with the alternatives approval process of
this section, may be categorized as Inspection Items C-1 or F-1, 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 the purpose of determining ISI frequencies, all
other butt welds that rely on Alloy 82/182 for structural integrity
shall 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 an alternative code
case Inspection Item is used based on conformance with an ASME
mitigation code case endorsed in NRC Regulatory Guide 1.147 with any
applying conditions specified in NRC Regulatory Guide 1.147, as
incorporated by reference in paragraph (a)(3)(ii) of this section.
Paragraph -1100(e) of ASME BPV Code Case N-770-2 shall not be used to
exempt welds that rely on Alloy 82/182 for structural integrity from
any requirement of paragraph (g)(6)(ii)(F) of this section.
(3) Baseline examinations. Baseline examinations for welds in Table
1 of ASME BPV Code Case N-770-2, Inspection Items A-1, A-2, and B, if
not
[[Page 32986]]
previously performed or currently scheduled to be performed in an
ongoing refueling outage as of August 17, 2017, in accordance with
paragraph (g)(6)(ii)(F) of this section, shall be completed by the end
of the next refueling outage. Previous examinations of these welds can
be credited for baseline examinations only if they were performed
within the re-inspection period for the weld item in Table 1 of ASME
BPV Code Case N-770-2 and the examination of each weld meets the
examination requirements of paragraphs -2500(a) or -2500(b) of ASME BPV
Code Case N-770-2 as conditioned in this section. Other previous
examinations that do not meet these requirements can be used to meet
the baseline examination requirement, provided NRC approval in
accordance with paragraph (z)(1) or (2) of this section, is granted
prior to the end of the next refueling outage.
(4) Examination coverage. When implementing Paragraph -2500(a) of
ASME BPV Code Case N-770-2, essentially 100 percent of the required
volumetric examination coverage shall be obtained, including greater
than 90 percent of the volumetric examination coverage for
circumferential flaws. Licensees are prohibited from using Paragraphs -
2500(c) and -2500(d) of ASME BPV Code Case N-770-2 to meet examination
requirements.
(5) Inlay/onlay inspection frequency. All hot-leg operating
temperature welds in Inspection Items G, H, J, and K shall be inspected
each inspection interval. A 25 percent sample of Inspection Items G, H,
J, and K cold-leg operating temperature welds shall be inspected
whenever the core barrel is removed (unless it has already been
inspected within the past 10 years) or within 20 years, whichever is
less.
(6) Reporting requirements. 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) Defining ``t''. 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) Optimized weld overlay examination. Initial inservice
examination of Inspection Item C-2 welds shall be performed between the
third refueling outage and no later than 10 years after application of
the overlay.
(9) Deferral. Note (11)(b)(1) in ASME BPV Code Case N-770-2 shall
not be used to defer the initial inservice examination of optimized
weld overlays (i.e., Inspection Item C-2 of ASME BPV Code Case N-770-
2).
(10) Examination technique. Note 14(b) of Table 1 and Note (b) of
Figure 5(a) of ASME BPV Code Case N-770-2 may only be implemented if
the requirements of Note 14(a) of Table 1 of ASME BPV Code Case N-770-2
cannot be met.
(11) Cast stainless steel. Examination of ASME BPV Code Class 1
piping and vessel nozzle butt welds involving cast stainless steel
materials, shall be performed with Appendix VIII, Supplement 9
qualifications, or qualifications similar to Appendix VIII, Supplement
2 or 10 using cast stainless steel mockups no later than the next
scheduled weld examination after January 1, 2022, in accordance with
the requirements of Paragraph -2500(a).
(12) Stress improvement inspection coverage. Under Paragraph I.5.1,
for cast stainless steel items, the required examination volume shall
be examined by Appendix VIII procedures to the maximum extent practical
including 100 percent of the susceptible material volume.
(13) Encoded ultrasonic examination. Ultrasonic examinations of
non-mitigated or cracked mitigated dissimilar metal butt welds in the
reactor coolant pressure boundary must be performed in accordance with
the requirements of Table 1 for Inspection Item A-1, A-2, B, E, F-2, J,
and K for 100 percent of the required inspection volume using an
encoded method.
* * * * *
Dated at Rockville, Maryland, this 30th day of June 2017.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2017-14166 Filed 7-17-17; 8:45 am]
BILLING CODE 7590-01-P