[Federal Register Volume 80, Number 181 (Friday, September 18, 2015)]
[Proposed Rules]
[Pages 56820-56864]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-23193]
[[Page 56819]]
Vol. 80
Friday,
No. 181
September 18, 2015
Part V
Nuclear Regulatory Commission
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10 CFR Part 50
Incorporation by Reference of American Society of Mechanical Engineers
Codes and Code Cases; Proposed Rule
Federal Register / Vol. 80 , No. 181 / Friday, September 18, 2015 /
Proposed Rules
[[Page 56820]]
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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: Proposed rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
amend its regulations to incorporate by reference seven 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 proposing to incorporate by reference four 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: Submit comments by December 2, 2015. Comments received after
this date will be considered if it is practical to do so, but the NRC
is able to ensure consideration only for comments received on or before
this date.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
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.
Email comments to: [email protected]. If you do
not receive an automatic email reply confirming receipt, then contact
us at 301-415-1677.
Fax comments to: Secretary, U.S. Nuclear Regulatory
Commission at 301-415-1101.
Mail comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and
Adjudications Staff.
Hand deliver comments to: 11555 Rockville Pike, Rockville,
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal
workdays; telephone: 301-415-1677.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
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 proposing to amend its regulations to incorporate by
reference seven recent editions and addenda to the ASME codes for
nuclear power plants and an ASME standard for quality assurance. The
NRC is also proposing to incorporate by reference four ASME code cases.
This proposed 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 periodically
revises and updates its codes for nuclear power plants by issuing new
editions and addenda, and 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. The ASME is a voluntary consensus standards
organization, and the ASME codes are voluntary consensus standards. The
NRC's use of the ASME codes is consistent with applicable requirements
of the National Technology Transfer and Advancement Act. Additional
discussion of voluntary consensus standards and the NRC's compliance
with the National Technology Transfer and Advancement Act (NTTAA) is
set forth in Section VIII of this notice, ``Voluntary Consensus
Standards.''
B. Major Provisions
Major provisions of the proposed rule include:
Incorporation by reference of ASME codes into NRC
regulations and delineation of NRC 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 and approval of four ASME Code
Cases.
C. Costs and Benefits
The NRC prepared a draft regulatory analysis to determine the
expected costs and benefits of the proposed rule. The regulatory
analysis identified costs and benefits in a quantitative fashion as
well as in a qualitative fashion.
The analysis concluded that the proposed rule would result in net
quantitative costs to the industry and the NRC. The proposed rule,
relative to the regulatory baseline, would result in a net cost for
industry of between $5.1 million based on a 7 percent net present value
and $4.3 million based on a 3 percent net present value. The estimated
incremental industry cost per reactor unit ranges from $49,000 based on
a 7 percent net present value to $41,000 based on a 3 percent net
present value. The NRC benefits from the proposed rulemaking
alternative because of the averted cost of not reviewing and approving
Code alternative requests on a plant-specific basis under Sec.
50.55a(z) of title 10 of the Code of Federal Regulations (10 CFR). The
NRC net benefit ranges from $1.4 million based on a 7 percent net
present value to $1.9 million based on a 3 percent net present value.
Qualitative factors which were considered include regulatory
stability and predictability, regulatory efficiency, and consistency
with the NTTAA Act of 1995, as amended. Table 44 in the draft
regulatory analysis includes a discussion of the costs and benefits
that were considered qualitatively. If the results of the regulatory
analysis were based solely on quantified costs and benefits, then the
regulatory analysis would show that the rulemaking is not justified
because the total quantified benefits of the proposed regulatory action
do not equal or exceed the costs of the proposed action. However, if
the qualitative benefits (including the safety benefit, cost savings,
and other non-quantified benefits) are considered together with the
quantified benefits, then the benefits outweigh the identified
quantitative and qualitative impacts.
With respect to regulatory stability and predictability, the NRC
has had a decades-long practice of approving and/
[[Page 56821]]
or mandating the use of certain parts of editions and addenda of these
ASME Codes in 10 CFR 50.55a through the rulemaking process of
``incorporation by reference.'' Retaining the practice of approving
and/or mandating the ASME Codes continues the regulatory stability and
predictability provided by the current practice. Retaining the practice
also assures consistency across the industry, and provides assurance to
the industry and the public that the NRC will continue to support the
use of the most updated and technically sound techniques developed by
the ASME to provide adequate protection to the public. In this regard,
these ASME Codes are voluntary consensus standards developed by
participants with broad and varied interests and have already undergone
extensive external review before being reviewed by the NRC. Finally,
the NRC's use of the ASME Codes is consistent with the NTTAA, which
directs Federal agencies to adopt voluntary consensus standards instead
of developing ``government-unique'' (i.e., Federal agency-developed)
standards, unless inconsistent with applicable law or otherwise
impractical.
For more information, please see the draft regulatory analysis
(Accession No. ML14170B104 in the NRC's Agencywide Documents Access and
Management System).
Table of Contents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
II. Background
III. Discussion
A. ASME BPV Code, Section III
B. ASME BPV Code, Section XI
C. ASME OM Code
D. ASME Code Cases
IV. Section-by-Section Analysis
V. Generic Aging Lessons Learned Report
VI. Specific Request for Comments
VII. Plain Writing
VIII. Voluntary Consensus Standards
IX. Incorporation by Reference--Reasonable Availability to
Interested Parties
X. Environmental Assessment and Final Finding of No Significant
Environmental Impact
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis: Availability
XIII. Backfitting and Issue Finality
XIV. Regulatory Flexibility Certification
XV. Availability of Documents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2011-0088 when contacting the NRC
about the availability of information for this proposed rule. You may
obtain information related to this proposed rule by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2011-0088.
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.
B. Submitting Comments
Please include Docket ID NRC-2011-0088 in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. 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 ASME 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 ASME OM Code. Starting in 2012, the ASME decided to
issue editions of its BPV and OM Codes (no addenda) every 2 years with
the BPV Code to be issued on the odd years (e.g., 2013, 2015, etc.) and
the OM Code to be issued on the even years (e.g., 2012, 2014, etc.).
The new editions and addenda typically revise provisions of the 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 Codes do not significantly change Code 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
BPV and OM Codes (ASME Codes) in Sec. 50.55a. 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 Codes' 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 NRC
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
[[Page 56822]]
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
conditions 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 ASME 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 notice. 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 notice.
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 VIII of this notice, ``Voluntary
Consensus Standards.''
This proposed rule contains changes from a November 5, 2014, NRC
final rule amending Sec. 50.55a to, among other things, re-designate
paragraphs within Sec. 50.55a (79 FR 65776). The re-designation of
paragraphs was needed to address the Office of the Federal Register's
requirements in 10 CFR part 51 applicable to incorporation by
reference. For additional information on the November 2014 final rule,
please consult the statement of considerations (preamble) for that
final rule.
III. 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 proposed 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. 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 subjects 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 reviewed changes to the Codes in the editions and addenda
of the Codes identified in this rulemaking. The NRC concluded, 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 the specified
conditions.
The NRC proposes to amend its regulations to incorporate by
reference:
The 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013
Edition to 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
to Division 1 of the ASME 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 proposed rule language Sec. 50.55a(a)(1)(v). More specifically, the
NRC proposes to incorporate 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-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 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 conditions identified in current
Sec. 50.55a(b)(2)(i) through (b)(2)(xxix). The proposed amendment
would revise 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 would also clarify the wording and add,
remove, or revise some of the conditions as explained in this notice.
The NRC proposes to revise Sec. 50.55a(a)(1)(iv) to incorporate by
reference the 2009 Edition, 2011 Addenda, and 2012 Edition of Division
1 of the ASME OM Code. Based on this revision, the NRC regulations
would
[[Page 56823]]
incorporate by reference in Sec. 50.55a the 1995 Edition through the
2012 Edition of the ASME OM Code.
Each of the proposed NRC conditions and the reasons for each
proposed condition are discussed below. The discussions are organized
under the applicable ASME Code and Section. Please note that 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 ASME OM Code--
because there are three proposed conditions related to NQA-1, one in
each of those areas (paragraph (b)(1)(iv) for Section III, paragraph
(b)(2)(x) for Section XI, and paragraph (b)(3)(i) for the OM Code).
A. ASME BPV Code, Section III
10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section
III
The NRC proposes to clarify 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 ``(excluding Non-mandatory Appendices)'' in 10 CFR
50.55a(a)(1)(i).
10 CFR 50.55a(b)(1)(ii) Section III Condition: Weld Leg Dimensions
The NRC proposes to identify prohibited subparagraphs and footnotes
for each BPV Code edition and addenda in tabular form as opposed to the
textual listing of the current 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.
Currently, Sec. 50.55a(b)(1)(ii) includes a condition prohibiting
the use of Footnote 11 from the 1989 Addenda through the 2003 Addenda
or Footnote 13 from the 2004 Edition through the 2008 Addenda to
Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg sizes less
than 1.09 tn when using the ASME BPV Code, Section III,
Division 1. These 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. This
is due to the fact that the current provisions would result in a weld
that would be weaker than the pipe to which it is adjoined under these
dimensions. The weld stress provisions in the version of the footnotes
contained in the 1989 Addenda have been relocated to different
subparagraphs in subsequent BPV Code editions and addenda. Therefore,
the current Code's reference in Footnote 11 to Figures NC-3673.2(b)-1
and ND-3673.2(b)-1 is not correct for BPV Code editions and addenda
after the 1989 Addenda, in applying the condition. The proposed rule
would correct this issue by clearly identifying the prohibited code
provisions in the editions and addenda in a tabular format.
As an editorial matter, this proposed rule identifies the
prohibited BPV Code provisions as ``notes,'' which is the term used by
the ASME, rather than ``footnotes.'' The NRC proposes to use the
terminology used by the ASME for clarity.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
The NRC proposes to approve 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 will allow 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.
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 proposes to revise 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, governs the applicant's or licensee's Section
III activities. The proposed clarification is consistent with Sec.
50.55a(b)(2)(x) and Sec. 50.55a(b)(3)(i). 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 requirement. Thus, 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 considering removing the reference in Sec.
50.55a(b)(1)(iv) to versions of NQA-1 older than the 1994 Edition. The
NRC requests public comment on whether any applicant or licensee is
committed to, and is using, a version of NQA-1 older than the 1994
Edition, and if so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification
and Demonstration of Function of Incompressible-Fluid Pressure-Relief
Valves
The NRC proposes to revise 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 proposing to add new paragraph, Sec.
50.55a(b)(1)(viii), to allow
[[Page 56824]]
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
certification designators and class designators as specified in the
2013 Edition through the latest edition and addenda incorporated by
reference in 10 CFR 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. ASME has confirmed that the
Certification Mark with designator is equivalent to the corresponding
BPV Code Symbol Stamp. Based on statements 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.
B. ASME BPV Code, Section XI
10 CFR 50.55a(a)(1)(ii) ASME Boiler and Pressure Vessel Code, Section
XI
The NRC proposes to revise Sec. 50.55a(a)(1)(ii) to clarify that
Section XI Non-mandatory Appendix U of the 2013 Edition of ASME BPV
Code Section XI is not incorporated by reference and therefore not
approved for use. The NRC is developing an integrated approach to the
issue of operational leakage. The NRC has not completed its
determination of how Appendix U fits into this integrated approach to
address the operational leakage issue at nuclear power plants. The
operational leakage issue has many factors that need to be considered
such as acceptance criteria, corrective actions, application of repair/
replacement requirements, component operability determination, concerns
related to continued operation, maximum acceptable leakage rates, flaw
growth rates, flaw measurement techniques, schedules for eliminating
leakage, and when or if the leakage requires authorization by the NRC.
The NRC plans to complete the development of the regulatory approach to
operational leakage and issue it 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 proposes to revise Sec. 50.55a(b)(2)(vi) to explicitly
state that the provision requiring the use of either the 1992 Edition
with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of
Subsection IWE and Subsection IWL when implementing the initial 120-
month containment inservice inspection program applies only to those
licensees that were required by previous versions \2\ of Sec. 50.55a
to develop and implement a containment inservice inspection 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.
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\2\ See the supplementary information and rule language for
Sec. 50.55a(b)(2)(vi), Sec. 50.55a(g)(4), and Sec.
50.55a(g)(6)(ii)(B) in Federal Register notices published on August
8, 1996 (61 FR 41303), and September 22, 1999 (64 FR 51370).
---------------------------------------------------------------------------
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 the operating reactors in the
above stated class would have gone past their initial 120-month
inspection interval by 2011. The proposed change removes the
possibility of misinterpretation of the provision as requiring plants
that do not fall in the above 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 inservice inspection program. Applicants and
licensees that do not fall in the above class must use Code editions
and addenda in accordance with Sec. 50.55a(g)(4)(i) and (g)(4)(ii),
respectively, for the initial and successive 120-month containment
inservice inspection intervals.
10 CFR 50.55a(b)(2)(viii) Section XI Condition: Concrete Containment
Examinations
The NRC proposes to revise 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) and adding a requirement to comply with Sec.
50.55a(b)(2)(viii)(H) and (I).
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 proposes to remove 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.'' The
reasons for requiring compliance with Sec. 50.55a(b)(2)(viii)(H) and
(I) are set forth in the next two sections.
[[Page 56825]]
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth
Provision
The NRC proposes to add a new paragraph, 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 would replace 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. Sec. 50.55a(b)(2)(viii)(E)(1), (E)(2),
and (E)(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 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 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) 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. Sec. 50.55a(b)(2)(viii)(E)(1), (E)(2), and (E)(3) of
the existing condition to be provided in the IWA-6000 ISI Summary
Report.
For these reasons, the NRC proposes to identify 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 would replace 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 requested 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 proposes to add 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
[[Page 56826]]
below-grade inaccessible area to satisfy the condition in Sec.
50.55a(b)(2)(viii)(E).
