[Federal Register Volume 78, Number 121 (Monday, June 24, 2013)]
[Proposed Rules]
[Pages 37886-37920]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-15022]
[[Page 37885]]
Vol. 78
Monday,
No. 121
June 24, 2013
Part II
Nuclear Regulatory Commission
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10 CFR Part 50
Approval of American Society of Mechanical Engineers' Code Cases;
Proposed Rule
Federal Register / Vol. 78, No. 121 / Monday, June 24, 2013 /
Proposed Rules
[[Page 37886]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2009-0359]
RIN 3150-AI72
Approval of American Society of Mechanical Engineers' 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 the latest revisions
of three regulatory guides (RGs) approving new and revised Code Cases
published by the American Society of Mechanical Engineers (ASME). This
proposed action would allow nuclear power plant licensees, and
applicants for construction permits (CPs), operating licenses (OLs),
combined licenses (COLs), standard design certifications, standard
design approvals and manufacturing licenses, to use the Code Cases
listed in these RGs as alternatives to engineering standards for the
construction, inservice inspection (ISI), and inservice testing (IST)
of nuclear power plant components.
This rulemaking also includes consideration of a petition for
rulemaking (PRM), PRM-50-89, submitted by Mr. Raymond West. This
rulemaking also proposes resequencing NRC's requirements governing
Codes and standards in order to comply with the Office of the Federal
Register's (OFR) guidelines for incorporation by reference.
DATES: Submit comments by September 9, 2013. Comments received after
this date will be considered if it is practical to do so, but the NRC
is able to ensure consideration only of 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-2009-0359. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected]. For technical questions, contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this proposed rule.
Email comments to: [email protected]. If you do
not receive an automatic email reply confirming receipt, contact the
NRC directly 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.
You may submit comments on the information collections by the
methods indicated in the Paperwork Reduction Act Statement.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Manash K. Bagchi, Office of Nuclear
Reactor Regulation, telephone: 301-415-2905; email:
[email protected]; or Wallace Norris, Office of Nuclear Regulatory
Research, telephone: 301-251-7506; email: [email protected]; U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION
Executive Summary
The NRC is proposing to amend its regulations to incorporate by
reference the latest revisions of three NRC RGs approving new and
revised Code Cases published by the ASME. The three RGs that would be
incorporated by reference are RG 1.84, ``Design, Fabrication, and
Materials Code Case Acceptability, ASME Section III,'' Revision 36,
(DG-1230 for this proposed rule); RG 1.147, ``Inservice Inspection Code
Case Acceptability, ASME Section XI, Division 1,'' Revision 17, (DG-
1231 for this proposed rule); and RG 1.192, ``Operation and Maintenance
[OM] Code Case Acceptability, ASME OM Code,'' Revision 1 (DG-1232 for
this proposed rule). This proposed action would allow nuclear power
plant licensees, and applicants for CPs, OLs, COLs, standard design
certifications, standard design approvals, and manufacturing licenses,
to use the Code Cases listed in these RGs as alternatives to
engineering standards for the construction, ISI, and IST of nuclear
power plant components.
This rulemaking also includes consideration of PRM-50-89, submitted
by Mr. Raymond West, requesting that the NRC amend its regulations to
allow consideration of alternatives to the ASME Boiler and Pressure
Vessel [BPV] and OM Code Cases. Lastly, this rulemaking proposes
resequencing the order of NRC's requirements, governing Codes and
standards in order to comply with the OFR guidelines for incorporating
by reference.
I. Accessing Information and Submitting Comments
A. Accessing Information
B. Submitting Comments
II. Background
III. Discussion
A. Code Cases Approved for Unconditional Use
B. Code Case Approved for Use With Conditions
Section III Code Cases (DG-1230/RG 1.84)
Section XI Code Cases (DG-1231/RG 1.147)
OM code Cases (DG-1232/RG 1.192)
C. NRC Proposals for Code Cases on Which the NRC Received Public
Comments in the 2009 Proposed ASME Code Case Rulemaking
Section III Code Cases (DG-1230/RG 1.84)
Section XI Code Cases (DG-1231/RG 1.147)
D. ASME Code Cases Not Approved for Use
IV. Petition for Rulemaking (PRM-50-89)
V. Changes Addressing Office of the Federal Register Guidelines on
Incorporation by Reference
VI. Addition of Headings to Paragraphs
VII. Paragraph-by-Paragraph Discussion
VIII. Plain Writing
IX. Availability of Documents
X. Voluntary Consensus Standards
XI. Finding of No Significant Environmental Impact: Environmental
Assessment
XII. Paperwork Reduction Act Statement
XIII. Regulatory Analysis
XIV. Regulatory Flexibility Certification
XV. Backfitting and Issue Finality
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2009-0359 when contacting the NRC
about the availability of information for this proposed rule. You may
access information related to this proposed rule, which the NRC
possesses and is publicly available, by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2009-0359.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library 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]. The
[[Page 37887]]
ADAMS Accession Number for each document referenced in this proposed
rule (if that document is available in ADAMS) is provided the first
time that a document is referenced. In addition, for the convenience of
the reader, the ADAMS Accession Numbers are provided in a table in
Section IX, ``Availability of Documents,'' of this document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2009-0359 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
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 submissions 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 ISI of nuclear power plant components, and the ASME
Code for Operation and Maintenance of Nuclear Power Plants (OM Code),
which contains requirements for IST of nuclear power plant components.
In response to BPV and OM Code user requests, the ASME develops ASME
Code Cases that provide alternatives to BPV and OM Code requirements
under special circumstances.
The NRC approves and/or mandates the use of the ASME BPV and OM
Code in Sec. 50.55a of Title 10 of the Code of Federal Regulations (10
CFR) through the process of incorporation by reference. As such, each
provision of the ASME Codes incorporated by reference into, and
mandated by, 10 CFR 50.55a, ``Codes and standards,'' constitutes a
legally-binding NRC requirement imposed by rule. As noted previously,
ASME Code Cases, for the most part, represent alternative approaches
for complying with provisions of the ASME BPV and OM Codes.
The NRC periodically amends 10 CFR 50.55a to incorporate by
reference NRC RGs listing approved ASME Code Cases that may be used as
alternatives to the BPV Code and the OM Code. See Federal Register
notice (FRN), ``Incorporation by Reference of ASME BPV and OM Code
Cases'' (68 FR 40469; July 8, 2003).
This rulemaking is the latest in a series of rulemakings that
incorporate by reference new versions of several RGs identifying new
and revised \1\ unconditionally or conditionally acceptable ASME Code
Cases that are approved for use. In developing these RGs, the NRC staff
reviews ASME BPV and OM Code Cases, determines the acceptability of
each Code Case, and publishes its findings in RGs. The RGs are revised
periodically as new Code Cases are published by the ASME. The NRC
incorporates by reference the RGs listing acceptable and conditionally
acceptable ASME Code Cases into 10 CFR 50.55a. Currently, NRC RG 1.84,
Revision 35, ``Design, Fabrication, and Materials Code Case
Acceptability, ASME Section III''; RG 1.147, Revision 16, ``Inservice
Inspection Code Case Acceptability, ASME Section XI, Division 1''; and
RG 1.192, Revision 0, ``Operation and Maintenance Code Case
Acceptability, ASME OM Code,'' are incorporated into the NRC's
regulations at 10 CFR 50.55a. A request for comment on the draft RGs is
published elsewhere in today's Federal Register (Docket ID NRC-2009-
0359).
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\1\ ASME Code Cases can be categorized as one of two types: new
or revised. A new Code Case provides for a new alternative to
specific ASME Code provisions or addresses a new need. A revised
Code Case is a revision (modification) to an existing Code Case to
address, for example, technological advancements in examination
techniques or to address NRC conditions imposed in one of the
regulatory guides that have been incorporated by reference into 10
CFR 50.55a.
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This rulemaking also addresses PRM-50-89 that was submitted to the
NRC on December 14, 2007, and revised on December 19, 2007, by Mr.
Raymond West (ADAMS Accession No. ML073600974). The petition requests
that the NRC amend 10 CFR 50.55a to allow NRC authorization of
alternatives to NRC-approved ASME BPV and OM Code Cases. This
rulemaking includes proposed provisions that address the PRM. A
detailed discussion of the PRM is provided in Section IV, ``Petition
for Rulemaking (PRM-50-89),'' of this document.
III. Discussion
This proposed rule would incorporate by reference the latest
revisions of the NRC regulatory guides that list ASME BPV and OM Code
Cases the NRC finds to be acceptable or ``conditionally acceptable''
(i.e., NRC-specified conditions). Draft Regulatory Guide (DG)-1230,
Regulatory Guide 1.84, Revision 36, (ADAMS Accession No. ML102590003)
would supersede the incorporation by reference of Revision 35; DG-1231,
RG 1.147, Revision 17, (ADAMS Accession No. ML102590004) would
supersede the incorporation by reference of Revision 16; and DG-1232,
RG 1.192, Revision 1, (ADAMS Accession No. ML102600001) would supersede
the incorporation by reference of Revision 0.
This proposed rule addresses two categories of ASME Code Cases. The
first category of Code Cases are the new and revised Section III and
Section XI Code Cases listed in Supplements 1 through 10 to the 2007
Edition of the BPV Code, and the OM Code Cases published with the 2002
Addenda through the 2006 Addenda. The second category is the Code Cases
that were not addressed in the final rule published on October 5, 2010
(75 FR 61321). The 2010 final rule addressed the new and revised
Section III and Section XI Code Cases listed in Supplements 2 through
11 to the 2004 Edition and Supplement 0 to the 2007 Edition of BPV
Code. Public comments were received during the proposed rule stage
(June 2, 2009; 74 FR 26303) requesting that the NRC include certain
revised Code Cases in the final guides that were not listed in the
draft guides. The NRC determined that the revised Code Cases
represented changes significant enough to warrant broader public
participation prior to the NRC making a final determination of them.
Accordingly, the NRC is requesting comment on these Code Cases in this
proposed rule.
The latest editions and addenda of the ASME BPV and OM Codes that
the NRC has approved for use are referenced in 10 CFR 50.55a. The ASME
also publishes Code Cases that provide alternatives to existing Code
requirements developed and approved by the ASME. The proposed rule
would incorporate by reference RGs 1.84, 1.147, and 1.192. The NRC, by
incorporating by reference these three RGs, would allow nuclear power
plant licensees and applicants for standard
[[Page 37888]]
design certifications, standard design approvals, manufacturing
licenses, applicants for Ols, CPs, and COLs under the regulations that
govern license certifications, to use the Code Cases listed in these
RGs as suitable alternatives to the ASME BPV and OM Codes for the
construction, ISI, and IST of nuclear power plant components. This
action would be consistent with the provisions of the National
Technology Transfer and Advancement Act of 1995, Public Law 104-113,
which encourages Federal regulatory agencies to consider adopting
industry consensus standards as an alternative to de novo agency
development of standards affecting an industry. This action would also
be consistent with the NRC policy of evaluating the latest versions of
consensus standards in terms of their suitability for endorsement by
regulations or regulatory guides.
The NRC follows a three-step process to determine the acceptability
of new and revised Code Cases and the need for regulatory positions on
the uses of these Code Cases. This process was employed in the review
of the Code Cases in Supplements 1 through 10 to the 2007 Edition of
the BPV Code and the 2002 Addenda through the 2006 Addenda of the OM
Code. The Code Cases in these supplements are the subject of this
proposed rule. First, the ASME develops Code Cases through a consensus
development process, as administered by the American National Standards
Institute (ANSI), which ensures that the various technical interests
(e.g., utility, manufacturing, insurance, regulatory) are represented
on standards development committees and that their viewpoints are
addressed fairly. This process includes development of a technical
justification in support of each new or revised Code Case. The ASME
committee meetings are open to the public, and attendees are encouraged
to participate. Task groups, working groups, and subgroups report to a
standards committee. The standards committee is the decisive consensus
committee and ensures that the development process fully complies with
the ANSI consensus process. The NRC actively participates through full
involvement in discussions and technical debates of the task groups,
working groups, subgroups, and standards committee regarding the
development of new and revised standards.
Second, the standards committee transmits to its members a first
consideration letter ballot requesting comment or approval of new and
revised Code Cases. To be approved, Code Cases from the first
consideration letter ballot must receive the following: (1) Approval
votes from at least two thirds of the eligible consensus committee
membership, (2) no disapprovals from the standards committee, and (3)
no substantive comments from ASME oversight committees such as the
Technical Oversight Management Committee (TOMC). The TOMC's duties, in
part, are to oversee various standards committees to ensure technical
adequacy and provide recommendations in the development of codes and
standards, as required. The Code Cases that are disapproved or receive
substantive comments from the first consideration ballot are reviewed
by the working level group(s) responsible for their development to
consider the comments received. These Code Cases may be approved by the
standards committee on second consideration with an approval vote by at
least two thirds of the eligible consensus committee membership, with
no more than three disapprovals from the consensus committee.
Third, the NRC reviews new and revised Code Cases to determine
their acceptability for incorporation by reference in 10 CFR 50.55a
through the subject RGs. This rulemaking process, when considered
together with the ANSI process for developing and approving ASME codes
and standards and ASME Code Cases, constitutes the NRC's basis that the
Code Cases (with conditions as necessary) provide reasonable assurance
of adequate protection to public health and safety.
The NRC reviewed the new and revised Code Cases identified in this
proposed rule and concluded, in accordance with the process previously
described, that the Code Cases are technically adequate (with
conditions as necessary) and consistent with current NRC regulations.
Thus, the new and revised Code Cases listed in the subject RGs are
approved for use subject to any specified conditions.
A. Code Cases Approved for Unconditional Use
The NRC determined, in accordance with the process previously
described for review of ASME Code Cases, that each ASME Code Case
listed in Table I is appropriate for incorporation by reference without
conditions into the NRC's regulations.
Table I
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Code Case No. Supplement Title
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Boiler and Pressure Vessel Code Section III (Addressed in DG-1230/RG
1.84, Table 1)
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N-4-13.......... 5 (07 Edition)..... Special Type 403 Modified
Forgings or Bars Class 1 and CS,
Section III, Division 1.
N-570-2......... 7 (07 Edition)..... Alternative Rules for Linear
Piping and Linear Standard
Supports for Classes 1, 2, 3,
and MC, Section III, Division 1.
N-580-2......... 4 (07 Edition)..... Use of Alloy 600 With Columbium
Added, Section III, Division 1.
N-655-1......... 2 (07 Edition)..... Use of SA-738, Grade B, for Metal
Containment Vessels, Class MC,
Section III, Division 1.
N-708........... 2 (07 Edition)..... Use of JIS G-4303, Grades SUS304,
SUS304L, SUS316, and SUS316L,
Section III, Division 1.
N-759-2......... 4 (07 Edition)..... Alternative Rules for Determining
Allowable External Pressure and
Comprehensive Stress for
Cylinders, Cones, Spheres, and
Formed Heads, Section III,
Division 1.
N-760-2......... 7 (07 Edition)..... Welding of Globe Valve Disks to
Valve Stem Retainers, Classes 1,
2, and 3, Section III, Division
1.
N-767........... 4 (07 Edition)..... Use of 21 Cr-6Ni-9Mn (Alloy UNS
S21904) Grade GXM-11 (Conforming
to SA[dash]182/SA-182M and SA-
336/SA-336M), Grade TPXM-11
(Conforming to SA[dash]312/SA-
312M) and Type XM-11 (Conforming
to SA-666) Material, for Class 1
Construction, Section III,
Division 1.
N-774........... 7 (07 Edition)..... Use of 13Cr-4Ni (Alloy UNS
S41500) Grade F6NM Forgings
Weighing in Excess of 10,000 lb
(4,540 kg) and Otherwise
conforming to the Requirements
of SA-336/SA-336M for Class 1,
2, and 3 Construction, Section
III, Division 1.
N-782........... 9 (07 Edition)..... Use of Editions, Addenda, and
Cases, Section III, Division 1.
[[Page 37889]]
N-801........... 4 (10 Edition)..... Rules for Repair of N-Stamped
Class 1, 2, and 3 Components by
Organization Other Than the N
Certificate Holder That
Originally Stamped the Component
Being Repaired, Section III,
Division 1.
N-802........... 4 (10 Edition)..... Rules for Repair of Stamped
Components by the N Certificate
Holder That Originally Stamped
the Component, Section III,
Division 1.
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Boiler and Pressure Vessel Code Section XI (Addressed in DG-1231/RG
1.147, Table 1)
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N-532-5......... 5 (10 Edition ).... Alternative Requirements to
Repair and Replacement
Documentation Requirements and
Inservice Summary Report
Preparation and Submission as
Required by IWA-4000 and IWA-
6000, Section XI, Division 1.
N-716-1......... 1 (13 Edition)..... Alternative Piping Classification
and Examination Requirements,
Section XI Division 1.
N-747........... 9 (04 Edition)..... Reactor Vessel Head-to-Flange
Weld Examinations Section XI,
Division 1.
N-762........... 1 (07 Edition)..... Temper Bead Procedure
Qualification Requirements for
Repair/Replacement Activities
Without Postweld Heat Treatment,
Section XI, Division 1.
N-765........... 8 (07 Edition)..... Alternative to Inspection
Interval Scheduling Requirements
of IWA-2430, Section XI,
Division 1.
N-769........... 8 (07 Edition)..... Roll Expansion of Class 1 In-Core
Housing Bottom Head Penetrations
in BWR's, Section XI, Division
1.
N-773........... 8 (07 Edition)..... Alternatives Qualification
Criteria for Eddy Current
Examinations of Piping Inside
Surfaces, Section XI Division 1.
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Code for Operation and Maintenance (Addressed in DG-1232/RG 1.192, Table
1).
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OMN-6........... 2006 Addenda....... Alternate Rules for Digital
Instruments.
OMN-8........... 2006 Addenda....... Alternative Rules for Preservice
and Inservice Testing of Power-
Operated Valves That Are Used
for System Control and Have a
Safety Function per OM-10.
OMN-14.......... 2004 Addenda....... Alternative Rules for Valve
Testing Operations and
Maintenance, Appendix I, Boiling
Water Reactor (BWR) Control Rod
Drive Rupture Disk Exclusion.
OMN-16.......... 2006 Addenda....... Use of a Pump Curve for Testing.
------------------------------------------------------------------------
B. Code Cases Approved for Use With Conditions
The NRC has determined that certain Code Cases, as issued by the
ASME, are generally acceptable for use, but that the alternative
requirements specified in those Code Cases must be supplemented to
provide an acceptable level of quality and safety. Accordingly, the NRC
proposes to impose conditions on the use of these Code Cases to modify,
limit or clarify their requirements. For each applicable Code Case, the
conditions would specify the additional activities that must be
performed, the limits on the activities specified in the Code Case,
and/or the supplemental information needed to provide clarity. These
ASME Code Cases are included in Table 2 of the following: DG-1230 (RG
1.84), DG-1231 (RG 1.147), and DG-1232 (RG 1.192). The NRC's evaluation
of the Code Cases and the reasons for the NRC's proposed conditions are
discussed in the following paragraphs. Notations have been made to
indicate the conditions duplicated from previous versions of the RGs.
The NRC requests public comment on these Code Cases and the
proposed conditions. It should also be noted that the following
paragraphs only address those Code Cases for which the NRC proposes to
impose a condition or conditions that are listed in the RG for the
first time (e.g., the conditions on OMN-4, 2004 are identical to those
listed in Revision 0 to RG 1.192 on OMN-4, 1999 Addenda).
Section III Code Cases (DG-1230/RG 1.84)
NRC-proposed changes to Tables 1 and 2 of DG-1230/RG 1.84 for Code
Cases N-520-2, N-655-1, N-757-1, N-759-2, and N-782, are discussed in
this notice under the heading, NRC Proposals for Code Cases on which
NRC Received Public Comments in the 2009 Proposed ASME Code Case
Rulemaking.
Code Case N-60-5
Type: Revised
Title: Material for Core Support Structures, Section III, Division
1
Published: Supplement 12, 2001 Edition
The NRC proposes to reinstate a condition on the use of ASME Code
Case N-60-5, which in a previous publication was inadvertently deleted.
Code Case N-60-5 was originally listed in RG 1.85, ``Materials Code
Case Acceptability, ASME Section III, Division 1.'' Two conditions were
listed in RG 1.85 for Code Case N-60-5: 1) welding of age-hardenable
Alloy SA-453 Grade 660 and SA-637 Grade 688 should be performed when
the material is in the solution-treated condition, and 2) the maximum
yield strength of strain-hardened austenitic stainless steel should not
exceed 90,000 psi in view of the susceptibility of this material to
environmentally assisted cracking. Revision 31 of RG 1.85 was last
published in May 1999. In June 18, 2004 (69 FR 34202), RG 1.85 was
merged into RG 1.84. The combined RG 1.84 now lists all Section III
Code Cases, and RG 1.85 is no longer published. When RG 1.85 was merged
into RG.1.84, the NRC inadvertently dropped the two conditions
applicable to Code Case N-60-5. The NRC is now proposing to reinstate
the second of the two conditions by reinstating Code Case N-60-5 in DG-
1230/RG 1.84, Table 2, ``Conditionally Acceptable Section III Code
Cases.''
The NRC has determined that the first condition, regarding age-
hardenable Alloy SA-453 Grade 660 and SA-637 Grade 688, is no longer
needed. These alloy materials are used for bolting and pins that are
not typically subjected to welding.
The second condition was instituted because operating experience
and laboratory testing showed that strain hardened (also known as cold-
worked), austenitic stainless steel in excess of 90,000 psi yield
strength, is susceptible to environmentally induced cracking. The
caution regarding the limit on the maximum yield strength of strain-
[[Page 37890]]
hardened austenitic stainless steels has been addressed in the Standard
Review Plan (SRP) for over 30 years and has been used as guidance by
the NRC staff in its review of reactor coolant pressure boundary
materials in all new reactors since the condition was inadvertently
dropped in RG 1.84. Specifically, the limit is addressed in NUREG-0800,
``Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants SRP Section 4.5.1, Control Rod Drive Structural
Materials, and Section 5.2.3, Reactor Coolant Pressure Boundary
Materials. In Section II, SRP Acceptance Criteria state the need for
such a limitation: ``Laboratory stress corrosion tests and service
experience provide the basis for the criterion that cold-worked
austenitic stainless steels used in the reactor coolant pressure
boundary should have an upper limit on the yield strength of 620 MPa
(90,000 psi).''
Thus, the technical basis for the condition is well-established and
continues to be valid because these materials are used in current
reactor designs and may be used in future reactor designs. Accordingly,
the NRC proposes to reinstate this condition on Code Case N-60-5 in
Table 2 of DG-1230/RG 1.84. A licensee that implemented Code Case N-60-
5 after RG 1.84 and RG 1.85 were combined (i.e., Code Case N-60-5
unconditionally approved) would not have to comply with the reinstated
condition limiting the maximum yield strength. Two of the five new
reactor designs, Economic Simplified Boiling Water Reactor [ESBWR] and
US-EPR, specified the use of Code Case N-60-5 during the time period
that no conditions were listed in RG 1.84. These new reactor design
certifications were reviewed by the NRC staff for conformance with this
condition using the guidelines of the SRP. The condition is included in
the Design Control Document for each of these two designs. Operating
reactor licensees, who specified Code Case N-60-5 during the time that
it was unconditionally approved, are required to meet the ISI
examinations in ASME Code Section XI, to ensure detection of
environmentally assisted cracking that might result from using strain
hardened austenitic stainless steels with yield strength in excess of
90,000 psi.
Reinstatement of this condition would not impact combined license
applications that are currently under review by the NRC or have been
approved. The condition would only apply to those applicants or
licensees in the future that implement Code Case N-60-5 in accordance
with Revision 36 (or later) of the final RG 1.84.
Code Case N-208-2
Type: Revised
Title: Fatigue Analysis for Precipitation Hardening Nickel Alloy
Bolting Material to Specification SB-637 N07718 for Class 1
Construction, Section III, Division 1
Published: Supplement 4, January 4, 2008
Figure A, ``Design Fatigue Curve for Nickel-Chromium Alloy 718,''
Code Case N-208-2, presents maximum mean stress curves. The upper-most
curve is labeled ``No mean stress or [sigma]max < 100 ksi.''
The words ``No mean stress'' may be confusing to users and should be
implemented with the condition that this means ``Maximum mean stress.''
In addition, the lower-most curve is labeled as
``[sigma]y,'' which may also be confusing to users. The
[sigma]y should be implemented with the condition that it
means [sigma]max. Therefore, the NRC proposes to add two
conditions to Code Case N-208-2 in Table 2 of DG-1230/RG 1.84 that
would provide definitions for ``no mean stress'' and
``[sigma]max'' with respect to Figure A.
Section XI Code Cases (DG-1231/RG 1.147)
NRC-proposed changes to Tables 1 and 2 of DG-1231/RG 1.147 for Code
Cases N-508-4, N-597-2, N-619, N-648-1, and N-702, are discussed in
this notice under the heading, NRC Proposals for Code Cases on which
NRC Received Public Comments in the 2009 Proposed ASME Code Case
Rulemaking.
Code Case N-561-2 [Supplement 1]
Type: Revised
Title: Alternative Requirements for Wall Thickness Restoration of
Class 2 and High Energy Class 3 Carbon Steel Piping, Section XI,
Division 1
Published: Supplement 1, 2007 Edition
The original version and first version of this Code Case were not
approved by the NRC for use. The NRC's basis for not approving the use
of this Code Case was that: 1) no criteria for determining the rate or
extent of degradation of the repair of the wall thickness restoration
or the surrounding base metal were provided, and 2) re-inspection
requirements were not provided to verify structural integrity since the
root cause may not be mitigated. The ASME made significant technical
revisions to previous versions of this Code Case by applying the
findings from a very similar application (i.e., Code Case N-661,
``Alternative Requirements for Wall Thickness Restoration of Class 2
and 3 Carbon Steel Piping for Raw Water Service'').
A request to apply Code Case N-661 at the Edwin I. Hatch Nuclear
Power Plant (Hatch Plant) was conditionally approved by the NRC in the
Hatch Safety Evaluation Report (SER) (ADAMS Accession No. ML033280037).
