[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]]
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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.
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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 http://www.regulations.gov and search for Docket ID NRC-2011-0088. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. For
the convenience of the reader, 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:
[email protected]; or Keith Hoffman, Office of Nuclear Reactor
Regulation, telephone: 301-415-1294, email: [email protected]. Both
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
The 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
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Alternative 2--
the rule
alternative
net benefits
Objective (costs)
(million) (Net
present value,
7% discount
rate)
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Industry................................................ $11.5
NRC..................................................... 3.28
Net benefit............................................. 14.7
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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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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.
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\1\ The editions and addenda of the ASME Code for Operation and
Maintenance of Nuclear Power Plants have had different titles from
2005 to 2012 and are referred to collectively in this rule as the
``OM Code.''
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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
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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
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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 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
[[Page 32951]]
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 http://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: [email protected].
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
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: [email protected].
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; http://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
http://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 http://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, http://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:
[email protected]; or at the National Archives and Records
Administration (NARA). For information on the availability of this
material at NARA, call 202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.
(1) * * *
(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