The technical evaluation requirements in IWL-2512(b) help to
determine 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 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 proposes to add a new 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, with regard to aging management of inaccessible below-grade
concrete components of the containment. The new IWL-2512(b) is
applicable to inaccessible below-grade concrete surfaces of the
containment cylindrical wall and basemat foundations, which are exposed
to foundation soil, backfill, or groundwater. 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.
Also, the condition requires the examination of representative samples
of the exposed portions of the below-grade concrete be performed when
excavated for any reason. 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 proposes to continue to apply the existing conditions in
Sec. Sec. 50.55a(b)(2)(ix)(A)(2), (b)(2)(ix)(B), and (b)(2)(ix)(J)
governing examinations of metal containments and the liners of concrete
containments under Subsection IWE to the 2007 Edition with 2009 Addenda
through the 2013 Edition (the code editions and addenda which are the
subject of this rulemaking). The NRC reviewed the code changes in
Subsection IWE of the 2009 Addenda through the 2013 Edition of ASME BPV
Code, Section XI, and notes 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. Although this continuation of the applicability of the
three conditions does not require a rule change, the NRC is discussing
this for the benefit of stakeholder understanding of the effect of the
proposed rule.
10 CFR 50.55a(b)(2)(x) Section XI Condition: Quality Assurance
The NRC proposes to approve 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 will allow licensees to use the 1994 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 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 proposes to amend Sec. 50.55a(b)(2)(x) to clarify that a
licensee's commitments addressing those areas where NQA-1 either does
not address an appendix B requirement or is less stringent than the
comparable appendix B requirement governs the licensee's Section XI
activities. The proposed clarification is consistent with Sec. 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 requirement. Thus, 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 considering removing the reference in Sec.
50.55a(b)(2)(x) to versions of NQA-1 older than the 1994 Edition. The
NRC requests public comment on whether any licensee is committed to,
and is using, a version of NQA-1 older than the 1994 Edition, and if
so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth
Provision
The NRC proposes to add a new paragraph, 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.
Section 50.55a(b)(2)(xviii) currently includes conditions on the
certification
[[Page 56827]]
of NDE personnel. In addition, the new paragraph would 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 would
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 proposing to prohibit the use of Appendix VII and
VIII-2200 in the 2011 Addenda and 2013 Edition, and instead require
applicants and licensees using the 2011 Addenda and 2013 Edition to use
Table VII-4110-1 in the 2010 Edition, and VIII-2200, Appendix VIII
prerequisites for ultrasonic examination personnel requirements in the
2010 Edition.
10 CFR 50.55a(b)(2)(xxi)(A) Table IWB-2500-1 Examination Requirements:
First Provision
The NRC proposes to revise 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 VT-1 equipment. This revision
essentially removes a requirement that was in 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, BWRVIP-03 was
changed to eliminate a \1/2\ mil wire from the Sensitivity Resolution
and Contrast Standards due to similar concerns.
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)(xxx) Section XI Condition: Steam Generator
Preservice Examinations
The NRC proposes to add Sec. 50.55a(b)(2)(xxx) to require a full
length examination of 100 percent of the tubing in each newly installed
steam generator prior to plant startup. This requirement would be
instead of the unapproved provisions in IWB-2200(c) pertaining to steam
generator tube preservice inspections (PSI).
Steam generator tubes, a significant portion of the reactor coolant
pressure boundary, are important to the safe operation of a pressurized
water reactor. As such, the NRC has established requirements pertaining
to the design, fabrication, erection, testing, and inspection of the
steam generator tubes. With respect to the performance of the PSI of
steam generator tubes, the NRC has indicated in NRC Regulatory Guide
(RG) 1.83, Revision 1, ``Inservice Inspection of Pressurized Water
Reactor Steam Generator Tubes,'' (withdrawn in 2009) that all tubes in
the steam generator should be inspected by eddy current or alternative
technique prior to service to establish a baseline condition of the
tubing. A similar position is articulated in NUREG-0800, Standard
Review Plan (SRP) Section 5.4.2.2, ``Steam Generator Tube Inservice
Inspection,'' Revision 1 and subsequent revisions. A PSI is important
since it ensures that the steam generator tubes are acceptable for
initial operation. In addition, the PSI provides the baseline condition
of the tubes. This data is essential in assessing the nature of
indications found in the tubes during subsequent inservice inspections.
Preservice requirements for ASME Class 1 components are provided in
IWB-2200, and IWB-2200(c) currently states, ``Steam generator tube
examination shall be governed by the plant Technical Specifications
(TS).'' However, there are no preservice examination requirements for
steam generators defined in plant TS. Preservice examination
requirements for steam generators are not within any of the categories
described in 50.36 for the content of TS. Because IWB-2200(c) requires
the steam generator tube examinations be performed in accordance with
plant TS, and TS contain no rules for PSI of steam generator tubing,
the NRC is clarifying the preservice inspection requirements for steam
generator tubes.
The proposed clarification is consistent with industry guidelines
and the NRC staff position outlined in SRP Section 5.4.2.2, ``Steam
Generator Program.'' The proposed requirement supersedes the
requirements of IWB-2200(c). These inspections must be performed with
the objective of finding and characterizing the types of preservice
flaws that may be present in the tubes and flaws that may occur during
operation.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping
Devices
The NRC proposes to add Sec. 50.55a(b)(2)(xxxi) to prohibit the
use of mechanical clamping devices on Class 1 piping and portions of
piping systems that form the containment boundary. 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
[[Page 56828]]
the use of small items exemption rules in the installation of
mechanical clamps. 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 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 clamps on Class 1 piping or portions of a piping system
that form the containment boundary as a permanent repair. Furthermore,
the NRC does not believe that the ASME has 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).
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report
Submittal
The NRC proposes to add 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 believes 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 PSI and ISI
programs. Therefore, the NRC proposes adding Sec. 50.55a(b)(2)(xxxii)
to ensure that preservice and inservice summary reports will continue
to be submitted within the timeframes currently 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 proposes to add 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 inservice inspection 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 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 proposing to prohibit 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 P-T
limits remains acceptable.
10 CFR 50.55a(b)(2)(xxxiv) Section XI Condition: Disposition of Flaws
in Class 3 Components
The NRC proposes to add Sec. 50.55a(b)(2)(xxxiv) to require that
when using the 2013 Edition of the ASME BPV Code, Section XI, the
licensee shall use the acceptance standards of IWD-3510 for the
disposition of flaws in Category D-A components (i.e., welded
attachments for vessels, piping, pumps, and valves).
The 2013 Edition of the ASME BPV Code, Section XI, IWD-3510,
``Standards for Examination Category D-A, Welded Attachments for
Vessels, Piping, Pumps, and Valves,'' states that the acceptance
standards are: ``In the course of preparation, the requirements of IWC-
3500 may be used.'' The ASME BPV Code, Section XI, IWD-3410,
``Acceptance Standards,'' states that the acceptance standards
referenced in Table IWD-3410-1 shall be applied to determine
acceptability for service. Table IWD-3410-1 states that the acceptance
standard for Examination Category D-A is IWB-3510.
A discrepancy exists between the provisions in IWD-3410, which
references Table IWD-3410-1, and the provisions in IWD-3510. The
provisions in IWD-3510 require the use of the acceptance standards of
IWC-3500 whereas Table IWD-3410-1 requires the use of the acceptance
standards of IWB-3510 to disposition flaws detected in the Examination
Category D-A components. Both IWD-3410 and IWD-3510 should reference
the same subarticle and acceptance standards. The NRC believes that
this discrepancy is due to an error in the publishing of the 2013
Edition because the code committee action which proposed the revised
Class 3 acceptance criteria and added Table IWD-3410-1 showed the
appropriate Acceptance Standard to be IWD-3510. The intent of the
condition is to provide clarification and consistency in requirements
between IWD-3410 and IWD-3510.
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0
in the KIa and KIc Equations
The NRC proposes to add 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.
Non-mandatory Appendix A provides a procedure based on linear
elastic
[[Page 56829]]
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 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 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 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 NRC staff 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, IN, USA). The NRC
staff recognized that the proposed insertion is consistent with 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 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 proposing to add 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 proposes to add 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 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 not accepted by the NRC. With this condition, users of
Appendix A can avoid 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 proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and
(g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i),
(g)(3)(ii), and (g)(3)(v) to distinguish the requirements for
accessibility and preservice examination from those for inservice
inspection in Sec. 50.55a(g). No substantive change to the
requirements is intended by these revisions.
C. ASME OM Code
10 CFR 50.55a(b)(3) Conditions on ASME OM Code
The NRC proposes to revise Sec. 50.55a(b)(3) to clarify that
Subsections ISTA, ISTB, ISTC, ISTD, ISTE, and ISTF; Mandatory
Appendices I, II, III, and V; and Non-mandatory Appendices A through H
and J through M of the ASME OM Code would be incorporated by reference
in Sec. 50.55a. The NRC is clarifying that the ASME OM Code non-
mandatory appendices, which are incorporated by reference into Sec.
50.55a are approved for use, but are not mandated. The non-mandatory
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 proposes to revise Sec. 50.55a(b)(3)(i) to allow use of
the 1983 Edition through 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
after the 1983 Edition and compared them to the previously approved
versions of NQA-1 and found that there were no significant differences.
The NRC is considering removing the reference in Sec.
50.55a(b)(3)(i) to versions of NQA-1 older than the 1994 Edition. The
NRC requests public comment on whether any licensee is committed to,
and is using, a version of NQA-1 older than the 1994 Edition and, if
so, what version the applicant or licensee is using.
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV)
Testing
The NRC proposes to revise 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 ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition.
Appendix III of the ASME 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 ASME OM Code reflects the
incorporation of ASME 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 proposes
to add four conditions in new Sec. Sec. 50.55a(b)(3)(ii)(A), (B), (C),
and (D) to address periodic verification of MOV design-basis
capability. These conditions are discussed in the next four sections.
[[Page 56830]]
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(A) to require 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 ASME OM Code, 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 64 FR 51386
(September 22, 1999) with respect to the use of ASME 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.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact On Risk
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(B) to require that
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 64 FR
51386 (September 22, 1999) with respect to the use of ASME 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 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
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(C) to require, when
applying Appendix III to the ASME OM Code, that licensees 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
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 ASME 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 risk ranking methodologies must have been accepted by the
NRC through RG 1.192 (which accepts ASME OM Code Case OMN-3 with the
specified conditions) or safety evaluations issued to address plant-
specific or industry-wide risk ranking methodologies.
Two conditions that were previously in RG 1.192 on the use of ASME
OM Code Case OMN-11 related to application of the test interval
criteria and grouping for low safety significant MOVs have been
incorporated in an acceptable manner in Appendix III to the ASME OM
Code. As noted in RG 1.192 on the use of ASME OM Code Case OMN-1, the
benefits of performing a particular test should be balanced against the
potential adverse effects placed on the valves or systems caused by
this testing.
10 CFR 50.55a(b)(3)(ii)(D) MOV Stroke Time
The NRC proposes to add 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
safety analyses. Previous editions and addenda of the ASME 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 ASME OM Code Case OMN-1. However, Paragraph III-3600
of Appendix III of the versions of the OM Code proposed to 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 proposing to adopt the new condition which will
effectively retain the need to verify MOV stroke time during periodic
exercising.
10 CFR 50.55a(b)(3)(iii) OM condition: New Reactors
The NRC proposes to add 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 ASME OM Code as
incorporated by reference with conditions in Sec. 50.55a. Licensees of
``new reactors'' are, as identified in the proposed paragraph: (i)
Holders of operating licenses for nuclear power reactors that received
construction permits under this part on or after the date 12 months
after the effective date of this rulemaking and (ii) holders of
combined licenses (COLs) issued under 10 CFR part 52, whose initial
fuel loading occurs on or after the date 12 months after the effective
date of this rulemaking. This implementation schedule for new reactors
is consistent with the NRC regulations in Sec. 50.55a(f)(4)(i).
The NRC is evaluating COL applications to construct and operate
nuclear power plants with certified designs under the process described
in 10 CFR part 52. 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
[[Page 56831]]
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 ASME
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 ASME 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 ASME OM
Code. The ASME is considering further modifications to the ASME 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.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves
The NRC proposes to add Sec. 50.55a(b)(3)(iii)(A) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) develop a program to
periodically verify the capability of power-operated valves (POVs) to
perform their design-basis safety functions. While Appendix III to the
ASME OM Code addresses this requirement for motor-operated valves
(MOVs) with applicable conditions specified in Sec. 50.55a, nuclear
power plant licensees will need to develop programs to periodically
verify the design-basis capability of other POVs. The NRC's Regulatory
Issue Summary (RIS) 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 during research
programs. Implementation of Appendix III to the ASME OM Code as
accepted in Sec. 50.55a(b)(3)(ii) is acceptable in satisfying Sec.
50.55a(b)(3)(iii)(A) for MOVs.
10 CFR 50.55a(b)(3)(iii)(B) Check Valves
The NRC proposes to add 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.
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 ASME 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.
10 CFR 50.55a(b)(3)(iii)(C) Flow-Induced Vibration
The NRC proposes to add 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
need 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 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. The initial test program can 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.
10 CFR 50.55a(b)(3)(iii)(D) High-Risk Non-Safety Systems
The NRC proposes to add 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 (RTNSS) for applicable reactor designs. 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 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 challenged. 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 together as
the RTNSS program. In a public memorandum 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 (ADAMS Accession No.
ML003708048). This new paragraph will specify the need for licensees to
assess the operational readiness of RTNSS pumps, valves, and dynamic
restraints.