Code Case N-661 was subsequently approved with the same conditions in
RG 1.147, Revision 15. The ASME used these same conditions in revising
Code Case N-561-1 resulting in Code Case N-561-2. Based on the NRC
staff's review of Code Cases N-561-2 and N-661, and on its experience
applying Code Case N-661 at the Hatch Plant, the NRC proposes to
approve Code Case N-561-2 with certain conditions. This is reflected in
Table 2 of DG-1231. Five proposed conditions on this Code Case will be
listed in Table 2 of DG-1231/RG 1.147. The proposed conditions are
discussed in this section.
The provisions of Code Cases N-561-2 and N-661-1 are similar in
that the Code Cases apply to similar systems (i.e., Class 2 and High
Energy Class 3 Carbon Steel Piping, Class 3 Moderate Energy Carbon
Steel Piping, and Carbon Steel Piping for Raw Water Service). The
provisions were developed by the ASME to perform an alternative repair
of degraded components, which involves the application of weld metal
overlay on the exterior of the piping system to restore the wall
thickness of the component. Accordingly, the conditions identified in
the SER regarding Code Case N-661 also apply to Code Case N-561-2.
One of the conditions in the SER addressed the time period for
which the repair would be considered acceptable. The definition
established by the NRC was modified when added to Code Case N-561-2. In
Code Case N-661, the repair is only acceptable until the ``next
refueling outage.'' In contrast, Code Case N-561-2 states that the
repair would be acceptable for ``one fuel cycle.'' The NRC believes
that it is unclear in Code Case N-561-2 what one fuel cycle actually
infers if a repair is performed at mid-cycle.
It could be interpreted that the repair is acceptable for the
remainder of the current fuel cycle plus the subsequent fuel cycle.
This interpretation could double the time period. The NRC established
this limitation on the acceptable life of the repair of the five
because the Code Case does not require that the root cause of the
degradation be determined. If the root cause of the degradation has not
been determined, a suitable reinspection frequency cannot be
established. In addition, the Code Case would allow repairs to be made
by welding on surfaces that are wet or exposed to water. Performing
through-
[[Page 37891]]
wall weld repairs on surfaces that are wet or exposed to water would
greatly increase the chances of producing welds that include weld
defects such as porosity, lack of fusion, and cracks. It is highly
unlikely that a weld can be made on an open root joint with water
present on the backside of the weld without having several weld
defects. These types of weld defects can, and many times do, lead to
premature failure of a weld joint.
Accordingly, the NRC is proposing on Code Case N-561-2 two of the
five conditions (identified as Conditions 1 and 3) in the DG-1231 to
address these concerns. The first proposed condition addresses those
situations where welds are fabricated with water present on the
backside, defects are likely, and the service life time would be
expected to be greatly reduced: ``Paragraph 5(b): [of Code Case N-561-
2] for repairs performed on a wet surface, the overlay is only
acceptable until the next refueling outage.'' A second proposed
condition is being added on Code Case N-561-2 that would not allow the
exemption in Paragraph (6)(c)(1). Paragraph (6)(c)(1) states that
``Class 3 weld overlays are exempt from volumetric examination when the
Construction Code does not require that full-penetration butt welds in
the same location be volumetrically examined.'' Many licensees are
mitigating stress corrosion cracking through the addition of a weld
overlay on the outside of the piping. The purpose of the overlay is to
restore wall thickness. The NRC has approved this mitigation technique
provided that the full thickness of the weld overlay as well as a
certain portion of the base material can be volumetrically examined.
The exemption in Paragraph (6)(c)(1) conflicts with the NRC position on
this matter, and thus the third condition is proposed requiring the
performance of a volumetric examination of the weld overlay.
The third proposed condition on Code Case N-561-2 is: ``Paragraph
7(c): if the cause of the degradation has not been determined, the
repair is only acceptable until the next refueling outage.''
The fourth condition on Code Case N-561-2 is proposed to address
the NRC's concern that a preexisting flaw could grow through-wall after
application of a weld overlay: ``The area where the weld overlay is to
be applied must be examined using ultrasonic methods to demonstrate
that no crack-like defects exist.'' The basis for this proposed
condition is discussed in detail here. Weld overlays have been used as
a mitigation method and as a repair method to address stress corrosion
cracking in piping butt welds. The basis for applying a weld overlay is
that it will result in compressive residual stresses on the inside
surface of the pipe, thus preventing a flaw from growing. Analytical
modeling has been used to predict post-weld repair residual stress
distributions for common piping configurations. Many times, however,
weld records are not available or are not complete with regard to weld
repairs made during construction. The investigations using modeling to
predict the residual stresses resulting from weld repairs have used
various assumptions to address the lack of data from weld records.
This raises a question whether a model can accurately predict
residual stresses if the extent of repairs is unknown. Factors such as
the number of weld passes, welding sequence, and heat input can greatly
influence stress patterns. Thus, analytical modeling of typical piping
weld configurations with a weld overlay has been used to determine
whether application of a weld overlay would result in compressive
residual stresses and impede the growth of a preexisting flaw. Because
of the many assumptions that might be required, configurations have
been analyzed with up to a 75 percent through thickness flaw.
While the results of the analyses performed have shown that a weld
overlay could produce compressive stresses on the inside diameter of
the piping for repairs as great as 75 percent through-wall, the NRC
continues to be concerned regarding the lack of repair information. For
example, an investigation into a leak that occurred several years ago
showed that at least four weld repairs had been performed. This case is
not believed to be unique. Thus, the NRC does not believe that the
analyses that have been conducted to date are bounding, nor that the
analyses have demonstrated that a preexisting flaw would not continue
to grow circumferentially and perhaps through-wall after application of
a weld overlay. Accordingly, the NRC proposes that it must be shown,
using ultrasonic methods that no flaws exist in the area where the weld
overlay is to be applied.
The fifth and last condition being proposed on Code Case N-561-2 is
``Paragraph 4(b): All systems must be depressurized before welding.''
The need for this condition is the same as that for the first proposed
condition, i.e., the Code Case would allow repairs to be made by
welding on surfaces that is wet or exposed to water. As previously
discussed, it is highly unlikely that a weld can be made on an open
root joint with water present on the backside of the weld without
having several weld defects, and these types of weld defects can lead
to premature failure of a weld joint. Thus, depressurizing the system
would decrease the chances of producing a suspect weld.
Code Case N-562-2
Type: Revised
Title: Alternative Requirements for Wall Thickness Restoration of
Class 3 Moderate Energy Carbon Steel Piping, Section XI, Division 1
Published: Supplement 1, 2007 Edition
Code Case N-562-2 is nearly identical to Code Case N-561-2, which
is discussed separately herein. The principal difference between the
Code Cases is that N-562-2 addresses lower energy piping. However, the
same concerns previously discussed regarding Code Case N-561-2 also
apply to Code Case N-562-2. Accordingly, the same five conditions are
being proposed for Code Case N-562-2.
Code Case N-661-2
Type: Revised
Title: Alternative Requirements for Wall Thickness Restoration of
Classes 2 and 3 Carbon Steel Piping for Raw Water Service, Section XI,
Division 1
Published: Supplement 1, 2007 Edition
As previously discussed with respect to Code Case N-561-2, Code
Case N-661-2 is very similar to the other two Code Cases addressing
restoration of wall thickness (namely N-561-1 and N-562-2), except that
N-661-2 addresses raw water service systems.
Conditions (1) and (3) in draft Revision 17 to RG 1.147 for Code
Case N-661-2 were listed in Revision 16 to RG 1.147. Those conditions
are: (1) Paragraph 4(b): for repairs performed on a wet surface, the
overlay is only acceptable until the next refueling outage; and (3)
paragraph 7(c): if the cause of the degradation has not been
determined, the repair is only acceptable until the next refueling
outage. As previously indicated in the discussion addressing Code Case
N-561-2, the ASME made significant technical revisions to Code Cases N-
561-1, N-562-1, and N-661-1. Consistent with the technical
justification addressing the proposed conditions for Code Case N-561-2,
the NRC is proposing three new conditions for Code Case N-661-2. Those
conditions are listed in draft Revision 17 to RG 1.147 as following:
(2) Paragraph 6(c)(1): this exemption is not permitted; (4) The area
where the weld overlay is to be applied must be examined using
ultrasonic methods to
[[Page 37892]]
demonstrate that no crack-like defects exist; and (5) All systems must
be depressurized before welding.
Code Case N-739-1 [Supplement 1]
Type: Revised
Title: Alternative Qualification Requirements for Personnel
Performing Class CC Concrete and Post-Tensioning System Visual
Examinations, Section XI, Division 1
Published: Supplement 1, 2007 Edition
The original version of this Code Case was not approved by the NRC
for use. The NRC had concerns regarding the lack of detail provided on
the instructional material to be covered in the qualification of
personnel performing these inspections. The revised Code Case includes
detailed instructional material regarding requirements for training.
The NRC finds the added requirements to be acceptable. However, the
reference in the Code Case to the American Concrete Institute (ACI)
standard has been printed incorrectly. To ensure that the correct
instructional material is used, the NRC is proposing to conditionally
approve Code Case N-739-1 to indicate that the correct ACI reference is
201.1.
OM Code Cases (DG-1232/RG 1.192)
Code Case OMN-1
Type: Revised
Title: Alternative Rules for Preservice and Inservice Testing of
Active Electric Motor-Operated Valve Assemblies in Light-Water Reactor
Power Plants
Published: 2006 Addenda
Proposed Revision 1 to RG 1.192 does not modify the conditions
imposed on the implementation of Code Case OMN-1 that were listed in
Revision 0 to RG 1.192, issued June 2003. The following discussion is
included in the proposed rule to emphasize that caution is required
when using risk insights to evaluate the performance of MOVs that have
exercise intervals extended from quarterly to every refueling outage.
In 1996, ASME issued Code Case OMN-1 that allows quarterly stroke-
time testing of motor-operated valves (MOVs) in the IST program to be
replaced by a program of exercising on a refueling outage frequency and
periodic diagnostic testing at intervals up to 10 years. In 1999, the
NRC accepted the use of Code Case OMN-1 with conditions in 10 CFR
50.55a(b)(3)(iii) as an alternative to the requirement in 10 CFR
50.55a(b)(3)(ii) that licensees shall comply with the provisions for
MOV stroke-time testing in the OM Code and shall establish a program to
ensure that MOVs continue to be capable of performing their design-
basis safety functions.
In June 2003, the NRC staff developed RG 1.192 and transferred the
acceptance of Revision 0 to Code Case OMN-1 from 10 CFR
50.55a(b)(3)(iii) to RG 1.192 with the following conditions. Those
conditions are:
(1) The adequacy of the diagnostic test interval for each MOV must
be evaluated and adjusted as necessary, but not later than 5 years or
three refueling outages (whichever is longer) from initial
implementation of OMN-1.
(2) When extending exercise test intervals for high risk MOVs
beyond a quarterly frequency, licensees must ensure that the potential
increase in Core Damage Frequency (CDF) and risk associated with the
extension is small and consistent with the intent of the Commission's
Safety Goal Policy Statement.
(3) When applying risk insights as part of the implementation of
OMN-1, licensees must categorize MOVs according to their safety
significance using the methodology described in Code Case OMN-3,
``Requirements for Safety Significance Categorization of Components
Using Risk Insights for Inservice Testing of LWR Power Plants,'' with
the conditions discussed in RG 1.192 or use other MOV risk-ranking
methodologies accepted by the NRC on a plant-specific or industry-wide
basis with the conditions in the applicable safety evaluations.
Licensees may use Code Case OMN-1 in lieu of the provisions for
stroke-time testing in Subsection ISTC of the 1995 Edition up to and
including the 2000 Addenda of the ASME OM Code when applied in
conjunction with the provisions for leakage rate testing in, as
applicable, ISTC 4.3 (1995 Edition with the 1996 and 1997 Addenda) and
ISTC-3600 (1998 Edition through the 1999 and 2000 Addenda). In
addition, licensees who continue to implement Section XI of the ASME
BPV Code as their Code of Record may use OMN-1 in lieu of the
provisions for stroke-time testing specified in Paragraph 4.2.1 of
ASME/ANSI OM Part 10 as required by 10 CFR 50.55a(b)(2)(vii) subject to
the conditions in this regulatory guide. Licensees who choose to apply
OMN-1 must apply all its provisions.
It should be noted that ASME issued Code Case OMN-11, ``Risk-
Informed Testing for Motor-Operated Valves,'' in the 2001 Edition to
provide more specific provisions for the application of risk insights
as part of the MOV diagnostic testing alternative allowed in Code Case
OMN-1. The NRC accepted the use of OMN-11 in Revision 0 of RG 1.192
with conditions related to determination of acceptable MOV test
intervals based on diagnostic data, evaluation of test results for
grouped low-risk MOVs, and extension of the exercise interval for high-
risk MOVs similar to the condition in RG 1.192 for Code Case OMN-1.
In the 2006 Addenda to the ASME OM Code, ASME issued an updated
version of Code Case OMN-1 to clarify the guidance for users of the
code case. In its updated version, Code Case OMN-1 incorporates the
provisions of Code Case OMN-11 for applying risk insights as well as
the conditions specified in the June 2003 version of RG 1.192 for the
use of Code Case OMN-11.
The NRC staff is not proposing to modify the conditions for the
acceptability of Code Case OMN-1 based on the incorporation of
provisions for applying risk insights from OMN-11. However, based on
operating experience at nuclear power plants, the NRC emphasizes the
importance of evaluating the performance of MOVs that have exercise
intervals extended from quarterly to every refueling outage. As
discussed in Federal Register Notice 51370 (dated September 22, 1999)
on page 51386, and which the NRC finds is still applicable when using
the 2006 version of Code Case OMN-1, the licensee should 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. This information may be
obtained by grouping similar MOVs and staggering the exercising of the
MOVs in the group equally over the refueling interval. Licensees are
cautioned that, when implementing OMN-1, the benefits of performing a
particular test should be balanced against the potential adverse
effects placed on the valves or systems caused by this testing.
Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants,'' to 10 CFR part 50 requires nuclear
power plant licensees to evaluate deficiencies in the performance of
safety-related MOVs. Where degradation in the performance of a high-
risk MOV is identified when exercised or tested at an extended
interval, licensees should reapply the quarterly frequency for the
exercise test interval for all high-risk MOVs and implement diagnostic
testing of those MOVs at an interval that provides assurance of their
design-basis capability throughout the test interval. Licensees should
also incorporate the performance results for all MOVs into the
probabilistic risk analysis to determine whether the risk ranking of
[[Page 37893]]
MOVs should be modified based on those results.
For additional information on OMN-1, see the discussion on OMN-4
and OMN-12 below.
Code Case OMN-3
Type: Revised
Title: Requirements for Safety Significance Categorization of
Components Using Risk Insights for Inservice Testing of LWR Power
Plants
Published: 2004 Edition
The NRC initially issued RG 1.192 in June 2003 accepting several
ASME OM Code Cases, including Code Case OMN-3. Subsequently, on
December 18, 2003, the Commission issued Staff Requirements Memorandum
(SRM) COMNJD-03-0002, ``Stabilizing the PRA Quality Expectations and
Requirements'' (ADAMS Accession No. ML033520457), which approved
implementation of a phased approach to achieving an appropriate quality
for probabilistic risk assessments (PRAs) for the NRC's risk-informed
decisionmaking. In SECY-04-0118 dated July 13, 2004 (ADAMS Accession
No. ML041470505), the NRC staff described its action plan to implement
the SRM, which the Commission subsequently approved in an SRM dated
October 6, 2004 (ADAMS Accession No. ML042800369).
The central concept of the action plan specifies the development of
consensus PRA standards and associated industry guidance documents, as
discussed in RG 1.200 (March 2009), ``An Approach for Determining the
Technical Adequacy of Probabilistic Risk Assessment Results for Risk-
Informed Activities.'' RG 1.200 clarifies that the staff anticipates
that current good practice, (i.e., Capability Category II (CCII)) as
explained in the appendices of RG 1.200, is the level of technical
adequacy that is sufficient for the majority of applications. RG 1.200
provides that licensees evaluate all deviations from CCII or higher and
document why the PRA is sufficient for the proposed application.
In a related action, the Commission published Section 69, ``Risk-
Informed Categorization and Treatment of Structures, Systems and
Components for Nuclear Power Reactors'' in 10 CFR part 50 on November
22, 2004. RG 1.201 (May 2006), ``Guidelines for Categorizing
Structures, Systems, and Components in Nuclear Power Plants According
to their Safety Significance,'' describes one acceptable method to
categorize the safety significance of active components. Section 50.69
specifies high level treatment requirements for low risk SSCs whereas
SSC treatment is prescribed in more detail in several risk-informed
ASME OM Code Cases.
Based on a consideration of the information in Section 69 in 10 CFR
part 50 and in the RG 1.201, the NRC proposes Conditions (5), (6) and
(7) in RG 1.192 to its acceptance of Code Case OMN-3 included in the
2004 Edition of the ASME OM Code. Licensees applying Code Case OMN-3,
2004 Edition, will need to apply the conditions specified in the
previous version of RG 1.192 issued in June 2003, and new Conditions
(5), (6) and (7) discussed in this section. As stated in RG 1.192, if a
licensee implements a Code Case and a later version of the Code Case is
incorporated by reference into 10 CFR 50.55a and listed in Tables 1 and
2 during the licensee's present 120-month IST program interval, that
licensee may use either the later version or the previous version. An
exception to this provision would be the inclusion of a limitation or
condition on the use of Code Case that is necessary, for example, to
enhance safety. The NRC staff has determined that a licensee currently
using Code Case OMN-3 must use the later version of the Code Case
listed in Table 2 of RG 1.192, Revision 1, after it is incorporated by
reference into 10 CFR 50.55a.
Condition (5) specifies that the implementation of Section 3.2,
``Plant Specific PRA,'' in Code Case OMN-3 must 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.
Condition (5) references RG 1.200 and 1.201 for guidance in satisfying
this condition. For example, RG 1.200 includes descriptions of
technical adequacy of PRA analyses beyond those modeling only internal
initiating events, (e.g., for seismic and internal fire initiating
events). RG 1.201 endorses the guidance described by the Nuclear Energy
Institute (NEI) in Revision 0 to NEI 00-04, ``10 CFR 50.69 SSC
Categorization Guideline,'' dated July 2005. This document describes
how the importance of components relied on for seismic, fires, and
other initiating events (and operating modes) should be included in the
categorization process, including if no plant-specific PRA is available
for the hazard.
Condition (6) specifies that paragraph (b) in Section 4.2.4,
``Reconciliation,'' in Code Case OMN-3 is not endorsed. Condition (6)
states that the expert panel may not classify components that are
ranked as a High Safety Significant Component (HSSC) by the results of
a qualitative or quantitative PRA evaluation (excluding the sensitivity
studies) or the defense-in-depth assessment to a Low Safety Significant
Component (LSSC). RG 1.201 clarifies that a component, identified as
high safety significant by any of the PRA (excluding the sensitivity
studies) or defense-in-depth evaluations may not be re-categorized to
low safety significant by the expert panel. The position in RG 1.201
that an expert panel may not decide the PRA or defense-in-depth
evaluations are in error and lower the safety significance assigned
according to these evaluations is applicable to OMN-3 deliberations.
Rather, the expert panel should provide information regarding its views
to the PRA analysts so that the evaluations can be re-performed, if
appropriate, to address the expert panel issue or document the
appropriateness of the current analysis results.
Condition (7) specifies that implementation of Section 3.3,
``Living PRA,'' in Code Case OMN-3 must be consistent with the
following: (1) to account for potential changes in the failure rates
and other changes that could affect the PRA, changes to the plant must
be reviewed, and, as appropriate, the PRA updated; (2) when the PRA is
updated, the SSC categorization must be reviewed and changed if
necessary to remain consistent with the categorization process; and (3)
the review of plant changes must be performed in a timely manner and
must be performed once every two refueling outages or as required by
50.71(h)(2) for COL holders. Changes to the plant, including potential
changes in failure rates, might affect the PRA evaluations, and changes
to the PRA evaluations might affect the safety significance of the
components developed from these evaluations. Therefore, the PRA must be
periodically updated and the risk categorization reviewed when the PRA
is updated. The period of two refueling outages as the maximum period
between determinations of whether a PRA update is needed is consistent
with the time span in 10 CFR 50.69.
Code Case OMN-3 addresses safety significance categorization of
components using risk insights as applied to inservice testing. Several
new conditions are proposed with respect to Code Case OMN-3 (discussed
earlier) that reflect current NRC regulatory positions on determining
PRA technical adequacy when using risk insights in regulatory
applications. Code Cases OMN-1, OMN-4, 2004 Edition, and OMN-12, 2004
Edition, also address the use of risk insights for inservice testing.
Accordingly, to ensure consistent
[[Page 37894]]
implementation among these Code Cases, a note has been added to Code
Case OMN-4 and OMN-12. Paragraph 3.1 of Code Case OMN-12 states that
``Valve assemblies shall be classified as either high safety
significant or low safety significant in accordance with Code Case OMN-
3.'' However, given the interdependence of Code Cases OMN-1, OMN-3,
OMN-4, and OMN-12, Note 3 has been added to Code Case OMN-12 as a
reminder of the dependence on Code Case OMN-3 (i.e., paragraph 3.1). In
addition, Note 2 has been added to Code Case OMN-4 as a reminder that
the conditions with respect to allowable methodologies for OMN-3 risk
ranking specified for the use of OMN-1 also apply to OMN-4.
C. NRC Proposals for Code Cases on Which NRC Received Public Comments
in the 2009 Proposed ASME Code Case Rulemaking
On June 2, 2009, the NRC published a proposed rule (74 FR 26303)
and a parallel notice of availability of draft RGs (74 FR 26440)
seeking public comments on incorporating by reference draft RG 1.84,
Revision 35, and draft RG 1.147, Revision 16. The NRC received public
comments on draft Revision 35 to RG 1.84 and draft Revision 16 to RG
1.147 requesting that certain revised Code Cases that were not listed
in those draft guides be approved in the final guides. These revised
Code Cases that were the subject of comment in 2009 are N-520-2, N-655-
1, N-757-1, N-759-2, and N-782 for RG 1.84; and Code Cases N-508-4, N-
597-2, N-619, N-648-1, and N-702 for RG 1.147. In that earlier
rulemaking, the NRC determined that the revised Code Cases represented
changes significant enough to warrant broader public participation
prior to the NRC making a final determination of them. Therefore, the
final RG 1.84 and RG 1.147 associated with the 2010 final rule (75 FR
61321; October 5, 2010) did not include these Code Cases.
The NRC has reviewed these Code Cases, and now proposes to approve
those Code Cases, in some cases with conditions. These Code Cases are
discussed in this section, under the applicable draft regulatory guide.
Section III Code Cases (DG-1230/RG 1.84)
Code Case N-520-2
Type: Revised
Title: Alternative Rules for Renewal of Active or Expired N-type
Certificates for Plants Not in Active Construction
Published: Supplement 4, 2007 Edition
Code Case N-520-1, the predecessor of Code Case N-520-2, was
unconditionally approved in Revision 34 to RG 1.84. The objective of
Code Case N-520-1 was to address situations where construction was
halted on a nuclear power plant, interrupting ASME Code activities, but
the Certificate Holder had maintained its certificate. Code Case N-520-
1 provided guidance on what a Certificate Holder had to do to document
and stamp the completed construction work. On June 2, 2009, the NRC
published a proposed rule (74 FR 26303) and a parallel notice of
availability of draft RGs (74 FR 26440) seeking public comment on draft
RG 1.84, Revision 35. The NRC received a public comment requesting that
the NRC approve Code Case N-520-2 for inclusion in final Revision 35,
noting that Code Case N-520-2 had been approved by the ASME on November
1, 2007, and published in Supplement 4 to the 2007 Edition. Code Case
N-520-2 was developed to allow an organization with an expired
certificate to secure an ASME Temporary Certificate of Authorization.
Because Code Case N-520-2 was not part of the June 2009 proposed rule
and the changes reflected in N-520-2 were significant, the NRC did not
adopt the public comment to list Code Case N-520-2 in final Revision 35
to RG 1.84 (incorporated by reference in the final rule published on
October 5, 2010 (75 FR 61321)).
The NRC has now determined that the provisions of Code Case N-520-2
are adequate for addressing a situation where a Certificate Holder has
let its N-type certificates expire. The basis for this determination is
that all completed in-process work must be clearly documented to ensure
that remaining activities and Code responsibilities are readily
identifiable. In addition, the ASME Temporary Certificate of
Authorization is for the sole purpose of completing the required
documentation and component stamping. Finally, this work must be
completed under a contract with an Authorized Nuclear Inspection Agency
(ANIA).
The NRC is proposing to conditionally approve Code Case N-520-2
because it believes that the wording of the Code Case may create
confusion regarding the relationship between the ANIA and the
Authorized Nuclear Inspector (ANI). The purpose of the condition in
Table 2 of DG1230/RG 1.84, Revision 36, is to clearly indicate that the
ANIA employs the ANI.
Code Cases N-655-1, N-757-1, N-759-2, N-782
A comment responding to the June 2, 2009, proposed rule (74 FR
26303) and a parallel notice of availability of draft RGs (74 FR
26440), requested that the following four Code Cases used in the AP-
1000 design that were not included in draft Revision 35 of RG 1.84 be
included in the final guide: Code Case N-655-1, ``Use of SA-738, Grade
B, for Metal Containment Vessels, Class MC, Section III, Division 1;''
Code Case N-757-1, ``Alternative Rules for Acceptability for Class 2
and 3 Valves, NPS 1 (DN25) and Smaller with Welded and Nonwelded End
Connections other than Flanges, Section III, Division 1;'' Code Case N-
759-2, ``Alternative Rules for Determining Allowable External Pressure
and Compressive Stresses for Cylinders, Cones, Spheres, and Formed
Heads, Section III, Division 1;'' and Code Case N-782, ``Use of Code
Editions, Addenda, and Cases Section III, Division 1.'' Draft Revision
35 of RG 1.84 considered Code Cases published up to Supplement 0 to the
2007 Edition. Code Cases N-655-1 and N-757-1 were published in
Supplement 2 to the 2007 Edition. Code Case N-759-2 was published in
Supplement 4 to the 2007 Edition. Code Case N-782 was published in
Supplement 9 to the 2007 Edition. These four Code Cases were beyond the
scope of the draft RG and thus had not been considered for inclusion in
the draft RG.
