83 FR 2331 - Approval of American Society of Mechanical Engineers' Code Cases
NUCLEAR REGULATORY COMMISSION
Federal Register Volume 83, Issue 11 (January 17, 2018)
Page Range
2331-2354
FR Document
2018-00112
The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to incorporate by reference (IBR) the latest revisions of three regulatory guides (RGs) approving new, revised, and reaffirmed Code Cases published by the American Society of Mechanical Engineers (ASME). This action allows nuclear power plant licensees and applicants for construction permits, operating licenses, combined licenses, standard design certifications, standard design approvals and manufacturing licenses to voluntarily 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. These engineering standards are set forth in the ASME's Boiler and Pressure Vessel (BPV) Codes and ASME Operation and Maintenance (OM) Codes, which are currently incorporated by reference into the NRC's regulations. This final rule announces the availability of the final versions of the three RGs that are being incorporated by reference. Further, the final rule announces the availability of a related RG, not incorporated by reference into the NRC's regulations that lists Code Cases that the NRC has not approved for use.
Federal Register, Volume 83 Issue 11 (Wednesday, January 17, 2018)
[Federal Register Volume 83, Number 11 (Wednesday, January 17, 2018)]
[Rules and Regulations]
[Pages 2331-2354]
From the Federal Register Online [www.thefederalregister.org]
[FR Doc No: 2018-00112]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2012-0059]
RIN 3150-AJ13
Approval of American Society of Mechanical Engineers' Code Cases
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its
regulations to incorporate by reference (IBR) the latest revisions of
three regulatory guides (RGs) approving new, revised, and reaffirmed
Code Cases published by the American Society of Mechanical Engineers
(ASME). This action allows nuclear power plant licensees and applicants
for construction permits, operating licenses, combined licenses,
standard design certifications, standard design approvals and
manufacturing licenses to voluntarily 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. These engineering standards are set
forth in the ASME's Boiler and Pressure Vessel (BPV) Codes and ASME
Operation and Maintenance (OM) Codes, which are currently incorporated
by reference into the NRC's regulations. This final rule announces the
availability of the final versions of the three RGs that are being
incorporated by reference. Further, the final rule announces the
availability of a related RG, not incorporated by reference into the
NRC's regulations that lists Code Cases that the NRC has not approved
for use.
DATES: This final rule is effective on February 16, 2018. The
incorporation by reference of certain publications listed in the
regulation is approved by the Director of the Federal Register as of
February 16, 2018.
ADDRESSES: Please refer to Docket ID NRC-2012-0059 when contacting the
NRC about the availability of information for this action. You may
obtain publicly-available information related to this action by any of
the following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0059. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. For
the convenience of the reader, instructions about obtaining materials
referenced in this document are provided in the ``Availability of
Documents'' section.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Jennifer Tobin, Office of Nuclear
Reactor Regulation, telephone: 301-415-2328, email:
[email protected]; or Giovanni Facco, Office of Nuclear Regulatory
Research, telephone: 301-415-6337; email: [email protected]. Both
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
The purpose of this regulatory action is to incorporate by
reference into the NRC's regulations the latest revisions of three RGs.
The three RGs identify new, revised, and reaffirmed Code Cases
published by the ASME, which the NRC has determined are acceptable for
use as alternatives to certain provisions of the ASME BPV Codes and
ASME OM Codes, currently incorporated by reference into the NRC's
regulations. The three RGs that the NRC is incorporating by reference
are RG 1.84, ``Design, Fabrication, and Materials Code Case
Acceptability, ASME Section III,'' Revision 37; RG 1.147, ``Inservice
Inspection Code Case Acceptability, ASME Section XI, Division 1,''
Revision 18; and RG 1.192, ``Operation and Maintenance Code Case
Acceptability, ASME OM Code,'' Revision 2. This regulatory action
allows nuclear power plant licensees and applicants for construction
permits, operating licenses, combined licenses, standard design
certifications, standard design approvals, and manufacturing licenses
to voluntarily use the Code Cases, newly listed in these revised RGs,
as
[[Page 2332]]
alternatives to engineering standards for the design, construction,
ISI, and IST, and repair/replacement of nuclear power plant components.
In this notice, the NRC also notifies the public of the availability of
RG 1.193, ``ASME Code Cases Not Approved for Use,'' Revision 5. The
regulatory guide lists Code Cases that the NRC has not approved for
generic use, and will not be incorporated by reference into the NRC's
regulations.
The NRC prepared a regulatory analysis (ADAMS Accession No.
ML16285A013) to identify the benefits and costs associated with this
final rule. The regulatory analysis prepared for this rulemaking was
used to determine if the rule is cost-effective, overall, and to help
the NRC evaluate potentially costly conditions placed on specific
provisions of the ASME Code Cases, which are the subject of this
rulemaking.