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
The NRC proposes to revise 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 ASME OM Code. In the
2003 Addenda of the ASME OM Code, ASME revised Appendix II to address
the conditions specified in Sec. 50.55a for older versions of the
appendix. Therefore, the NRC considers Appendix
[[Page 56832]]
II in the 2003 Addenda through the 2012 Edition of the ASME OM Code to
be acceptable for use without conditions. In accepting the recent
versions of Appendix II, the NRC proposes to clarify that (1) the
maximum test interval allowed by Appendix II for individual check
valves in a group of two valves or more must be supported by periodic
testing of a sample of check valves in the group during the allowed
interval and (2) the periodic testing plan must be designed to test
each valve of a group at approximate equal intervals not to exceed the
maximum requirement interval. The NRC notes that ASME has provided
additional improvements to Appendix II since issuance of the 2003
Addenda. Therefore, where a licensee plans to voluntarily implement
Appendix II to the ASME OM Code, the licensee should apply Appendix II
in the most recent addenda and edition of ASME OM Code incorporated by
reference in Sec. 50.55a. 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 proposed rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
The NRC proposes to add 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 ASME OM Code. In the
2011 Addenda to the ASME 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 ASME 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 ASME OM Code. However,
the 2011 Addenda to the ASME OM Code did not include the requirement
for a pump periodic verification test program as an oversight. Since
then, the 2012 Edition to the ASME OM Code has incorporated Mandatory
Appendix V, ``Pump Periodic Verification Test Program,'' that supports
the changes to the acceptable and required action ranges for
comprehensive pump testing in Subsection ISTB. Therefore, proposed new
Sec. 50.55a(b)(3)(vii) would prohibit the use of Subsection ISTB in
the 2011 Addenda of the ASME OM Code. Licensees will be allowed to
apply Subsection ISTB with the revised acceptable and required action
ranges in the 2012 Edition of the ASME OM Code as incorporated by
reference in Sec. 50.55a.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
The NRC proposes to add Sec. 50.55a(b)(3)(viii) to specify that
licensees proposing to implement Subsection ISTE, ``Risk-Informed
Inservice Testing of Components in Light-Water Reactor Nuclear Power
Plants,'' of the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012
Edition, must request and obtain NRC authorization 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
ASME OM Code.
In the 2009 Edition of the ASME 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 ASME OM Code continue to apply, except those
identified in Subsection ISTE. The ASME selected risk-informed guidance
from ASME 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 ASME OM Code.
During development of Subsection ISTE, the NRC staff participating
on the ASME OM Code committees indicated that the conditions specified
in RG 1.192 for the use of the applicable ASME 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,
new Sec. 50.55a(b)(3)(viii) requires that licensees proposing to
implement Subsection ISTE of the ASME 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
ASME 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 ASME OM Code
for NRC review in their applications.
The NRC will evaluate Sec. 50.55a(z) requests for approval to
implement Subsection ISTE in accordance with the following
considerations:
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 ASME 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 ASME 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 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 ASME 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
[[Page 56833]]
conditions in RG 1.192 for the use of ASME 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 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 Non-mandatory Appendix L, ``Acceptance Guidelines,'' of
the ASME OM Code for evaluating the acceptability of aggregate risk
effects (i.e., for Core Damage Frequency [CDF] and Large Early Release
Frequency [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 ASME 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 ASME 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 ASME 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 ASME 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 ASME 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 ASME 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 ASME OM Code (and Code Cases OMN-1 and 11).
Therefore, licensees proposing to implement Subsection ISTE should
address the provisions in paragraph III-3700, ``Risk-Informed MOV
Inservice Testing,'' of Appendix III to the ASME 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 ASME 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 proposing to implement
Subsection ISTE should describe the planned IST provisions for AOVs and
HOVs in its request for authorization to implement Subsection ISTE.
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 ASME OM Code, 2012 Edition. The licensee should address the
consideration of a pump periodic verification test program in its risk-
informed IST program proposed as part
[[Page 56834]]
of the authorization request to implement Subsection ISTE.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
The NRC proposes to add Sec. 50.55a(b)(3)(ix) for two purposes.
First, the proposed 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. The
proposed condition also states that Subsection ISTF, 2011 Addenda, is
not acceptable for use. As previously discussed regarding new 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 ASME OM Code
was raised to higher values in combination with the incorporation of
Mandatory Appendix V, ``Pump Periodic Verification Test Program.''
However, Subsection ISTF in the 2011 Addenda and 2012 Edition to the
ASME OM Code does not include a requirement for a pump periodic
verification test program. Therefore, new Sec. 50.55a(b)(3)(ix) would
require 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. The proposed paragraph would
prohibit the use of Subsection ISTF in the 2011 Addenda of the OM Code,
which does not include Appendix V.
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
The NRC proposes to add a new paragraph, Sec. 50.55a(b)(3)(xi),
containing a new condition that would specify that when implementing
ASME OM Code, Subsection ISTC-3700, ``Position Verification Testing,''
licensees shall supplement the ASME OM Code provisions as necessary to
verify that valve operation is accurately indicated. Subsection ISTC-
3700 of the ASME OM Code requires that 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 ASME OM Code, Subsection ISTC-3700, licensees
shall develop and implement a method to 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.
This is not a new requirement but rather a clarification of the intent
of the existing ASME OM Code. The ASME 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
(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 ASME OM Code. The
NRC's position is further elaborated in NUREG-1482 Revision 2,
paragraph 4.2.7.
10 CFR 50.55a(f): Inservice Testing Requirements
The NRC proposes to revise 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 proposed change clarifies that the ASME
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.
10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iii)(A) to ensure
that the paragraph is applicable to pumps and valves that are within
the scope of the ASME OM Code. Paragraph ISTA-1100, ``Scope,'' in
Subsection ISTA, ``General Requirements,'' of the ASME OM Code states
that the requirements for preservice and inservice testing and
examination of components in light-water reactor nuclear power plants
apply to (a) pumps and valves that are required to perform a specific
function in shutting down a reactor to the safe shutdown condition, in
maintaining the safe shutdown condition, or in mitigating the
consequences of an accident; (b) pressure relief devices that protect
systems or portions of systems that perform one or more of these three
functions; and (c) dynamic restraints (snubbers) used in systems that
perform one of more of these three functions, or to ensure the
integrity of the reactor coolant pressure boundary. This revision will
align the scope of pumps and valves for inservice testing with the
scope defined in the ASME OM Code and in SRP Section 3.9.6,
``Functional Design, Qualification, and Inservice Testing Programs for
Pumps, Valves, and Dynamic Restraints.''
10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iii)(B) to clarify
that this paragraph is applicable to pumps and valves that are within
the scope of the ASME OM Code. This revision will align the scope of
pumps and valves for inservice testing with the scope defined in the
ASME OM Code and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First
Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iv)(A) to clarify
that this paragraph is applicable to pumps and valves that are within
the scope of the ASME OM Code and not covered by paragraph
(f)(3)(iii)(A) for Class 1 pumps and valves. This revision will align
the scope of pumps and valves for inservice testing with the scope
defined in the ASME OM Code and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second
Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iv)(B) to clarify
that this paragraph is applicable to pumps and valves that are within
the scope of the ASME OM Code and not covered by paragraph
(f)(3)(iii)(B) for Class 1 pumps
[[Page 56835]]
and valves. This revision will align the scope of pumps and valves for
inservice testing with the scope defined in the ASME OM Code and in SRP
Section 3.9.6.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for
Operating Plants
The NRC proposes to revise Sec. 50.55a(f)(4) to clarify that this
paragraph is applicable to pumps and valves that are within the scope
of the ASME OM Code. This revision will align the scope of pumps and
valves for inservice testing with the scope defined in the ASME OM Code
and in SRP Section 3.9.6.
D. ASME Code Cases
The NRC proposes to remove 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. The revision numbers for the RGs approved for
incorporation by reference (currently, RGs 1.84, 1.147, and 1.192)
would be retained in paragraph (a)(3)(i) through (a)(3)(iii) of Sec.
50.55a, where the RGs are listed by full title, including revision
number. These proposed changes would 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.
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 Code (73 FR 52730). As part of
the final rule, Sec. 50.55a(g)(6)(ii)(D) implemented an augmented
inservice inspection program for the examination of reactor pressure
vessel (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 welds by primary water stress corrosion
cracking (PWSCC). The safety consequences of inadequate inspections can
be significant. The NRC's determination that the ASME 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
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 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 proposes 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 NRC conditions have been modified to
address the changes in ASME BPV Code Case N-729-4. The NRC's proposed
revisions to the conditions on ASME BPV Code Case N-729-1 are discussed
in the next four sections.
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
The NRC proposes to revise 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 the final rule.
10 CFR 50.55a(g)(6)(ii)(D)(2) Through (6)
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(D)(2) through (6)
to remove the conditions currently in Sec. 50.55a(g)(6)(ii)(D)(2)
through (5) and to redesignate the condition currently in Sec.
50.55a(g)(6)(ii)(D)(6) as Sec. 50.55a(g)(6)(ii)(D)(2). The conditions
currently in Sec. 50.55a(g)(6)(ii)(D)(2) to Sec.
50.55a(g)(6)(ii)(D)(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. The NRC proposes to include new conditions
in Sec. 50.55a(g)(6)(ii)(D)(3) and (4) as described in the following
discussion.
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency
The NRC proposes to adopt a new condition (to be included in
proposed Sec. 50.55a(g)(6)(ii)(D)(3)) to modify the option to extend
bare metal visual inspections of the reactor pressure vessel upper head
surface beyond the frequency listed in Table 1 of ASME BPV Code Case N-
729-4. 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 said to have
``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.
[[Page 56836]]
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 no flaw 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 temperature, 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 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 on 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 temperature 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 proposes to
increase the frequency of the bare metal visual examinations of ``cold
heads'' to identify potential leakage as soon as reasonably possible
because of the volumetric examination limitations. Therefore, the NRC
proposes to condition Note 4 of ASME BPV Code Case N-729-4 to require a
bare metal visual exam each outage in which a volumetric exam is not
performed. The NRC also proposes to 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 Code meetings, at the annual Materials
Programs Technical Information Exchange public meeting 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
The NRC proposes to adopt a new condition (to be included in
proposed 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 Code Committee approved an interpretation of the
language in Paragraph -3132(b) that implied 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 stance 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
stance in implementing Paragraph -3132(b), even after the recent ASME
Code Committee interpretation approval over NRC objection.
Therefore, in order to ensure compliance with the previous and
ongoing requirement, the NRC proposes to revise 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, the NRC issued a final rule including Sec.
50.55a(g)(6)(ii)(F) requiring 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 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 finds that the updates and improvements in N-770-2 are
sufficient to update Sec. 50.55a(g)(6)(ii)(F).
[[Page 56837]]
The NRC therefore proposes to update 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
proposed changes to the conditions on ASME BPV Code Case N-770-2.
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
The NRC proposes to revise 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 proposes that licensees
begin using N-770-2 on the effective date of this rule.
10 CFR 50.55a(g)(6)(ii)(F)(2) Categorization
The NRC proposes to revise 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 proposes
to 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 proposes to revise 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. The 2011 rule required the use of ASME 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 current 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 proposes to revise 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 proposes to establish an essentially 100 percent
volumetric examination coverage requirement for circumferential flaws
to provide reasonable assurance of structural integrity of all ASME
Code Class 1 butt welds susceptible to PWSCC. Therefore, the NRC
proposes to adopt 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(5) to add the
explanatory heading, ``Inlay/onlay inspection frequency,'' and to make
minor editorial corrections.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
The NRC proposes to revise 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 proposes to revise 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(8) 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 stresses. 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 proposes to continue
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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(9) 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 proposes to adopt 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(10) 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
[[Page 56838]]
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 full examination requirement should be implemented,
if possible, before the option of reduced examination coverage is
allowed. Therefore, the NRC proposes to modify 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 proposes to add 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 Class 1 welds, the NRC has established
the requirement for examination qualification under the Appendix VIII.
Thus, the NRC proposes to adopt 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, 2020.
The development of a sufficient number of mockups would be required
to establish an Appendix VIII program for examination of ASME 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 proposes that
licensees be required to use these Appendix VIII qualifications no
later than their first scheduled weld examinations involving cast
stainless steel materials occurring after January 1, 2020.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
The NRC proposes to add 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 proposes to adopt a condition
changing the inspection volume for stress-improved dissimilar metal
welds with cast stainless steel from the ASME Code Section XI
requirements to ``the maximum extent practical including 100 percent of
the susceptible material volume.'' This will remain applicable until an
Appendix VIII qualified procedure for the inspection through cast
stainless steel materials is available in accordance with the proposed
condition in Sec. 50.55a(g)(6)(ii)(F)(11).
10 CFR 50.55a(g)(6)(ii)(F)(13) Encoded Ultrasonic Examination
The NRC proposes to add 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 recent examinations that
failed to identify significant flaws at North Anna Power Station, Unit
1, in 2012 (Licensee Event Report 50-338/2012-001-00, ADAMS Accession
No. ML12151A441) and at Diablo Canyon Nuclear Power Plant in 2013
(Relief Request REP-1 U2, Revision 2, ADAMS Accession No. ML13232A308)
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 limiting 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 rate. Therefore, the NRC proposes to adopt 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 proposes to add 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 NRC-proposed
conditions of Sec. 50.55a(b)(2)(xxxvii)(A) through (E), 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
[[Page 56839]]
adequately. To illustrate that ASME Code does not provide adequate
guidance, ASME 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 Code requirements for inservice inspection of CASS
components to the NRC, resulting in a significant regulatory burden.
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.
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. These NUREG/CR reports show that CASS materials less than 1.6
inches thick can be reliably inspected for flaws 10 percent through-
wall or deeper if encoded phased-array examinations are performed using
low ultrasonic frequencies and a sufficient number of inspection
angles. Additionally, for thicker welds, flaws greater than 30 percent
through-wall in depth can be detected using low frequency encoded
phased-array ultrasonic inspections.
The NRC, using 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
finds that encoding CASS inspection data provides significant detection
benefits. The NRC proposes to add 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. The use of dual
element phased-array search units showed the most promise in obtaining
meaningful responses from flaws. The NRC proposes to add 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. The
optimum inspection frequencies for examining CASS components of various
thicknesses as described in NUREG/CR-6933 and NUREG/CR-7122 are
reflected in proposed conditions Sec. 50.55a(b)(2)(xxxvii)(C) and (D).