The NRC did not include these four Code Cases in final Revision 35
of RG 1.84 because it would have been inappropriate to include them in
the final RG without providing the public an opportunity for comment.
In addition, these Code Cases were not referenced in the latest AP-1000
Design Control Document.
Code Cases N-655-1, N-759-2, and N-782 have been reviewed by the
NRC and have been found to be acceptable. Accordingly, these Code Cases
are listed in Table 1 of DG-1230/RG 1.84, Revision 36, and the NRC
proposes to unconditionally approve them, as presented in Table I under
``Code Cases Approved for ``Unconditional Use''.
Code Case N-757-1 was reviewed and found to be conditionally
acceptable. It is listed in Table 2 of the DG-1230/RG 1.84. The
proposed condition for Code Case N-757-1 is discussed in the following
discussion.
Code Case N-757-1 [Supplement 2]
Type: Revised
Title: Alternative Rules for Acceptability for Class 2 and 3 Valves
NPS 1 (DN 25) and Smaller with Welded and Nonwelded End Connections
Other than Flanges, Section III, Division 1
Published: Supplement 2, 2007 Edition
The NRC proposes to impose a condition on Code Case N-757-1 in
Table 2 of RG 1.84 to prohibit the use
[[Page 37895]]
of the design provisions in ASME Section III, Division 1, Appendix
XIII, for Class 3 valves. This would be accomplished by adding the
condition to Table 2 of DG-1230/RG 1.84. The Code Case addresses the
use of instrument, control, and sampling line valves, NPS 1 (DN 25) and
smaller, with nonwelded end connections other than ASME B16.5 flanges
for Section III, Division 1, Class 2 and Class 3 construction. The Code
Case provides three options for the design of Class 2 and Class 3
valves that do not meet the minimum thickness requirements in ASME
B16.34. These options include the following: 1) the pressure design
rules of Section III, paragraphs NC-3324 and ND-3324; 2) the
experimental stress analysis rules in Section III, Appendix II; or 3)
design based on the stress analysis rules in Section III, Appendix
XIII.
The NRC finds that the first option provides an acceptable
alternative basis for the design of ASME Class 2 and Class 3 valves
because it provides adequate design margin by using the vessel design
rules accepted by the NRC in 10 CFR 50.55a. The second option is also
acceptable for the design of ASME Class 2 and Class 3 valves because it
allows the designer to use experimental stress analysis techniques to
establish that the design provides acceptable ASME Code margins for
parts in which theoretical stress analysis might not be possible or
practical. The third option, however, is not acceptable to the NRC.
Option 3 would allow a designer to use the criteria provided in
Section III, Division 1, Appendix XIII. As defined by the scope of
Appendix XIII, these Code rules are only applicable to the design of
Class 2 vessels meeting the requirements of NC-3200. Further, Appendix
XIII provides for design based on a stress analysis that uses criteria
similar to that used for the design of ASME Class 1 components
(including the ASME Class 1 stress intensity allowable limits). The
stress intensity values in the acceptance criteria are greater than the
allowable stress intensity values specified for the design of ASME
Class 3 components. The NRC concludes that the criteria in Appendix
XIII are not intended for the design of ASME Class 3 components,
including the valves within the scope of N-757, and that a condition
should be added to Table 2 of DG-1230/RG 1.84 that prohibits the use of
these design provisions for Class 3 valves.
It should be noted that the NRC staff approved this Code Case as it
was considered by the cognizant ASME committees. However, upon further
consideration as Code Cases were reviewed for inclusion in the subject
RGs, the NRC determined that use of the Code Case was inappropriate for
ASME Class 3 components. Therefore, the NRC proposes to impose a
condition that would prohibit the use of the design provisions in ASME
Section III, Division 1, Appendix XIII, for Class 3 valves.
Section XI Code Cases (DG-1231/RG 1.147)
Code Case N-508-4
Type: New
Title: Rotation of Snubbers and Pressure Retaining Items for the
Purpose of Testing or Preventive Maintenance, Section XI, Division 1
Published: Supplement 8, 2007 Edition
Code Case N-508-3, the predecessor of Code Case N-508-4, was
unconditionally approved in Revision 15 to RG 1.147. The objective of
Code Case N-508-3 was to provide guidance on rotating snubbers and
relief valves from stock for the purpose of testing or preventive
maintenance. On June 2, 2009, the NRC published a proposed rule (74 FR
26303) and a parallel notice of availability of draft RGs (74 FR 26440)
seeking public comment on draft RG 1.147, Revision 16. The NRC received
a public comment noting that Code Case N-508-4 had been approved by the
ASME on January 26, 2009, and published in Supplement 8 to the 2007
Edition, and requesting that the NRC approve Code Case N-508-4 in final
Revision 16 rather than cease approval at Code Case N-508-3. Code Case
N-508-4 significantly expands the list of components that may be
rotated from stock for the purpose of testing or preventive maintenance
(adds pumps, control rod drive mechanisms, and pump seal packages).
Because Code Case N-508-4 was not part of the June 2009 proposed
rule and the changes reflected in N-508-4 were significant, the NRC did
not adopt the public comment to list Code Case N-508-4 in final
Revision 16 to RG 1.147 (incorporated by reference in the final rule
published on October 5, 2010 (75 FR 61321)). Instead, this Code Case is
addressed in draft Revision 17 to RG 1.147.
The NRC has not identified any technical reasons why additional
components may be considered for the purpose of testing or preventive
maintenance as described in the Code Case N-508-4. However, the NRC has
identified an issue and proposes to condition Code Case N-508-4 to
ensure that there is no conflict regarding the application of this Code
Case. When Section XI is used to govern snubber examination and
testing, Footnote 1 (which was later added to the Code Case) conflicts
with Subsection IWF, Section XI, up to and including the 2004 Edition
through the 2005 Addenda. Footnote 1 directs the user to implement the
OM Code for snubber examination and testing. The OM Code was developed
in order to have a separate Code for the development and maintenance of
provisions for the IST of pumps and valves. In 1990, the ASME published
the initial edition of the OM Code, thereby transferring responsibility
for these provisions from Section XI to the OM Code Committee. While
the use of the OM Code is an option under paragraph (b)(3)(v)(A), the
examination and testing requirements for snubbers are also provided in
the 2005 Addenda and earlier editions and addenda of Section XI. Thus,
there is a conflict for editions and addenda up to the 2005 Addenda of
Section XI, but there is no conflict for licensees who have adopted the
2006 Addenda or later editions and addenda of Section XI.
To resolve the conflict, the NRC is proposing to include in DG-
1231/RG 1.147, Revision 17, a condition to Code Case N-508-4 stating
that Footnote 1 to the Code Case would not apply when the ISI Code of
record is earlier than Section XI, 2006 Addenda, and Section XI
requirements are used to govern the examination and testing of
snubbers.
Code Case N-597-2
Type: Revised
Title: Requirements for Analytical Evaluation of Pipe Wall Thinning
Listed: Revision 15 to RG 1.147
Published: November 18, 2003
Code Case N-597-2 was conditionally approved in Revision 15 to RG
1.147. Two comments responding to the proposed rule published on June
2, 2009 (74 FR 26303), and a parallel notice of availability of draft
RGs (74 FR 26440) seeking public comment on draft RG 1.147, Revision
16, suggested that the method in Code Case N-513-2, ``Evaluation
Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2
or 3 Piping,'' used to evaluate local degradation, should be approved
by the NRC for application to Code Case, N-597-2. The comments argued
that the NRC has conditionally approved Code Case N-513-2 with an
evaluation methodology to allow licensees to temporarily accept flaws
in moderate energy Class 2 or 3 piping, whereas condition (2) on Code
Case N-597-2 requires NRC approval for any amount of local degradation
beyond that calculated by the hoop stress equation.
[[Page 37896]]
Because Code Case N-513-2 was not part of the June 2, 2009,
proposed rule and the changes reflected in N-513-2 are significant, the
NRC did not adopt the public comments to allow the Code Case N-513-2
evaluation to also be used with respect to Code Case N-597-2. While the
NRC agrees that the flaw evaluation methodology for analyzing piping
degradation contained in Code Case N-513-2 could under certain
circumstances be applied for a Code Case N-597-2 evaluation (i.e., both
Code Cases address the analytical evaluation of pipe wall thinning),
the NRC disagrees with the comments that through-wall leakage should be
included in the scope of such an evaluation. Code Case N-597-2 was not
developed to address leakage; it is focused only on analytical
evaluation of wall thinning. The comments discuss local degradation up
to and including through-wall leakage and believe it would be
appropriate to allow such leakage for all ASME Code class components.
This implies that such leakage from high temperature, high pressure
systems is no different from leakage from low temperature, low pressure
systems. Permitting degradation up to and including through-wall
leakage in certain systems would violate 10 CFR part 50, appendix A,
Criterion General Design Criteria (GDC) 14, ``Reactor coolant pressure
boundary,'' and/or similar provisions in the licensing basis for these
facilities, which require that the reactor coolant pressure boundary be
tested to ensure an extremely low probability of abnormal leakage, of
propagating failure, and of gross rupture. In addition, there have been
pipe breaks and leakage in high temperature, high pressure lines
throughout the world and some have been sudden and catastrophic. Code
Case N-597-2 is applicable to all ASME Code class piping, including
high energy piping; whereas, Code Case N-513-2 is limited to Class 2
and 3 moderate energy piping. The NRC has only approved temporary
acceptance of flaws for moderate energy Class 2 or 3 piping (maximum
operating temperature does not exceed 200[emsp14][deg]F (93 [deg]C) and
maximum operating pressure does not exceed 275 psig (1.9 MPa)). The
comments' requested change would redefine the defense-in-depth concept.
Rather than performing inspections to detect flaws before structural
integrity is compromised, degradation would be managed in effect after
leakage is discovered.
The NRC agrees, however, that it should be permissible under
certain circumstances for licensees to evaluate local thinning using
the acceptance criteria of the Code Case without NRC review and
acceptance. Thus, a sixth condition is being proposed for Code Case N-
597-2 in DG-1231/RG 1.147, Revision 17. The condition would propose
that, on moderate-energy Class 2 and 3 piping, wall thinning acceptance
criteria may be used on a temporary basis based on the provisions of
Code Case N-513-2, and that Code Case N-597-2 cannot be used to
evaluate through-wall leakage conditions.
Code Cases N-619
Type: Conditionally approved
Title: Alternative Requirements for Nozzle Inner Radius Inspections
for Class 1 Pressurizer and Steam Generator Nozzles Published
Published: April 8, 2002
Code Case N-648-1
Type: Conditionally approved for the first time
Title: Alternative Requirements for Inner Radius Examination of
Class 1 Reactor Pressure Vessel Nozzles
Published: September 18, 2001
A comment on the proposed rule published on June 2, 2009 (74 FR
26303), and a parallel notice of availability of draft RGs (74 FR
26440) seeking public comment on draft RG 1.147, Revision 16, requested
that the NRC reconsider the conditions placed on Code Case N-619,
``Alternative Requirements for Nozzle Inner Radius Inspections for
Class 1 Pressurizer and Steam Generator Nozzles,'' and Code Case N-648-
1, ``Alternative Requirements for Inner Radius Examination of Class 1
Reactor Pressure Vessel Nozzles.'' The comment states that the
conditions on the two Code Cases requiring a wire standard to
demonstrate the resolution capability of remote visual examination
systems should be changed to the ASME 0.044-inch characters because
those characters have been recognized to be a better resolution
standard than the wire standard.
Because Code Case N-619 and Code Case N-648-1 were not part of the
June 2, 2009, proposed rule and the changes reflected in N-619 and N-
648-1 are significant; the NRC did not adopt the public comment to use
characters rather than the wire standard.
The NRC is addressing the comment as part of this rulemaking. The
NRC agrees with the 2009 comment that characters have been demonstrated
to be a better resolution standard than the wire standard. However, the
NRC believes that the shift to characters should be part of broader
changes to the visual testing standards. Visual examinations are used
in certain situations as alternatives to volumetric and/or surface
examination tests where it is not possible to conduct volumetric
examination (e.g., where there are limitations due to access or
geometry) or to reduce occupational exposure in high radiation fields.
Visual testing experts had believed that if the camera and lighting
were sufficient to see a 12 [mu]m (0.0005 in.) diameter wire, then the
camera system had a resolution sufficiently high for the inspection.
Subsequent investigation of the effectiveness and reliability of visual
examinations has shown that the wire resolution standard is not
sufficient to determine the visual acuity of a remote system (i.e.,
there are important differences between visually detecting a wire and a
crack). Research conducted at the Pacific Northwest National Laboratory
showed that other calibration standards should be adapted for visual
testing such as reading charts and resolution targets. Results
supporting this recommendation were published in NUREG/CR-6943, ``A
Study of Remote Visual Methods to Detect Cracking in Reactor
Components'' (ADAMS Accession No. ML073110060). As also discussed in
the report, other parameters such as crack size, lighting conditions,
camera resolution, and surface conditions were assessed. The NRC
concluded from the investigation that a significant fraction of the
cracks that have been reported in nuclear power plant components are at
the lower end of the capabilities of the visual testing equipment
currently being used. Code Case N-619 addresses the examination of the
nozzle inner radius of Class 1 pressurizers and steam generators. Code
Case N-648-1 provides an alternative for examining the inner radius of
Class 1 reactor vessel nozzles. The NRC investigation of crack opening
dimensions of service-induced cracks in nuclear components included
thermal fatigue, mechanical fatigue, and stress corrosion cracks. The
NRC concluded that current visual testing systems may not reliably
detect a significant number of these cracks, and the research results
showed that detection of these cracks under field conditions is
strongly dependent on camera magnification, lighting, inspector
training, and inspector vigilance. While this research supports the use
of characters in lieu of a wire standard, the research also showed that
other changes should be considered to visual testing as related to
these two Code Cases. The NRC and the Electric Power Research Institute
(EPRI) are currently conducting a collaborative
[[Page 37897]]
research project investigating these parameters. The results of the
collaborative research will be assessed by the NRC and the industry to
determine what changes should be made to visual testing requirements in
the future.
The comment also indicated that it is unclear how allowable flaw
lengths would be determined from Table IWB-3512-1. The condition on the
two Code Cases states that ``licensees may perform a visual examination
with enhanced magnification that has a resolution sensitivity to detect
a 1-mil width wire or crack, utilizing the allowable flaw length
criteria of Table IWB-3512-1 with limiting assumptions on the flaw
aspect ratio.'' Table IWB-3512-1 does not specifically provide
allowable flaw length criteria. The commenter recommended that the
acceptance criteria be modified as following: ``Crack-like surface
flaws exceeding the acceptance criteria of Table IWB-3510-3 are
unacceptable for continued service unless the vessel meets the
requirements of IWB-2142.2, IWB-3142.3, or IWB-3142.4. The component
thickness, t, to be applied in calculating the allowable surface flaw,
I, in Table IWB-3510-3 shall be selected as specified in Table IWB-
3512-2.''
The NRC does not agree with the suggestion. Table IWB-3512-1 was
selected because it is the only table that considers the inside corner
region. In determining an acceptable flaw size, the limiting aspect
ratio is assumed, which is 0.5. The surface flaw allowable size divided
by the limiting aspect ratio yields the limiting surface flaw size in
terms of the l/t. In the case of wall thickness sizes provided in Table
IWB-3512-1, the acceptance criteria are the same as those in Table IWB-
3510-3. The NRC does not intend to make any changes to the table
referred to for acceptance criteria, because Table IWB-3512-1 is the
only table to refer to the inside corner region.
Finally, the commenter believes that the condition on Code Case N-
648-1 describing the surfaces to be examined is unnecessary because the
Code Case describes the same examination surfaces. The NRC agrees and
proposes to eliminate this condition in Table 2 of DG-1231/RG 1.147,
Revision 17.
Code Case N-702
Type: New
Title: Alternative Requirements for Boiling Water Reactor (BWR)
Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division 1
Published: Supplement 12, 2001 Edition
Two comments on the proposed rule published on June 2, 2009 (74 FR
26303), and a parallel notice of availability of draft RGs (74 FR
26440) seeking public comment on draft RG 1.147, Revision 16, requested
that Code Case N-702, ``Alternative Requirements for Boiling Water
Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section
XI, Division 1,'' be conditionally approved in the final guide. Code
Case N-702 had been listed in draft RG 1.193, Revision 3, ``ASME Code
Cases Not Approved for Use,'' because at the time that draft Revision
16 to RG 1.147 was published (October 2007), the NRC staff was
considering the industry response to the NRC staff's request for
additional information relative to the acceptability of ``BWRVIP-108:
BWR Vessel and Internals Project (VIP), Technical Basis for the
Reduction of Inspection Requirements for the Boiling Water Reactor
Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,'' EPRI Technical
Report 1003557, October 2002 (ADAMS Accession No. ML023330203). BWRVIP-
108 provides the technical basis supporting Code Case N-702.
Subsequently, the NRC conditionally approved a licensee's request to
use the Code Case on the basis of the NRC's Safety Evaluation (ADAMS
Accession No. ML073600374; December 18, 2007).
The Safety Evaluation discussed the NRC's review of BWRVIP-108 and
the conditions under which it could be used. The commenters believed
that the conditions in the Safety Evaluation provided a basis for the
NRC to conditionally approve Code Case N-702 in final RG 1.147,
Revision 16. The NRC did not adopt the public comment to approve the
Code Case in final Revision 16 to RG 1.147. Code Case N-702 is an
alternative to provisions in the ASME Code to reduce the inspection
requirements of BWR reactor vessel nozzle-to-shell welds and nozzle
blend radii. BWRVIP-108 discusses the probabilistic fracture mechanics
evaluation that was performed to demonstrate that the probability of
failure considering these inspection changes meets NRC requirements.
While the NRC believes that the Safety Evaluation and BWRVIP report
provide a basis for conditionally approving the Code Case on a generic
basis, the NRC did not believe that it would have been appropriate to
move the Code Case from RG 1.193 to RG 1.147 without first having
sought public comment. Thus, the NRC is proposing to conditionally
approve Code Case N-702 in DG-1231/RG 1.147, Revision 17, based on the
conditions that were discussed in the Safety Evaluation. The
applicability of Code Case N-702 must be shown by demonstrating that
the criteria in Section 5.0 of the NRC Safety Evaluation regarding
BWRVIP-108 dated December 18, 2007, are met. The evaluation
demonstrating the applicability of the Code Case must be reviewed and
approved by the NRC prior to the application of the Code Case.
Code Case N-747
Type: New
Title: Reactor Vessel Head-to Flange Weld Examinations, Section XI,
Division 1
Published: Supplement 9, 2004 Edition
A comment on the proposed rule published on June 2, 2009 (74 FR
26303), and a parallel notice of availability of draft RGs (74 FR
26440) seeking public comment on draft RG 1.147, Revision 16, suggested
that the basis for listing Code Case N-747, ``Reactor Vessel Head-to
Flange Weld Examinations, Section XI, Division 1,'' in draft RG 1.193
was flawed, and that the Code Case should be unconditionally accepted
in final Revision 16. Additional technical information to support
approval of the Code Case was provided in the comment letter (ADAMS
Accession No. ML092190138). The NRC did not adopt the public comment to
list Code Case N-747 in final Revision 16 to RG 1.147 (incorporated by
reference in the final rule published on October 5, 2010, (75 FR
61321)), because the NRC determined that the public should have an
opportunity to comment on the additional information that was submitted
by the commenter.
The NRC has reviewed the information provided in the comment, which
deals with the expected fluence levels of reactor vessel head-to-flange
welds. Based on this information, the NRC believes that an adequate
technical basis has been provided to support a conclusion that the
fracture toughness will remain high. The key points discussed in the
additional information are that calculations show that the fluence in
the upper head region will be low, even after 60 years of service.
Therefore, there will be no irradiation induced change in
RTNDT. In addition, the industry has calculated
RTNDT for the upper head region for early Westinghouse plant
designs using the Standard Review Plan (NUREG-0800) and determined that
the fracture toughness is high. Therefore, the NRC proposes to
unconditionally approve
[[Page 37898]]
Code Case N-747 in Table 1 of DG-1231/RG 1.147, Revision 17.
D. ASME Code Cases Not Approved for Use
The ASME Code Cases that are currently issued by the ASME but not
approved for generic use by the NRC are listed in RG 1.193, ``ASME Code
Cases Not Approved for Use.'' In addition to ASME Code Cases that the
NRC has found to be technically or programmatically unacceptable, RG
1.193 includes Code Cases on reactor designs for high-temperature gas-
cooled reactors and liquid metal reactors, reactor designs not
currently licensed by the NRC, and certain requirements in Section III,
Division 2, for submerged spent fuel waste casks, that are not endorsed
by the NRC. Regulatory Guide 1.193 complements RGs 1.84, 1.147, and
1.192. It should be noted that RG 1.193 is not part of this rulemaking
because the NRC is not proposing to adopt any of the Code Cases listed
in that RG. Comments have been submitted in the past, however, on
certain Code Cases listed in RG 1.193 where the commenter believed that
additional technical information was available that might not have been
considered by the NRC in its determination not to approve the use of
these Code Cases. While the NRC will consider those comments, any
changes in the NRC's non-approval of such Code Cases will be the
subject of an additional opportunity for public comment.
IV. Petition for Rulemaking (PRM-50-89)
On December 14, 2007, Mr. Raymond West (the petitioner) submitted a
PRM requesting the NRC to amend 10 CFR 50.55a to allow consideration of
alternatives to the ASME BPV and OM Code Cases. The petitioner
submitted an amended petition on December 19, 2007 (ADAMS Accession No.
ML073600974). The petition was docketed by the NRC as PRM-50-89. The
petitioner requested that the regulations be amended to provide
applicants and licensees a process for requesting NRC approval of
changes or modifications to ASME Code Cases that are listed in the
relevant NRC-approved RGs cited in the current regulations. The
petitioner stated that the current requirements do not allow changes or
modifications to be proposed as alternatives to NRC-approved ASME Code
Cases, and asserted that such changes or modifications should be
allowed as alternatives to NRC Code Cases. Overall, the petitioner
requested that the regulations be amended to allow applicants and
licensees to request authorization of NRC-approved Code Cases with
proposed modifications directly through Sec. 50.55a(a)(3).
The NRC believes that Code Cases often provide alternatives that
have technical merit and, in many instances, are incorporated into
future ASME Code editions. The ASME Code Case process itself
constitutes a method of how an applicant or licensee can seek to obtain
ASME approval for a variation of a previously-approved Code provision.
Section 50.55a(a)(3) currently provides specific approaches for
obtaining NRC authorization of alternatives to ASME Code provisions.
Inasmuch as ASME Code Cases are analogous to ASME Code provisions, it
is not unreasonable to provide an analogous regulatory approach for
obtaining NRC authorization of alternatives to ASME Code Cases. For
these reasons, the NRC determined that the issues raised in this PRM
should be considered in the NRC's rulemaking process, and the NRC
published a FRN with this determination on April 22, 2009 (74 FR
18303). Accordingly, the NRC is addressing PRM-50-89 in this proposed
rule.
On the basis of the previous discussion, the NRC is proposing to
include language in proposed 10 CFR 50.55a(z) (existing 50.55a(a)(3))
that would allow applicants and licensees to request authorization of
alternatives for changes to conditions on NRC-approved ASME Code Cases
in current paragraphs (b)(4), (b)(5), and (b)(6) of Sec. 50.55a. In
addition, the NRC proposes extending the scope of the petitioner's
request for allowing alternatives to NRC-approved Code Case conditions
to allow applicants and licensees to request authorization of
alternatives for changes to conditions on Section III and XI of the
ASME BPV Code and OM Code in current paragraphs (b)(1), (b)(2), and
(b)(3).
V. Changes Addressing Office of the Federal Register Guidelines on
Incorporation by Reference
This proposed rule includes changes to 10 CFR 50.54, 50.55, and
50.55a. These changes were made in accordance with the guidance for
incorporation by reference of multiple standards that is included in
Chapter 6 of the OFR's ``Federal Register Document Drafting Handbook,''
January 2011 Revision. This latest revision of the OFR's guidance
provides several options for incorporating by reference multiple
standards into regulations.
The NRC proposes to incorporate by reference, in a single
paragraph, the multiple standards mentioned in 10 CFR 50.55a. For the
least disruption to the existing structure of the section, the NRC
proposes to incorporate by reference the multiple standards into 10 CFR
50.55a(a), the first paragraph of the section. Each national consensus
standard that is being incorporated by reference in 10 CFR 50.55a has
been listed separately. Accordingly, the regulatory language of 10 CFR
50.54, 50.55, and 50.55a has been reorganized by moving existing
paragraphs, creating new paragraphs, and revising introductory and
regulatory texts.
The NRC has made conforming changes to references throughout 10 CFR
50.55a to reflect this reorganization. A detailed discussion of the
affected paragraphs, other than the aforementioned reference changes,
is provided in Section VII, ``Paragraph-by-Paragraph Discussion,'' of
this document. The regulatory text of 10 CFR 50.55a has been set out in
its entirety for the convenience of the reader. The staff has also
developed reader aids to help users understand these changes (see
Section VI of this document.)
VI. Addition of Headings to Paragraph
The NRC is proposing to add headings (explanatory titles) to
paragraphs and all lower-level subparagraphs of 10 CFR 50.55a. These
headings are intended to enhance the readers' ability to identify the
paragraphs (e.g., paragraphs (a), (b), (c)) and subparagraphs with the
same subject matter. The NRC's proposal addresses longstanding
complaints by external and internal stakeholders on the readability and
complex structure of the requirements in 10 CFR 50.55a. To address this
concern, the NRC evaluated a range of solutions, including the creation
of new regulations and relocation of existing requirements from 10 CFR
50.55a to the new regulations.