Table 1--Cost-Benefit Summary
------------------------------------------------------------------------
Alternative 2--
the rule
alternative net
benefits (costs)
Objective (net present
value, 7%
discount rate)
($ million)
------------------------------------------------------------------------
Industry............................................. 2.42
NRC.................................................. 2.52
Net Benefit.......................................... 4.94
------------------------------------------------------------------------
Table 1 summarizes the benefits and costs for the alternative of
proceeding with the final rule (Alternative 2) and shows that the final
rule is quantitatively cost-beneficial with a net benefit of $4.94
million to both the industry and the NRC when compared to the
regulatory baseline (Alternative 1). The regulatory analysis shows that
implementing the final rule is quantitatively cost-effective and an
efficient use of the NRC's and Industry's resources. Uncertainty
analysis shows that the net benefit ranges from $2.86 million to $6.90
million with a mean of $4.94 million. Because the rulemaking
alternative is cost-effective, the rulemaking approach is recommended.
There are several benefits associated with this final rule. Under
this final rule, a licensee of a nuclear power plant would no longer be
required to submit a Code Case alternative request under the new Sec.
50.55a(z) of Title 10 of the Code of Federal Regulations (10 CFR),
which would provide an averted cost of $7.75 million (7[dash]percent
net present value) to the licensee. Additionally, the NRC would not
receive Code Case alternative request submittals, which would provide
an averted cost of $2.52 million (7[dash]percent net present value) to
the NRC.
Table of Contents
I. Background
II. Discussion
A. ASME Code Cases Approved for Unconditional Use
B. ASME Code Cases Approved for Use with Conditions
ASME BPV Code, Section III Code Cases (RG 1.84)
ASME BPV Code, Section XI Code Cases (RG 1.147)
OM Code Cases (RG 1.192)
C. ASME Code Cases not Approved for Use (RG 1.193)
III. Opportunities for Public Participation
IV. Public Comment Analysis
V. Section-by-Section Analysis
VI. Regulatory Flexibility Certification
VII. Regulatory Analysis
VIII. Backfitting and Issue Finality
IX. Plain Writing
X. Environmental Assessment and Final Finding of No Significant
Environmental Impact
XI. Paperwork Reduction Act
XII. Congressional Review Act
XIII. Voluntary Consensus Standards
XIV. Incorporation by Reference--Reasonable Availability to
Interested Parties
XV. Availability of Documents
I. Background
The ASME develops and publishes the ASME BPV Code, which contains
requirements for the design, construction, and ISI and examination of
nuclear power plant components, and ASME's Nuclear Power Plants (OM)
Code,\1\ which contains requirements for IST of nuclear power plant
components. In response to BPV Code and OM Code user requests, the ASME
develops Code Cases that provide alternatives to BPV Code and OM Code
requirements under special circumstances.
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\1\ The editions and addenda of the ASME Code for Operation and
Maintenance of Nuclear Power Plants have had different titles from
2005 to 2012, and are referred to collectively in this rule as the
``OM Code.''
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The NRC approves and can mandate the use of the ASME BPV Codes and
OM Codes in Sec. 50.55a, ``Codes and standards,'' through the process
of incorporation by reference. As such, each provision of the ASME
Codes incorporated by reference into and mandated by Sec. 50.55a
constitutes a legally[dash]binding NRC requirement imposed by the
regulations. As noted previously, ASME Code Cases, for the most part,
represent alternative approaches for complying with provisions of the
ASME BPV Codes and OM Codes. Accordingly, the NRC periodically amends
Sec. 50.55a to incorporate by reference the NRC's RGs listing approved
ASME Code Cases that may be used as alternatives to the BPV Codes and
OM Codes.\2\
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\2\ See Federal Register notice, ``Incorporation by Reference of
ASME BPV and OM Code Cases'' (68 FR 40469; July 8, 2003).
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This rulemaking is the latest in a series of rulemakings that
incorporates by reference new versions of several RGs identifying new,
revised, and reaffirmed,\3\ and unconditionally or conditionally
acceptable ASME Code Cases that the NRC approves for use. In developing
these RGs, the staff reviews ASME BPV and OM Code Cases, determines the
acceptability of each Code Case, and publishes its findings in the RGs.
The RGs are revised periodically, as new Code Cases and are published
by the ASME. The NRC incorporates by reference the RGs, listing
acceptable and conditionally acceptable ASME Code Cases into Sec.
50.55a. Currently, NRC RG 1.84, ``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 OM Code,'' Revision 1, are incorporated
into the NRC's regulations in Sec. 50.55a.
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\3\ Code Cases are categorized by ASME as one of three types:
new, revised, or reaffirmed. A new Code Case provides for a new
alternative to specific ASME Code provisions or addresses a new
need. The ASME defines a revised Code Case to be 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 RGs that have been incorporated
by reference into Sec. 50.55a. The ASME defines ``reaffirmed'' as
an OM Code Case to be one that does not have any change to technical
content, but includes editorial changes.