The NRC proposes to add 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 less than or equal to 1.6 inches thick shall
use a phased array search unit with a center frequency of 500 kHz to 1
MHz. The NRC proposes to add a condition in Sec.
50.55a(b)(2)(xxxvii)(D) to require that ultrasonic examinations
performed to implement ASME BPV Code Case N-824 on piping greater than
1.6 inches thick shall use a phased array search unit with a center
frequency of 500 kHz. As NUREG/CR-6933 shows that the grain structure
of CASS can reduce the effectiveness of some inspection angles, the NRC
finds sufficient technical basis to condition the code case for the use
of phased-array ultrasound using angles from 30 to 70 degrees with a
maximum increment of 5 degrees. The NRC proposes to add a condition in
Sec. 50.55a(b)(2)(xxxvii)(E) to require that ultrasonic examinations
performed to implement ASME BPV Code Case N-824 shall use a phased
array search unit which produces angles from 30 to 70 degrees with a
maximum increment of 5 degrees.
Obtaining effective examination results of CASS components requires
using 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.
The recent advances in inspection technology are driving renewed
work at ASME 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 Code Section XI,
Appendix III, Supplement 1 which states that inspection of CASS
materials meeting the ASME 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.
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,
as conditioned by Sec. 50.55a(b)(2)(xxxvii)(A) through (E), 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.
ASME OM Code Case OMN-20
10 CFR 50.55a(b)(3)(x) OM Condition: ASME OM Code Case OMN-20
The NRC proposes to add new paragraph Sec. 50.55a(b)(3)(x) to
allow the use of ASME OM Code Case OMN-20, ``Inservice Test
Frequency,'' which provides inservice test frequencies for pumps and
valves which a licensee may voluntarily use in place of the frequencies
specified in the 2012 Edition of the ASME OM Code. Paragraph Sec.
50.55a(a)(1)(iii)(E) would be added to incorporate ASME OM Code Case
OMN-20 by reference into Sec. 50.55a. Surveillance Requirement (SR)
3.0.3 from Technical Specification (TS) 5.5.6, ``Inservice Testing
Program,''
[[Page 56840]]
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 ASME 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 intends to 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)(E).
IV. Section-by-Section Analysis
The NRC proposes to remove 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. The revision numbers for the RGs approved for
incorporation by reference would be retained in paragraph (a) of Sec.
50.55a, 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. No substantive change is intended by the NRC by this
proposed amendment. Readers would need to refer to paragraph (a) of
Sec. 50.55a to determine the specific revision of the relevant RG
which is approved for incorporation by reference by Office of the
Federal Register.
10 CFR 50.55a(a) Documents Approved for Incorporation by Reference
The NRC proposes to revise the incorporation by reference language
to update the contact information for the NRC Technical Library.
10 CFR 50.55a(a)(1)(i) ASME Boiler and Pressure Vessel Code, Section
III
The NRC proposes to revise Sec. 50.55a(a)(1)(i) to clarify 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 ``(excluding Nonmandatory Appendices)'' in 10 CFR
50.55a(a)(1)(i).
10 CFR 50.55a(a)(1)(i)(E) ``Rules for Construction of Nuclear Facility
Components--Division 1''
The NRC proposes to revise 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 proposes to revise Sec. 50.55a(a)(1)(ii) to include a
minor editorial change and to clarify that Nonmandatory Appendix U is
not incorporated by reference.
10 CFR 50.55a(a)(1)(ii)(C) ``Rules for Inservice Inspection of Nuclear
Power Plant Components--Division 1''
The NRC proposes to revise Sec. 50.55a(a)(1)(ii)(C) to add ASME
BPV Code, Section XI 2009 Addenda, 2010 Edition, 2011 Addenda, and 2013
Edition.
10 CFR 50.55a(a)(1)(iii)(B) ASME BPV Code Case N-729-4
The NRC proposes to revise Sec. 50.55a(a)(1)(iii)(B) 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)(C) ASME BPV Code Case N-770-2
The NRC proposes to revise Sec. 50.55a(a)(1)(iii)(C) 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)(D) ASME BPV Code Case N-824
The NRC proposes to add Sec. 50.55a(a)(1)(iii)(D) to add the title
``ASME BPV Code Case N-824,'' and include information for the standard
that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iii)(E) ASME OM Code Case OMN-20
The NRC proposes to add Sec. 50.55a(a)(1)(iii)(E) to add the title
``ASME OM Code Case OMN-20,'' and include information for the standard
that is being incorporated by reference.
10 CFR 50.55a(a)(1)(iv) ASME Operation and Maintenance Code
The NRC proposes to revise Sec. 50.55a(a)(1)(iv) to correct the
title of the OM Code.
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 proposes to revise Sec. 50.55a(a)(1)(iv)(B) to add 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 proposes to add Sec. 50.55a(a)(1)(iv)(C) to add ASME OM
Code 2012 Edition.
10 CFR 50.55a(a)(1)(v) ASME Quality Assurance Requirements
The NRC proposes to add Sec. 50.55a(a)(1)(v) to add the title
``ASME Quality Assurance Requirements'' for ASME NQA-1 Code as part of
NRC titling convention and include information regarding NQA-1
standards.
10 CFR 50.55a(b) Use and Conditions on the Use of Standards
The NRC proposes to revise Sec. 50.55a(b) to correct the title of
the OM Code.
10 CFR 50.55a(b)(1) Conditions on ASME BPV Code Section III
The NRC proposes to revise 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 proposes to revise Sec. 50.55a(b)(1)(ii) to clarify rule
language and add Table 1, which clarifies prohibited Section III
provisions in tabular form for welds with leg size less than 1.09
tn.
10 CFR 50.55a(b)(1)(iv) Section III Condition: Quality Assurance
The NRC proposes to revise 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 2011 Addenda.
[[Page 56841]]
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 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. 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. Regulatory Guide 1.28, ``Quality
Assurance Criteria (Design and Construction),'' provides additional
guidance and regulatory positions on how to meet appendix B when using
NQA-1.
Section 50.55a(b)(1)(iv) clarifies that applicants and licensees
are required to meet appendix B to 10 CFR part 50 and that the
commitments contained in their QA program descriptions that are more
stringent than those contained in NQA-1 or are not addressed in NQA-1
apply to Section III activities.
10 CFR 50.55a(b)(1)(vii) Section III Condition: Capacity Certification
and Demonstration of Function of Incompressible-Fluid Pressure-Relief
Valves
The NRC proposes to revise Sec. 50.55a(b)(1)(vii) to reflect the
latest edition incorporated by reference, the 2013 Edition.
10 CFR 50.55a(b)(1)(viii) Section III Condition: Use of ASME
Certification Marks
The NRC proposes to add 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 10 CFR 50.55a.
10 CFR 50.55a(b)(2) Conditions on ASME BPV Code, Section XI
The NRC proposes to revise Sec. 50.55a(b)(2) to reflect the latest
edition incorporated by reference, the 2013 Edition, and to clarify
that Nonmandatory Appendix U is not incorporated by reference.
10 CFR 50.55a(b)(2)(vi) Section XI Condition: Effective Edition and
Addenda of Subsection IWE and Subsection IWL
The NRC proposes to revise 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 inservice inspection 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 proposes to revise Sec. 50.55a(b)(2)(viii) by removing the
condition for using the 2009 Addenda up to and including the 2013
Edition of Subsection IWL requiring compliance with Sec.
50.55a(b)(2)(viii)(E).
10 CFR 50.55a(b)(2)(viii)(H) Concrete Containment Examinations: Eighth
Provision
The NRC proposes to add Sec. 50.55a(b)(2)(viii)(H) to require
licensees to provide the applicable information specified in paragraphs
(b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and (b)(2)(viii)(E)(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 proposes to add Sec. 50.55a(b)(2)(viii)(I) containing 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 proposed
condition would apply 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 2013 Edition.
10 CFR 50.55a(b)(2)(ix) Section XI Condition: Metal Containment
Examinations
The NRC proposes to revise Sec. 50.55a(b)(2)(ix) to continue to
apply the existing conditions in Sec. 50.55a(b)(2)(ix)(A)(2), Sec.
50.55a(b)(2)(ix)(B) and Sec. 50.55a(b)(2)(ix)(J) with respect to the
metal containment examination requirements in Subsection IWE to the
2009 Addenda up to and including the 2013 Edition and to make minor
editorial corrections.
10 CFR 50.55a(b)(2)(ix)(D) Metal Containment Examinations: Fourth
Provision
The NRC proposes to revise the rule text in Sec.
50.55a(b)(2)(ix)(D) to improve clarity. Paragraphs Sec.
50.55a(b)(2)(ix)(D) and Sec. 50.55a(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 proposes to revise Sec. 50.55a(b)(2)(x) to clarify that it
allows, but does not require, licensees to use the 1994 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. To meet the requirements
of appendix B, when using NQA-1 during inservice inspection 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 is
provided in RG 1.28, ``Quality Assurance Criteria (Design and
Construction).''
[[Page 56842]]
Section 50.55a(b)(2)(x) clarifies that licensees are required to
meet appendix B to 10 CFR part 50 and that the commitments contained in
their QA program descriptions that are more stringent than those
contained in NQA-1 or are not addressed in NQA-1 apply to Section XI
activities.
10 CFR 50.55a(b)(2)(xviii)(D) NDE Personnel Certification: Fourth
Provision
The NRC proposes to add 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 would be 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 proposes to revise 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 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. This revision
removes a requirement that was in addition to 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)(xxx) Section XI Condition: Steam Generator
Preservice Examinations
The NRC proposes to add Sec. 50.55a(b)(2)(xxx) to provide a new
condition requiring that instead of the preservice inspection
requirements of Section XI, IWB-2200(c), a full length examination of
100 percent of the tubing in each newly installed steam generator shall
be performed prior to plant startup. These inspections shall be
performed with the objective of finding the types of flaws that may
potentially be present in the tubes and that may potentially occur
during operation.
10 CFR 50.55a(b)(2)(xxxi) Section XI Condition: Mechanical Clamping
Devices
The NRC proposes to add Sec. 50.55a(b)(2)(xxxi) to provide a new
condition prohibiting 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.
10 CFR 50.55a(b)(2)(xxxii) Section XI Condition: Summary Report
Submittal
The NRC proposes to add 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.
10 CFR 50.55a(b)(2)(xxxiii) Section XI Condition: Risk-Informed
Allowable Pressure
The NRC proposes to add 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: Disposition of Flaws
in Class 3 Components
The NRC proposes to add Sec. 50.55a(b)(2)(xxxiv) to provide a new
condition to require that when using the 2013 Edition of the ASME BPV
Code, Section XI, the licensee shall use the acceptance standards of
IWD-3510 for the disposition of flaws in Category D-A components (i.e.,
welded attachments for vessels, piping, pumps, and valves).
10 CFR 50.55a(b)(2)(xxxv) Section XI Condition: Use of RTT0
in the KIa and KIc Equations
The NRC proposes to add 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 proposes to add 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 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 proposes to add Sec. 50.55a(b)(2)(xxxvii) with
subparagraphs (A) through (E) 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,'' as conditioned by subparagraphs (A) through
(E).
10 CFR 50.55a(b)(2)(xxxvii)(A) Section XI Condition: ASME BPV Code Case
N-824
The NRC proposes to add 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.
10 CFR 50.55a(b)(2)(xxxvii)(B) Section XI Condition: ASME BPV Code Case
N-824
The NRC proposes to add 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 proposes to add 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 less than or equal to 1.6
inches thick shall use a phased array search unit with a center
frequency of 500 kHz to 1 MHz instead of the requirements of Paragraph
1(c)(1)(-c)(-1).
10 CFR 50.55a(b)(2)(xxxvii)(D) Section XI Condition: ASME BPV Code Case
N-824
The NRC proposes to add Sec. 50.55a(b)(2)(xxxvii)(D) to add a new
[[Page 56843]]
condition that requires that ultrasonic examinations performed to
implement ASME BPV Code Case N-824 on piping greater than 1.6 inches
thick shall use a phased array search unit with a center frequency of
500 kHz instead of the requirements of Paragraph 1(c)(1)(-c)(-2).
10 CFR 50.55a(b)(2)(xxxvii)(E) Section XI Condition: ASME BPV Code Case
N-824
The NRC proposes to add Sec. 50.55a(b)(2)(xxxvii)(E) 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 from 30 to 70 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 proposes to revise Sec. 50.55a(b)(3) to require that the
2012 Edition of the ASME OM Code be used during the initial 120-month
inservice test interval under Sec. 50.55a(f)(4)(i) and during
mandatory 120-month IST program updates under Sec. 50.55a(f)(4)(ii).
The proposed revision would also allow users to voluntarily update
their IST programs to the 2009 Edition, 2011 Addenda, or 2012 Edition
of the ASME OM Code (with the exceptions and conditions specified in
this notice) under Sec. 50.55a(f)(4)(iv).
10 CFR 50.55a(b)(3)(i) OM Condition: Quality Assurance
The NRC proposes to revise Sec. 50.55a(b)(3)(i) to allow licensees
to use the 1983 Edition through the 1994 Edition, 2008 Edition, and
2009-1a Addenda of NQA-1 when using the 1995 Edition through the 2012
Edition of the ASME 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. 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.
Regulatory Guide 1.28, ``Quality Assurance Criteria (Design and
Construction),'' provides additional guidance on how to meet appendix B
when using NQA-1.
Paragraph 50.55a(b)(3)(i) clarifies that licensees are required to
meet appendix B to 10 CFR part 50 and that the commitments contained in
their QA program descriptions that are more stringent than those
contained in NQA-1 or are not addressed in NQA-1 apply to OM Code
activities.
10 CFR 50.55a(b)(3)(ii) OM Condition: Motor-Operated Valve (MOV)
Testing
The NRC proposes to revise Sec. 50.55a(b)(3)(ii) to reflect
Appendix III, ``Preservice and Inservice Testing of Active Electric
Motor Operated Valve Assemblies in Light-Water Reactor Power Plants,''
in the ASME OM Code, 2009 Edition, 2011 Addenda, and 2012 Edition.