Some alternatives the NRC considered were a new regulation adjacent
to 10 CFR 50.55a (e.g., Sec. Sec. 50.55b, 50.55c, 50.55d), a new
subpart containing a new series of regulations at the end of 10 CFR
part 50 (e.g., subpart B beginning at Sec. 50.200, and continuing with
Sec. Sec. 50.201, 50.202, 50.203), or a new part (designated for codes
and standards) containing a new series of regulations addressing codes
and standards approved for incorporation by reference by the OFR. The
relocation of each existing requirement to a new regulation (or set of
regulations) would follow a set of organizing principles established by
the NRC after consideration of stakeholder's views.
Upon consideration of these alternatives, the NRC decided that
these alternatives should not be adopted--at least not at this time
without further stakeholder input--and instead that the
[[Page 37899]]
NRC should develop and adopt headings for paragraphs and subparagraphs.
The primary reason for the NRC's decision is external stakeholders'
objections to a previous attempt by the NRC to re-designate paragraphs
in Sec. 50.55a (75 FR 24324; May 4, 2010). As the NRC understands it,
many nuclear power plant licensees' procedures reference specific
paragraphs and subparagraphs of Sec. 50.55a. It would require
substantial rewriting of these procedures and documents to correct the
references to the old (superseded) section, paragraphs and
subparagraphs. In addition, currently-approved design certification
rules may require conforming amendments to be made to correct
references to ASME Code provisions on design (and possibly ISI and
IST).
The NRC requests public comment on whether the NRC should adopt one
of these approaches, either as a follow-on activity to the addition of
headings, or as a substitute for the addition of headings. The most
helpful comments would identify a specific approach, and set forth the
reasons why the proposed approach should be adopted, taking into
account the factors considered by the NRC in selecting the headings
approach.
NRC's Proposal: Convention for Headings and Subheadings
The NRC is proposing to add headings to all first, second, third,
fourth, and some fifth-level paragraphs for certain sections of 10 CFR
50.55a to add clarity and a user-friendly method for following sublevel
contents within a regulation. The proposed heading for a fourth-level
would follow the same convention, but may designate the provision
number only. Fifth-level paragraphs are proposed for only newly
incorporated Code Cases. Each first-level paragraph (designated using
letters, (e.g., (a), (b), (c))) would have a heading that concisely
describes the general subject matter addressed in that paragraph. Each
second-level paragraph (designated using numbers, e.g., (1), (2), (3))
would have a heading comprised of a summary of the first-level
paragraph's heading and a semicolon (``;''), followed by a concise
description of the subject matter addressed in the second paragraph.
The proposed heading for a third-level paragraph would follow the same
convention (i.e., a heading comprised of a summary level of the higher-
level paragraph's title and a semicolon, followed by a concise
description of the subject matter addressed in that subparagraph). The
proposed heading for a fourth-level paragraph would follow the same
convention, but may designate the provision number only. The proposed
fifth-level paragraph is applied to only paragraph (a) for
incorporation by reference of approved editions and addenda to the ASME
BPV and OM Codes.
Reader Aids
The staff has developed a table showing the proposed structure of
10 CFR 50.55a. This table, ``Proposed Reorganization of Paragraphs and
Subparagraphs in 10 CFR 50.55a, Codes and standards'' (ADAMS Accession
No. ML12289A121) is available in a separate document and outlines the
section showing all paragraph designations, including the new paragraph
headings. The staff has also developed Cross-Reference tables showing
the current designations for 10 CFR 50.54, 50.55, and 50.55a
regulations and the proposed designations for these sections. These
tables contain the new headings and a description of each change and
are available in a separate document (ADAMS Accession No. ML12289A114).
VII. Paragraph-by-Paragraph Discussion
Section 50.54
In Sec. 50.54, the introductory statement would be revised to
include a reference to Sec. 50.55a. This revision would clarify that
nuclear power plant licensees, as described in the introductory
paragraph of Sec. 50.54, also are subject to the applicable
requirements delineated in Sec. 50.55a. In addition, the NRC proposes
to revise the introductory text of this section, add and reserve
paragraph (ii), and add paragraph (jj) to include a condition of every
license. This requirement is currently contained in Sec. 50.55a(a)(1),
and no change to the requirement is intended by the transfer of this
requirement from Sec. 50.55a(a)(1) to Sec. 50.54(jj).
Section 50.55
In Sec. 50.55, the introductory text would be revised to include
references to existing Sec. 50.55a, and paragraphs (g) and (h) would
be added and reserved for future use. Further, existing Sec.
50.55a(a)(1) would be moved to a newly created Sec. 50.55(i).
Section 50.55a
In Sec. 50.55a, the current introductory statement would be
relocated to Sec. 50.54(jj), 50.55(i), and 50.55a.
Paragraph (a)
A new paragraph (a) would be created in Sec. 50.55a to incorporate
by reference the multiple standards currently identified in existing
Sec. 50.55a. The heading would be revised to read ``Documents approved
for incorporation by reference.''
Paragraph (a)(1): This paragraph ``American Society of Mechanical
Engineers (ASME)'' would be added to group all ASME Sections.
Paragraph (a)(1)(i): This paragraph, ``ASME Boiler and Pressure
Vessel Code, Section III,'' would be added to discuss the availability
of standards referenced in current paragraph (b)(1). This change would
bring the NRC's requirements into compliance with the OFR's revised
guidelines for incorporating by reference consensus standards in
regulations.
Paragraph (a)(1)(i)(A): This paragraph, ``Rules for Construction of
Nuclear Vessels,'' would be added to group all the individual standards
referenced regarding the subject matter included in current paragraph
(b)(1). This change would bring the NRC's requirements into compliance
with the OFR's revised guidelines for incorporating by reference
consensus standards in regulations.
Paragraph (a)(1)(i)(B): This paragraph, ``Rules for Construction of
Nuclear Power Plant Components,'' would be added to group all the
individual standards referenced regarding the subject matter included
in current paragraph (b)(1). This change would bring the NRC's
requirements into compliance with the OFR's revised guidelines for
incorporating by reference consensus standards in regulations.
Paragraph (a)(1)(i)(C): This paragraph, ``Division I Rules for
Construction of Nuclear Power Plant Components,'' would be added to
group all the individual standards referenced regarding the subject
matter included in current paragraph (b)(1). This change would bring
the NRC's requirements into compliance with the OFR's revised
guidelines for incorporating by reference consensus standards in
regulations.
Paragraph (a)(1)(i)(D): This paragraph, ``Rules for Construction of
Nuclear Power Plant Components--Division 1,'' would be added to group
all the individual standards referenced regarding the subject matter
included in current paragraph (b)(1). This change would bring the NRC's
requirements into compliance with the OFR's revised guidelines for
incorporating by reference consensus standards in regulations.
Paragraph (a)(1)(i)(E): This paragraph, ``Rules for Construction of
Nuclear Facility Components--Division 1,'' would be added to group all
the individual standards referenced regarding the subject matter
included in
[[Page 37900]]
current paragraph (b)(1). This change would bring the NRC's
requirements into compliance with the OFR's revised guidelines for
incorporating by reference consensus standards in regulations.
Paragraph (a)(1)(ii)(A): This paragraph, ``Rules for Inservice
Inspection of Nuclear Reactor Coolant Systems,'' would be added to
discuss the availability of individual standards referenced regarding
the subject matter included in current paragraph (b)(2). This change
would bring the NRC's requirements into compliance with the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations.
Paragraph (a)(1)(ii)(B): This paragraph, ``Rules for Inservice
Inspection of Nuclear Power Plant Components,'' would be added to
discuss the availability of individual standards referenced regarding
the subject matter included in current paragraph (b)(2). This change
would bring the NRC's requirements into compliance with the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations.
Paragraph (a)(1)(ii)(C): This paragraph, ``Rules for Inservice
Inspection of Nuclear Power Plant Components--Division 1,'' would be
added to discuss the availability of individual standards referenced
regarding the subject matter included in current paragraph (b)(2). This
change would bring the NRC's requirements into compliance with the
OFR's revised guidelines for incorporating by reference consensus
standards in regulations.
Paragraph (a)(1)(iii): This paragraph, ``ASME Code Cases: Nuclear
Components,'' would be added to discuss the newly approved Code Cases
referenced regarding the subject matter in current paragraph (b). This
change would bring the NRC's requirements into compliance with the
OFR's revised guidelines for incorporating by reference consensus
standards in regulations.
Paragraph (a)(1)(iii)(A): This paragraph, ``ASME Code Case N-722-
1,'' would be added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b). This change
would bring the NRC's requirements into compliance with the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations.
Paragraph (a)(1)(iii)(B): This paragraph, ``ASME Code Case N-729-
1,'' would be added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b). This change
would bring the NRC's requirements into compliance with the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations.
Paragraph (a)(1)(iii)(C): This paragraph, ``ASME Code Case N-770-
1,'' would be added to discuss the newly approved Code Case referenced
regarding the subject matter in current paragraph (b). This change
would bring the NRC's requirements into compliance with the OFR's
revised guidelines for incorporating by reference consensus standards
in regulations.
Paragraph (a)(1)(iv): This paragraph, ``ASME Operation and
Maintenance Code,'' would be added to group all the individual
standards referenced in current paragraph (b). This change would bring
the NRC's requirements into compliance with the OFR's revised
guidelines for incorporating by reference consensus standards in
regulations.
Paragraph (a)(1)(iv)(A): This paragraph, ``Code for Operation and
Maintenance of Nuclear Power Plants,'' would be added to group all the
individual standards referenced in current paragraph (b).
Paragraph (a)(1)(iv)(B): This paragraph would be added and reserved
for future use.
Paragraph (a)(2): This paragraph, ``Institute of Electrical and
Electronics Engineers (IEEE) Service Center,'' would be added to list
all IEEE sections.
Paragraph (a)(2)(i): This paragraph, ``IEEE Standard 279-1971,''
would be added to discuss the availability of standards referenced in
current paragraph (h)(2). This would be done in compliance with OFR
revised guidelines for incorporation by reference standards in
regulations.
Paragraph (a)(2)(ii): This paragraph, ``IEEE Standard 603-1991,''
would be added to discuss the availability of the standard referenced
in current paragraph (h)(2) and (h)(3). This would be done in
compliance with OFR revised guidelines for incorporation by reference
standards in regulations.
Paragraph (a)(2)(iii): This paragraph, ``IEEE Standard 603-1991
correction sheet,'' would be added to discuss the availability of the
standard referenced in current paragraphs (h)(2) and (h)(3). This would
be done in compliance with OFR revised guidelines for incorporation by
reference standards in regulations.
Paragraph (a)(3): This paragraph, ``U.S. Nuclear Regulatory
Commission (NRC) Reproduction and Distribution Services Section,''
lists all regulatory guides being incorporated by reference. This would
be done in compliance with OFR revised guidelines for incorporation by
reference standards in regulations.
Paragraph (a)(3)(i): This paragraph, ``NRC Regulatory Guide 1.84,
Revision 36,'' would be added to discuss the availability of the
standard. This would be done in compliance with OFR revised guidelines
for incorporation by reference standards in regulations.
Paragraph (a)(3)(ii): This paragraph, ``NRC Regulatory Guide 1.147,
Revision 17,'' would be added to discuss the availability of the
standard. This would be done in compliance with OFR revised guidelines
for incorporation by reference standards in regulations.
Paragraph (a)(3)(iii): This paragraph, ``NRC Regulatory Guide
1.192, Revision 1,'' would be added to discuss the availability of the
standard. This would be done in compliance with OFR revised guidelines
for incorporation by reference standards in regulations.
Paragraph (b): The paragraph heading would be revised to ``Use and
conditions on the use of standards.'' The contents would be moved, in
part, to 50.55a(a) for compliance with OFR revised guidelines for
incorporation by reference standards in regulations.
Paragraph (c): Introductory text would be added to the existing
paragraph (c). Explanatory headings would be added for subparagraphs.
Paragraph (d): The new paragraph would add introductory text to
``Quality Group B components,'' as part of the NRC initiative of adding
headings and providing clarity. Explanatory headings would be added for
subparagraphs.
Paragraph (e): The new paragraph would add introductory text to
``Quality Group C components,'' as part of the NRC initiative of adding
headings and providing clarity. Explanatory headings would be added for
subparagraphs.
Paragraph (f): Introductory text would be revised and expanded in
``Inservice testing requirements,'' as part of the NRC initiative of
adding headings and providing clarity. Explanatory headings would be
added for subparagraphs.
Paragraph (g): Introductory text would be revised and expanded in
``Inservice inspection requirements,'' as part of the NRC initiative of
adding headings and providing clarity. Explanatory headings would be
added for subparagraphs.
Paragraphs (b)(5), (f)(2), (f)(3)(iii)(A), (f)(3)(iv)(A),
(f)(4)(ii), (g)(2), (g)(3)(i), (g)(3)(ii), (g)(4)(i), and (g)(4)(ii):
References to the revision number for RG 1.147 would be changed from
``Revision 16'' to ``Revision 17.''
Paragraph (h)(1): This paragraph would be designated as reserved
because the informational content from
[[Page 37901]]
current (h)(1) would be moved to proposed paragraph (a)(2).
Paragraphs (i)-(y): The paragraphs would be added and reserved for
future use.
Paragraph (z): This paragraph would be added to contain information
that would be relocated from the introductory text of current paragraph
(a)(3) and current subparagraphs (a)(3)(i)-(ii) as a result of the
NRC's compliance with the OFR's revised guidelines for incorporating by
reference. Paragraph (z) would also be revised to allow applicants and
licensees to request alternatives to the requirements in paragraph (b)
of this section.
Overall Considerations on the Use of ASME Code Cases
This rulemaking would amend 10 CFR 50.55a to incorporate by
reference RG 1.84, Revision 36, which would supersede Revision 35; RG
1.147, Revision 17, which would supersede Revision 16; and RG 1.192,
Revision 1, which would supersede Revision 0. The following general
guidance applies to the use of the ASME Code Cases approved in the
latest versions of the RGs that are incorporated by reference into 10
CFR 50.55a as part of this rulemaking.
The approval of a Code Case in the NRC RGs constitutes acceptance
of its technical position for applications that are not precluded by
regulatory or other requirements or by the recommendations in these or
other RGs. The applicant and/or licensee are responsible for ensuring
that use of the Code Case does not conflict with regulatory
requirements or licensee commitments. The Code Cases listed in the RGs
are acceptable for use within the limits specified in the Code Cases.
If the RG states an NRC condition on the use of a Code Case, then the
NRC condition supplements and does not supersede any condition(s)
specified in the Code Case, unless otherwise stated in the NRC
condition.
The ASME Code Cases may be revised for many reasons, (e.g., to
incorporate operational examination and testing experience and to
update material requirements based on research results). On occasion,
an inaccuracy in an equation is discovered or an examination, as
practiced, is found not to be adequate to detect a newly discovered
degradation mechanism. Hence, when an applicant or a licensee initially
implements a Code Case, 10 CFR 50.55a requires that the applicant or
the licensee implement the most recent version of that Code Case as
listed in the RGs incorporated by reference. Code Cases superseded by
revision are no longer acceptable for new applications unless otherwise
indicated.
Section III of the ASME BPV Code applies only to new construction
(i.e., the edition and addenda to be used in the construction of a
plant are selected based on the date of the construction permit and are
not changed thereafter, except voluntarily by the applicant or the
licensee). Hence, if a Section III Code Case is implemented by an
applicant or a licensee and a later version of the Code Case is
incorporated by reference into 10 CFR 50.55a and listed in the RGs, the
applicant or the licensee may use either version of the Code Case
(subject, however, to whatever change requirements apply to its
licensing basis, (e.g., 10 CFR 50.59)).
A licensee's ISI and IST programs must be updated every 10 years to
the latest edition and addenda of Section XI and the OM Code,
respectively, that were incorporated by reference into 10 CFR 50.55a
and in effect 12 months prior to the start of the next inspection and
testing interval. Licensees who were using a Code Case prior to the
effective date of its revision may continue to use the previous version
for the remainder of the 120-month ISI or IST interval. This relieves
licensees of the burden of having to update their ISI or IST program
each time a Code Case is revised by the ASME and approved for use by
the NRC. Code Cases apply to specific editions and addenda, and Code
Cases may be revised if they are no longer accurate or adequate, so
licensees choosing to continue using a Code Case during the subsequent
ISI or IST interval must implement the latest version incorporated by
reference into 10 CFR 50.55a and listed in the RGs.
The ASME may annul Code Cases that are no longer required, are
determined to be inaccurate or inadequate, or have been incorporated
into the BPV or OM Codes. If an applicant or a licensee applied a Code
Case before it was listed as annulled, the applicant or the licensee
may continue to use the Code Case until the applicant or the licensee
updates its construction Code of Record (in the case of an applicant,
updates its application) or until the licensee's 120-month ISI or IST
update interval expires, after which the continued use of the Code Case
is prohibited unless NRC authorization is given under the current 10
CFR 50.55a(a)(3). If a Code Case is incorporated by reference into 10
CFR 50.55a and later annulled by the ASME because experience has shown
that the design analysis, construction method, examination method, or
testing method is inadequate; the NRC will amend 10 CFR 50.55a and the
relevant RG to remove the approval of the annulled Code Case.
Applicants and licensees should not begin to implement such annulled
Code Cases in advance of the rulemaking.
A Code Case may be revised, for example, to incorporate user
experience. The older or superseded version of the Code Case cannot be
applied by the licensee or applicant for the first time.
If an applicant or a licensee applied a Code Case before it was
listed as superseded, the applicant or the licensee may continue to use
the Code Case until the applicant or the licensee updates its
construction Code of Record (in the case of an applicant, updates its
application) or until the licensee's 120-month ISI or IST update
interval expires, after which the continued use of the Code Case is
prohibited unless NRC authorization is given under proposed 10 CFR
50.55a(z). If a Code Case is incorporated by reference into 10 CFR
50.55a and later a revised version is issued by the ASME because
experience has shown that the design analysis, construction method,
examination method, or testing method is inadequate; the NRC will amend
10 CFR 50.55a and the relevant RG to remove the approval of the
superseded Code Case. Applicants and licensees should not begin to
implement such superseded Code Cases in advance of the rulemaking.
VIII. Plain Writing
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 the proposed rule with respect to the clarity and
effectiveness of the language used.
IX. Availability of Documents
The NRC is making the documents identified in Table II available to
interested persons through one or more of the following methods, as
indicated. To access documents related to this action, see the
ADDRESSES section of this document.
[[Page 37902]]
Table II
----------------------------------------------------------------------------------------------------------------
Document PDR WEB NRC Library.
----------------------------------------------------------------------------------------------------------------
Proposed Rule--Regulatory Analysis............. X X ML103060189.
Proposed Rule Federal Register Notice.......... X X ML103060003.
Proposed Reorganization of Paragraphs and X X ML12289A121.
Subparagraphs.
Cross-Reference Tables......................... X X ML12289A114.
RG 1.84, Revision 36 (DG-1230)................. X X ML102590003.
RG 1.147, Revision 17 (DG-1231)................ X X ML102590004.
RG 1.192, Revision 1 (DG-1232)................. X X ML102600001.
RG 1.200, Revision 2, An Approach for X X ML090410014.
Determining the Technical Adequacy of
Probabilistic Risk Assessment Results for Risk-
informed Activities.
RG 1.201, Revision 1, Guidelines for X X ML061090627.
Categorizing Structures, Systems, and
Components in Nuclear Power Plants According
to Their Safety Significance.
2007/12/19--Petition for Rulemaking PRM-50-89 X X ML073600974.
submitted by Ray West regarding, ``To Amend
CFR 5-.55a--Codes and Standards--Revision 1''.
Hatch Plant Report--Hatch, Units 1 & 2, Farley, X X ML033280037.
Units 1 & 2, Vogtle, Units 1 & 2, Safety
Evaluation Re. Request to Use ASME Code Case N-
661.
EPRI Technical Report--Project No. 704--BWRVIP- X X ML023330203.
108: BWR Vessel & Internals Project, Technical
Basis for Reduction of Inspection Requirements
for Boiling Water Reactor Nozzle-to-Vessel
Shell Welds & Nozzle Blend Radii.
Safety Evaluation of Proprietary EPRI Report-- X X ML073600374.
BWR Vessel and Internals Project, Technical
Basis for the Reduction of Inspection
Requirements for the Boiling Water Reactor
Nozzle-to-Vessel Shell Welds and Nozzle Inner
Radius (BWRVIP-108).
Comment Letter--Comment (4) of Bryan A. Erler X X ML092190138.
on Behalf of ASME Supporting Draft Regulatory
Guides DG-1191, DG-1192, DG-1193, and the
Proposed Rule Incorporating the Final
Revisions of these Regulatory Guides into 10
CFR 50.55a.
SRM-COMNJD-03-0002--Stabilizing the PRA Quality X X ML033520457.
Expectations and Requirements.
SECY-04-0118--Plan for the Implementation of X X ML041470505.
the Commission's Phased Approach to
Probabilistic Risk Assessment Quality.
SRM-SECY-04-0118--Plan for the Implementation X X ML042800369.
of the Commission's Phased Approach to
Probabilistic Risk Assessment Quality.
NUREG-0800--Chapter 4, Section 4.5.1, Revision X X ML070230007.
3, Control Rod Drive Structural Materials,
dated March 2007.
NUREG-0800--Chapter 5, Section 5.2.3, Revision X X ML063190006.
3, Reactor Coolant Pressure Boundary
Materials, dated March 2007.
NUREG/CR-6943--A Study of Remote Visual Methods X X ML073110060.
to Detect Cracking in Reactor Components.
----------------------------------------------------------------------------------------------------------------
X. Voluntary Consensus Standards
Section 12(d)(3) of the National Technology Transfer and
Advancement Act (NTTAA) of 1995, Public Law 104-113, and implementing
guidance in U.S. Office of Management and Budget (OMB) Circular A-119
(February 10, 1998), require each Federal government agency (should it
decide that regulation is necessary) to use a voluntary consensus
standard instead of developing a government-unique standard. An
exception to using a voluntary consensus standard is allowed where the
use of such a standard is inconsistent with applicable law or is
otherwise impractical. The NTTAA requires Federal agencies to use
industry consensus standards to the extent practical; it does not
require Federal agencies to endorse a standard in its entirety. Neither
the NTTAA nor OMB Circular A-119 prohibit an agency from adopting a
voluntary consensus standard while taking exception to specific
portions of the standard, if those provisions are deemed to be
``inconsistent with applicable law or otherwise impractical.''
Furthermore, taking specific exceptions furthers the Congressional
intent of Federal reliance on voluntary consensus standards because it
allows the adoption of substantial portions of consensus standards
without the need to reject the standards in their entirety because of
limited provisions that are not acceptable to the agency.
In this rulemaking, the NRC is continuing its existing practice of
approving the use of ASME BPV and OM Code Cases, which are ASME-
approved alternatives to compliance with various provisions of the ASME
BPV and OM Code. The NRC's approval of the ASME Code Cases is
accomplished by amending the NRC's regulations to incorporate by
reference the latest revisions of the following, which are the subject
of this rulemaking, into 10 CFR 50.55a: RG 1.84, ``Design, Fabrication,
and Materials Code Case Acceptability, ASME Section III,'' Revision 36;
RG 1.147, ``Inservice Inspection Code Case Acceptability, ASME Section
XI, Division 1,'' Revision 17; and RG 1.192, ``Operation and
Maintenance Code Case Acceptability, ASME Code,'' Revision 1. These RGs
list the ASME Code Cases that the NRC has approved for use. The ASME
Code Cases are national consensus standards as defined in the NTTAA and
OMB Circular A-119. The ASME Code Cases constitute voluntary consensus
standards, in which all interested parties (including the NRC and
licensees of nuclear power plants) participate. Therefore, the NRC's
approval of the use of the ASME Code Cases identified in RGs 1.84,
Revision 36; RG 1.147, Revision 17; and RG 1.192, Revision 1, which are
the subject of this rulemaking, is consistent with the overall
objectives of the NTTAA and OMB Circular A-119.
The NRC reviews each Section III, Section XI, and OM Code Case
published by the ASME to ascertain whether it is consistent with the
safe operation of nuclear power plants. The Code Cases found to be
generically acceptable are listed in the RGs that are incorporated by
reference in 10 CFR 50.55a. The Code Cases found to be unacceptable are
listed in RG 1.193, but licensees may still seek the NRC's approval to
apply these Code Cases through the processes in 10 CFR 50.55a for
requesting the approval of alternatives or for relief. Code Cases that
the NRC finds to be conditionally acceptable are also listed in RGs
1.84,
[[Page 37903]]
1.147, and 1.192, which are the subject of this rulemaking, together
with the conditions that must be used if the Code Case is applied. The
NRC believes that this rule complies with the NTTAA and OMB Circular A-
119 despite these conditions. If the NRC did not conditionally accept
ASME Code Cases, it would disapprove these Code Cases entirely. The
effect would be that licensees and applicants would submit a larger
number of requests for use of alternatives under the current 10 CFR
50.55a(a)(3), requests for relief under 10 CFR 50.55a(f) and (g), or
requests for exemptions under 10 CFR 50.12 and/or 10 CFR 52.7. For
these reasons, the treatment of ASME Code Cases and any conditions
proposed to be placed on them in this proposed rule do not conflict
with any policy on agency use of consensus standards specified in OMB
Circular A-119.
The NRC did not identify any other voluntary consensus standards
developed by the United States voluntary consensus standards bodies for
use within the United States that the NRC could approve instead of the
ASME Code Cases.
The NRC also did not identify any voluntary consensus standards
developed by multinational voluntary consensus standards bodies for use
on a multinational basis that the NRC could incorporate by reference
instead of the ASME Code Cases. This is because no other multinational
voluntary consensus body would develop alternatives to a voluntary
consensus standard (i.e., either the ASME BPV Code or the ASME OM Code)
for which they did not develop and do not maintain.
In summary, this proposed rule satisfies the requirements of
Section 12(d)(3) of the NTTAA and OMB Circular A-119.
XI. Finding of No Significant Environmental Impact: Environmental
Assessment
This proposed action stems from the Commission's practice of
incorporating by reference the RGs listing the most recent set of NRC-
approved ASME Code Cases. The purpose of this proposed action is to
allow licensees to use the Code Cases listed in the RGs as alternatives
to requirements in the ASME BPV and OM Codes for the construction, ISI,
and IST of nuclear power plant components. This proposed action is
intended to advance the NRC's strategic goal of ensuring adequate
protection of public health and safety and the environment. It also
demonstrates the agency's commitment to participate in the national
consensus standards process under the National Technology Transfer and
Advancement Act of 1995, Pub. L.104-113.