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II. Discussion
This rule incorporates by reference the latest revisions of the NRC
RGs that list ASME BPV and OM Code Cases that the NRC finds to be
acceptable, or acceptable with NRC[dash]specified conditions
(``conditionally acceptable''). Regulatory Guide 1.84, Revision 37,
supersedes Revision 36; RG 1.147, Revision 18, supersedes Revision 17;
and RG 1.192, Revision 2, supersedes Revision 1. The NRC also publishes
a document (RG 1.193, ``ASME Code Cases Not Approved for Use'') that
lists Code Cases that the NRC has not approved for generic use. The RG
1.193 is not incorporated by reference into the NRC's regulations;
however, in this final rule, the NRC notes the availability of RG
1.193, Revision 5.
The ASME Code Cases that are the subject of this rulemaking are the
new, revised, and reaffirmed Section III and Section XI Code Cases
listed in
[[Page 2333]]
Supplement 11 to the 2007 BPV Code through Supplement 10 to the 2010
BPV Code, and the OM Code Cases published with the 2009 Edition through
the 2012 Edition.
The latest editions and addenda of the ASME BPV and OM Codes that
the NRC has approved for use are referenced in Sec. 50.55a. The ASME
also publishes Code Cases that provide alternatives to existing Code
requirements that the ASME developed and approved. This rule
incorporates by reference the latest revisions of RGs 1.84, 1.147, and
1.192. This rule allows nuclear power plant licensees and applicants
for construction permits, operating licenses, combined licenses,
standard design certifications, standard design approvals, and
manufacturing licenses, under the regulations that govern license
certifications, to voluntarily use the Code Cases listed in these RGs
as suitable alternatives to certain provisions of the ASME BPV and OM
Codes for the construction, ISI, and IST of nuclear power plant
components. This action is consistent with the provisions of the
National Technology Transfer and Advancement Act of 1995 (NTTAA),
Public Law 104-113, which encourages Federal regulatory agencies to
consider adopting industry consensus standards as an alternative to de
novo agency development of standards affecting an industry. This action
is also consistent with the NRC 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 acceptability of
new, revised, and reaffirmed Code Cases, and the need for regulatory
positions on the use of these Code Cases. This process was employed in
the review of the Code Cases in Supplement 11 to the 2007 Edition
through Supplement 10 to the 2010 Edition of the BPV Code and the 2009
Edition through the 2012 Edition of the OM Code. The Code Cases in
these supplements and OM Editions and Addenda are the subject of this
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 view points are
addressed fairly. The NRC staff actively participates in discussions
and technical debates of the task groups, working groups, subgroups,
and standards committees regarding the development of new and revised
standards. The Code Case process includes the 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 respective standards committees. The standards committee is
the decisive consensus committee in that it ensures that the
development process fully complies with the ANSI consensus process.
Second, the standards committee transmits a first consideration
letter ballot to every member of the standards committee, requesting
comment or approval of new and revised Code Cases. Code Cases are
approved by the standards committee from the first consideration letter
ballot when: (1) At least two thirds of the eligible consensus
committee membership vote approved; (2) there are no disapprovals from
the standards committee; and (3) no substantive comments are received
from the 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 to
provide recommendations in the development of codes and standards, as
required. Code Cases that were disapproved or received 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 are approved by the standards
committee on second consideration when at least two thirds of the
eligible consensus committee membership vote approved, and there are no
more than three disapprovals from the consensus committee.
Third, the NRC reviews new, revised, and reaffirmed Code Cases to
determine their acceptability for incorporation by reference in Sec.
50.55a through the subject RGs. This rulemaking process, when
considered together with the ANSI process for developing and approving
the ASME codes and standards, and 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 staff concludes, in accordance with the process described, that
the Code Cases are technically adequate (with conditions as necessary)
and consistent with current NRC regulations, and the staff is
referencing these Code Cases in the applicable RGs, thereby approving
them for voluntary use, without conditions as addressed in Section A of
this document; subject to the specified conditions, or as identified in
Section B of this document. The staff reviewed the new, revised, and
reaffirmed Code Cases identified in the three RGs being incorporated by
reference into Sec. 50.55a in this rulemaking. Therefore, the NRC
approves revising the Sec. 50.55a regulations to incorporate by
reference the latest revisions of RGs 1.84, 1.147, and 1.192.
Additionally, the NRC announces the availability of the latest revision
of RG 1.193.
A. ASME Code Cases Approved for Unconditional Use
The Code Cases that are discussed in Table I are new, revised, or
reaffirmed Code Cases that the NRC is approving for use without
conditions. The NRC concludes, in accordance with the process described
for review of ASME Code Cases, that each of the ASME Code Cases listed
in Table I are acceptable for use without conditions. Therefore, the
NRC is approving for unconditional use the Code Cases listed in Table
I. This table identifies the regulatory guide the applicable Code Case
that the NRC is approving for use.