10 CFR 50.55a(b)(3)(ii)(A) MOV Diagnostic Test Interval
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(A) to require 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 Appendix III of the ASME OM Code.
10 CFR 50.55a(b)(3)(ii)(B) MOV Testing Impact on Risk
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(B) 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.
10 CFR 50.55a(b)(3)(ii)(C) MOV Risk Categorization
The NRC proposes to add Sec. 50.55a(b)(3)(ii)(C) to require, when
applying Appendix III to the ASME OM Code, that licensees categorize
MOVs according to their safety significance using the methodology
described in ASME 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 proposes to add 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 the MOV satisfies the assumptions in the plant safety analyses.
10 CFR 50.55a(b)(3)(iii) OM Condition: New Reactors
The NRC proposes to add 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 the effective date of this rulemaking and holders of COLs issued
under 10 CFR part 52, whose initial fuel loading occurs on or after the
date 12 months after the effective date of this rulemaking, shall also
comply with specified conditions, as applicable.
10 CFR 50.55a(b)(3)(iii)(A) Power-Operated Valves
The NRC proposes to add Sec. 50.55a(b)(3)(iii)(A) to require that
licensees subject to Sec. 50.55a(b)(3)(iii) develop a program to
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
The NRC proposes to add 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
The NRC proposes to add 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.
10 CFR 50.55a(b)(3)(iii)(D) High Risk Non-Safety Systems
The NRC proposes to add 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 (RTNSS) for applicable reactor designs. The proposed rule
language refers to such components using the term, ``high risk non-
safety systems.''
[[Page 56844]]
10 CFR 50.55a(b)(3)(iv) OM Condition: Check Valves (Appendix II)
The NRC proposes to revise Sec. 50.55a(b)(3)(iv) to specify that
Appendix II in the 2003 Addenda through the 2012 Edition of the OM Code
is acceptable for use without conditions with the clarifications that
(1) the maximum test interval allowed by Appendix II for individual
check valves in a group of two valves or more must be supported by
periodic testing of a sample of check valves in the group during the
allowed interval and (2) the periodic testing plan must be designed to
test each valve of a group at approximate equal intervals not to exceed
the maximum requirement interval. 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 proposed rule.
10 CFR 50.55a(b)(3)(vii) OM Condition: Subsection ISTB
The NRC proposes to add Sec. 50.55a(b)(3)(vii) to prohibit the use
of Subsection ISTB in the 2011 Addenda to the ASME OM Code.
10 CFR 50.55a(b)(3)(viii) OM Condition: Subsection ISTE
The NRC proposes to add 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 ASME 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
ASME OM Code.
10 CFR 50.55a(b)(3)(ix) OM Condition: Subsection ISTF
The NRC proposes to add 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 proposed paragraph will also state 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 proposes to add Sec. 50.55a(b)(3)(x) to allow licensees to
implement ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' in
the ASME OM Code, 2012 Edition.
10 CFR 50.55a(b)(3)(xi) OM Condition: Valve Position Indication
The NRC proposes to add Sec. 50.55a(b)(3)(xi) to require that
licensees supplement the ASME OM Code provisions in Subsection ISTC-
3700, ``Position Verification Testing,'' as necessary to verify that
valve operation is accurately indicated. The ASME 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.
10 CFR 50.55a(f): Inservice Testing Requirements
The NRC proposes to revise Sec. 50.55a(f) to clarify that the ASME
OM Code includes provisions for preservice testing of components as
part of its overall provisions for IST programs.
10 CFR 50.55a(f)(3)(iii)(A) Class 1 Pumps and Valves: First Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iii)(A) to state that
the paragraph is applicable to pumps and valves that are within the
scope of the ASME OM Code. This will align the scope of pumps and
valves for inservice testing with the scope defined in the ASME Code
and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iii)(B) Class 1 Pumps and Valves: Second Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iii)(B) to ensure
that the paragraph is applicable to pumps and valves that are within
the scope of the ASME OM Code. This will align the scope of pumps and
valves for inservice testing with the scope defined in the ASME Code
and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(A) Class 2 and 3 Pumps and Valves: First
Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iv)(A) to ensure that
the paragraph is applicable to pumps and valves that are within the
scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(A)
for Class 1 pumps and valves. This will align the scope of pumps and
valves for inservice testing with the scope defined in the ASME Code
and in SRP Section 3.9.6.
10 CFR 50.55a(f)(3)(iv)(B) Class 2 and 3 Pumps and Valves: Second
Provision
The NRC proposes to revise Sec. 50.55a(f)(3)(iv)(B) to ensure that
the paragraph is applicable to pumps and valves that are within the
scope of the ASME OM Code and not covered by paragraph (f)(3)(iii)(B)
for Class 1 pumps and valves. This will align the scope of pumps and
valves for inservice testing with the scope defined in the ASME Code
and in SRP Section 3.9.6.
10 CFR 50.55a(f)(4) Inservice Testing Standards Requirement for
Operating Plants
The NRC proposes to revise Sec. 50.55a(f)(4) to ensure that the
paragraph is applicable to pumps and valves that are within the scope
of the ASME OM Code. This will align the scope of pumps and valves for
inservice testing with the scope defined in the ASME Code and in SRP
Section 3.9.6.
10 CFR 50.55a(g) Inservice and Preservice Inspection Requirements
The NRC proposes to add new paragraphs (g)(2)(i), (g)(2)(ii), and
(g)(2)(iii) and to revise paragraphs (g), (g)(2), (g)(3), (g)(3)(i),
(g)(3)(ii), and (g)(3)(v) to distinguish the requirements for
accessibility and preservice examination from those for inservice
inspection in Sec. 50.55a(g). No substantive change to the
requirements is intended by these revisions.
10 CFR 50.55a(g)(6)(ii)(D)(1) Implementation
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(D)(1) to require
licensees to implement an augmented inservice inspection program for
the examination of the 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.
10 CFR 50.55a(g)(6)(ii)(D)(2) Through (5) of the Current Regulation
The NRC proposes to remove the conditions in existing Sec.
50.55a(g)(6)(ii)(D)(2) through (5) of the current regulation, inasmuch
as these conditions have been included in or reflected in other Code
requirements. In their place, the NRC proposes to adopt new conditions
in Sec. 50.55a(g)(6)(ii)(D)(2) through (4).
10 CFR 50.55a(g)(6)(ii)(D)(2) Appendix I Use
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(D)(2) to require
NRC approval prior to implementing Appendix I of ASME BPV Code Case N-
729-4. This requirement is currently located in Sec.
50.55a(g)(6)(ii)(D)(6) for implementation of N-729-1.
[[Page 56845]]
10 CFR 50.55a(g)(6)(ii)(D)(3) Bare Metal Visual Frequency
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(D)(3) to add a
new condition which requires cold head plants (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
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.
The condition that is in the current Sec. 50.55a(g)(6)(ii)(D)(3) was
incorporated into N-729-4 by the ASME Code committees.
10 CFR 50.55a(g)(6)(ii)(D)(4) Surface Exam Acceptance Criteria
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(D)(4) to add a
new condition which clarifies 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 Section III of the current
edition and addenda for the licensee's ongoing 10-year inservice
inspection interval. The condition that is in the current Sec.
50.55a(g)(6)(ii)(D)(4) was incorporated into N-729-4 by the ASME Code
committees.
10 CFR 50.55a(g)(6)(ii)(F)(1) Implementation
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(F)(1) to require
licensees to implement an augmented inservice inspection 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 proposes to revise 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 proposes to revise 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. Additionally, the NRC proposes to add 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(3) to clarify
the current requirement in this paragraph to complete baseline
examinations. Additionally, this condition clarifies that the
examination coverage requirements, for a licensee to count previous
inspections as baseline examinations, are the same examination coverage
requirements described in 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(4) to clarify
that licensees are required to ensure 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.
10 CFR 50.55a(g)(6)(ii)(F)(5) Inlay/Onlay Inspection Frequency
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(F)(5) to add
explanatory heading and to make minor editorial corrections.
10 CFR 50.55a(g)(6)(ii)(F)(6) Reporting Requirements
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(F)(6) to add
explanatory heading.
10 CFR 50.55a(g)(6)(ii)(F)(7) Defining ``t''
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(F)(7) to add
explanatory heading.
10 CFR 50.55a(g)(6)(ii)(F)(8) Optimized Weld Overlay Examination
The NRC proposes to revise Sec. 50.55a(g)(6)(ii)(F)(8) 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(9) 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 proposes to revise Sec. 50.55a(g)(6)(ii)(F)(10) to modify
the current condition to allow the previously prohibited alternate
examination requirements of Note (b) of Figure 5(a) of ASME BPV Code
Case 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 ASME
BPV Code Case N-770-2 cannot be met.
10 CFR 50.55a(g)(6)(ii)(F)(11) Cast Stainless Steel
The NRC proposes to add 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 and through
cast stainless steel to meet the examination requirements of paragraph
-2500(a) of
[[Page 56846]]
ASME BPV Code Case N-770-2 by January 1, 2020.
10 CFR 50.55a(g)(6)(ii)(F)(12) Stress Improvement Inspection Coverage
The NRC proposes to add Sec. 50.55a(g)(6)(ii)(F)(12) to provide a
new condition that would allow licenses 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 proposes to add Sec. 50.55a(g)(6)(ii)(F)(13) to provide a
new condition requiring licensees to perform encoded examinations of
essentially 100 percent of the inspection surface area when required to
perform volumetric examinations of all non-mitigated and cracked
mitigated butt welds in accordance with N-770-2.
V. Generic Aging Lessons Learned Report
Background
In December 2010, the NRC issued ``Generic Aging Lessons Learned
(GALL) Report,'' 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 10 CFR 54.21(a)(3). In addition, ``Standard Review
Plan for Review of License Renewal Applications for Nuclear Power
Plants,'' NUREG-1800, Revision 2 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 also issued ``Disposition of Public Comments and
Technical Bases for Changes in the License Renewal Guidance Documents
NUREG-1801 and NUREG-1800,'' NUREG-1950, 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 aging management programs 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 10 CFR 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 also 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 10 CFR 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
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 Code Class 1 Small-Bore Piping,'' includes the augmentation of
the requirements in ASME BPV Code, Section XI, Subsection IWB to
perform a one-time inspection of a sample of ASME 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
[[Page 56847]]
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.
VI. Specific Request for Comments
The NRC requests specific comments on the following questions:
NRC Question 1. NQA-1. The NRC is considering removing the
references to versions of NQA-1 older than the 1994 Edition in Sec.
50.55a(b)(1)(iv), Sec. 50.55a(b)(2)(x), and Sec. 50.55a(b)(3)(i). The
NRC requests public comment on whether any applicant or licensee is
committed to, and is using, a version of NQA-1 older than the 1994
Edition, and if so, what version the applicant or licensee is using.
NRC Question 2. ASME BPV Code Case N-824. The NRC is proposing to
make ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast
Austenitic Piping Welds From the Outside Surface Section XI, Division
1,'' acceptable for use with conditions. The use of N-824, as
conditioned, is considered a stop-gap improvement until ASME Section XI
Appendix VIII Supplement 9 is developed and implemented. The NRC is
considering whether ASME BPV Code Case N-824, as conditioned, should be
mandatory because of the potential that licensees may continue to use
less effective ASME Code Section XI Appendix III techniques for
examinations of welds next to CASS material. Should ASME BPV Code Case
N-824, as conditioned, be mandatory? What are the possible advantages
and disadvantages of making N-824, as conditioned, mandatory?
VII. 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). The NRC requests comment on this document with respect to the
clarity and effectiveness of the language used.
VIII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113 (NTTAA), and implementing guidance in U.S. Office of
Management and Budget (OMB) Circular A-119 (February 10, 1998),
requires that Federal agencies use technical standards that are
developed or adopted by voluntary consensus standards bodies unless
using such a standard is inconsistent with applicable law or is
otherwise impractical. 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 Circular A-119 prohibit an agency from adopting a
voluntary consensus standard while taking exception to specific
portions of the standard, if those provisions are deemed to be
``inconsistent with applicable law or otherwise impractical.''
Furthermore, taking specific exceptions furthers the Congressional
intent of Federal reliance on voluntary consensus standards because it
allows the adoption of substantial portions of consensus standards
without the need to reject the standards in their entirety because of
limited provisions that are not acceptable to the agency.
In this rulemaking, the NRC is continuing its existing practice of
establishing requirements for the design, construction, operation,
inservice inspection (examination) and inservice testing of nuclear
power plants by approving the use of the latest editions and addenda of
the ASME BPV and OM Codes (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.
As discussed in Section III of this statement of considerations, in
this proposed rule the NRC is conditioning the use of certain
provisions of the 2009 Addenda, 2010 Edition, 2011 Addenda, and the
2013 Edition to the ASME BPV Code, Section III, Division 1 and the 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
to the ASME OM Code and Code Cases N-770-2, N-729-4, and N-824. In
addition, the proposed rule does not adopt (``excludes'') certain
provisions of the ASME Codes and this statement of considerations, and
in the regulatory and backfit analysis for this rulemaking. The NRC
believes that this proposed 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 as proposed
in this rulemaking (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 that are the subject of this rulemaking are acceptable for
use (in some cases with conditions). For these reasons, the NRC
concludes that this proposed 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 U.S. that the NRC could incorporate by reference instead of the
ASME Codes. The NRC also did not identify any voluntary consensus
standards developed by multinational voluntary consensus standards
bodies for use on a multinational basis that the NRC could incorporate
by reference instead of the ASME 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. Evaluation by the NRC of the countries' codes to determine
whether each code provides a comparable or enhanced level of safety
[[Page 56848]]
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 proposed rulemaking satisfies the requirements of
the NTTAA and OMB Circular A-119.