The National Environmental Policy Act (NEPA), as amended, requires
Federal government agencies to study the impacts of their ``major
Federal actions significantly affecting the quality of the human
environment'' and prepare detailed statements on the environmental
impacts of the action and alternatives to the action (United States
Code (U.S.C), Volume 42, Section 4332(C) [42 U.S.C. Sec. 4332(C)]; NEPA
Sec. 102(C)).
The Commission has determined under NEPA, as amended, and the
Commission's regulations in subpart A of 10 CFR part 51, that this
proposed rule would not be a major Federal action significantly
affecting the quality of the human environment. Therefore, an
environmental impact statement is not required.
As alternatives to the ASME Code, NRC-approved Code Cases provide
an equivalent level of safety. Therefore, the probability or
consequences of accidents is not changed. There are also no
significant, non-radiological impacts associated with this action
because no changes would be made affecting non-radiological plant
effluents and because no changes would be made in activities that would
adversely affect the environment. The determination of this
environmental assessment is that there will be no significant offsite
impact to the public from this proposed action.
XII. Paperwork Reduction Act Statement
This proposed rule contains new or amended information collection
requirements that are subject to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq). This rule has been submitted to the Office of
Management and Budget (OMB) for review and approval of the information
collection requirements.
Type of submission, new or revision: Revision.
The title of the information collection: Domestic Licensing of
Production and Utilization Facilities: Updates to Incorporation by
Reference and Regulatory Guides.
The form number if applicable: Not applicable.
How often the collection is required: On occasion.
Who will be required or asked to report: Power reactor licensees
and applicants for power reactors under construction.
An estimate of the number of annual responses: -185.
The estimated number of annual respondents: 109.
An estimate of the total number of hours needed annually to
complete the requirement or request: A reduction of 14,800 reporting
hours.
Abstract: This proposed rule is the latest in a series of
rulemakings that incorporate by reference the latest versions of
several Regulatory Guides identifying new and revised unconditionally
or conditionally acceptable ASME Code Cases that are approved for use.
The incorporation by reference of these Code Cases will reduce the
number of alternative requests submitted by licensees under proposed 10
CFR 50.55a(z) by an estimated 185 requests annually.
The U.S. Nuclear Regulatory Commission is seeking public comment on
the potential impact of the information collections contained in this
proposed rule (or proposed policy statement) 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 burden 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 information collection be minimized,
including the use of automated collection techniques?
The public may examine and have copied for a fee publicly available
documents, including the draft supporting statement at the NRC's PDR,
One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville,
Maryland 20852. The OMB clearance requests are available at the NRC's
Web site: http://www.nrc.gov/public-involve/doc-comment/omb/. The
document will be available on the NRC home page site for 60 days after
the signature date of this notice.
Send comments on any aspect of these proposed information
collections, including suggestions for reducing the burden and on the
above issues, by July 24, 2013 to the Information Services Branch (T-5
F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or
by email to [email protected] and to the Desk Officer, Chad
Whiteman, Office of Information and Regulatory Affairs, NEOB-10202,
(3150-0011), Office of Management and Budget, Washington, DC 20503.
Comments on the proposed information collections may also be submitted
via the Federal eRulemaking Portal http://www.regulations.gov, docket
NRC-2009-0359. Comments received after this date will be
considered if it is
[[Page 37904]]
practical to do so, but assurance of consideration cannot be given to
comments received after this date. Comments can also be emailed to
[email protected] or submitted by telephone at 202-395-
4718.
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
unless the requesting document displays a currently valid OMB control
number.
XIII. Regulatory Analysis
The ASME Code Cases listed in the RGs to be incorporated by
reference provide voluntary alternatives to the provisions in the ASME
BPV and OM Codes for design, construction, ISI, and IST of specific
structures, systems, and components used in nuclear power plants.
Implementation of these Code Cases is not required. Licensees and
applicants use NRC-approved ASME Code Cases to reduce unnecessary
regulatory burden or gain additional operational flexibility. It would
be difficult for the NRC to provide these advantages independently of
the ASME Code Case publication process without expending considerable
additional resources. The NRC has prepared a regulatory analysis
addressing the qualitative benefits of the alternatives considered in
this proposed rulemaking and comparing the costs associated with each
alternative (ADAMS Accession No. ML103060189). The NRC invites public
comment on this draft regulatory analysis. Copies of the regulatory
analysis are available to the public as indicated in Section IX,
``Availability of Documents,'' of this document.
In addition to the general opportunity to submit comments on the
proposed rule, the NRC also requests comments on the NRC's cost and
benefit estimates as shown in the proposed rule regulatory analysis.
XIV. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act of 1980 (5 U.S.C. 605(b)), the
Commission certifies that this proposed rule would not impose a
significant economical impact on a substantial number of small
entities. This proposed rule would affect only the licensing and
operation of nuclear power plants. The companies that own these plants
are not ``small entities'' as defined in the Regulatory Flexibility Act
or the size standards established by the NRC (10 CFR 2.810).
XV. Backfitting and Issue Finality
The provisions in this proposed rulemaking would allow licensees
and applicants to voluntarily apply NRC-approved Code Cases, sometimes
with NRC-specified conditions. The approved Code Cases are listed in
three regulatory guides that are incorporated by references into 10 CFR
50.55a.
An applicant's and/or licensees voluntary application of an
approved Code Cases does not constitute backfitting, inasmuch as there
is no imposition of a new requirement or new position. Similarly,
voluntary application of an approved Code Case by a part 52 applicant
or licensee does not represent NRC imposition of a requirement or
action which is inconsistent with any issue finality provision in part
52. For these reasons the NRC finds that this proposed rule does not
involve any provisions requiring the preparation of a backfit analysis
or documentation demonstrating that one or more of the issue finality
criteria in part 52 are met.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Radiation protection, Reactor siting
criteria, Reporting and recordkeeping requirements.
For the reasons set forth in the preamble, and under the authority
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 553, the NRC is proposing 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 secs. 102, 103, 104, 105, 147,
149, 161, 181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133,
2134, 2135, 2167, 2169, 2201, 2231, 2232, 2233, 2236, 2239, 2273,
2282); Energy Reorganization Act secs. 201, 202, 206 (42 U.S.C.
5841, 5842, 5846); Nuclear Waste Policy Act sec. 306 (42 U.S.C.
10226); Government Paperwork Elimination Act sec. 1704 (44 U.S.C.
3504 note); Energy Policy Act of 2005, Pub. L. No. 109-58, 119 Stat.
194 (2005). Section 50.7 also issued under Pub. L. 95-601, sec. 10,
as amended by Pub. L. 102-486, sec. 2902 (42 U.S.C. 5851). Section
50.10 also issued under Atomic Energy Act secs. 101, 185 (42 U.S.C.
2131, 2235); National Environmental Policy Act sec. 102 (42 U.S.C.
4332). Sections 50.13, 50.54(dd), and 50.103 also issued under
Atomic Energy Act sec. 108 (42 U.S.C. 2138).
Sections 50.23, 50.35, 50.55, and 50.56 also issued under Atomic
Energy Act sec. 185 (42 U.S.C. 2235). Appendix Q also issued under
National Environmental Policy Act sec. 102 (42 U.S.C. 4332).
Sections 50.34 and 50.54 also issued under sec. 204 (42 U.S.C.
5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L.
97-415 (42 U.S.C. 2239). Section 50.78 also issued under Atomic
Energy Act sec. 122 (42 U.S.C. 2152). Sections 50.80--50.81 also
issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).
0
2. In Sec. 50.54, revise the introductory text of the section, add and
reserve paragraph (ii), and add paragraph (jj) to read as follows:
Sec. 50.54 Conditions of licenses.
The following paragraphs of this section, with the exception of
paragraphs (r) and (gg), and the applicable requirements of 10 CFR
50.55a, are conditions in every nuclear power reactor operating license
issued under this part. The following paragraphs with the exception of
paragraph (r), (s), and (u) of this section are conditions in every
combined license issued under part 52 of this chapter, provided,
however, that paragraphs (i), (i-1), (j), (k), (l), (m), (n), (w), (x),
(y), and (z) of this section are only applicable after the Commission
makes the finding under Sec. 52.103(g) of this chapter.
* * * * *
(ii) [Reserved]
(jj) Structures, systems, and components must be designed,
fabricated, erected, constructed, tested, and inspected to quality
standards commensurate with the importance of the safety function to be
performed.
0
3. In Sec. 50.55, revise the introductory text of the section, add and
reserve paragraphs (g) and (h), and add paragraph (i) to read as
follows:
Sec. 50.55 Conditions of construction permits, early site permits,
combined licenses, and manufacturing licenses.
Each construction permit is subject to the following terms and
conditions and the applicable requirements of 10 CFR 50.55a; each early
site permit is subject to the terms and conditions in paragraph (f) of
this section; each manufacturing license is subject to the terms and
conditions in paragraphs (e) and (f) of this section and the applicable
requirements of 10 CFR 50.55a; and each combined license is subject to
the terms and conditions in paragraphs (e) and (f) of this section and
the applicable requirements of 10 CFR 50.55a until the date that the
Commission makes the
[[Page 37905]]
finding under Sec. 52.103(g) of this chapter:
* * * * *
(g) [Reserved]
(h) [Reserved]
(i) Structures, systems, and components must be designed,
fabricated, erected, constructed, tested, and inspected to quality
standards commensurate with the importance of the safety function to be
performed.
0
4. Revise Sec. 50.55a to read as follows:
Sec. 50.55a Codes and standards.
(a) Documents approved for incorporation by reference. The
standards listed in this paragraph have been approved for incorporation
by reference by the Director of the Federal Register pursuant to 5
U.S.C. 552(a) and 1 CFR Part 51. The standards are available for
inspection at the NRC Technical Library, 11545 Rockville Pike,
Rockville, Maryland 20852; or at the National Archives and Records
Administration (NARA). For information on the availability of this
material at NARA, call 202-741-6030 or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.
(1) American Society of Mechanical Engineers (ASME), Three Park
Avenue, New York, NY 10016 (telephone 800-843-2763), http://www.asme.org/Codes/.
(i) ASME Boiler and Pressure Vessel Code, Section III. The editions
and addenda for Section III of the ASME Boiler and Pressure Vessel Code
are listed below, but limited to those provisions identified in
paragraph (b)(1) of this section.
(A) ``Rules for Construction of Nuclear Vessels:''
(1) 1963 Edition,
(2) Summer 1964 Addenda,
(3) Winter 1964 Addenda,
(4) 1965 Edition
(5) 1965 Summer Addenda,
(6) 1965 Winter Addenda,
(7) 1966 Summer Addenda,
(8) 1966 Winter Addenda,
(9) 1967 Summer Addenda,
(10) 1967 Winter Addenda,
(11) 1968 Edition,
(12) 1968 Summer Addenda,
(13)1968 Winter Addenda,
(14) 1969 Summer Addenda,
(15) 1969 Winter Addenda,
(16) 1970 Summer Addenda, and
(17) 1970 Winter Addenda.
(B) ``Rules for Construction of Nuclear Power Plant Components:''
(1) 1971 Edition,
(2) 1971 Summer Addenda,
(3) 1971 Winter Addenda,
(4) 1972 Summer Addenda,
(5) 1972 Winter Addenda,
(6) 1973 Summer Addenda, and
(7) 1973 Winter Addenda.
(C) ``Division 1 Rules for Construction of Nuclear Power Plant
Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda,
(4) 1975 Summer Addenda,
(5) 1975 Winter Addenda,
(6) 1976 Summer Addenda, and
(7) 1976 Winter Addenda;
(D) ``Rules for Construction of Nuclear Power Plant Components--
Division 1;''
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Summer Addenda,
(10) 1980 Winter Addenda,
(11) 1981 Summer Addenda,
(12) 1981 Winter Addenda,
(13) 1982 Summer Addenda,
(14) 1982 Winter Addenda,
(15) 1983 Edition,
(16) 1983 Summer Addenda,
(17) 1983 Winter Addenda,
(18) 1984 Summer Addenda,
(19) 1984 Winter Addenda,
(20) 1985 Summer Addenda,
(21) 1985 Winter Addenda,
(22) 1986 Edition,
(23) 1986 Addenda,
(24) 1987 Addenda,
(25) 1988 Addenda,
(26) 1989 Edition,
(27) 1989 Addenda,
(28) 1990 Addenda,
(29) 1991 Addenda,
(30) 1992 Edition,
(31) 1992 Addenda,
(32) 1993 Addenda,
(33) 1994 Addenda,
(34) 1995 Edition,
(35)1995 Addenda,
(36)1996 Addenda, and
(37) 1997 Addenda.
(E) ``Rules for Construction of Nuclear Facility Components--
Division 1:''
(1) 1998 Edition,
(2) 1998 Addenda,
(3) 1999 Addenda,
(4) 2000 Addenda,
(5) 2001 Edition,
(6) 2001 Addenda,
(7) 2002 Addenda,
(8) 2003 Addenda,
(9) 2004 Edition,
(10) 2005 Addenda,
(11) 2006 Addenda,
(12) 2007 Edition, and
(13) 2008 Addenda.
(ii) ASME Boiler and Pressure Vessel Code, Section XI. The editions
and addenda for Section XI of the ASME Boiler and Pressure Vessel Code
are listed below, but limited to those provisions identified in
paragraph (b)(2) of this section.
(A) ``Rules for Inservice Inspection of Nuclear Reactor Coolant
Systems:''
(1) 1970 Edition,
(2) 1971 Edition,
(3) 1971 Summer Addenda,
(4) 1971 Winter Addenda,
(5) 1972 Summer Addenda,
(6) 1972 Winter Addenda,
(7) 1973 Summer Addenda, and
(8) 1973 Winter Addenda.
(B) ``Rules for Inservice Inspection of Nuclear Power Plant
Components:''
(1) 1974 Edition,
(2) 1974 Summer Addenda,
(3) 1974 Winter Addenda, and
(4) 1975 Summer Addenda.
(C) ``Rules for Inservice Inspection of Nuclear Power Plant
Components--Division 1:''
(1) 1977 Edition,
(2) 1977 Summer Addenda,
(3) 1977 Winter Addenda,
(4) 1978 Summer Addenda,
(5) 1978 Winter Addenda,
(6) 1979 Summer Addenda,
(7) 1979 Winter Addenda,
(8) 1980 Edition,
(9) 1980 Winter Addenda,
(10) 1981 Summer Addenda,
(11) 1981 Winter Addenda,
(12) 1982 Summer Addenda,
(13) 1982 Winter Addenda,
(14) 1983 Edition,
(15) 1983 Summer Addenda,
(16) 1983 Winter Addenda,
(17) 1984 Summer Addenda,
(18) 1984 Winter Addenda,
(19) 1985 Summer Addenda,
(20) 1985 Winter Addenda,
(21) 1986 Edition,
(22) 1986 Addenda,
(23) 1987 Addenda,
(24) 1988 Addenda,
(25) 1989 Edition,
(26) 1989 Addenda,
(27) 1990 Addenda,
(28) 1991 Addenda,
(28) 1992 Edition,
(30) 1992 Addenda,
(31) 1993 Addenda,
(32) 1994 Addenda,
(33) 1995 Edition,
(34) 1995 Addenda,
(35) 1996 Addenda,
(36) 1997 Addenda,
(37) 1998 Edition,
(38) 1998 Addenda,
(39) 1999 Addenda,
(40) 2000 Addenda,
(41) 2001 Edition,
(42) 2001 Addenda,
(43) 2002 Addenda,
(44) 2003 Addenda,
[[Page 37906]]
(45) 2004 Edition,
(46) 2005 Addenda,
(47) 2006 Addenda,
(48) 2007 Edition, and
(49) 2008 Addenda.
(iii) ASME Code Cases: Nuclear Components
(A) ASME Code Case N-722-1. ASME Code Case N-722-1, ``Additional
Examinations for PWR Pressure Retaining Welds in Class 1 Components
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1''
(Approval Date: January 26, 2009), with the conditions in paragraph
(g)(6)(ii)(E) of this section.
(B) ASME Code Case N-729-1. ASME Code Case N-729-1, ``Alternative
Examination Requirements for PWR Reactor Vessel Upper Heads With
Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section
XI, Division 1'' (Approval Date: March 28, 2006), with the conditions
in paragraph (g)(6)(ii)(D) of this section.
(C) ASME Code Case N-770-1. ASME Code Case N-770-1, ``Additional
Examinations for PWR Pressure Retaining Welds in Class 1 Components
Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1''
(Approval Date: December 25, 2009), with the conditions in paragraph
(g)(6)(ii)(F) of this section.
(iv) ASME Operation and Maintenance Code. The editions and addenda
for the ASME Code for Operation and Maintenance of Nuclear Power Plants
are listed below, but limited to those provisions identified in
paragraph (b)(3) of this section.
(A) ``Code for Operation and Maintenance of Nuclear Power Plants:''
(1) 1995 Edition,
(2) 1996 Addenda,
(3) 1997 Addenda,
(4) 1998 Edition,
(5) 1999 Addenda,
(6) 2000 Addenda,
(7) 2001 Edition,
(8) 2002 Addenda,
(9) 2003 Addenda,
(10) 2004 Edition,
(11) 2005 Addenda, and
(12) 2006 Addenda.
(B) [Reserved]
(2) Institute of Electrical and Electronics Engineers (IEEE)
Service Center, 445 Hoes Lane, Piscataway, NJ 08855.
(i) IEEE standard 279-1971. (IEEE Std 279-1971), ``Criteria for
Protection Systems for Nuclear Power Generating Stations'' (Approval
Date: June 3, 1971), referenced in paragraphs (h)(2) of this section.
(ii) IEEE Standard 603-1991. (IEEE Std 603-1991), ``Standard
Criteria for Safety Systems for Nuclear Power Generating Stations''
(Approval Date: June 27, 1971), referenced in paragraphs (h)(2) and
(h)(3) of this section. All other standards that are referenced in IEEE
Std 603-1991 are not approved incorporation by reference.
(iii) IEEE standard 603-1991, correction sheet. (IEEE Std 603-1991
correction sheet), ``Standard Criteria for Safety Systems for Nuclear
Power Generating Stations, Correction Sheet, Issued January 30, 1995,
'' referenced in paragraphs (h)(2) and (h)(3) of this section. (Copies
of this correction sheet may be purchased from Thomson Reuters, 3916
Ranchero Dr., Ann Arbor, MI 48108, http://www.techstreet.com.)
(3) U.S. Nuclear Regulatory Commission (NRC) Reproduction and
Distribution Services Section, Washington, DC 20555- 0001; fax: 301-
415-2289; email: [email protected].
(i) NRC Regulatory Guide 1.84, Revision 36. NRC Regulatory Guide
1.84, Revision 36, ``Design, Fabrication, and Materials Code Case
Acceptability, ASME Section III,'' [INSERT DATE OF FINAL RULE
PUBLICATION IN THE Federal Register], with the requirements in
paragraph (b)(4) of this section.
(ii) NRC Regulatory Guide 1.147, Revision 17. NRC Regulatory Guide
1.147, Revision 17, ``Inservice Inspection Code Case Acceptability,
ASME Section XI, Division 1,'' [INSERT DATE OF FINAL RULE PUBLICATION
IN THE Federal Register], which lists ASME Code Cases that the NRC has
approved in accordance with the requirements in paragraph (b)(5) of
this section.
(iii) NRC Regulatory Guide 1.192, Revision 1. NRC Regulatory Guide
1.192, Revision 1, ``Operation and Maintenance Code Case Acceptability,
ASME OM Code,'' [INSERT DATE OF FINAL RULE PUBLICATION IN THE Federal
Register], which lists ASME Code Cases that the NRC has approved in
accordance with the requirements in paragraph (b)(6) of this section.
(b) Use and conditions on the use of standards. Systems and
components of boiling and pressurized water-cooled nuclear power
reactors must meet the requirements of the ASME Boiler and Pressure
Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance
of Nuclear Power Plants (OM Code) as specified in this paragraph. Each
combined license for a utilization facility is subject to the following
conditions.
(1) Conditions on ASME BPV Code Section III. Each manufacturing
license, standard design approval, and design certification under Part
52 of this chapter is subject to the following conditions. As used in
this section, references to Section III refer to Section III of the
ASME Boiler and Pressure Vessel Code and include the 1963 Edition
through 1973 Winter Addenda and the 1974 Edition (Division 1) through
the 2008 Addenda (Division 1), subject to the following conditions:
(i) Section III condition: Section III materials. When applying the
1992 Edition of Section III, applicants or licensees must apply the
1992 Edition with the 1992 Addenda of Section II of the ASME Boiler and
Pressure Vessel Code.
(ii) Section III condition: Weld leg dimensions. When applying the
1989 Addenda through the latest edition and addenda, applicants or
licensees may not apply subparagraphs NB-3683.4(c)(1) and NB-
3683.4(c)(2) or Footnote 11 from the 1989 Addenda through the 2003
Addenda, or Footnote 13 from the 2004 Edition through the 2008 Addenda
to Figures NC-3673.2(b)-1 and ND-3673.2(b)-1 for welds with leg size
less than 1.09 tn.
(iii) Section III condition: Seismic design of piping. Applicants
or licensees may use Subarticles NB-3200, NB-3600, NC-3600, and ND-3600
for seismic design of piping, up to and including the 1993 Addenda,
subject to the condition specified in paragraph (b)(1)(ii) of this
section. Applicants or licensees may not use these subarticles for
seismic design of piping in the 1994 Addenda through the 2005 Addenda
incorporated by reference in paragraph (a)(1) of this section, except
that Subarticle NB-3200 in the 2004 Edition through the 2008 Addenda
may be used by applicants and licensees, subject to the condition in
paragraph (b)(1)(iii)(A) of this section. Applicants or licensees may
use Subarticles NB-3600, NC-3600, and ND-3600 for the seismic design of
piping in the 2006 Addenda through the 2008 Addenda, subject to the
conditions of this paragraph corresponding to those subarticles.
(A) Seismic design of piping: first provision. When applying Note
(1) of Figure NB-3222-1 for Level B service limits, the calculation of
Pb stresses must include reversing dynamic loads (including
inertia earthquake effects) if evaluation of these loads is required by
NB-3223(b).
(B) Seismic design of piping: second provision. For Class 1 piping,
the material and Do/t requirements of NB-3656(b) must be met
for all Service Limits when the Service Limits include reversing
dynamic loads, and the alternative rules for reversing dynamic loads
are used.
[[Page 37907]]
(iv) Section III condition: Quality assurance. When applying
editions and addenda later than the 1989 Edition of Section III, the
requirements of NQA-1, ``Quality Assurance Requirements for Nuclear
Facilities,'' 1986 Edition through the 1994 Edition, are acceptable for
use, provided that the edition and addenda of NQA-1 specified in NCA-
4000 is used in conjunction with the administrative, quality, and
technical provisions contained in the edition and addenda of Section
III being used.
(v) Section III condition: Independence of inspection. Applicants
or licensees may not apply NCA-4134.10(a) of Section III, 1995 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1) of this section.
(vi) Section III condition: Subsection NH. The provisions in
Subsection NH, ``Class 1 Components in Elevated Temperature Service,''
1995 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1) of this section, may only be used for the
design and construction of Type 316 stainless steel pressurizer heater
sleeves where service conditions do not cause the components to reach
temperatures exceeding 900[emsp14][deg]F.
(vii) Section III condition: Capacity certification and
demonstration of function of incompressible-fluid pressure-relief
valves. When applying the 2006 Addenda through the 2007 Edition up to
and including the 2008 Addenda, applicants and licensees may use
paragraph NB-7742, except that paragraph NB-7742(a)(2) may not be used.
For a valve design of a single size to be certified over a range of set
pressures, the demonstration of function tests under paragraph NB-7742
must be conducted as prescribed in NB-7732.2 on two valves covering the
minimum set pressure for the design and the maximum set pressure that
can be accommodated at the demonstration facility selected for the
test.
(2) Conditions on ASME BPV Code Section XI. As used in this
section, references to Section XI refer to Section XI, Division 1, of
the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition
through the 1976 Winter Addenda and the 1977 Edition through the 2007
Edition with the 2008 Addenda, subject to the following conditions:
(i) [Reserved]
(ii) Section XI condition: Pressure-retaining welds in ASME Code
Class 1 piping (applies to Table IWB-2500 and IWB-2500-1 and Category
B-J). If the facility's application for a construction permit was
docketed prior to July 1, 1978, the extent of examination for Code
Class 1 pipe welds may be determined by the requirements of Table IWB-
2500 and Table IWB-2600 Category B-J of Section XI of the ASME BPV Code
in the 1974 Edition and Addenda through the Summer 1975 Addenda or
other requirements the NRC may adopt.
(iii) [Reserved]
(iv) [Reserved]
(v) [Reserved]
(vi) Section XI condition: Effective edition and addenda of
Subsection IWE and Subsection IWL. Applicants or licensees may use
either the 1992 Edition with the 1992 Addenda or the 1995 Edition with
the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned
by the requirements in paragraphs (b)(2)(viii) and (b)(2)(ix) of this
section, when implementing the initial 120-month inspection interval
for the containment inservice inspection requirements of this section.
Successive 120-month interval updates must be implemented in accordance
with paragraph (g)(4)(ii) of this section.
(vii) Section XI condition: Section XI references to OM Part 4, OM
Part 6, and OM Part 10 (Table IWA-1600-1). When using Table IWA-1600-1,
``Referenced Standards and Specifications,'' in the Section XI,
Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified
``Revision Date or Indicator'' for ASME/ANSI OM part 4, ASME/ANSI part
6, and ASME/ANSI part 10 must be the OMa--1988 Addenda to the OM-1987
Edition. These requirements have been incorporated into the OM Code,
which is incorporated by reference in paragraph (a)(1)(iv) of this
section.
(viii) Section XI condition: Concrete containment examinations.
Applicants or licensees applying Subsection IWL, 1992 Edition with the
1992 Addenda, must apply paragraphs (b)(2)(viii)(A) through
(b)(2)(viii)(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 (b)(2)(viii)(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 (b)(2)(viii)(G) of this section. Applicants or licensees
applying Subsection IWL, 2007 Edition through the latest edition and
addenda incorporated by reference in paragraph (a)(1)(ii) of this
section, must apply paragraph (b)(2)(viii)(E) of this section.