The NRC revised RG 1.147, Revision 18 to approve Code Case N-786-1
in Table 1 to address inconsistencies that were identified between the
NRC's position in the proposed rule regarding the acceptability of Code
Case N-786 and several licensee requests for alternatives to ASME Code
requirements, in accordance with Title 10 of the Code of Federal
Regulations (10 CFR) 50.55a(z), that have utilized Code Case N-786. The
NRC had authorized the use of Code Case N-786 with modifications. The
NRC erred in not listing N-786 in DG-1296, Table 2 ``Conditionally
Acceptable Section XI Code Cases'' with appropriate conditions, in
order to be consistent with modifications that the NRC has required for
requested alternatives based on Code Case N-786. In response to
modifications to N-786 by licensees requesting to use this code case as
an alternative to ASME Code, ASME revised the code case. The revised
Code Case, N-786-1 ``Alternative Requirements for Sleeve Reinforcement
of Class 2 and 3 Moderate[dash]Energy Carbon Steel Piping Section XI,
Division 1,'' includes modifications that address all of the NRC's
concerns that the NRC identified in previously approved alternatives
that were based on N-786. Therefore, the NRC has listed Code Case N-
786-1 in Table 1 of RG 1.147 Revision 18 in lieu of code Case N-786.
There were no public comments
[[Page 2334]]
received on the inclusion of N-786 in the RG. Code Case N-786-1 is
included in this final rule because it includes the latest ASME
guidance and the NRC conditions on the use of this method of repair.
Table I--ASME Code Cases Approved for Unconditional Use
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Code Case No. Supplement Title
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Boiler and Pressure Vessel Code Section III
(addressed in RG 1.84, Revision 37, Table 1)
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N-284-3....................... 7 (10 Edition)... Metal Containment
Shell Buckling
Design Methods,
Class MC, TC, and SC
Construction,
Section III,
Divisions 1 and 3.
N-500-4....................... 8 (10 Edition)... Alternative Rules for
Standard Supports
for Classes 1, 2, 3,
and MC, Section III,
Division 1.
N-520-5....................... 10 (10 Edition).. Alternative Rules for
Renewal of Active or
Expired N-type
Certificates for
Plants Not in Active
Construction,
Section III,
Division 1.
N-594-1....................... 8 (10 Edition)... Repairs to P-4 and P-
5A Castings without
Postweld Heat
Treatment Class 1,
2, and 3
Construction,
Section III,
Division 1.
N-637-1....................... 3 (10 Edition)... Use of 44Fe-25Ni-21Cr-
Mo (Alloy UNS
N08904) Plate, Bar,
Fittings, Welded
Pipe, and Welded
Tube, Classes 2 and
3, Section III,
Division 1.
N-655-2....................... 4 (10 Edition)... Use of SA-738, Grade
B, for Metal
Containment Vessels,
Class MC, Section
III, Division 1.
N-763......................... 2 (10 Edition)... ASTM A 709-06, Grade
HPS 70W (HPS 485W)
Plate Material
Without Postweld
Heat Treatment as
Containment Liner
Material or
Structural
Attachments to the
Containment Liner,
Section III,
Division 2.
N-777......................... 4 (10 Edition)... Calibration of Cv
Impact Test
Machines, Section
III, Divisions 1, 2,
and 3.
N-785......................... 11 (07 Edition).. Use of SA-479/SA-
479M, UNS S41500 for
Class 1 Welded
Construction,
Section III,
Division 1.
N-811......................... 7 (10 Edition)... Alternative
Qualification
Requirements for
Concrete Level III
Inspection
Personnel, Section
III, Division 2.
N-815......................... 8 (10 Edition)... Use of SA-358/SA-358M
Grades Fabricated as
Class 3 or Class 4
Welded Pipe, Class
CS Core Support
Construction,
Section III,
Division 1.
N-816......................... 8 (10 Edition)... Use of Temper Bead
Weld Repair Rules
Adopted in 2010
Edition and Earlier
Editions, Section
III, Division 1.
N-817......................... 8 (10 Edition)... Use of Die Forgings,
SB-247, UNS A96061
Class T6, With
Thickness <= 4.000
in. Material, Class
2 Construction (1992
Edition or Later),
Section III,
Division 1.
N-819......................... 8 (10 Edition)... Use of Die Forgings,
SB-247, UNS A96061
Class T6, With
Thickness <= 4.000
in. Material, Class
2 Construction (1989
Edition with the
1991 Addenda or
Earlier), Section
III, Division 1.
N-822......................... 8 (10 Edition)... Application of the
ASME Certification
Mark, Section III,
Divisions 1, 2, 3,
and 5.
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Boiler and Pressure Vessel Code Section XI
(addressed in RG 1.147, Revision 18, Table 1)
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N-609-1....................... 3 (10 Edition)... Alternative
Requirements to
Stress-Based
Selection Criteria
for Category B-J
Welds, Section XI,
Division 1.
N-613-2....................... 4 (10 Edition)... Ultrasonic
Examination of Full
Penetration Nozzles
in Vessels,
Examination Category
B-D, Reactor
Nozzle[dash]To[dash]
Vessel Welds, and
Nozzle Inside Radius
Section Figs. IWB-
2500-7(a), (b), (c),
and (d), Section XI,
Division 1.