IX. Incorporation by Reference--Reasonable Availability to Interested
Parties
The NRC proposes to incorporate by reference seven recent editions
and addenda to the ASME codes for nuclear power plants and a standard
for quality assurance. The NRC is also proposing to incorporate by
reference four ASME code cases. As described in the ``Background'' and
``Discussion'' sections of this notice, 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 proposed rule a discussion of the
ways that the materials the agency proposes to incorporate by reference
are reasonably available to interested parties or how it worked to make
those materials reasonably available to interested parties. The
discussion in this section complies with the requirement for proposed
rules as set forth in 10 CFR 51.5(a)(1).
The NRC considers ``interested parties'' to include all potential
NRC stakeholders, not only the individuals and entities regulated or
otherwise subject to the NRC's regulatory oversight. These NRC
stakeholders are not a homogenous group but vary with respect to the
considerations for determining reasonable availability. Therefore, the
NRC distinguishes between different classes of interested parties for
purposes of determining whether the material is ``reasonably
available.'' The NRC considers the following to be classes of
interested parties in NRC rulemakings with regard to the material to be
incorporated by reference:
Individuals and small entities regulated or otherwise
subject to the NRC's regulatory oversight (this class also includes
applicants and potential applicants for licenses and other NRC
regulatory approvals) and who are subject to the material to be
incorporated by reference by rulemaking. In this context, ``small
entities'' has the same meaning as a ``small entity'' under 10 CFR
2.810.
Large entities otherwise subject to the NRC's regulatory
oversight (this class also includes applicants and potential applicants
for licenses and other NRC regulatory approvals) and who are subject to
the material to be incorporated by reference by rulemaking. In this
context, ``large entities'' are those which do not qualify as a ``small
entity'' under 10 CFR 2.810.
Non-governmental organizations with institutional
interests in the matters regulated by the NRC.
Other Federal agencies, states, local governmental bodies
(within the meaning of 10 CFR 2.315(c)).
Federally-recognized and State-recognized \3\ Indian
tribes.
---------------------------------------------------------------------------
\3\ State-recognized Indian tribes are not within the scope of
10 CFR 2.315(c). However, for purposes of the NRC's compliance with
1 CFR 51.5, ``interested parties'' includes a broad set of
stakeholders, including State-recognized Indian tribes.
---------------------------------------------------------------------------
Members of the general public (i.e., individual,
unaffiliated members of the public who are not regulated or otherwise
subject to the NRC's regulatory oversight) who may wish to gain access
to the materials which the NRC proposes to incorporate by reference by
rulemaking 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 requesting that they consider enhancing public
access to these materials during the public comment period (ADAMS
Accession No. ML15085A206). 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 (ADAMS
Accession No. ML15112A064). Therefore, the seven editions and addenda
to the ASME codes for nuclear power plants, the ASME standard for
quality assurance, and the four ASME code cases which the NRC proposes
to incorporate by reference in this rulemaking are available in read-
only format at the ASME Web site http://go.asme.org/NRC.
The NRC concludes that the materials the NRC proposes to
incorporate by reference in this rulemaking are reasonably available to
all interested parties because the materials are available to all
interested parties in multiple ways and in a manner consistent with
their interest in the materials.
X. Environmental Assessment and Final Finding of No Significant
Environmental Impact
This proposed rule action 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. Sec. 4332(C); NEPA Sec. 102(C)).
The NRC has determined under NEPA, as amended, and the NRC's
[[Page 56849]]
regulations in subpart A of 10 CFR part 51, that this proposed 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. The proposed rule does
not involve non-radiological plant effluents and has no other
environmental impact. 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.
XI. Paperwork Reduction Act Statement
This proposed rule contains new or amended collections of
information subject to the Paperwork Reduction Act of 1995 (44 U.S.C.
3501 et seq.). This proposed rule has been submitted to the Office of
Management and Budget for review and approval of the information
collections.
Type of submission, new or revision: Revision.
The title of the information collection: Domestic Licensing of
Production and Utilization Facilities: Incorporation by Reference of
American Society of Mechanical Engineers Codes and Code Cases.
The form number if applicable: Not applicable.
How often the collection is required or requested: On occasion.
Who will be required or asked to respond: Power reactor licensees
and applicants for power reactors under construction.
An estimate of the number of annual responses: 320.
The estimated number of annual respondents: 104.
An estimate of the total number of hours needed annually to comply
with the information collection requirement or request: 121,600.
Abstract: This proposed 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 number of
operating nuclear power plants has decreased and the NRC has increased
its estimate of the burden associated with developing alternative
requests. Overall, the reporting burden for 10 CFR 50.55a has
increased.
The U.S. Nuclear Regulatory Commission is seeking public comment on
the potential impact of the information collections contained in this
proposed rule and on the following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
2. Is the estimate of the burden of the proposed information
collection accurate?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the proposed information collection on
respondents be minimized, including the use of automated collection
techniques or other forms of information technology?
A copy of the OMB clearance package and proposed rule is available
in ADAMS (Accession Nos. ML14141A281 and ML14258B191) or may be viewed
free of charge at the NRC's PDR, One White Flint North, 11555 Rockville
Pike, Room O-1 F21, Rockville, MD 20852. You may obtain information and
comment submissions related to the OMB clearance package by searching
on http://www.regulations.gov under Docket ID NRC-2011-0088.
You may submit comments on any aspect of these proposed information
collection(s), including suggestions for reducing the burden and on the
previously stated issues, 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: FOIA, Privacy, and Information
Collections Branch, Office of Information Services, Mail Stop: T-5 F53,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or to
Vlad Dorjets, Desk Officer, Office of Information and Regulatory
Affairs (3150-0011), NEOB-10202, Office of Management and Budget,
Washington, DC 20503; telephone 202-395-7315, email:
[email protected].
Submit comments by October 19, 2015. Comments received after this
date will be considered if it is practical to do so, but the NRC staff
is able to ensure consideration only for comments received on or before
this date.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
XII. Regulatory Analysis: Availability
The NRC has prepared a draft regulatory analysis on this proposed
rule. The analysis examines the costs and benefits of the alternatives
considered by the Commission. The NRC requests public comments on the
draft regulatory analysis. Comments on the draft analysis may be
submitted to the NRC by any method provided in the ADDRESSES section of
this notice.
XIII. 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
[[Page 56850]]
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 ASME
OM Code.
This rulemaking proposes to incorporate by reference the 2009
Addenda, 2010 Edition, 2011 Addenda, and the 2013 Edition to 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 to the ASME 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 National Technology Transfer and Advancement Act, 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. Thus, 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 the final rule incorporating
these new editions and addenda. Thus, 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 ASME OM Code
Incorporation by reference of more recent editions and addenda of
Section XI of the ASME BPV Code and the ASME 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 ASME OM Code incorporated by reference into Sec.
50.55a 12 months before the start of a new 120-month ISI and IST
interval. Thus, 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. Modifications and limitations imposed during previous routine
updates of Sec. 50.55a have established a precedent for determining
which modifications or limitations are backfits, or require a backfit
analysis (e.g., final rule dated September 10, 2008 [73 FR 52731], and
a correction dated October 2, 2008 [73 FR 57235]). The application of
the backfit requirements to modifications and limitations in the
current rule are consistent with the application of backfit
requirements to modifications and limitations in previous rules.
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: Proposed 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
proposed 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 1, which clarifies prohibited
Section III provisions in tabular form for welds with leg size less
than 1.09 tn. This proposed change would not alter the
original intent of this requirement and, therefore, would not impose a
new requirement. Therefore, this proposed change is not a backfit.
2. Revise Sec. 50.55a(b)(1)(iv), ``Section III condition: Quality
assurance,'' to require that when applying editions and addenda later
than the 1989 Edition of
[[Page 56851]]
Section III, the requirements of NQA-1, 1983 Edition through the 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 proposed 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 proposed change is
not a backfit.
3. Add a new proposed 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 10 CFR 50.55a. This proposed condition would 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, Section XI;'' 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 inservice inspection 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 proposed 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 proposed change is not a
backfit.
2. Revise Sec. 50.55a(b)(2)(viii), ``Examination of concrete
containments,'' so that when using the 2007 Edition with 2009 Addenda
through the 2013 Edition of Subsection IWL, the conditions in 10 CFR
50.55a(b)(2)(viii)(E) do not apply, but the proposed conditions in new
10 CFR 50.55a(b)(2)(viii)(H) and 10 CFR 50.55a(b)(2)(viii)(I) do apply.
This proposed revision would not require 10 CFR 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
proposed change is not a backfit because the requirements have not
changed. The revision to add the condition in 10 CFR
50.55a(b)(2)(viii)(H) captures the reporting requirements of the
current 10 CFR 50.55a(b)(2)(viii)(E) which were not included in IWL-
2512. Therefore, this proposed change is not a backfit because the
requirements have not changed. The revision to add the condition in 10
CFR 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 and applies to plants going forward
using the 2007 Edition with 2009 Addenda through the 2013 Edition of
Subsection IWL. The requirements would remain unchanged from those of
the GALL Report and, therefore, this change is not a backfit.
3. Revise Sec. 50.55a(b)(2)(ix), ``Examination of metal
containments,'' to extend the applicability of the existing conditions
in Sec. 50.55a(b)(2)(ix)(A)(2), Sec. 50.55a(b)(2)(ix)(B), and Sec.
50.55a(b)(2)(ix)(J) to the 2007 Edition with 2009 Addenda through the
2013 Edition of Subsection IWE. This proposed condition would not
result in a change to current requirements, and is therefore not a
backfit.
4. Revise Sec. 50.55a(b)(2)(x), ``Section XI condition: 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, 1983 Edition through the 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 quality assurance
program in conjunction with Section XI requirements. This proposed
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 proposed change is not a backfit.
5. Add a new proposed 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 would be required
to implement Appendix VII 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 proposed
condition is not a backfit.
6. 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 proposed revision removes a condition that was in
addition to the ASME Code requirements and does not impose a new
requirement. Therefore, this change is not a backfit.
7. Add a new proposed condition as Sec. 50.55a(b)(2)(xxx), ``Steam
Generator Preservice Examinations;'' to require that instead of the
preservice inspection requirements of Section XI, IWB-2200(c), a full
length examination of 100 percent of the tubing in each newly installed
steam generator shall be performed prior to plant startup. This
proposed condition provides a clarification consistent with industry
guidelines and the NRC staff position in SRP Section 5.4.2.2.
Therefore, the addition of this new proposed condition is not a
backfit.
8. Add a new proposed 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 would not affect licensees because the use of the
subject provisions is not currently allowed by Sec. 50.55a. Therefore,
the addition of this new proposed condition is not a backfit.
9. Add a new proposed 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 proposed condition
would not result in a change in NRC's requirements insomuch as these
reports have been required in the 2009 Addenda of Section XI and all
previous editions and
[[Page 56852]]
addenda. Therefore, the addition of this new proposed condition is not
a backfit.
10. Add a new proposed 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 proposed
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 proposed condition is not a backfit.
11. Add a new proposed condition as Sec. 50.55a(b)(2)(xxxiv),
``Disposition of flaws in Class 3 components;'' to require that when
using the 2013 Edition of the ASME BPV Code, Section XI, the licensee
shall use the acceptance standards of IWD-3510 for the disposition of
flaws in Category D-A components. The condition is imposed to provide
clarification and consistency in requirements between IWD-3410 and IWD-
3510. This proposed change would not alter the original intent of this
requirement and, therefore, would not impose a new requirement. This
proposed change is not a backfit.
12. Add a new proposed 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
proposed rule.
13. Add a new proposed 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 proposed rule.
14. Add a new proposed condition as Sec. 50.55a(b)(2)(xxxvii),
ASME BPV Code Case N-824, ``Ultrasonic Examination of Cast Austenitic
Piping Welds From the Outside Surface Section XI, Division 1,'' 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'').
ASME OM Code
1. Add a new proposed condition as Sec. 50.55a(b)(3)(ii)(A) to
require 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 Appendix III of the ASME OM Code. This
proposed condition represents an exception to a later OM Code provision
but merely retains the current NRC requirement in RG 1.192, and is
therefore not a backfit because the NRC is not imposing a new
requirement.
2. Add a new proposed condition as Sec. 50.55a(b)(3)(ii)(B) 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 proposed
condition represents an exception to a later OM Code provision but
merely retains the current NRC requirement in RG 1.192, and is
therefore not a backfit because the NRC is not imposing a new
requirement.
3. Add a new proposed condition as Sec. 50.55a(b)(3)(ii)(C) to
require, when applying Appendix III to the ASME OM Code, that licensees
categorize MOVs according to their safety significance using the
methodology described in ASME 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 proposed condition represents an exception to a later
OM Code provision but merely retains the current NRC requirement in RG
1.192, and is therefore not a backfit because the NRC is not imposing a
new requirement.
4. Add a new proposed condition as Sec. 50.55a(b)(3)(ii)(D) 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 MOV satisfies the assumptions in the plant
safety analyses. This proposed condition retains the MOV stroke time
requirement that was specified in previous editions and addenda of the
ASME OM Code. The retention of this requirement is not a backfit.
5. Add new proposed conditions as Sec. 50.55a(b)(3)(iii)(A)
through Sec. 50.55a(b)(3)(iii)(D), ``OM condition: New Reactors;'' to
apply specific conditions for IST programs applicable to licensees of
new nuclear power plants in addition to the provisions of the ASME OM
Code as incorporated by reference with conditions in Sec. 50.55a.
Licensees of ``new reactors'' are, as identified in the proposed
paragraph: (i) Holders of operating licenses for nuclear power reactors
that received construction permits under this part on or after the date
12 months after the effective date of this rulemaking and (ii) holders
of COLs issued under 10 CFR part 52, whose initial fuel loading occurs
on or after the date 12 months after the effective date of this
rulemaking. This implementation schedule for new reactors is consistent
with the NRC regulations in Sec. 50.55a(f)(4)(i). These proposed
conditions represent an exception to a later OM Code provision but
merely retain the 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), ``OM condition: 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 without conditions with the
clarifications that (1) the maximum test interval allowed by Appendix
II for individual check valves in a group of two valves or more must be
supported by periodic testing of a sample of check valves in the group
during the allowed interval and (2) the periodic testing plan must be
designed to test each valve of a group at approximate equal intervals
not to exceed the maximum requirement interval. 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. This does not represent
a change in the NRC's position that the condition is needed with
respect to the OM Code. Therefore, this proposed condition is not a
backfit.