(A) Concrete containment examinations: first provision. Grease caps
that are accessible must be visually examined to detect grease leakage
or grease cap deformations. Grease caps must be removed for this
examination when there is evidence of grease cap deformation that
indicates deterioration of anchorage hardware.
(B) Concrete containment examinations: second provision. When
evaluation of consecutive surveillances of prestressing forces for the
same tendon or tendons in a group indicates a trend of prestress loss
such that the tendon force(s) would be less than the minimum design
prestress requirements before the next inspection interval, an
evaluation must be performed and reported in the Engineering Evaluation
Report as prescribed in IWL-3300.
(C) Concrete containment examinations: third provision. When the
elongation corresponding to a specific load (adjusted for effective
wires or strands) during retensioning of tendons differs by more than
10 percent from that recorded during the last measurement, an
evaluation must be performed to determine whether the difference is
related to wire failures or slip of wires in anchorage. A difference of
more than 10 percent must be identified in the ISI Summary Report
required by IWA-6000.
(D) Concrete containment examinations: fourth provision. The
applicant or licensee must report the following conditions, if they
occur, in the ISI Summary Report required by IWA-6000:
(1) The sampled sheathing filler grease contains chemically
combined water exceeding 10 percent by weight or the presence of free
water;
(2) The absolute difference between the amount removed and the
amount replaced exceeds 10 percent of the tendon net duct volume; and
(3) Grease leakage is detected during general visual examination of
the containment surface.
(E) Concrete containment examinations: fifth provision. For Class
CC applications, the applicant or licensee must evaluate the
acceptability of inaccessible areas when conditions exist in accessible
areas that could indicate the presence of or the result in degradation
to such inaccessible areas. For each inaccessible area identified, the
applicant or licensee must provide the following in the ISI Summary
Report required by IWA-6000:
(1) A description of the type and estimated extent of degradation,
and the conditions that led to the degradation;
(2) An evaluation of each area, and the result of the evaluation;
and
[[Page 37908]]
(3) A description of necessary corrective actions.
(F) Concrete containment examinations: sixth provision. Personnel
that examine containment concrete surfaces and tendon hardware, wires,
or strands must meet the qualification provisions in IWA-2300. The
``owner-defined'' personnel qualification provisions in IWL-2310(d) are
not approved for use.
(G) Concrete containment examinations: seventh provision. Corrosion
protection material must be restored following concrete containment
post-tensioning system repair and replacement activities in accordance
with the quality assurance program requirements specified in IWA-1400.
(ix) Section XI condition: Metal containment examinations.
Applicants or licensees applying Subsection IWE, 1992 Edition with the
1992 Addenda, or the 1995 Edition with the 1996 Addenda, must satisfy
the requirements of paragraphs (b)(2)(ix)(A) through (b)(2)(ix)(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), (b)(2)(ix)(B), and
(b)(2)(ix)(F) through (b)(2)(ix)(I) of this section. Applicants or
licensees applying Subsection IWE, 2004 Edition, up to and including
the 2005 Addenda, must satisfy the requirements of paragraphs
(b)(2)(ix)(A), (b)(2)(ix)(B), and (b)(2)(ix)(F) through (b)(2)(ix)(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 addenda incorporated by reference in paragraph (a)(1)(ii) of
this section, must satisfy the requirements of paragraphs
(b)(2)(ix)(A)(2), (b)(2)(ix)(B), and (b)(2)(ix)(J) of this section.
(A) Metal containment examinations: first provision. For Class MC
applications, the following apply to inaccessible areas.
(1) The applicant or licensee must evaluate the acceptability of
inaccessible areas when conditions exist in accessible areas that could
indicate the presence of or could result in degradation to such
inaccessible areas.
(2) For each inaccessible area identified for evaluation, the
applicant or licensee must provide the following in the ISI Summary
Report as required by IWA-6000:
(i) A description of the type and estimated extent of degradation,
and the conditions that led to the degradation;
(ii) An evaluation of each area, and the result of the evaluation;
and
(iii) A description of necessary corrective actions.
(B) Metal containment examinations: second provision. When
performing remotely the visual examinations required by Subsection IWE,
the maximum direct examination distance specified in Table IWA-2210-1
may be extended and the minimum illumination requirements specified in
Table IWA-2210-1 may be decreased provided that the conditions or
indications for which the visual examination is performed can be
detected at the chosen distance and illumination.
(C) Metal containment examinations: third provision. The
examinations specified in Examination Category E-B, Pressure Retaining
Welds, and Examination Category E-F, Pressure Retaining Dissimilar
Metal Welds, are optional.
(D) Metal containment examinations: fourth provision. This
paragraph (b)(2)(ix)(D) may be used as an alternative to the
requirements of IWE-2430.
(1) If the examinations reveal flaws or areas of degradation
exceeding the acceptance standards of Table IWE-3410-1, an evaluation
must be performed to determine whether additional component
examinations are required. For each flaw or area of degradation
identified that exceeds acceptance standards, the applicant or licensee
must provide the following in the ISI Summary Report required by IWA-
6000:
(i) A description of each flaw or area, including the extent of
degradation, and the conditions that led to the degradation;
(ii) The acceptability of each flaw or area and the need for
additional examinations to verify that similar degradation does not
exist in similar components; and
(iii) A description of necessary corrective actions.
(2) The number and type of additional examinations to ensure
detection of similar degradation in similar components.
(E) Metal containment examinations: fifth provision. A general
visual examination as required by Subsection IWE must be performed once
each period.
(F) Metal containment examinations: sixth provision. VT-1 and VT-3
examinations must be conducted in accordance with IWA-2200. Personnel
conducting examinations in accordance with the VT-1 or VT-3 examination
method must be qualified in accordance with IWA-2300. The ``owner-
defined'' personnel qualification provisions in IWE-2330(a) for
personnel that conduct VT-1 and VT-3 examinations are not approved for
use.
(G) Metal containment examinations: seventh provision. The VT-3
examination method must be used to conduct the examinations in Items
E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method
must be used to conduct the examination in Item E4.11 of Table IWE-
2500-1. An examination of the pressure-retaining bolted connections in
Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must
be conducted once each interval. The ``owner-defined'' visual
examination provisions in IWE-2310(a) are not approved for use for VT-1
and VT-3 examinations.
(H) Metal containment examinations: eighth provision. Containment
bolted connections that are disassembled during the scheduled
performance of the examinations in Item E1.11 of Table IWE-2500-1 must
be examined using the VT-3 examination method. Flaws or degradation
identified during the performance of a VT-3 examination must be
examined in accordance with the VT-1 examination method. The criteria
in the material specification or IWB-3517.1 must be used to evaluate
containment bolting flaws or degradation. As an alternative to
performing VT-3 examinations of containment bolted connections that are
disassembled during the scheduled performance of Item E1.11, VT-3
examinations of containment bolted connections may be conducted
whenever containment bolted connections are disassembled for any
reason.
(I) Metal containment examinations: ninth provision. The ultrasonic
examination acceptance standard specified in IWE-3511.3 for Class MC
pressure-retaining components must also be applied to metallic liners
of Class CC pressure-retaining components.
(J) Metal containment examinations: tenth provision. In general, a
repair/replacement activity such as replacing a large containment
penetration, cutting a large construction opening in the containment
pressure boundary to replace steam generators, reactor vessel heads,
pressurizers, or other major equipment; or other similar modification
is considered a major containment modification. When applying IWE-5000
to Class MC pressure-retaining components, any major containment
modification or repair/replacement must be followed by a Type A test to
provide assurance of
[[Page 37909]]
both containment structural integrity and leaktight integrity prior to
returning to service, in accordance with 10 CFR Part 50, Appendix J,
Option A or Option B on which the applicant's or licensee's Containment
Leak-Rate Testing Program is based. When applying IWE-5000, if a Type
A, B, or C Test is performed, the test pressure and acceptance standard
for the test must be in accordance with 10 CFR Part 50, Appendix J.
(x) Section XI condition: Quality assurance. When applying Section
XI editions and addenda later than the 1989 Edition, the requirements
of NQA-1, ``Quality Assurance Requirements for Nuclear Facilities,''
1979 Addenda through the 1989 Edition, are acceptable as permitted by
IWA-1400 of Section XI, if the licensee uses its 10 CFR Part 50,
Appendix B, quality assurance program, in conjunction with Section XI
requirements. Commitments contained in the licensee's quality assurance
program description that are more stringent than those contained in
NQA-1 must govern Section XI activities. Further, where NQA-1 and
Section XI do not address the commitments contained in the licensee's
Appendix B quality assurance program description, the commitments must
be applied to Section XI activities.
(xi) [Reserved]
(xii) Section XI condition: Underwater welding. The provisions in
IWA-4660, ``Underwater Welding,'' of Section XI, 1997 Addenda through
the latest edition and addenda incorporated by reference in paragraph
(a)(1)(ii) of this section, are not approved for use on irradiated
material.
(xiii) [Reserved]
(xiv) Section XI condition: Appendix VIII personnel qualification.
All personnel qualified for performing ultrasonic examinations in
accordance with Appendix VIII must receive 8 hours of annual hands-on
training on specimens that contain cracks. Licensees applying the 1999
Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section may use the annual
practice requirements in VII-4240 of Appendix VII of Section XI in
place of the 8 hours of annual hands-on training provided that the
supplemental practice is performed on material or welds that contain
cracks, or by analyzing prerecorded data from material or welds that
contain cracks. In either case, training must be completed no earlier
than 6 months prior to performing ultrasonic examinations at a
licensee's facility.
(xv) Section XI condition: Appendix VIII specimen set and
qualification requirements. Licensees using Appendix VIII in the 1995
Edition through the 2001 Edition of the ASME Boiler and Pressure Vessel
Code may elect to comply with all of the provisions in paragraphs
(b)(2)(xv)(A) through (b)(2)(xv)(M) of this section, except for
paragraph (b)(2)(xv)(F) of this section, which may be used at the
licensee's option. Licensees using editions and addenda after 2001
Edition through the 2006 Addenda must use the 2001 Edition of Appendix
VIII and may elect to comply with all of the provisions in paragraphs
(b)(2)(xv)(A) through (b)(2)(xv)(M) of this section, except for
paragraph (b)(2)(xv)(F) of this section, which may be used at the
licensee's option.
(A) Specimen set and qualification: first provision. When applying
Supplements 2, 3, and 10 to Appendix VIII, the following examination
coverage criteria requirements must be used:
(1) Piping must be examined in two axial directions, and when
examination in the circumferential direction is required, the
circumferential examination must be performed in two directions,
provided access is available. Dissimilar metal welds must be examined
axially and circumferentially.
(2) Where examination from both sides is not possible, full
coverage credit may be claimed from a single side for ferritic welds.
Where examination from both sides is not possible on austenitic welds
or dissimilar metal welds, full coverage credit from a single side may
be claimed only after completing a successful single-sided Appendix
VIII demonstration using flaws on the opposite side of the weld.
Dissimilar metal weld qualifications must be demonstrated from the
austenitic side of the weld, and the qualification may be expanded for
austenitic welds with no austenitic sides using a separate add-on
performance demonstration. Dissimilar metal welds may be examined from
either side of the weld.
(B) Specimen set and qualification: second provision. The following
conditions must be used in addition to the requirements of Supplement 4
to Appendix VIII:
(1) Paragraph 3.1, Detection acceptance criteria--Personnel are
qualified for detection if the results of the performance demonstration
satisfy the detection requirements of ASME Section XI, Appendix VIII,
Table VIII-S4-1, and no flaw greater than 0.25 inch through-wall
dimension is missed.
(2) Paragraph 1.1(c), Detection test matrix--Flaws smaller than the
50 percent of allowable flaw size, as defined in IWB-3500, need not be
included as detection flaws. For procedures applied from the inside
surface, use the minimum thickness specified in the scope of the
procedure to calculate a/t. For procedures applied from the outside
surface, the actual thickness of the test specimen is to be used to
calculate a/t.
(C) Specimen set and qualification: third provision. When applying
Supplement 4 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.15 inch RMS must be used in
lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a
length sizing requirement of 0.75 inch RMS must be used in lieu of the
requirement in Subparagraph 3.2(b).
(2) In lieu of the location acceptance criteria requirements of
Subparagraph 2.1(b), a flaw will be considered detected when reported
within 1.0 inch or 10 percent of the metal path to the flaw, whichever
is greater, of its true location in the X and Y directions.
(3) In lieu of the flaw type requirements of Subparagraph
1.1(e)(1), a minimum of 70 percent of the flaws in the detection and
sizing tests must be cracks. Notches, if used, must be limited by the
following:
(i) Notches must be limited to the case where examinations are
performed from the clad surface.
(ii) Notches must be semielliptical with a tip width of less than
or equal to 0.010 inches.
(iii) Notches must be perpendicular to the surface within 2 degrees.
(4) In lieu of the detection test matrix requirements in paragraphs
1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain
a representative distribution of flaw orientations, sizes, and
locations.
(D) Specimen set and qualification: fourth provision. The following
conditions must be used in addition to the requirements of Supplement 6
to Appendix VIII:
(1) Paragraph 3.1, Detection Acceptance Criteria--Personnel are
qualified for detection if:
(i) No surface connected flaw greater than 0.25 inch through-wall
has been missed.
(ii) No embedded flaw greater than 0.50 inch through-wall has been
missed.
(2) Paragraph 3.1, Detection Acceptance Criteria--For procedure
qualification, all flaws within the scope of the procedure are
detected.
(3) Paragraph 1.1(b) for detection and sizing test flaws and
locations--Flaws smaller than the 50 percent of allowable flaw size, as
defined in IWB-3500, need not be included as detection flaws. Flaws
that are less than the allowable flaw size, as defined in IWB-3500, may
be used as detection and sizing flaws.
[[Page 37910]]
(4) Notches are not permitted.
(E) Specimen set and qualification: fifth provision. When applying
Supplement 6 to Appendix VIII, the following conditions must be used:
(1) A depth sizing requirement of 0.25 inch RMS must be used in
lieu of the requirements of subparagraphs 3.2(a), 3.2(c)(2), and
3.2(c)(3).
(2) In lieu of the location acceptance criteria requirements in
Subparagraph 2.1(b), a flaw will be considered detected when reported
within 1.0 inch or 10 percent of the metal path to the flaw, whichever
is greater, of its true location in the X and Y directions.
(3) In lieu of the length sizing criteria requirements of
Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch
RMS must be used.
(4) In lieu of the detection specimen requirements in Subparagraph
1.1(e)(1), a minimum of 55 percent of the flaws must be cracks. The
remaining flaws may be cracks or fabrication type flaws, such as slag
and lack of fusion. The use of notches is not allowed.
(5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test
matrix, personnel demonstration test sets must contain a representative
distribution of flaw orientations, sizes, and locations.
(F) Specimen set and qualification: sixth provision. The following
conditions may be used for personnel qualification for combined
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII
qualification. Licensees choosing to apply this combined qualification
must apply all of the provisions of Supplements 4 and 6 including the
following conditions:
(1) For detection and sizing, the total number of flaws must be at
least 10. A minimum of 5 flaws must be from Supplement 4, and a minimum
of 50 percent of the flaws must be from Supplement 6. At least 50
percent of the flaws in any sizing must be cracks. Notches are not
acceptable for Supplement 6.
(2) Examination personnel are qualified for detection and length
sizing when the results of any combined performance demonstration
satisfy the acceptance criteria of Supplement 4 to Appendix VIII.
(3) Examination personnel are qualified for depth sizing when
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws
are sized within the respective acceptance criteria of those
supplements.
(G) Specimen set and qualification: seventh provision. When
applying Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII,
or combined Supplement 4 and Supplement 6 qualification, the following
additional conditions must be used, and examination coverage must
include:
(1) The clad-to-base-metal-interface, including a minimum of 15
percent T (measured from the clad-to-base-metal-interface), must be
examined from four orthogonal directions using procedures and personnel
qualified in accordance with Supplement 4 to Appendix VIII.
(2) If the clad-to-base-metal-interface procedure demonstrates
detectability of flaws with a tilt angle relative to the weld
centerline of at least 45 degrees, the remainder of the examination
volume is considered fully examined if coverage is obtained in one
parallel and one perpendicular direction. This must be accomplished
using a procedure and personnel qualified for single-side examination
in accordance with Supplement 6. Subsequent examinations of this volume
may be performed using examination techniques qualified for a tilt
angle of at least 10 degrees.
(3) The examination volume not addressed by paragraph
(b)(2)(xv)(G)(1) of this section is considered fully examined if
coverage is obtained in one parallel and one perpendicular direction,
using a procedure and personnel qualified for single sided examination
when the conditions in paragraph (b)(2)(xv)(G)(2) are met.
(H) Specimen set and qualification: eighth provision. When applying
Supplement 5 to Appendix VIII, at least 50 percent of the flaws in the
demonstration test set must be cracks and the maximum misorientation
must be demonstrated with cracks. Flaws in nozzles with bore diameters
equal to or less than 4 inches may be notches.
(I) Specimen set and qualification: ninth provision. When applying
Supplement 5, Paragraph (a), to Appendix VIII, the number of false
calls allowed must be D/10, with a maximum of 3, where D is the
diameter of the nozzle.
(J) [Reserved]
(K) Specimen set and qualification: eleventh provision. When
performing nozzle-to-vessel weld examinations, the following conditions
must be used when the requirements contained in Supplement 7 to
Appendix VIII are applied for nozzle-to-vessel welds in conjunction
with Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or
combined Supplement 4 and Supplement 6 qualification.
(1) For examination of nozzle-to-vessel welds conducted from the
bore, the following conditions are required to qualify the procedures,
equipment, and personnel:
(i) For detection, a minimum of four flaws in one or more full-
scale nozzle mock-ups must be added to the test set. The specimens must
comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for
flaw locations specified in Table VIII S6-1. Flaws may be notches,
fabrication flaws, or cracks. Seventy-five (75) percent of the flaws
must be cracks or fabrication flaws. Flaw locations and orientations
must be selected from the choices shown in paragraph (b)(2)(xv)(K)(4)
of this section, Table VIII-S7-1--Modified, with the exception that
flaws in the outer eighty-five (85) percent of the weld need not be
perpendicular to the weld. There may be no more than two flaws from
each category, and at least one subsurface flaw must be included.
(ii) For length sizing, a minimum of four flaws as in paragraph
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set.
The length sizing results must be added to the results of combined
Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The
combined results must meet the acceptance standards contained in
paragraph (b)(2)(xv)(E)(3) of this section.
(iii) For depth sizing, a minimum of four flaws as in paragraph
(b)(2)(xv)(K)(1)(i) of this section must be included in the test set.
Their depths must be distributed over the ranges of Supplement 4,
Paragraph 1.1, to Appendix VIII, for the inner 15 percent of the wall
thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the
remainder of the wall thickness. The depth sizing results must be
combined with the sizing results from Supplement 4 to Appendix VIII for
the inner 15 percent and to Supplement 6 to Appendix VIII for the
remainder of the wall thickness. The combined results must meet the
depth sizing acceptance criteria contained in paragraphs
(b)(2)(xv)(C)(1), (b)(2)(xv)(E)(1), and (b)(2)(xv)(F)(3) of this
section.
(2) For examination of reactor pressure vessel nozzle-to-vessel
welds conducted from the inside of the vessel, the following conditions
are required:
(i) The clad-to-base-metal-interface and the adjacent examination
volume to a minimum depth of 15 percent T (measured from the clad-to-
base-metal-interface) must be examined from four orthogonal directions
using a procedure and personnel qualified in accordance with Supplement
4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and
(b)(2)(xv)(C) of this section.
(ii) When the examination volume defined in paragraph
(b)(2)(xv)(K)(2)(i)
[[Page 37911]]
of this section cannot be effectively examined in all four directions,
the examination must be augmented by examination from the nozzle bore
using a procedure and personnel qualified in accordance with paragraph
(b)(2)(xv)(K)(1) of this section.
(iii) The remainder of the examination volume not covered by
paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of
paragraphs (b)(2)(xv)(K)(2)(i) and (b)(2)(xv)(K)(2)(ii) of this
section, must be examined from the nozzle bore using a procedure and
personnel qualified in accordance with paragraph (b)(2)(xv)(K)(1) of
this section, or from the vessel shell using a procedure and personnel
qualified for single sided examination in accordance with Supplement 6
to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through
(b)(2)(xv)(G) of this section.
(3) For examination of reactor pressure vessel nozzle-to-shell
welds conducted from the outside of the vessel, the following
conditions are required:
(i) The clad-to-base-metal-interface and the adjacent metal to a
depth of 15 percent T (measured from the clad-to-base-metal-interface)
must be examined from one radial and two opposing circumferential
directions using a procedure and personnel qualified in accordance with
Supplement 4 to Appendix VIII, as conditioned by paragraphs
(b)(2)(xv)(B) and (b)(2)(xv)(C) of this section, for examinations
performed in the radial direction, and Supplement 5 to Appendix VIII,
as conditioned by paragraph (b)(2)(xv)(J) of this section, for
examinations performed in the circumferential direction.
(ii) The examination volume not addressed by paragraph
(b)(2)(xv)(K)(3)(i) of this section must be examined in a minimum of
one radial direction using a procedure and personnel qualified for
single sided examination in accordance with Supplement 6 to Appendix
VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (b)(2)(xv)(G)
of this section.
(4) Table VIII-S7-1, ``Flaw Locations and Orientations,''
Supplement 7 to Appendix VIII, is conditioned as follows:
Table VIII-S7-1--Modified
------------------------------------------------------------------------
Flaw locations and orientations
-------------------------------------------------------------------------
Parallel Perpendicular to
to weld weld
------------------------------------------------------------------------
Inner 15 percent.......................... X X
Outside Diameter Surface.................. X ................
Subsurface................................ X ................
------------------------------------------------------------------------
(L) Specimen set and qualification: twelfth provision. As a
condition to the requirements of Supplement 8, Subparagraph 1.1(c), to
Appendix VIII, notches may be located within one diameter of each end
of the bolt or stud.
(M) Specimen set and qualification: thirteenth provision. When
implementing Supplement 12 to Appendix VIII, only the provisions
related to the coordinated implementation of Supplement 3 to Supplement
2 performance demonstrations are to be applied.
(xvi) Section XI condition: Appendix VIII single side ferritic
vessel and piping and stainless steel piping examinations. When
applying editions and addenda prior to the 2007 Edition of Section XI,
the following conditions apply.
(A) Ferritic and stainless steel piping examinations: first
provision. Examinations performed from one side of a ferritic vessel
weld must be conducted with equipment, procedures, and personnel that
have demonstrated proficiency with single side examinations. To
demonstrate equivalency to two sided examinations, the demonstration
must be performed to the requirements of Appendix VIII, as conditioned
by this paragraph and paragraphs (b)(2)(xv)(B) through (b)(2)(xv)(G) of
this section, on specimens containing flaws with non-optimum sound
energy reflecting characteristics or flaws similar to those in the
vessel being examined.
(B) Ferritic and stainless steel piping examinations: second
provision. Examinations performed from one side of a ferritic or
stainless steel pipe weld must be conducted with equipment, procedures,
and personnel that have demonstrated proficiency with single side
examinations. To demonstrate equivalency to two sided examinations, the
demonstration must be performed to the requirements of Appendix VIII,
as conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this
section.
(xvii) Section XI condition: Reconciliation of quality
requirements. When purchasing replacement items, in addition to the
reconciliation provisions of IWA-4200, 1995 Addenda through 1998
Edition, the replacement items must be purchased, to the extent
necessary, in accordance with the licensee's quality assurance program
description required by 10 CFR 50.34(b)(6)(ii).
(xviii) Section XI condition: NDE personnel certification.
(A) NDE personnel certification: first provision. Level I and II
nondestructive examination personnel must be recertified on a 3-year
interval in lieu of the 5-year interval specified in the 1997 Addenda
and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the
1999 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section.
(B) NDE personnel certification: second provision. When applying
editions and addenda prior to the 2007 Edition of Section XI, paragraph
IWA-2316 may only be used to qualify personnel that observe leakage
during system leakage and hydrostatic tests conducted in accordance
with IWA 5211(a) and (b).
(C) NDE personnel certification: third provision. When applying
editions and addenda prior to the 2005 Addenda of Section XI,
licensee's qualifying visual examination personnel for VT-3 visual
examination under paragraph IWA-2317 of Section XI must demonstrate the
proficiency of the training by administering an initial qualification
examination and administering subsequent examinations on a 3-year
interval.
(xix) Section XI condition: Substitution of alternative methods.
The provisions for substituting alternative examination methods, a
combination of methods, or newly developed techniques in the 1997
Addenda of IWA-2240 must be applied when using the 1998 Edition through
the 2004 Edition of Section XI of the ASME BPV Code. The provisions in
IWA-4520(c), 1997 Addenda through the 2004 Edition, allowing the
substitution of alternative methods, a combination of methods, or newly
developed techniques for the methods specified in the Construction
Code, are not approved for use. The provisions in IWA-4520(b)(2) and
IWA-4521 of the 2008 Addenda through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section,
allowing the substitution of ultrasonic examination for radiographic
examination specified in the Construction Code, are not approved for
use.
(xx) Section XI condition: System leakage tests.
(A) System leakage tests: first provision. When performing system
leakage tests in accordance with IWA-5213(a), 1997 through 2002
Addenda,
[[Page 37912]]
the licensee must maintain a 10-minute hold time after test pressure
has been reached for Class 2 and Class 3 components that are not in use
during normal operating conditions. No hold time is required for the
remaining Class 2 and Class 3 components provided that the system has
been in operation for at least 4 hours for insulated components or 10
minutes for uninsulated components.
(B) System leakage tests: second provision. The NDE provision in
IWA-4540(a)(2) of the 2002 Addenda of Section XI must be applied when
performing system leakage tests after repair and replacement activities
performed by welding or brazing on a pressure retaining boundary using
the 2003 Addenda through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section.
(xxi) Section XI condition: Table IWB-2500-1 examination
requirements.