N-652-2....................... 9 (10 Edition)... Alternative
Requirements to
Categorize B-G-1, B-
G-2, and C-D Bolting
Examination Methods
and Selection
Criteria, Section
XI, Division 1.
N-653-1....................... 9 (10 Edition)... Qualification
Requirements for
Full Structural
Overlaid Wrought
Austenitic Piping
Welds, Section XI,
Division 1.
N-694-2 \4\................... 1 (13 Edition)... Evaluation Procedure
and Acceptance
Criteria for
[pressurized water
reactors] (PWR)
Reactor Vessel Head
Penetration Nozzles,
Section XI, Division
1.
N-730-1....................... 10 (10 Edition).. Roll Expansion of
Class 1 Control Rod
Drive Bottom Head
Penetrations in
[boiling water
reactors] BWRs,
Section XI, Division
1.
N-769-2....................... 10 (10 Edition).. Roll Expansion of
Class 1 In[dash]Core
Housing Bottom Head
Penetrations in
BWRs, Section XI,
Division 1.
N-771......................... 7 (10 Edition)... Alternative
Requirements for
Additional
Examinations of
Class 2 or 3 Items,
Section XI, Division
1.
N-775......................... 2 (10 Edition)... Alternative
Requirements for
Bolting Affected by
Borated Water
Leakage, Section XI,
Division 1.
N-776......................... 1 (10 Edition)... Alternative to IWA-
5244 Requirements
for Buried Piping,
Section XI, Division
1.
N-786-1....................... 5 (10 Edition)... Alternative
Requirements for
Sleeve Reinforcement
of Class 2 and 3
Moderate-Energy
Carbon Steel Piping,
Section XI, Division
1.
[[Page 2335]]
N-798......................... 4 (10 Edition)... Alternative Pressure
Testing Requirements
for Class 1 Piping
Between the First
and Second Vent,
Drain, and Test
Isolation Devices,
Section XI, Division
1.
N-800......................... 4 (10 Edition)... Alternative Pressure
Testing Requirements
for Class 1 Piping
Between the First
and Second Injection
Valves, Section XI,
Division 1.
N-803......................... 5 (10 Edition)... Similar and
Dissimilar Metal
Welding Using
Ambient Temperature
Automatic or Machine
Dry Underwater Laser
Beam Welding (ULBW)
Temper Bead
Technique, Section
XI, Division 1.
N-805......................... 6 (10 Edition)... Alternative to Class
1 Extended Boundary
End of Interval or
Class 2 System
Leakage Testing of
the Reactor Vessel
Head Flange O-Ring
Leak-Detection
System, Section XI,
Division 1.
N-823......................... 9 (10 Edition)... Visual Examination,
Section XI, Division
1.
N-825 \5\..................... 3 (13 Edition)... Alternative
Requirements for
Examination of
Control Rod Drive
Housing Welds,
Section XI, Division
1.
N-845 \6\..................... 6 (13 Edition)... Qualification
Requirements for
Bolts and Studs,
Section XI, Division
1.
------------------------------------------------------------------------
Operation and Maintenance Code (OM)
(addressed in RG 1.192, Revision 2, Table 1)
------------------------------------------------------------------------
OMN-2......................... 2012 Edition..... Thermal Relief Valve
Code Case, OM Code-
1995, Appendix I.
OMN-5......................... 2012 Edition..... Testing of Liquid
Service Relief
Valves without
Insulation.
OMN-6......................... 2012 Edition..... Alternative Rules for
Digital Instruments.
OMN-7......................... 2012 Edition..... Alternative
Requirements for
Pump Testing.
OMN-8......................... 2012 Edition..... Alternative Rules for
Preservice and
Inservice Testing of
Power-Operated
Valves That Are Used
for System Control
and Have a Safety
Function per OM-10,
ISTC-1.1, or ISTA-
1100.
OMN-13, Revision 2............ 2012 Edition..... Performance-Based
Requirements for
Extending Snubber
Inservice Visual
Examination Interval
at [light water
reactor] LWR Power
Plants.
OMN-14........................ 2012 Edition..... Alternative Rules for
Valve Testing
Operations and
Maintenance,
Appendix I: BWR
[control rod drive]
CRD Rupture Disk
Exclusion.
OMN-15, Revision 2............ 2012 Edition..... Performance-Based
Requirements for
Extending the
Snubber Operational
Readiness Testing
Interval at LWR
Power Plants.
OMN-17........................ 2012 Edition..... Alternative Rules for
Testing ASME Class 1
Pressure Relief/
Safety Valves.