7. Add a new proposed condition as Sec. 50.55a(b)(3)(vii), ``OM
condition: Subsection ISTB;'' to prohibit the use of Subsection ISTB in
the 2011 Addenda to the ASME OM Code because the complete set of
planned Code
[[Page 56853]]
modifications to support the changes to the comprehensive pump test
acceptance criteria was not made in that addenda. This proposed
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 proposed condition as Sec. 50.55a(b)(3)(viii), ``OM
condition: Subsection ISTE;'' to allow licensees to implement
Subsection ISTE, ``Risk-Informed Inservice Testing of Components in
Light-Water Reactor Nuclear Power Plants,'' in the ASME 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 ASME OM Code on a case-by-case basis
in accordance with Sec. 50.55a(z). This proposed 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 proposed condition as Sec. 50.55a(b)(3)(ix), ``OM
Condition: 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 ASME OM Code, 2012 Edition. The proposed condition also specifies
that Subsection ISTF, 2011 Addenda, is not acceptable for use. This
proposed 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 proposed condition as Sec. 50.55a(b)(3)(x), ``OM
condition: ASME OM Code Case OMN-20,'' to allow licensees to implement
ASME OM Code Case OMN-20, ``Inservice Test Frequency,'' in the ASME OM
Code, 2012 Edition. This proposed condition allows voluntary action
initiated by the licensee to use the code case and is, therefore, not a
backfit.
11. Add a new proposed condition as Sec. 50.55a(b)(3)(xi), ``OM
condition: Valve Position Indication,'' to specify that when
implementing ASME OM Code, Subsection ISTC-3700, ``Position
Verification Testing,'' licensees shall supplement the ASME OM Code
provisions as necessary to verify that valve operation is accurately
indicated. This proposed condition 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. As such, this proposed condition is not a
backfit.
12. Revise Sec. 50.55a(f), ``Inservice testing requirements,'' to
clarify that the ASME 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 proposed condition would not result in a change in
requirements previously approved in the Code and is, therefore, not a
backfit.
13. Revise Sec. 50.55a(f)(3)(iii)(A), ``Class 1 pumps and valves:
First provision,'' to state that the paragraph is applicable to pumps
and valves that are within the scope of the ASME OM Code. This will
align the scope of pumps and valves for inservice testing with the
scope defined in the ASME OM Code and in SRP Section 3.9.6. This
proposed condition would not result in a change in requirements
previously approved in the Code and is, therefore, not a backfit.
14. Revise Sec. 50.55a(f)(3)(iii)(B), ``Class 1 pumps and valves:
Second provision,'' to state that the paragraph is applicable to pumps
and valves that are within the scope of the ASME OM Code. This will
align the scope of pumps and valves for inservice testing with the
scope defined in the ASME OM Code and in SRP Section 3.9.6. This
proposed condition would not result in a change in requirements
previously approved in the Code and is, therefore, not a backfit.
15. Revise Sec. 50.55a(f)(3)(iv)(A), ``Class 2 and 3 pumps and
valves: First provision;'' to state that the paragraph is applicable to
pumps and valves that are within the scope of the ASME OM Code and not
covered by paragraph (f)(3)(iii)(A) for Class 1 pumps and valves. This
will align the scope of pumps and valves for inservice testing with the
scope defined in the ASME OM Code and in SRP Section 3.9.6. This
proposed condition would not result in a change in requirements
previously approved in the Code and is, therefore, not a backfit.
16. Revise Sec. 50.55a(f)(3)(iv)(B), ``Class 2 and 3 pumps and
valves: Second provision,'' to state that the paragraph is applicable
to pumps and valves that are within the scope of the ASME OM Code and
not covered by paragraph (f)(3)(iii)(B) for Class 1 pumps and valves.
This will align the scope of pumps and valves for inservice testing
with the scope defined in the ASME OM Code and in SRP Section 3.9.6.
This proposed condition would not result in a change in requirements
previously approved in the Code, and is therefore not a backfit.
17. 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 ASME OM Code. This will
align the scope of pumps and valves for inservice testing with the
scope defined in the ASME OM Code and in SRP Section 3.9.6. This
proposed condition would 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 proposes 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, NRC
concluded that imposition of ASME BPV Code Case N-729-1, 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 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
[[Page 56854]]
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 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 reactor pressure vessel 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 reactor pressure vessel 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 reactor pressure
vessel 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 organization 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 proposed rule in accordance with Sec. 50.109(a)(4)(ii)
and Sec. 50.109(a)(4)(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 proposes to update 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 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
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 organization 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,
[[Page 56855]]
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 proposed rule in
accordance with Sec. 50.109(a)(4)(ii) and Sec. 50.109(a)(4)(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 ASME 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.
XIV. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the
NRC certifies that this proposed rule does not impose a significant
economical impact on a substantial number of small entities. This
proposed rule affects only the licensing and operation of commercial
nuclear power plants. A licensee who is a subsidiary of a large entity
does not qualify as a small entity. The companies that own these plants
are not ``small entities'' as defined in the Regulatory Flexibility Act
or the size standards established by the NRC (10 CFR 2.810), as the
companies:
Provide services that are not engaged in manufacturing,
and have average gross receipts of more than $6.5 million over their
last 3 completed fiscal years, and have more than 500 employees;
Are not governments of a city, county, town, township or
village;
Are not school districts or special districts with
populations of less than 50; and
Are not small educational institutions.
XV. Availability of Documents
The NRC is making the documents identified in Table 1 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 notice.
Table 1--Availability of Documents
------------------------------------------------------------------------
Document ADAMS Accession No.
------------------------------------------------------------------------
Proposed Rule Documents:
Regulatory Analysis (includes ML14170B104.
backfitting discussion in Appendix A).
Related Documents:
Fatigue and Fracture Mechanics: 33rd ...........................
Volume, ASTM STP 1417, W.G. Reuter and
R.S. Piascik, Eds., ASTM
International, West Conshohocken, PA,
2002.
``Final Results from the CARINA Project ...........................
on Crack Initiation and Arrest of
Irradiated German RPV Steels for
Neutron Fluences in the Upper Bound,''
by AREVA at the 26th Symposium on
Effects of Radiation on Nuclear
Materials (June 12-13, 2013,
Indianapolis, IN, USA).
Letter from Brian Thomas, NRC, to ML15085A206.
Michael Merker, ASME; ``Public Access
to Material the NRC Seeks to
Incorporate by Reference into its
Regulations;'' April 9, 2015.
Letter from Michael Merker, ASME, to ML15112A064.
Brian Thomas, NRC; April 21, 2015.
Licensee Event Report 50-338/2012-001- ML12151A441.
00.
NUREG-0800, ``Standard Review Plan for ML070660036.
the 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-0800, Section 5.4.2.2, Revision ML052340627.
1, ``Steam Generator Tube Inservice
Inspection,'' July 1981.
NUREG-1482, Revision 2, ``Guidelines ML13295A020.
for Inservice Testing at Nuclear Power
Plants: Inservice Testing of Pumps and
Valves and Inservice Examination and
Testing of Dynamic Restraints
(Snubbers) at Nuclear Power Plants,''
October 2013.
NUREG-1801, Revision 2, ``Generic Aging ML103490041.
Lessons Learned (GALL) Report,''
December 2010.
NUREG-1950, ``Disposition of Public ML11116A062.
Comments and Technical Bases for
Changes in the License Renewal
Guidance Documents NUREG-1801 and
NUREG-1800,'' April 2011.
NUREG/CR-6860, ``An Assessment of ML043630040.
Visual Testing,'' November 2004.
NUREG/CR-6933, ``Assessment of Crack ML071020410 and
Detection in Heavy-Walled Cast ML071020414.
Stainless Steel Piping Welds Using
Advanced Low-Frequency Ultrasonic
Methods,'' March 2007.
NUREG/CR-7122, ``An Evaluation of ML12087A004.
Ultrasonic Phased Array Testing for
Cast Austenitic Stainless Steel
Pressurizer Surge Line Piping Welds,''
March 2012.
NRC Generic Letter 90-05, ``Guidance ML031140590.
for Performing Temporary Non-Code
Repair of ASME Code Class 1, 2, and 3
Piping (Generic Letter 90-05),'' June
1990.
NRC Meeting Summary of June 5-7, 2013, ML14003A230.
Annual Materials Programs Technical
Information Exchange Public Meeting.
NRC Memorandum, ``Consolidation of SECY- ML003708048.
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 1.83, Revision 1, ML003740256.
``Inservice Inspection of Pressurized
Water Reactor Steam Generator Tubes,''
July 1975 (withdrawn in 2009).
NRC Regulatory Guide 1.147, Revision ML13339A689.
17, ``Inservice Inspection Code Case
Acceptability, ASME Section XI,
Division 1,'' August 2014.
[[Page 56856]]
NRC Regulatory Guide 1.174, Revision 2, ML100910006.
``An Approach for Using Probabilistic
Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to
the Licensing Basis,'' May 2011.
NRC Regulatory Guide 1.175, ``An ML003740149.
Approach 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, ML090410014.
``An Approach for Determining the
Technical Adequacy of Probabilistic
Risk Assessment Results for Risk-
Informed Activities,'' March 2009.
NRC Regulatory Guide 1.201, Revision 1, 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/public-
Recent Operating Reactors Materials involve/conference-
Issues, Presentation Materials, 2013. symposia/ric/past/2013/
docs/abstracts/
sessionabstract-19.html.
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 ML003708068.
Issues Associated with the Regulatory
Treatment of Non-Safety Systems in
Passive Plant Designs''.
SECY-95-132, ``Policy and Technical ML003708005.
Issues Associated with the Regulatory
Treatment of Non-Safety Systems
(RTNSS) in Passive Plant Designs (SECY-
94-084)''.
ASME Codes, Standards, and Code Cases:
ASME BPV Code, Section III, Division 1: http://go.asme.org/NRC.
2009 Addenda, 2010 Edition, 2011
Addenda, and 2013 Edition.
ASME BPV Code, Section XI, Division 1: http://go.asme.org/NRC.
2009 Addenda, 2010 Edition, 2011
Addenda, and 2013 Edition.
ASME OM Code, Division 1: 2009 Edition, http://go.asme.org/NRC.
2011 Addenda, and 2012 Edition.
ASME Standard NQA-1: 1983 Edition http://go.asme.org/NRC.
through 1994 Edition, 2008 Edition,
and 2009-1a Addenda.
ASME BPV Code Case N-729-4............. http://go.asme.org/NRC.
ASME BPV Code Case N-770-2............. http://go.asme.org/NRC.
ASME BPV Code Case N-824............... http://go.asme.org/NRC.
ASME OM Code Case OMN-20............... http://go.asme.org/NRC.
------------------------------------------------------------------------
Throughout the development of this rulemaking, the NRC may post
documents related to this rule, including public comments, on the
Federal rulemaking Web site at http://www.regulations.gov under Docket
ID NRC-2011-0088. The Federal rulemaking Web site allows you to receive
alerts when changes or additions occur in a docket folder. To
subscribe: (1) Navigate to the docket folder for NRC-2011-0088; (2)
click the ``Sign up for Email Alerts'' link; and (3) enter your email
address and select how frequently you would like to receive emails
(daily, weekly, or monthly).
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. 553, the NRC proposes to adopt
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, Public Law 96-295, 94 Stat. 783.
0
2. In Sec. 50.55a:
0
a. Revise paragraphs (a) introductory text, (a)(1)(i) introductory
text, (a)(1)(i)(E)(12), (a)(1)(i)(E)(13) and add paragraphs
(a)(1)(i)(E)(14) through (a)(1)(i)(E)(17);
0
b. Revise paragraph (a)(1)(ii) introductory text, (a)(1)(ii)(C)(48) and
(a)(1)(ii)(C)(49) and add paragraphs (a)(1)(ii)(C)(50) through
(a)(1)(ii)(C)(53);
0
c. Revise paragraphs (a)(1)(iii)(B) and (a)(1)(iii)(C) and add
paragraphs (a)(1)(iii)(D), (a)(1)(iii)(E);
0
d. Revise paragraphs (a)(1)(iv) introductory text and add paragraphs
(a)(1)(iv)(B) and (a)(1)(iv)(C);
0
e. Add paragraph (a)(1)(v);
0
f. Revise paragraphs (b) introductory text, (b)(1) introductory text,
(b)(1)(ii), (b)(1)(iv), and (b)(1)(vii) and add paragraph (b)(1)(viii);
0
g. Revise paragraphs (b)(2) introductory text, (b)(2)(vi);
0
h. Revise paragraph (b)(2)(viii) introductory text and add paragraphs
(b)(2)(viii)(H) and (b)(2)(viii)(I);
0
i. Revise paragraphs (b)(2)(ix) introductory text, (b)(2)(ix)(D),
(b)(2)(x), add paragraph (b)(2)(xviii)(D), revise paragraph
(b)(2)(xxi)(A), and add paragraphs (b)(2)(xxx) through (b)(2)(xxxvii);
0
j. Revise paragraphs (b)(3) introductory text, (b)(3)(i), and
(b)(3)(ii), add paragraph (b)(3)(iii), revise paragraphs (b)(3)(iv)
introductory text and (b)(3)(iv)(A) though (b)(3)(iv)(D), and add
paragraphs (b)(3)(vii) through (b)(3)(xi);
[[Page 56857]]
0
k. Revise paragraphs (b)(4) introductory text, (b)(5), and (b)(6);
0
l. Revise paragraphs (f) introductory text, (f)(2), (f)(3)(iii)(A),
(f)(3)(iii)(B), (f)(3)(iv)(A), (f)(3)(iv)(B), (f)(4) introductory text,
(f)(4)(i), (f)(4)(ii);
0
m. Revise paragraphs (g) introductory text, (g)(2), (g)(3) introductory
text, (g)(3)(i), (g)(3)(ii), (g)(3)(v), (g)(4)(i), (g)(4)(ii), and
(g)(6)(ii)(D)(1) through (g)(6)(ii)(D)(4), remove paragraphs
(g)(6)(ii)(D)(5) and (g)(6)(ii)(D)(6), revise paragraphs
(g)(6)(ii)(F)(1) through (g)(6)(ii)(F)(10), and add paragraphs
(g)(6)(ii)(F)(11) through (g)(6)(ii)(F)(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 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) are listed below, but limited by
those provisions identified in paragraph (b)(1) of this section.