(A) Table IWB-2500-1 examination requirements: first provision. The
provisions of Table IWB-2500-1, Examination Category B-D, Full
Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60
(Inspection Program A) and Items B3.120 and B3.140 (Inspection Program
B) of the 1998 Edition must be applied when using the 1999 Addenda
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(ii) of this section. A visual examination with
magnification that has a resolution sensitivity to detect a 1-mil width
wire or crack, utilizing the allowable flaw length criteria in Table
IWB-3512-1, 1997 Addenda through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section, with
a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may
be performed instead of an ultrasonic examination.
(B) [Reserved]
(xxii) Section XI condition: Surface examination. The use of the
provision in IWA-2220, ``Surface Examination,'' of Section XI, 2001
Edition through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section, that allows use of
an ultrasonic examination method is prohibited.
(xxiii) Section XI condition: Evaluation of thermally cut surfaces.
The use of the provisions for eliminating mechanical processing of
thermally cut surfaces in IWA-4461.4.2 of Section XI, 2001 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(ii) of this section, is prohibited.
(xxiv) Section XI condition: Incorporation of the performance
demonstration initiative and addition of ultrasonic examination
criteria. The use of Appendix VIII and the supplements to Appendix VIII
and Article I-3000 of Section XI of the ASME BPV Code, 2002 Addenda
through the 2006 Addenda, is prohibited.
(xxv) Section XI condition: Mitigation of defects by modification.
The use of the provisions in IWA-4340, ``Mitigation of Defects by
Modification,'' Section XI, 2001 Edition through the latest edition and
addenda incorporated by reference in paragraph (a)(1)(ii) of this
section are prohibited.
(xxvi) Section XI condition: Pressure testing Class 1, 2 and 3
mechanical joints. The repair and replacement activity provisions in
IWA-4540(c) of the 1998 Edition of Section XI for pressure testing
Class 1, 2, and 3 mechanical joints must be applied when using the 2001
Edition through the latest edition and addenda incorporated by
reference in paragraph (a)(1)(ii) of this section.
(xxvii) Section XI condition: Removal of insulation. When
performing visual examination in accordance with IWA-5242 of Section XI
of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-
5241 of the 2007 Edition through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section,
insulation must be removed from 17-4 PH or 410 stainless steel studs or
bolts aged at a temperature below 1100[emsp14][deg]F or having a
Rockwell Method C hardness value above 30, and from A-286 stainless
steel studs or bolts preloaded to 100,000 pounds per square inch or
higher.
(xxviii) Section XI condition: Analysis of flaws. Licensees using
ASME BPV Code, Section XI, Appendix A, must use the following
conditions when implementing Equation (2) in A-4300(b)(1):
For R < 0, [Delta]KI depends on the crack depth (a), and
the flow stress ([sigma]f). The flow stress is defined by
[sigma]f = 1/2([sigma]ys+ [sigma]ult),
where [sigma]ys is the yield strength and
[sigma]ult is the ultimate tensile strength in units ksi
(MPa) and (a) is in units in. (mm). For -2 <= R <= 0 and
Kmax- Kmin <= 0.8 x 1.12
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI =
Kmax. For R < -2 and Kmax- Kmin <= 0.8
x 1.12 [sigma]f[radic]([pi]a), S = 1 and
[Delta]KI= (1 - R) Kmax/3. For R < 0 and
Kmax - Kmin > 0.8 x 1.12
[sigma]f[radic]([pi]a), S = 1 and [Delta]KI=
Kmax- Kmin.
(xxix) Section XI condition: Nonmandatory Appendix R. Nonmandatory
Appendix R, ``Risk-Informed Inspection Requirements for Piping,'' of
Section XI, 2005 Addenda through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(ii) of this section, may
not be implemented without prior NRC authorization of the proposed
alternative in accordance with paragraph (z) of this section.
(3) Conditions on ASME OM Code. As used in this section, references
to the OM Code refer to the ASME Code for Operation and Maintenance of
Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory
Appendices I and II, and Nonmandatory Appendices A through H and J,
including the 1995 Edition through the 2006 Addenda, subject to the
following conditions:
(i) OM condition: Quality assurance. When applying editions and
addenda of the OM Code, the requirements of NQA-1, ``Quality Assurance
Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as
permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-
1500 of the 1998 Edition through the latest edition and addenda
incorporated by reference in paragraph (a)(1)(iv) of this section,
provided the licensee uses its 10 CFR Part 50, Appendix B, quality
assurance program in conjunction with the OM Code requirements.
Commitments contained in the licensee's quality assurance program
description that are more stringent than those contained in NQA-1
govern OM Code activities. If NQA-1 and the OM Code do not address the
commitments contained in the licensee's Appendix B quality assurance
program description, the commitments must be applied to OM Code
activities.
(ii) OM condition: Motor-Operated Valve (MOV) testing. Licensees
must comply with the provisions for MOV testing in OM Code ISTC 4.2,
1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition
through the latest edition and addenda incorporated by reference in
paragraph (a)(1)(iv) of this section, and must establish a program to
ensure that motor-operated valves continue to be capable of performing
their design basis safety functions.
(iii) [Reserved]
(iv) OM condition: Check valves (Appendix II). Licensees applying
Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM
Code, 1995 Edition with the 1996 and 1997 Addenda, must satisfy the
requirements of paragraphs (b)(3)(iv)(A), (b)(3)(iv)(B), and
(b)(3)(iv)(C) of this section. Licensees applying Appendix II, 1998
Edition through the 2002 Addenda, must satisfy the requirements of
paragraphs (b)(3)(iv)(A), (b)(3)(iv)(B), and (b)(3)(iv)(D) of this
section.
(A) Check valves: first provision. Valve opening and closing
functions
[[Page 37913]]
must be demonstrated when flow testing or examination methods
(nonintrusive, or disassembly and inspection) are used;
(B) Check valves: second provision. The initial interval for tests
and associated examinations may not exceed two fuel cycles or 3 years,
whichever is longer; any extension of this interval may not exceed one
fuel cycle per extension with the maximum interval not to exceed 10
years. Trending and evaluation of existing data must be used to reduce
or extend the time interval between tests.
(C) Check valves: third provision. If the Appendix II condition
monitoring program is discontinued, then the requirements of ISTC 4.5.1
through 4.5.4 must be implemented.
(D) Check valves: fourth provision. The applicable provisions of
subsection ISTC must be implemented if the Appendix II condition
monitoring program is discontinued.
(v) OM condition: Snubbers ISTD. Article IWF-5000, ``Inservice
Inspection Requirements for Snubbers,'' of the ASME BPV Code, Section
XI, must be used when performing inservice inspection examinations and
tests of snubbers at nuclear power plants, except as conditioned in
paragraphs (b)(3)(v)(A) and (b)(3)(v)(B) of this section.
(A) Snubbers: first provision. Licensees may use Subsection ISTD,
``Preservice and Inservice Examination and Testing of Dynamic
Restraints (Snubbers) in Light-Water Reactor Power Plants,'' ASME OM
Code, 1995 Edition through the latest edition and addenda incorporated
by reference in paragraph (a)(1)(iv) of this section, in place of the
requirements for snubbers in the editions and addenda up to the 2005
Addenda of the ASME BPV Code, Section XI, IWF-5200(a) and (b) and IWF-
5300(a) and (b), by making appropriate changes to their technical
specifications or licensee-controlled documents. Preservice and
inservice examinations must be performed using the VT-3 visual
examination method described in IWA-2213.
(B) Snubbers: second provision. Licensees must comply with the
provisions for examining and testing snubbers in Subsection ISTD of the
ASME OM Code and make appropriate changes to their technical
specifications or licensee-controlled documents when using the 2006
Addenda and later editions and addenda of Section XI of the ASME BPV
Code.
(vi) OM condition: Exercise interval for manual valves. Manual
valves must be exercised on a 2-year interval rather than the 5-year
interval specified in paragraph ISTC-3540 of the 1999 through the 2005
Addenda of the ASME OM Code, provided that adverse conditions do not
require more frequent testing.
(4) Conditions on Design, Fabrication, and Materials Code Cases.
Each manufacturing license, standard design approval, and design
certification application under Part 52 of this chapter is subject to
the following conditions. Licensees may apply the ASME BPV Code Cases
listed in NRC Regulatory Guide 1.84, Revision 36, without prior NRC
approval, subject to the following conditions:
(i) Design, Fabrication, and Materials Code Case condition:
Applying Code Cases. When an applicant or licensee initially applies a
listed Code Case, the applicant or licensee must apply the most recent
version of that Code Case incorporated by reference in paragraph (a) of
this section.
(ii) Design, Fabrication, and Materials Code Case condition:
Applying different revisions of Code Cases. If an applicant or licensee
has previously applied a Code Case and a later version of the Code Case
is incorporated by reference in paragraph (a) of this section, the
applicant or licensee may continue to apply the previous version of the
Code Case as authorized or may apply the later version of the Code
Case, including any NRC-specified conditions placed on its use, until
it updates its Code of Record for the component being constructed.
(iii) Design, Fabrication, and Materials Code Case condition:
Applying annulled Code Cases. Application of an annulled Code Case is
prohibited unless an applicant or licensee applied the listed Code Case
prior to it being listed as annulled in Regulatory Guide 1.84. If an
applicant or licensee has applied a listed Code Case that is later
listed as annulled in Regulatory Guide 1.84, the applicant or licensee
may continue to apply the Code Case until it updates its Code of Record
for the component being constructed.
(5) Conditions on inservice inspection Code Cases. Licensees may
apply the ASME BPV Code Cases listed in Regulatory Guide 1.147,
Revision 17, without prior NRC approval, subject to the following:
(i) ISI Code Case condition: Applying Code Cases. When a licensee
initially applies a listed Code Case, the licensee must apply the most
recent version of that Code Case incorporated by reference in paragraph
(a) of this section.
(ii) ISI Code Case condition: Applying different revisions of Code
Cases. If a licensee has previously applied a Code Case and a later
version of the Code Case is incorporated by reference in paragraph (a)
of this section, the licensee may continue to apply, to the end of the
current 120-month interval, the previous version of the Code Case, as
authorized, or may apply the later version of the Code Case, including
any NRC-specified conditions placed on its use. Licensees who choose to
continue use of the Code Case during subsequent 120-month ISI program
intervals will be required to implement the latest version incorporated
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of
Regulatory Guide 1.147, Revision 17.
(iii) ISI Code Case condition: Applying annulled Code Cases.
Application of an annulled Code Case is prohibited unless a licensee
previously applied the listed Code Case prior to it being listed as
annulled in Regulatory Guide 1.147. If a licensee has applied a listed
Code Case that is later listed as annulled in Regulatory Guide 1.147,
the licensee may continue to apply the Code Case to the end of the
current 120-month interval.
(6) Conditions on Operation and Maintenance of Nuclear Power Plants
Code Cases. Licensees may apply the ASME Operation and Maintenance Code
Cases listed in Regulatory Guide 1.192, Revision 1, without prior NRC
approval, subject to the following:
(i) OM Code Case condition: Applying Code Cases. When a licensee
initially applies a listed Code Case, the licensee must apply the most
recent version of that Code Case incorporated by reference in paragraph
(a) of this section.
(ii) OM Code Case condition: Applying different revisions of Code
Cases. If a licensee has previously applied a Code Case and a later
version of the Code Case is incorporated by reference in paragraph (a)
of this section, the licensee may continue to apply, to the end of the
current 120-month interval, the previous version of the Code Case, as
authorized, or may apply the later version of the Code Case, including
any NRC-specified conditions placed on its use. Licensees who choose to
continue use of the Code Case during subsequent 120-month ISI program
intervals will be required to implement the latest version incorporated
by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of
Regulatory Guide 1.192, Revision 1.
(iii) OM Code Case condition: Applying annulled Code Cases.
Application of an annulled Code Case is prohibited unless a licensee
previously applied the listed Code Case prior to it being listed as
annulled in Regulatory Guide 1.192. If a licensee has applied a listed
Code Case that is later listed as
[[Page 37914]]
annulled in Regulatory Guide 1.192, the licensee may continue to apply
the Code Case to the end of the current 120-month interval.
(c) Reactor coolant pressure boundary. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code as specified in this paragraph.
Each manufacturing license, standard design approval, and design
certification application under Part 52 of this chapter and each
combined license for a utilization facility is subject to the following
conditions:
(1) Standards requirement for reactor coolant pressure boundary
components. Components that are part of the reactor coolant pressure
boundary must meet the requirements for Class 1 components in Section
III\4,5\ of the ASME BPV Code, except as provided in paragraphs (c)(2),
(c)(3), and (c)(4) of this section.
(2) Exceptions to reactor coolant pressure boundary standards
requirement. Components that are connected to the reactor coolant
system and are part of the reactor coolant pressure boundary as defined
in Sec. 50.2 need not meet the requirements of paragraph (c)(1) of
this section, provided that:
(i) Exceptions: Shutdown and cooling capability. In the event of
postulated failure of the component during normal reactor operation,
the reactor can be shut down and cooled down in an orderly manner,
assuming makeup is provided by the reactor coolant makeup system; or
(ii) Exceptions: Isolation capability. The component is or can be
isolated from the reactor coolant system by two valves in series (both
closed, both open, or one closed and the other open). Each open valve
must be capable of automatic actuation and, assuming the other valve is
open, its closure time must be such that, in the event of postulated
failure of the component during normal reactor operation, each valve
remains operable and the reactor can be shut down and cooled down in an
orderly manner, assuming makeup is provided by the reactor coolant
makeup system only.
(3) Applicable Code and Code Cases and conditions on their use. The
Code edition, addenda, and optional ASME Code Cases to be applied to
components of the reactor coolant pressure boundary must be determined
by the provisions of paragraph NCA-1140, Subsection NCA of Section III
of the ASME BPV Code, subject to the following conditions:
(i) Reactor coolant pressure boundary condition: Code edition and
addenda. The edition and addenda applied to a component must be those
that are incorporated by reference in paragraph (a)(1)(i) of this
section;
(ii) Reactor coolant pressure boundary condition: Earliest edition
and addenda for pressure vessel. The ASME Code provisions applied to
the pressure vessel may be dated no earlier than the summer 1972
Addenda of the 1971 Edition;
(iii) Reactor coolant pressure boundary condition: Earliest edition
and addenda for piping, pumps, and valves. The ASME Code provisions
applied to piping, pumps, and valves may be dated no earlier than the
Winter 1972 Addenda of the 1971 Edition; and
(iv) Reactor coolant pressure boundary condition: Use of Code
Cases. The optional Code Cases applied to a component must be those
listed in NRC Regulatory Guide 1.84 that is incorporated by reference
in paragraph (a)(3)(i) of this section.
(4) Standards requirement for components in older plants. For a
nuclear power plant whose construction permit was issued prior to May
14, 1984, the applicable Code edition and addenda for a component of
the reactor coolant pressure boundary continue to be that Code edition
and addenda that were required by Commission regulations for such a
component at the time of issuance of the construction permit.
(d) Quality Group B components. Systems and components of boiling
and pressurized water-cooled nuclear power reactors must meet the
requirements of the ASME BPV Code as specified in this paragraph. Each
manufacturing license, standard design approval, and design
certification application under Part 52 of this chapter, and each
combined license for a utilization facility is subject to the following
conditions:
(1) Standards requirement for Quality Group B components. For a
nuclear power plant whose application for a construction permit under
this part, or a combined license or manufacturing license under Part 52
of this chapter, docketed after May 14, 1984, or for an application for
a standard design approval or a standard design certification docketed
after May 14, 1984, components classified Quality Group B\9\ must meet
the requirements for Class 2 Components in Section III of the ASME BPV
Code.
(2) Quality Group B: Applicable Code and Code Cases and conditions
on their use. The Code edition, addenda, and optional ASME Code Cases
to be applied to the systems and components identified in paragraph
(d)(1) of this section must be determined by the rules of paragraph
NCA-1140, Subsection NCA of Section III of the ASME BPV Code, subject
to the following conditions:
(i) Quality Group B condition: Code edition and addenda. The
edition and addenda must be those that are incorporated by reference in
paragraph (a)(1)(i) of this section;
(ii) Quality Group B condition: Earliest edition and addenda for
components. The ASME Code provisions applied to the systems and
components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group B condition: Use of Code Cases. The optional
Code Cases must be those listed in NRC Regulatory Guide 1.84 that is
incorporated by reference in paragraph (a)(3)(i) of this section.
(e) Quality Group C components. Systems and components of boiling
and pressurized water-cooled nuclear power reactors must meet the
requirements of the ASME BPV Code as specified in this paragraph. Each
manufacturing license, standard design approval, and design
certification application under Part 52 of this chapter and each
combined license for a utilization facility is subject to the following
conditions.
(1) Standards requirement for Quality Group C components. For a
nuclear power plant whose application for a construction permit under
this part, or a combined license or manufacturing license under Part 52
of this chapter, docketed after May 14, 1984, or for an application for
a standard design approval or a standard design certification docketed
after May 14, 1984, components classified Quality Group C\9\ must meet
the requirements for Class 3 components in Section III of the ASME BPV
Code.
(2) Quality Group C applicable Code and Code Cases and conditions
on their use. The Code edition, addenda, and optional ASME Code Cases
to be applied to the systems and components identified in paragraph
(e)(1) of this section must be determined by the rules of paragraph
NCA-1140, subsection NCA of Section III of the ASME BPV Code, subject
to the following conditions:
(i) Quality Group C condition: Code edition and addenda. The
edition and addenda must be those incorporated by reference in
paragraph (a)(1)(i) of this section;
(ii) Quality Group C condition: Earliest edition and addenda for
components. The ASME Code provisions applied to the systems and
components may be dated no earlier than the 1980 Edition; and
(iii) Quality Group C condition: Use of Code Cases. The optional
Code Cases
[[Page 37915]]
must be those listed in NRC Regulatory Guide 1.84 that is incorporated
by reference in paragraph (a)(3)(i) of this section.
(f) Inservice testing requirements. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code and ASME Code for Operation and
Maintenance of Nuclear Power Plants as specified in this paragraph.
Each operating license for a boiling or pressurized water-cooled
nuclear facility is subject to the following conditions. Each combined
license for a boiling or pressurized water-cooled nuclear facility is
subject to the following conditions, but the conditions in paragraphs
(f)(4), (f)(5), and (f)(6) of this section must be met only after the
Commission makes the finding under Sec. 52.103(g) of this chapter.
Requirements for inservice inspection of Class 1, Class 2, Class 3,
Class MC, and Class CC components (including their supports) are
located in Sec. 50.55a(g).
(1) Inservice testing requirements for older plants (pre-1971 CPs).
For a boiling or pressurized water-cooled nuclear power facility whose
construction permit was issued prior to January 1, 1971, pumps and
valves must meet the test requirements of paragraphs (f)(4) and (f)(5)
of this section to the extent practical. Pumps and valves that are part
of the reactor coolant pressure boundary must meet the requirements
applicable to components that are classified as ASME Code Class 1.
Other pumps and valves that perform a function to shut down the reactor
or maintain the reactor in a safe shutdown condition, mitigate the
consequences of an accident, or provide overpressure protection for
safety-related systems (in meeting the requirements of the 1986
Edition, or later, of the BPV or OM Code) must meet the test
requirements applicable to components that are classified as ASME Code
Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice
testing in plants with CPs issued between 1971 and 1974. For a boiling
or pressurized water-cooled nuclear power facility whose construction
permit was issued on or after January 1, 1971, but before July 1, 1974,
pumps and valves that are classified as ASME Code Class 1 and Class 2
must be designed and provided with access to enable the performance of
inservice tests for operational readiness set forth in editions and
addenda of Section XI of the ASME BPV incorporated by reference in
paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases
listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide
1.192, Revision 1, that are incorporated by reference in paragraphs
(a)(3)(ii) and (a)(3)(iii) of this section, respectively) in effect 6
months before the date of issuance of the construction permit. The
pumps and valves may meet the inservice test requirements set forth in
subsequent editions of this Code and addenda that are incorporated by
reference in paragraph (a)(1)(ii) of this section (or the optional ASME
Code Cases listed in NRC Regulatory Guide 1.147, Revision 17; or
Regulatory Guide 1.192, Revision 1, that are incorporated by reference
in paragraphs (a)(3)(ii) and (a)(3)(iii) of this section,
respectively), subject to the applicable conditions listed therein.
(3) Design and accessibility requirements for performing inservice
testing in plants with CPs issued after 1974. For a boiling or
pressurized water-cooled nuclear power facility whose construction
permit under this part or design approval, design certification,
combined license, or manufacturing license under Part 52 of this
chapter was issued on or after July 1, 1974:
(i)-(ii) [Reserved]
(iii) IST design and accessibility requirements: Class 1 pumps and
valves.
(A) Class 1 pumps and valves: first provision. In facilities whose
construction permit was issued before November 22, 1999, pumps and
valves that are classified as ASME Code Class 1 must be designed and
provided with access to enable the performance of inservice testing of
the pumps and valves for assessing operational readiness set forth in
the editions and addenda of Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1)(ii) of this section (or
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147,
Revision 17, or Regulatory Guide 1.192, Revision 1, that are
incorporated by reference in paragraphs (a)(3)(ii) and (a)(3)(iii) of
this section, respectively) applied to the construction of the
particular pump or valve or the summer 1973 Addenda, whichever is
later.
(B) Class1 pumps and valves: second provision. In facilities whose
construction permit under this part, or design certification, design
approval, combined license, or manufacturing license under Part 52 of
this chapter, issued on or after November 22, 1999, pumps and valves
that are classified as ASME Code Class 1 must be designed and provided
with access to enable the performance of inservice testing of the pumps
and valves for assessing operational readiness set forth in editions
and addenda of the ASME OM Code (or the optional ASME Code Cases listed
in NRC Regulatory Guide 1.192, Revision 1, that are incorporated by
reference in paragraph (a)(3)(iii) of this section), incorporated by
reference in paragraph (a)(1)(iv) of this section at the time the
construction permit, combined license, manufacturing license, design
certification, or design approval is issued.
(iv) IST design and accessibility requirements: Class 2 and 3 pumps
and valves.
(A) Class 2 and 3 pumps and valves: first provision. In facilities
whose construction permit was issued before November 22, 1999, pumps
and valves that are classified as ASME Code Class 2 and Class 3 must be
designed and be provided with access to enable the performance of
inservice testing of the pumps and valves for assessing operational
readiness set forth in the editions and addenda of Section XI of the
ASME BPV Code incorporated by reference in paragraph (a)(1)(ii) of this
section (or the optional ASME Code Cases listed in NRC Regulatory Guide
1.147, Revision 17, that are incorporated by reference in paragraph
(a)(3)(ii) of this section) applied to the construction of the
particular pump or valve or the Summer 1973 Addenda, whichever is
later.
(B) Class 2 and 3 pumps and valves: second provision. In facilities
whose construction permit under this part, or design certification,
design approval, combined license, or manufacturing license under Part
52 of this chapter, issued on or after November 22, 1999, pumps and
valves that are classified as ASME Code Class 2 and 3 must be designed
and provided with access to enable the performance of inservice testing
of the pumps and valves for assessing operational readiness set forth
in editions and addenda of the ASME OM Code (or the optional ASME OM
Code Cases listed in NRC Regulatory Guide 1.192, Revision 1, that are
incorporated by reference in paragraph (a)(3)(iii) of this section),
incorporated by reference in paragraph (a)(1)(iv) of this section at
the time the construction permit, combined license, or design
certification is issued.
(v) IST design and accessibility requirements: Meeting later IST
requirements. All pumps and valves may meet the test requirements set
forth in subsequent editions of codes and addenda or portions thereof
that are incorporated by reference in paragraph (a) of this section,
subject to the conditions listed in paragraph (b) of this section.
[[Page 37916]]
(4) Inservice testing standards requirement for operating plants.
Throughout the service life of a boiling or pressurized water-cooled
nuclear power facility, pumps and valves that are classified as ASME
Code Class 1, Class 2, and Class 3 must meet the inservice test
requirements (except design and access provisions) set forth in the
ASME OM Code and addenda that become effective subsequent to editions
and addenda specified in paragraphs (f)(2) and (f)(3) of this section
and that are incorporated by reference in paragraph (a)(1)(iv) of this
section, to the extent practical within the limitations of design,
geometry, and materials of construction of the components.
(i) Applicable IST Code: Initial 120-month interval. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during the initial 120-month
interval must comply with the requirements in the latest edition and
addenda of the OM Code incorporated by reference in paragraph
(a)(1)(iv) of this section on the date 12 months before the date of
issuance of the operating license under this part, or 12 months before
the date scheduled for initial loading of fuel under a combined license
under Part 52 of this chapter (or the optional ASME Code Cases listed
in NRC Regulatory Guide 1.192, Revision 1, that is incorporated by
reference in paragraph (a)(3)(iii) of this section, subject to the
conditions listed in paragraph (b) of this section.
(ii) Applicable IST Code: Successive 120-month intervals. Inservice
tests to verify operational readiness of pumps and valves, whose
function is required for safety, conducted during successive 120-month
intervals must comply with the requirements of the latest edition and
addenda of the OM Code incorporated by reference in paragraph
(a)(1)(iv) of this section 12 months before the start of the 120-month
interval (or the optional ASME Code Cases listed in NRC Regulatory
Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that
are incorporated by reference in paragraphs (a)(3)(ii) and (a)(3)(iii)
of this section, respectively), subject to the conditions listed in
paragraph (b) of this section.
(iii) [Reserved]
(iv) Applicable IST Code: Use of later Code editions and addenda.
Inservice tests of pumps and valves may meet the requirements set forth
in subsequent editions and addenda that are incorporated by reference
in paragraph (a)(1)(iv) of this section, subject to the conditions
listed in paragraph (b) of this section, and subject to NRC approval.
Portions of editions or addenda may be used, provided that all related
requirements of the respective editions or addenda are met.
(5) Requirements for updating IST programs.
(i) IST program update: Applicable IST Code editions and addenda.
The inservice test program for a boiling or pressurized water-cooled
nuclear power facility must be revised by the licensee, as necessary,
to meet the requirements of paragraph (f)(4) of this section.