------------------------------------------------------------------------
B. ASME Code Cases Approved for Use With Conditions
The Code Cases that are discussed in Table II, below, are new,
revised or reaffirmed Code Cases, which the NRC is approving 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 in order to provide an acceptable level of quality and
safety. Accordingly, the NRC is imposing conditions on the use of these
Code Cases to modify, limit, or clarify their requirements. The
conditions specify, for each applicable Code Case, the additional
activities that must be performed, the limits on the activities
specified in the Code Case, and the supplemental information needed to
provide clarity. These ASME Code Cases with conditions are included in
Table 2 of each RG (i.e., RG 1.84, RG 1.147, and RG 1.192). It is noted
that both RG 1.147 and RG 1.192 have new ASME Code Cases with
conditions; however, there are no new ASME Code Cases with conditions
for RG 1.84.
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\4\ Code Case published in Supplement 1 to the 2013 Edition;
included at the request of ASME.
\5\ Code Case published in Supplement 3 to the 2013 Edition;
included at the request of ASME.
\6\ Code Case published in Supplement 6 to the 2013 Edition;
included at the request of ASME.
Table II--Code Cases Approved for Conditional Use
------------------------------------------------------------------------
Code Case No. Supplement Title
------------------------------------------------------------------------
Boiler and Pressure Vessel Code Section III
(addressed in RG 1.84, Revision 37, Table 2)
------------------------------------------------------------------------
No ASME Section III Code Cases are approved for conditional use in this
rule.
Boiler and Pressure Vessel Code Section XI
(addressed in RG 1.147, Revision 18, Table 2)
------------------------------------------------------------------------
N-552-1....................... 10 (10 Edition).. Alternative Methods--
Qualification for
Nozzle Inside Radius
Section from the
Outside Surface,
Section XI, Division
1.
N-576-2....................... 9 (10 Edition)... Repair of Class 1 and
2 SB-163, UNS N06600
Steam Generator
Tubing, Section XI,
Division 1.
N-593-2....................... 8 (10 Edition)... Examination
Requirements for
Steam Generator
Nozzle-to-Vessel
Welds, Section XI,
Division 1.
[[Page 2336]]
N-638-6....................... 6 (10 Edition)... Similar and
Dissimilar Metal
Welding Using
Ambient Temperature
Machine GTAW Temper
Bead Technique,
Section XI, Division
1.
N-662-1....................... 6 (10 Edition)... Alternative Repair/
Replacement
Requirements for
Items Classified in
Accordance with Risk-
Informed Processes,
Section XI, Division
1.
N-666-1....................... 9 (10 Edition)... Weld Overlay of
Classes 1, 2, and 3
Socket Welded
Connections, Section
XI, Division 1.
N-749......................... 9 (10 Edition)... Alternative
Acceptance Criteria
for Flaws in
Ferritic Steel
Components Operating
in the Upper Shelf
Temperature Range,
Section XI, Division
1.
N-754......................... 6 (10 Edition)... Optimized Structural
Dissimilar Metal
Weld Overlay for
Mitigation of PWR
Class 1 Items,
Section XI, Division
1.
N-778......................... 6 (10 Edition)... Alternative
Requirements for
Preparation and
Submittal of
Inservice Inspection
Plans, Schedules,
and Preservice and
Inservice Summary
Reports, Section XI,
Division 1.
N-789......................... 6 (10 Edition)... Alternative
Requirements for Pad
Reinforcement of
Class 2 and 3
Moderate Energy
Carbon Steel Piping
for Raw Water
Service, Section XI,
Division 1.
N-795......................... 3 (10 Edition)... Alternative
Requirements for BWR
Class 1 System
Leakage Test
Pressure Following
Repair/Replacement
Activities, Section
XI, Division 1.
N-799......................... 4 (10 Edition)... Dissimilar Metal
Welds Joining Vessel
Nozzles to
Components, Section
XI, Division 1.
------------------------------------------------------------------------
Operation and Maintenance Code (OM)
(addressed in RG 1.192, Revision 2, Table 2)
------------------------------------------------------------------------
OMN-1 Revision 1.............. 2012 Edition..... Alternative Rules for
Preservice and
Inservice Testing of
Active Electric
Motor Operated-Valve
Assemblies in
Light[dash]Water
Reactor Power
Plants.
OMN-3......................... 2012 Edition..... Requirements for
Safety Significance
Categorization of
Components Using
Risk Insights for
Inservice Testing of
LWR Power Plants.
OMN-4......................... 2012 Edition..... Requirements for Risk
Insights for
Inservice Testing of
Check Valves at LWR
Power Plants.
OMN-9......................... 2012 Edition..... Use of a Pump Curve
for Testing.
OMN-12........................ 2012 Edition..... Alternative
Requirements for
Inservice Testing
Using Risk Insights
for Pneumatically
and Hydraulically
Operated Valve
Assemblies in Light-
Water Reactor Power
Plants (OM-Code
1998, Subsection
ISTC).
OMN-16 Revision 1............. 2012 Edition..... Use of a Pump Curve
for Testing.
OMN-18........................ 2012 Edition..... Alternate Testing
Requirements for
Pumps Tested
Quarterly Within
20% of
Design Flow.
OMN-19........................ 2012 Edition..... Alternative Upper
Limit for the
Comprehensive Pump
Test.