* * * * *
(E) * * *
(12) 2007 Edition,
(13) 2008 Addenda,
(14) 2009 Addenda,
(15) 2010 Edition,
(16) 2011 Addenda, and
(17) 2013 Edition.
(ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions
and addenda for Section XI of the ASME Boiler and Pressure Vessel Code
(excluding Nonmandatory Appendix U) are listed below, but limited by
those provisions identified in paragraph (b)(2) of this section.
* * * * *
(C) * * *
(48) 2007 Edition,
(49) 2008 Addenda,
(50) 2009 Addenda,
(51) 2010 Edition,
(52) 2011 Addenda, and
(53) 2013 Edition.
(iii) * * *
(B) 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.
(C) 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.
(D) 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 (E)
of this section.
(E) 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).
(iv) ASME Operation and Maintenance Code. The editions and addenda
for the ASME Operation and Maintenance of Nuclear Power Plants are
listed below, 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, and
(8) NQA-1c-1988 Addenda.
(9) NQA-1-1989 Edition,
(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-1a-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 Boiler and Pressure
Vessel Code (BPV Code) and the ASME Operation and Maintenance of
Nuclear Power Plants (OM Code) as specified in this paragraph. Each
combined license for a utilization facility is subject to the following
conditions.
(1) Conditions on ASME BPV Code Section III. Each manufacturing
license, standard design approval, and design certification under part
52 of this chapter is subject to the following conditions. As used in
this section, references to Section III refer to Section III of the
ASME Boiler and Pressure Vessel Code and include the 1963 Edition
through 1973 Winter Addenda and the 1974 Edition (Division 1) through
the 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 1 of this
section for welds with leg size less than 1.09 tn
[[Page 56858]]
Table 1 of Sec. 50.55a--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,'' 1983 Edition through the 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 10 CFR part 50 quality assurance program; and that commitments
contained in the applicant's or licensee's quality assurance program
description which are either more stringent than those contained in
NQA-1 or have no comparable provision in NQA-1 or Section III, govern
the applicant's or licensee's 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.
(2) Conditions on ASME BPV Code, Section XI. As used in this
section, references to Section XI refer to Section XI, Division 1, of
the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition
through the 1976 Winter Addenda and the 1977 Edition through the 2013
Edition (excluding Nonmandatory Appendix U), 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 paragraph
(b)(2)(viii)(H) and (b)(2)(viii)(I) of this section.
* * * * *
(H) Concrete containment examinations: Eighth provision. For each
inaccessible area of concrete identified for evaluation under IWL-2512,
the licensee must provide the applicable information specified in
paragraphs (b)(2)(viii)(E)(1), (b)(2)(viii)(E)(2), and
(b)(2)(viii)(E)(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 (b)(2)(ix)(F)
through (I) of this section. Applicants or licensees applying
Subsection IWE, 2004 Edition, up to and including the 2005 Addenda,
must
[[Page 56859]]
satisfy the requirements of paragraphs (b)(2)(ix)(A) and (B) and
(b)(2)(ix)(F) through (H) of this section. Applicants or licensees
applying Subsection IWE, 2004 Edition with the 2006 Addenda, must
satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and
(b)(2)(ix)(B) of this section. Applicants 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,'' 1983 Edition through
the 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 10 CFR part 50 quality assurance program in
conjunction with Section XI requirements. Commitments contained in the
licensee's quality assurance program description that are more
stringent than those contained in NQA-1 must govern Section XI
activities. Further, where NQA-1 and Section XI do not address the
commitments contained in the licensee's appendix B quality assurance
program description, the commitments must be applied to Section XI
activities.
* * * * *
(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.
* * * * *
(xxx) Section XI condition: Steam generator preservice
examinations. Prior to plant start up with a newly installed steam
generator, a 100 percent full length examination will be conducted of
the tubing in each new steam generator instead of the preservice
inspection requirements of IWB-2200(c).
(xxxi) Section XI condition: Mechanical clamping devices. The use
of mechanical clamping devices on Class 1 piping and portions of piping
systems that form the containment boundary is prohibited.
(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. 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: Disposition of flaws in Class 3
components. When using the 2013 Edition of the ASME BPV Code, Section
XI, to disposition flaws in Examination Category D-A components (i.e.,
welded attachments for vessels, piping, pumps, and valves), the
acceptance standards of IWD-3510 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
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) (-1), licensees shall use a
phased array search unit with a center frequency between 500 kHz and 1
MHz.
(D) Instead of Paragraph 1(c)(1)(-c) (-2), licensees shall use a
phased array search unit with a center frequency of 500 kHz.
(E) Instead of Paragraph 1(c)(1)(-d), the phased array search unit
must
[[Page 56860]]
produce angles from 30 to 70 degrees with a maximum increment of 5
degrees.
(3) Conditions on ASME OM Code. As used in this section, references
to the 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).
The following conditions are applicable when implementing the ASME OM
Code:
(i) OM condition: Quality assurance. When applying editions and
addenda of the OM Code, the requirements of ASME Standard NQA-1,
``Quality Assurance Requirements for Nuclear Facility Applications,''
1983 Edition through the 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 OM Code incorporated by reference
in paragraph (a)(1)(iv) of this section, provided the licensee uses its
appendix B to 10 CFR part 50 quality assurance program in conjunction
with the OM Code requirements. Commitments contained in the licensee's
quality assurance program description that are more stringent than
those contained in NQA-1 govern OM Code activities. If NQA-1 and the OM
Code do not address the commitments contained in the licensee's
appendix B quality assurance program description, the commitments must
be applied to OM Code activities.
(ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees
must comply with the provisions for testing MOVs in OM Code, ISTC 4.2,
1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(iv) of this section, and must establish a program to
ensure that MOVs continue to be capable of performing their design
basis safety functions. Licensees implementing 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 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,
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 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 OM Code, licensees
shall verify that the stroke time of the MOV satisfies the assumptions
in the plant safety analyses.
(iii) OM condition: New Reactors. In addition to complying with the
provisions in the OM Code with the conditions specified in paragraph
(b)(3) of this section, holders of operating licenses for nuclear power
reactors that received construction permits under this part on or after
the date 12 months after [the effective date of the final rule], and
holders of combined licenses issued under 10 CFR part 52, whose initial
fuel loading occurs on or after the date 12 months after [the effective
date of the final rule] 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 and 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). Appendix II, ``Check
Valve Condition Monitoring Program,'' of the OM Code, 2003 Addenda
through the 2012 Edition, is acceptable for use without conditions with
the clarifications that (1) the maximum test interval allowed by
Appendix II for individual check valves in a group of two valves or
more must be supported by periodic testing of a sample of check valves
in the group during the allowed interval and (2) the periodic testing
plan must be designed to test each valve of a group at approximate
equal intervals not to exceed the maximum requirement interval.
Licensees applying Appendix II of the 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.
* * * * *
(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 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 OM Code pursuant to Sec.
50.55a(z).
(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 not acceptable
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)(E) of this section.
(xi) OM condition: Valve Position Indication. When implementing
ASME OM Code, Subsection ISTC-3700, ``Position Verification Testing,''
licensees shall develop and implement a method to verify that valve
operation is accurately indicated by supplementing valve position
indicating lights with other indications, such as flow meters or other
suitable
[[Page 56861]]
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
incorporate 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 10 CFR 50.55a 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 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 Operation and Maintenance of Nuclear Power Plants
Code Cases. Licensees may apply the ASME Operation and Maintenance Code
Cases listed in NRC Regulatory Guide 1.192, as incorporated by
reference in paragraph (a)(3)(iii), without prior NRC approval, subject
to the following:
(i) OM Code Case condition: Applying Code Cases. When a licensee
initially applies a listed Code Case, the licensee must apply the most
recent version of that Code Case incorporated by reference in paragraph
(a) of this section.
(ii) OM Code Case condition: Applying different revisions of Code
Cases. If a licensee has previously applied a Code Case and a later
version of the Code Case is incorporated by reference in paragraph (a)
of this section, the licensee may continue to apply, to the end of the
current 120-month interval, the previous version of the Code Case, as
authorized, or may apply the later version of the Code Case, including
any NRC-specified conditions placed on its use. Licensees who choose to
continue use of the Code Case during subsequent 120-month ISI program
intervals will be required to implement the latest version incorporated
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of 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) 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 collectively as inservice testing) of the ASME BPV Code
and ASME OM 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 (f)(4) through (6) of this
section must be met only after the Commission makes the finding under
Sec. 52.103(g) of this chapter. Requirements for inservice inspection
of Class 1, Class 2, Class 3, Class MC, and Class CC components
(including their supports) are located in Sec. 50.55a(g).
* * * * *
(2) Design and accessibility requirements for performing inservice
testing in plants with CPs issued between 1971 and 1974. For a boiling
or pressurized water-cooled nuclear power facility whose construction
permit was issued on or after January 1, 1971, but before July 1, 1974,
pumps and valves that are classified as ASME Code Class 1 and Class 2
must be designed and provided with access to enable the performance of
inservice tests for operational readiness set forth in editions and
addenda of Section XI of the ASME BPV incorporated by reference in
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases
listed in NRC Regulatory Guide 1.147 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 Code Class 1 must be designed and
provided with access to enable the performance of inservice testing of
those pumps and valves within the scope of the ASME OM Code for
assessing operational readiness, as set forth in either 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) which are 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,
[[Page 56862]]
pumps and valves that are classified as ASME Code Class 1 must be
designed and provided with access to enable the performance of
inservice testing of those pumps and valves within the scope of the
ASME OM Code for assessing operational readiness, as set forth in
editions and addenda of the ASME OM Code (or the optional ASME 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 Code Class 2 and Class 3 that
are within the scope of the ASME OM Code and are not covered by
paragraph (f)(3)(iii)(A) of this section must be designed and be
provided with access to enable the performance of inservice testing of
the pumps and valves for assessing operational readiness set forth in
the editions and addenda of Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1)(ii) of this section (or
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, 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 Code Class 2 and 3 that are within
the scope of the ASME OM Code and are not covered by paragraph
(f)(3)(iii)(B) of this section 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, 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.
(i) Applicable IST Code: Initial 120-month interval. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during the initial 120-month
interval must comply with the requirements in the latest edition and
addenda of the OM Code incorporated by reference in paragraph
(a)(1)(iv) of this section on the date 12 months before the date of
issuance of the operating license under this part, or 12 months before
the date scheduled for initial loading of fuel under a combined license
under part 52 of this chapter (or the optional ASME Code Cases listed
in NRC Regulatory Guide 1.192, 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 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 Sec. 50.55a(f).
* * * * *
(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 Code Class 1 and Class 2 and
supports for components that are classified as ASME 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 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 Code Class 1, Class 2, and Class 3 and
supports for components that are classified as ASME Code Class 1, Class
2, and Class 3 must be designed and provided with 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 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
[[Page 56863]]
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 Code Class 1 and
Class 2 and supports for components that are classified as ASME 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 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 Code Class 1, Class 2, and
Class 3 and supports for components that are classified as ASME Code
Class 1, Class 2, and Class 3 must meet the preservice examination
requirements set forth in the editions and addenda of Section III or
Section XI of the ASME BPV Code incorporated by reference in paragraph
(a)(1) of this section (or the optional ASME Code Cases listed in NRC
Regulatory Guide 1.147, 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 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
Section XI, or NRC Regulatory Guide 1.192, when using the 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.
(ii) Applicable ISI Code: Successive 120-month intervals. Inservice
examination of components and system pressure tests conducted during
successive 120-month inspection intervals must comply with the
requirements of the latest edition and addenda of the Code incorporated
by reference in paragraph (a) of this section 12 months before the
start of the 120-month inspection interval (or the optional ASME Code
Cases listed in NRC Regulatory Guide 1.147, when using Section XI, or
NRC Regulatory Guide 1.192, when using the 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 July 21, 2011, may delay the update of
their Appendix VIII program by up to 18 months after July 21, 2011.
* * * * *
(6) * * *
(ii) * * *
(D) * * *
(1) Implementation: Holders of operating licenses or combined
licenses for pressurized-water reactors as of or after [the effective
date of the final rule] 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 [the effective
date of the final rule].
(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 EDY<8 and if no
flaws are found that are attributed to PWSCC; (a) A bare metal visual
examination is not required during refueling outages when a volumetric
or surface examination is performed, (b) 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 [the effective
date of the final rule] 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 [the
effective date of the final rule].
(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
[[Page 56864]]
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 previously performed or currently scheduled to be performed in an
ongoing refueling outage at the time this rule becomes effective, 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. Other previous examinations
that do not meet these requirements can be used to meet the baseline
examination requirement, provided NRC approval in accordance with
paragraphs (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 Code Case N-770-2, essentially 100 percent volumetric examination
coverage shall be obtained, including greater than 90 percent
volumetric examination coverage for circumferential flaws. Licensees
are prohibited from using Paragraph-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 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, 2020, 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
performed in accordance with the requirements of Table 1 for Inspection
Item A-1, A-2, B, E, F-2, J, and K shall be performed for essentially
100 percent of the inspection surface area using an encoded method.
* * * * *
Dated at Rockville, Maryland, this 21st day of August 2015.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2015-23193 Filed 9-17-15; 8:45 am]
BILLING CODE 7590-01-P