(ii) IST program update: Conflicting IST Code requirements with
technical specifications. If a revised inservice test program for a
facility conflicts with the technical specifications for the facility,
the licensee must apply to the Commission for amendment of the
technical specifications to conform the technical specifications to the
revised program. The licensee must submit this application, as
specified in Sec. 50.4, at least 6 months before the start of the
period during which the provisions become applicable, as determined by
paragraph (f)(4) of this section.
(iii) IST program update: Notification of impractical IST Code
requirements. If the licensee has determined that conformance with
certain Code requirements is impractical for its facility, the licensee
must notify the Commission and submit, as specified in Sec. 50.4,
information to support the determination.
(iv) IST program update: Schedule for completing impracticality
determinations. Where a pump or valve test requirement by the Code or
addenda is determined to be impractical by the licensee and is not
included in the revised inservice test program (as permitted by
paragraph (f)(4) of this section), the basis for this determination
must be submitted for NRC review and approval not later than 12 months
after the expiration of the initial 120-month interval of operation
from the start of facility commercial operation and each subsequent
120-month interval of operation during which the test is determined to
be impractical.
(6) Actions by the Commission for evaluating impractical and
augmented IST Code requirements.
(i) Impractical IST requirements: Granting of relief. The
Commission will evaluate determinations under paragraph (f)(5) of this
section that code requirements are impractical. The Commission may
grant relief and may impose such alternative requirements as it
determines are authorized by law, will not endanger life or property or
the common defense and security, and are otherwise in the public
interest, giving due consideration to the burden upon the licensee that
could result if the requirements were imposed on the facility.
(ii) Augmented IST requirements. The Commission may require the
licensee to follow an augmented inservice test program for pumps and
valves for which the Commission deems that added assurance of
operational readiness is necessary.
(g) Inservice inspection requirements. Systems and components of
boiling and pressurized water-cooled nuclear power reactors must meet
the requirements of the ASME BPV Code as specified in this paragraph.
Each operating license for a boiling or pressurized water-cooled
nuclear facility is subject to the following conditions. Each combined
license for a boiling or pressurized water-cooled nuclear facility is
subject to the following conditions, but the conditions in paragraphs
(g)(4), (g)(5), and (g)(6) of this section must be met only after the
Commission makes the finding under Sec. 52.103(g) of this chapter.
Requirements for inservice testing of Class 1, Class 2, and Class 3
pumps and valves are located in Sec. 50.55a(f).
(1) Inservice inspection requirements for older plants (pre-1971
CPs). For a boiling or pressurized water-cooled nuclear power facility
whose construction permit was issued before January 1, 1971, components
(including supports) must meet the requirements of paragraphs (g)(4)
and (g)(5) of this section to the extent practical. Components that are
part of the reactor coolant pressure boundary and their supports must
meet the requirements applicable to components that are classified as
ASME Code Class 1. Other safety-related pressure vessels, piping, pumps
and valves, and their supports must meet the requirements applicable to
components that are classified as ASME Code Class 2 or Class 3.
(2) Design and accessibility requirements for performing inservice
inspection in plants with CPs issued between 1971 and 1974. For a
boiling or pressurized water-cooled nuclear power facility whose
construction permit was issued on or after January 1, 1971, but before
July 1, 1974, components (including supports) that are classified as
ASME Code Class 1 and Class 2 must be designed and be provided with
access to enable the performance of inservice examination of such
components (including supports) and must meet the preservice
examination requirements set forth in editions and addenda of Section
III or Section XI of the ASME BPV Code incorporated by reference in
paragraph (a)(1) of this section (or the optional ASME Code
[[Page 37917]]
Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are
incorporated by reference in paragraph (a)(3)(ii) of this section) in
effect 6 months before the date of issuance of the construction permit.
The components (including supports) may meet the requirements set forth
in subsequent editions and addenda of this Code that are incorporated
by reference in paragraph (a) of this section (or the optional ASME
Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are
incorporated by reference in paragraph (a)(3)(ii) of this section),
subject to the applicable limitations and modifications.
(3) Design and accessibility requirements for performing inservice
inspection in plants with CPs issued after 1974. For a boiling or
pressurized water-cooled nuclear power facility, whose construction
permit under this part, or design certification, design approval,
combined license, or manufacturing license under Part 52 of this
chapter, was issued on or after July 1, 1974, the following are
required:
(i) ISI design and accessibility requirements: Class 1 components
and supports. Components (including supports) that are classified as
ASME Code Class 1 must be designed and be provided with access to
enable the performance of inservice examination of these components and
must meet the preservice examination requirements set forth in the
editions and addenda of Section III or Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1) of this section (or the
optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision
17, that are incorporated by reference in paragraph (a)(3)(ii) of this
section) applied to the construction of the particular component.
(ii) ISI design and accessibility requirements: Class 2 and 3
components and supports. Components that are classified as ASME Code
Class 2 and Class 3 and supports for components that are classified as
ASME Code Class 1, Class 2, and Class 3 must be designed and provided
with access to enable the performance of inservice examination of these
components and must meet the preservice examination requirements set
forth in the editions and addenda of Section XI of the ASME BPV Code
incorporated by reference in paragraph (a)(1)(ii) of this section (or
the optional ASME Code Cases listed in NRC Regulatory Guide 1.147,
Revision 17, that are incorporated by reference in paragraph (a)(3)(ii)
of this section) applied to the construction of the particular
component.
(iii)-(iv) [Reserved]
(v) ISI design and accessibility requirements: Meeting later ISI
requirements. All components (including supports) may meet the
requirements set forth in subsequent editions of codes and addenda or
portions thereof that are incorporated by reference in paragraph (a) of
this section, subject to the conditions listed therein.
(4) Inservice inspection standards requirement for operating
plants. Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) that are
classified as ASME Code Class 1, Class 2, and Class 3 must meet the
requirements, except design and access provisions and preservice
examination requirements, set forth in Section XI of editions and
addenda of the ASME BPV Code (or ASME OM Code for snubber examination
and testing) that become effective subsequent to editions specified in
paragraphs (g)(2) and (g)(3) of this section and that are incorporated
by reference in paragraph (a)(1)(ii) or (a)(1)(iv) for snubber
examination and testing of this section, to the extent practical within
the limitations of design, geometry, and materials of construction of
the components. Components that are classified as Class MC pressure
retaining components and their integral attachments, and components
that are classified as Class CC pressure retaining components and their
integral attachments, must meet the requirements, except design and
access provisions and preservice examination requirements, set forth in
Section XI of the ASME BPV Code and addenda that are incorporated by
reference in paragraph (a)(1)(ii) of this section, subject to the
condition listed in paragraph (b)(2)(vi) of this section and the
conditions listed in paragraphs (b)(2)(viii) and (b)(2)(ix) of this
section, to the extent practical within the limitation of design,
geometry, and materials of construction of the components.
(i) Applicable ISI Code: Initial 120-month interval. Inservice
examination of components and system pressure tests conducted during
the initial 120-month inspection interval must comply with the
requirements in the latest edition and addenda of the Code incorporated
by reference in paragraph (a) of this section on the date 12 months
before the date of issuance of the operating license under this part,
or 12 months before the date scheduled for initial loading of fuel
under a combined license under Part 52 of this chapter (or the optional
ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when
using Section XI, or Regulatory Guide 1.192, Revision 1, when using the
OM Code, that are incorporated by reference in paragraphs (a)(3)(ii)
and (a)(3)(iii) of this section, respectively), subject to the
conditions listed in paragraph (b) of this section.
(ii) Applicable ISI Code: Successive 120-month intervals. Inservice
examination of components and system pressure tests conducted during
successive 120-month inspection intervals must comply with the
requirements of the latest edition and addenda of the Code incorporated
by reference in paragraph (a) of this section 12 months before the
start of the 120-month inspection interval (or the optional ASME Code
Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using
Section XI, or Regulatory Guide 1.192, Revision 1, when using the OM
Code, that are incorporated by reference in paragraphs (a)(3)(ii) and
(a)(3)(iii) of this section), subject to the conditions listed in
paragraph (b) of this section. However, a licensee whose inservice
inspection interval commences during the 12 through 18-month period
after July 21, 2011, may delay the update of their Appendix VIII
program by up to 18 months after July 21, 2011.
(iii) Applicable ISI Code: Optional surface examination
requirement. When applying editions and addenda prior to the 2003
Addenda of Section XI of the ASME BPV Code, licensees may, but are not
required to, perform the surface examinations of high-pressure safety
injection systems specified in Table IWB-2500-1, Examination Category
B-J, Item Numbers B9.20, B9.21, and B9.22.
(iv) Applicable ISI Code: Use of subsequent Code editions and
addenda. Inservice examination of components and system pressure tests
may meet the requirements set forth in subsequent editions and addenda
that are incorporated by reference in paragraph (a) of this section,
subject to the conditions listed in paragraph (b) of this section, and
subject to Commission approval. Portions of editions or addenda may be
used, provided that all related requirements of the respective editions
or addenda are met.
(v) Applicable ISI Code: Metal and concrete containments. For a
boiling or pressurized water-cooled nuclear power facility whose
construction permit under this part or combined license under Part 52
of this chapter was issued after January 1, 1956, the following are
required:
(A) Metal and concrete containments: first provision. Metal
containment pressure retaining components and their
[[Page 37918]]
integral attachments must meet the inservice inspection, repair, and
replacement requirements applicable to components that are classified
as ASME Code Class MC;
(B) Metal and concrete containments: second provision. Metallic
shell and penetration liners that are pressure retaining components and
their integral attachments in concrete containments must meet the
inservice inspection, repair, and replacement requirements applicable
to components that are classified as ASME Code Class MC; and
(C) Metal and concrete containments: third provision. Concrete
containment pressure retaining components and their integral
attachments, and the post-tensioning systems of concrete containments,
must meet the inservice inspections, repair, and replacement
requirements applicable to components that are classified as ASME Code
Class CC.
(5) Requirements for updating ISI programs.
(i) ISI program update: Applicable ISI Code editions and addenda.
The inservice inspection program for a boiling or pressurized water-
cooled nuclear power facility must be revised by the licensee, as
necessary, to meet the requirements of paragraph (g)(4) of this
section.
(ii) ISI program update: Conflicting ISI Code requirements with
technical specifications. If a revised inservice inspection program for
a facility conflicts with the technical specifications for the
facility, the licensee must apply to the Commission for amendment of
the technical specifications to conform the technical specifications to
the revised program. The licensee must submit this application, as
specified in Sec. 50.4, at least six months before the start of the
period during which the provisions become applicable, as determined by
paragraph (g)(4) of this section.
(iii) ISI program update: Notification of impractical ISI Code
requirements. If the licensee has determined that conformance with a
Code requirement is impractical for its facility the licensee must
notify the NRC and submit, as specified in Sec. 50.4, information to
support the determinations. Determinations of impracticality in
accordance with this section must be based on the demonstrated
limitations experienced when attempting to comply with the Code
requirements during the inservice inspection interval for which the
request is being submitted. Requests for relief made in accordance with
this section must be submitted to the NRC no later than 12 months after
the expiration of the initial or subsequent 120-month inspection
interval for which relief is sought.
(iv) ISI program update: Schedule for completing impracticality
determinations. Where the licensee determines that an examination
required by Code edition or addenda is impractical, the basis for this
determination must be submitted for NRC review and approval not later
than 12 months after the expiration of the initial or subsequent 120-
month inspection interval for which relief is sought.
(6) Actions by the Commission for evaluating impractical and
augmented ISI Code requirements.
(i) Impractical ISI requirements: Granting of relief. The
Commission will evaluate determinations under paragraph (g)(5) of this
section that code requirements are impractical. The Commission may
grant such relief and may impose such alternative requirements as it
determines are authorized by law, will not endanger life or property or
the common defense and security, and are otherwise in the public
interest giving due consideration to the burden upon the licensee that
could result if the requirements were imposed on the facility.
(ii) Augmented ISI program. The Commission may require the licensee
to follow an augmented inservice inspection program for systems and
components for which the Commission deems that added assurance of
structural reliability is necessary.
(A) [Reserved]
(B) Augmented ISI requirements: Submitting containment ISI
programs. Licensees do not have to submit to the NRC for approval of
their containment inservice inspection programs that were developed to
satisfy the requirements of Subsection IWE and Subsection IWL with
specified conditions. The program elements and the required
documentation must be maintained on site for audit.
(C) Augmented ISI requirements: Implementation of Appendix VIII to
Section XI.
(1) Appendix VIII and the supplements to Appendix VIII to Section
XI, Division 1, 1995 Edition with the 1996 Addenda of the ASME BPV Code
must be implemented in accordance with the following schedule: Appendix
VIII and Supplements 1, 2, 3, and 8--May 22, 2000; Supplements 4 and
6--November 22, 2000; Supplement 11--November 22, 2001; and Supplements
5, 7, and 10--November 22, 2002.
(2) Licensees implementing the 1989 Edition and earlier editions
and addenda of IWA-2232 of Section XI, Division 1, of the ASME BPV Code
must implement the 1995 Edition with the 1996 Addenda of Appendix VIII
and the supplements to Appendix VIII of Section XI, Division 1, of the
ASME BPV Code.
(D) Augmented ISI requirements: Reactor vessel head inspections.
(1) All licensees of pressurized water reactors must augment their
inservice inspection program with ASME Code Case N-729-1, subject to
the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of
this section. Licensees of existing operating reactors as of September
10, 2008, must implement their augmented inservice inspection program
by December 31, 2008. Once a licensee implements this requirement, the
First Revised NRC Order EA-03-009 no longer applies to that licensee
and must be deemed to be withdrawn.
(2) Note 9 of ASME Code Case N-729-1 must not be implemented.
(3) Instead of the specified ``examination method'' requirements
for volumetric and surface examinations in Note 6 of Table 1 of Code
Case N-729-1, the licensee must perform volumetric and/or surface
examination of essentially 100 percent of the required volume or
equivalent surfaces of the nozzle tube, as identified by Figure 2 of
ASME Code Case N-729-1. A demonstrated volumetric or surface leak path
assessment through all J-groove welds must be performed. If a surface
examination is being substituted for a volumetric examination on a
portion of a penetration nozzle that is below the toe of the J-groove
weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface
examination must be of the inside and outside wetted surface of the
penetration nozzle not examined volumetrically.
(4) By September 1, 2009, ultrasonic examinations must be performed
using personnel, procedures, and equipment that have been qualified by
blind demonstration on representative mockups using a methodology that
meets the conditions specified in paragraphs (g)(6)(ii)(D)(4)(i)
through (g)(6)(ii)(D)(4)(iv), instead of the qualification requirements
of Paragraph -2500 of ASME Code Case N-729-1. References herein to
Section XI, Appendix VIII, must be to the 2004 Edition with no addenda
of the ASME BPV Code.
(i) The specimen set must have an applicable thickness
qualification range of +25 percent to -40 percent for nominal depth
through-wall thickness. The specimen set must include geometric and
material conditions that normally require discrimination from
[[Page 37919]]
primary water stress corrosion cracking (PWSCC) flaws.
(ii) The specimen set must have a minimum of ten (10) flaws that
provide an acoustic response similar to PWSCC indications. All flaws
must be greater than 10 percent of the nominal pipe wall thickness. A
minimum of 20 percent of the total flaws must initiate from the inside
surface and 20 percent from the outside surface. At least 20 percent of
the flaws must be in the depth ranges of 10-30 percent through-wall
thickness and at least 20 percent within a depth range of 31-50 percent
through-wall thickness. At least 20 percent and no more than 60 percent
of the flaws must be oriented axially.
(iii) Procedures must identify the equipment and essential
variables and settings used for the qualification, in accordance with
Subarticle VIII-2100 of Section XI, Appendix VIII. The procedure must
be requalified when an essential variable is changed outside the
demonstration range as defined by Subarticle VIII-3130 of Section XI,
Appendix VIII, and as allowed by Articles VIII-4100, VIII-4200, and
VIII-4300 of Section XI, Appendix VIII. Procedure qualification must
include the equivalent of at least three personnel performance
demonstration test sets. Procedure qualification requires at least one
successful personnel performance demonstration.
(iv) Personnel performance demonstration test acceptance criteria
must meet the personnel performance demonstration detection test
acceptance criteria of Table VIII--S10-1 of Section XI, Appendix VIII,
Supplement 10. Examination procedures, equipment, and personnel are
qualified for depth sizing and length sizing when the RMS error, as
defined by Subarticle VIII-3120 of Section XI, Appendix VIII, of the
flaw depth measurements, as compared to the true flaw depths, do not
exceed \1/8\ inch (3 mm) and the root mean square (RMS) error of the
flaw length measurements, as compared to the true flaw lengths, do not
exceed \3/8\ inch (10 mm), respectively.
(5) If flaws attributed to PWSCC have been identified, whether
acceptable or not for continued service under Paragraphs -3130 or -3140
of ASME Code Case N-729-1, the re-inspection interval must be each
refueling outage instead of the re-inspection intervals required by
Table 1, Note (8), of ASME Code Case N-729-1.
(6) Appendix I of ASME Code Case N-729-1 must not be implemented
without prior NRC approval.
(E) Augmented ISI requirements: Reactor coolant pressure boundary
visual inspections.
(1) All licensees of pressurized water reactors must augment their
inservice inspection program by implementing ASME Code Case N-722-1,
subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2)
through (g)(6)(ii)(E)(4) of this section. The inspection requirements
of ASME Code Case N-722-1 do not apply to components with pressure
retaining welds fabricated with Alloy 600/82/182 materials that have
been mitigated by weld overlay or stress improvement.
(2) If a visual examination determines that leakage is occurring
from a specific item listed in Table 1 of ASME Code Case N-722-1 that
is not exempted by the ASME Code, Section XI, IWB-1220(b)(1),
additional actions must be performed to characterize the location,
orientation, and length of a crack or cracks in Alloy 600 nozzle
wrought material and location, orientation, and length of a crack or
cracks in Alloy 82/182 butt welds. Alternatively, licensees may replace
the Alloy 600/82/182 materials in all the components under the item
number of the leaking component.
(3) If the actions in paragraph (g)(6)(ii)(E)(2) of this section
determine that a flaw is circumferentially oriented and potentially a
result of primary water stress corrosion cracking, licensees must
perform non-visual NDE inspections of components that fall under that
ASME Code Case N-722-1 item number. The number of components inspected
must equal or exceed the number of components found to be leaking under
that item number. If circumferential cracking is identified in the
sample, non-visual NDE must be performed in the remaining components
under that item number.
(4) If ultrasonic examinations of butt welds are used to meet the
NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (g)(6)(ii)(E)(3) of
this section, they must be performed using the appropriate supplement
of Section XI, Appendix VIII, of the ASME BPV Code.
(F) Augmented ISI requirements: Examination requirements for Class
1 piping and nozzle dissimilar-metal butt welds.
(1) Licensees of existing, operating pressurized-water reactors as
of July 21, 2011, must implement the requirements of ASME Code Case N-
770-1, subject to the conditions specified in paragraphs
(g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) of this section, by the
first refueling outage after August 22, 2011.
(2) Full structural weld overlays authorized by the NRC staff may
be categorized as Inspection Items C or F, as appropriate. Welds that
have been mitigated by the Mechanical Stress Improvement Process
(MSIPTM) may be categorized as Inspection Items D or E, as
appropriate, provided the criteria in Appendix I of the Code Case have
been met. For ISI frequencies, all other butt welds that rely on Alloy
82/182 for structural integrity must be categorized as Inspection Items
A-1, A-2 or B until the NRC staff has reviewed the mitigation and
authorized an alternative Code Case Inspection Item for the mitigated
weld, or until an alternative Code Case Inspection Item is used based
on conformance with an ASME mitigation Code Case endorsed in Regulatory
Guide 1.147 with conditions, if applicable, and incorporated by
reference in this section.
(3) Baseline examinations for welds in Table 1, Inspection Items A-
1, A-2, and B, must be completed by the end of the next refueling
outage after January 20, 2012. Previous examinations of these welds can
be credited for baseline examinations if they were performed within the
re-inspection period for the weld item in Table 1 using Section XI,
Appendix VIII, requirements and met the Code required examination
volume of essentially 100 percent. Other previous examinations that do
not meet these requirements can be used to meet the baseline
examination requirement, provided NRC approval of alternative
inspection requirements in accordance with paragraphs (z)(1) or (z)(2)
of this section is granted prior to the end of the next refueling
outage after January 20, 2012.
(4) The axial examination coverage requirements of Paragraph--
2500(c) may not be considered to be satisfied unless essentially 100
percent coverage is achieved.
(5) All hot-leg operating temperature welds in Inspection Items G,
H, J, and K must be inspected each inspection interval. A 25 percent
sample of Inspection Items G, H, J, and K cold-leg operating
temperature welds must be inspected whenever the core barrel is removed
(unless it has already been inspected within the past 10 years) or 20
years, whichever is less.
(6) For any mitigated weld whose volumetric examination detects
growth of existing flaws in the required examination volume that exceed
the previous IWB-3600 flaw evaluations or new flaws, a report
summarizing the evaluation, along with inputs, methodologies,
assumptions, and causes of the new flaw or flaw growth is to be
provided to the NRC prior to the weld being placed in service other
than modes 5 or 6.
(7) For Inspection Items G, H, J, and K, when applying the
acceptance
[[Page 37920]]
standards of ASME BPV Code, Section XI, IWB-3514, for planar flaws
contained within the inlay or onlay, the thickness ``t'' in IWB-3514 is
the thickness of the inlay or onlay. For planar flaws in the balance of
the dissimilar metal weld examination volume, the thickness ``t'' in
IWB-3514 is the combined thickness of the inlay or onlay and the
dissimilar metal weld.
(8) Welds mitigated by optimized weld overlays in Inspection Items
D and E are not permitted to be placed into a population to be examined
on a sample basis and must be examined once each inspection interval.
(9) Replace the first two sentences of Extent and Frequency of
Examination for Inspection Item D in Table 1 of Code Case N-770-1 with,
``Examine all welds no sooner than the third refueling outage and no
later than 10 years following stress improvement application.'' Replace
the first two sentences of Note (11)(b)(2) in Code Case N-770-1 with,
``The first examination following weld inlay, onlay, weld overlay, or
stress improvement for Inspection Items D through K must be performed
as specified.''
(10) General Note (b) to Figure 5(a) of Code Case N-770-1
pertaining to alternative examination volume for optimized weld
overlays may not be applied unless NRC approval is authorized under
paragraphs (z)(1) or (z)(2) of this section.
(h) Protection and safety systems. Protection systems of nuclear
power reactors of all types must meet the requirements specified in
this paragraph. Each combined license for a utilization facility is
subject to the following conditions.
(1) [Reserved]
(2) Protection systems. For nuclear power plants with construction
permits issued after January 1, 1971, but before May 13, 1999,
protection systems must meet the requirements stated in either IEEE
Std. 279, ``Criteria for Protection Systems for Nuclear Power
Generating Stations,'' or in IEEE Std. 603-1991, ``Criteria for Safety
Systems for Nuclear Power Generating Stations,'' and the correction
sheet dated January 30, 1995. For nuclear power plants with
construction permits issued before January 1, 1971, protection systems
must be consistent with their licensing basis or may meet the
requirements of IEEE Std. 603-1991 and the correction sheet dated
January 30, 1995.
(3) Safety systems. Applications filed on or after May 13, 1999,
for construction permits and operating licenses under this part, and
for design approvals, design certifications, and combined licenses
under Part 52 of this chapter, must meet the requirements for safety
systems in IEEE Std. 603-1991 and the correction sheet dated January
30, 1995.
(i) through (y) [Reserved]
(z) Alternatives to codes and standards requirements. Alternatives
to the requirements of paragraphs (b), (c), (d), (e), (f), (g), and (h)
of this section or portions thereof may be used when authorized by the
Director, Office of Nuclear Reactor Regulation, or Director, Office of
New Reactors, as appropriate. A proposed alternative must be submitted
and authorized prior to implementation. The applicant or licensee must
demonstrate that:
(1) Acceptable level of quality and safety. The proposed
alternative would provide an acceptable level of quality and safety; or
(2) Hardship without a compensating increase in quality and safety.
Compliance with the specified requirements of this section would result
in hardship or unusual difficulty without a compensating increase in
the level of quality and safety.
Footnotes to Sec. 50.55a:
\1\ For inspections to be conducted once per interval, the
inspections must be performed in accordance with the schedule in
Section XI, paragraph IWB-2400, except for plants with inservice
inspection programs based on a Section XI edition or addenda prior
to the 1994 Addenda. For plants with inservice inspection programs
based on a Section XI edition or addenda prior to the 1994 Addenda,
the inspection must be performed in accordance with the schedule in
Section XI, paragraph IWB-2400, of the 1994 Addenda.
2-3 [Reserved]
\4\ USAS and ASME Code addenda issued prior to the winter 1977
Addenda are considered to be ``in effect'' or ``effective'' 6 months
after their date of issuance and after they are incorporated by
reference in paragraph (a) of this section. Addenda to the ASME Code
issued after the summer 1977 Addenda are considered to be ``in
effect'' or ``effective'' after the date of publication of the
addenda and after they are incorporated by reference in paragraph
(a) of this section.
\5\ For ASME Code editions and addenda issued prior to the
winter 1977 Addenda, the Code edition and addenda applicable to the
component is governed by the order or contract date for the
component, not the contract date for the nuclear energy system. For
the winter 1977 Addenda and subsequent editions and addenda the
method for determining the applicable Code editions and addenda is
contained in Paragraph NCA 1140 of Section III of the ASME Code.
6-8 [Reserved]
\9\ Guidance for quality group classifications of components
that are to be included in the safety analysis reports pursuant to
Sec. 50.34(a) and Sec. 50.34(b) may be found in Regulatory Guide
1.26, ``Quality Group Classifications and Standards for Water-,
Steam-, and Radiological-Waste-Containing Components of Nuclear
Power Plants,'' and in Section 3.2.2 of NUREG-0800, ``Standard
Review Plan for Review of Safety Analysis Reports for Nuclear Power
Plants.''
Dated at Rockville, Maryland, this 7th day of June 2013.
For the Nuclear Regulatory Commission.
Jennifer L. Uhle,
Deputy Director, Reactor Safety Programs, Office of Nuclear Reactor
Regulation.
[FR Doc. 2013-15022 Filed 6-21-13; 8:45 am]
BILLING CODE 7590-01-P