OMN-20........................ 2012 Edition..... Inservice Test
Frequency.
------------------------------------------------------------------------
The NRC's evaluation of the Code Cases and the reasons for the
NRC's conditions are discussed in the following paragraphs. Notations
have been made to indicate the conditions duplicated from previous
versions of the RG.
ASME BPV Code, Section III Code Cases (RG 1.84)
There are no new or revised Section III Code Cases in Supplement 11
to the 2007 Edition through Supplement 10 to the 2010 Edition that the
NRC is conditionally approving in Revision 37 of RG 1.84.
ASME BPV Code, Section XI Code Cases (RG 1.147)
Code Case N-552-1 [Supplement 10, 2010 Edition]
Type: Revised.
Title: Alternative Methods--Qualification for Nozzle Inside Radius
Section from the Outside Surface, Section XI, Division 1.
The conditions on Code Case N-552-1 are identical to the conditions
on N-552 that were approved by the NRC in Revision 16 of RG 1.147 in
October 2010. The reasons for imposing these conditions in Code Case N-
576 continue to apply to N-576-2. Therefore, these conditions have been
retained for this Code Case in Revision 18 of RG 1.147.
Code Case N-576-2 [Supplement 9, 2010 Edition]
Type: Revised.
Title: Repair of Class 1 and 2 SB-163, UNS N06600 Steam Generator
Tubing, Section XI, Division 1.
The conditions on Code Case N-576-2 are identical to the conditions
on N-576-1 that were approved by the NRC in Revision 17 of RG 1.147 in
October 2014. The reasons for imposing these conditions are not
resolved by Code Case N-576-2 and, therefore, these conditions have
been retained in Revision 18 of RG 1.147.
Public comments on N-576-2 requested that the NRC revise the
proposed condition to follow IWA-4200 in their code of record. In
response, the NRC revised the ``note'' in the condition in Revision 18
of RG 1.147 to eliminate the portion regarding reconciliation. The
revised ``note'' will read: ``Note: Steam generator tube repair methods
require prior NRC approval through the Technical Specifications. This
Code Case does not address certain aspects of this repair, e.g., the
qualification of the inspection and plugging criteria necessary for
staff approval of the repair method.''
[[Page 2337]]
Code Case N-593-2 [Supplement 8, 2010 Edition]
Type: Revised.
Title: Examination Requirements for Steam Generator Nozzle-to-
Vessel Welds, Section XI, Division 1.
The first condition on Code Case N-593-2 is identical to the
condition on Code Case N-593 that was first approved by the NRC in
Revision 13 of RG 1.147 in June 2003. The condition stated that,
``Essentially 100 percent (not less than 90 percent) of the examination
volume A-B-C-D-E-F-G-H [in Figure 1 of the Code Case] must be
examined.'' The reasons for imposing this condition in Code Case N-593
continue to apply to Code Case N-593-2. Therefore, this condition has
been retained for this Code Case in Revision 18 of RG 1.147.
The second condition on Code Case N-593-2 is new. Revision 2 of the
Code Case reduces the weld examination volume by reducing the width
examined on either side of the weld from ts/2 to \1/2\ in.
The basis for this change in inspection volume is to revise the
examination volume for steam generator nozzle[dash]to[dash]vessel welds
(under Code Case N-593-2) to be consistent with that specified in Code
Case N-613-1 for similar vessel nozzles.
The NRC identified an issue with respect to Code Case N-593-2
regarding its inconsistency with Code Case N-613-1. Code Case N-593-2
and Code Case N-613-1 address certain types of nozzle-to-vessel welds.
Code Case N-613-1 states that ``. . . Category B-D nozzle-to-vessel
welds previously ultrasonically examined using the examination volumes
of Figs. IWB-2500-7(a), (b), and (c) may be examined using the reduced
examination volume (A-B-C-D-E-F-G-H) of Figs. 1, 2, and 3.'' The
keywords are ``previously examined.'' Code Case N-613-1 requires the
larger volume to have been previously examined before examinations
using the reduced volume can be performed. This ensures that there are
no detrimental flaws in the component adjacent to the weld that would
be missed if the inspection was performed only on the reduced volume.
However, Code Case N-593-2 allows a licensee to immediately implement
the reduced volume. Accordingly, the NRC is approving Code Case N-593-2
with a condition to require that the examination volume specified in
Section XI, Table IWB-2500-1, Examination Category B-D, be used for the
examination of steam generator nozzle[dash]to[dash]vessel welds at
least once prior to use of the reduced volume, as allowed by the Code
Case.
Code Case N-638-6 [Supplement 6, 2010 Edition]
Type: Revised.
Title: Similar and Dissimilar Metal Welding Using Ambient
Temperature Machine GTAW Temper Bead Technique, Section XI, Division 1.
Code Case N-638-6 allows the use of the automatic or machine gas-
tungsten arc welding (GTAW) temper bead technique. The GTAW is a proven
method that can produce high-quality welds because it affords greater
control over the weld area than many other welding processes.
The NRC first approved Code Case N-638 (Revision 0) in 2003
(Revision 13 of RG 1.147). Code Case N-638-4 was approved by the NRC in
Revision 16 of RG 1.147 with two conditions. Code Case N-638-5 was not
approved in RG 1.147 for generic use but has been approved through
requests for an alternative to Sec. 50.55a. Code Case N-638-6 resolves
one of the NRC's concerns that were raised when Code Case N-638-4 was
considered for approval and, therefore, the NRC is deleting that
condition from RG 1.147.
Many of the provisions for developing and qualifying welding
procedure specifications for the temper bead technique that were
contained in earlier versions of the Code Case have been incorporated
into ASME Section IX, ``Welding and Brazing Qualifications,'' QW-290,
``Temper Bead Welding.'' Code Case N-638-6 retains the provisions not
addressed by QW-290 and references QW-290 in lieu of specifying them
directly in the Code Case.
In addition to retaining one of the two conditions on Code Case N-
638-4, the NRC considered adding a new condition to address technical
issues raised by certain provisions of Code Case N-638-6.
The retained condition on Code Case N-638-6 pertains to the
qualification of nondestructive evaluation (NDE) and is identical to
the condition on N-638-4 that was approved by the NRC in Revision 17 of
RG 1.147 in October 2014. The reasons for imposing this condition in
Code Case N-638 continue to apply to N-638-6. Therefore, this condition
has been retained in Revision 18 of RG 1.147.
The new proposed condition (2) states that section 1(b)(1) of the
Code Case shall not be used. Section 1(b)(1) would allow through-wall
circumferential repair welds to be made using the temper bead technique
without heat treatment. Revisions 1 through 5 of N-638 limited the
depth of the weld to one-half of the ferritic base metal thickness and
the previously stated condition will limit repairs to this previously
approved value. Repairs exceeding one-half of the ferritic base metal
thickness may represent significant repairs (e.g., replacement of an
entire portion of the reactor coolant loop). At the time that this
revision of the Code Case was approved by ASME, the NRC staff had
concerns related to through-wall repairs. Subsequently, through further
evaluation related to a separate rulemaking, the NRC resolved its
concerns related to through-wall repairs. Therefore, the NRC determined
that proposed Condition (2) is unnecessary and has removed this
condition from the final RG 1.147, Revision 18.
Code Case N-662-1 [Supplement 6, 2010 Edition]
Type: Revised.
Title: Alternative Repair/Replacement Requirements for Items
Classified in Accordance with Risk-Informed Processes, Section XI,
Division 1.
The condition on Code Case N-662-1 is identical to the condition on
N-662 that was approved by the NRC in Revision 16 of RG 1.147 in
October 2010. The reasons for imposing this condition were not resolved
by Code Case N-662-1. Therefore, this condition has been retained for
this Code Case in Revision 18 of RG 1.147.
Code Case N-666-1 [Supplement 9, 2010 Edition]
Type: Revised.
Title: Weld Overlay of Classes 1, 2, and 3 Socket Welded
Connections, Section XI, Division 1.
Code Case N-666 was unconditionally approved in Revision 17 of RG
1.147. The NRC approves Code Case N-666-1 with one condition.
The condition is that a surface examination must be performed on
the completed weld overlay for Class 1 and Class 2 piping socket welds.
Code Case N-666-1 contains provisions for the design, installation,
evaluation, pressure testing, and examination of the weld overlays on
Class 1, 2, and 3 socket welds. Section 5(a)(1) of the Code Case
requires NDE of the completed weld overlay in accordance with the
Construction Code. However, various Construction Codes have been used
in the design and fabrication of the nuclear power plant fleet. The
requirements for NDE have changed over the years, as more effective and
reliable methods and techniques have been developed. In addition,
Construction Code practices have evolved based on design and
construction experience. The NRC is concerned that some of the
Construction
[[Page 2338]]
Codes would not require a surface examination of the weld overlay and
would, therefore, be inadequate for NDE of the completed weld overlay.
The NRC believes that a VT-1 examination alone would not be adequate
and that a surface or volumetric examination must be performed on the
completed weld overlay for Class 1 and Class 2 piping socket welds.
Fabrication defects must be dispositioned using the surface or
volumetric examination criteria of the Construction Code, as identified
in the Repair/Replacement Plan.
Public commenters requested that the words ``and seal weld'' be
removed from the condition because the phrase implies that the seal
weld requires surface examination in addition to surface examination of
the final overlay. The Code Case requires a visual examination of the
seal weld, remaining socket weld, and adjacent base material before the
weld overlay can be applied, which the NRC has determined is the
appropriate examination prior to the application of the weld overlay.
Therefore, proposed Condition (1) has been revised to remove ``and seal
weld.''
In the proposed rule, the NRC included a second condition, which
required that if a surface or volumetric examination of the completed
weld overlay was not required by the plant-specific Construction Code,
that a VT-1 visual examination be performed of the weld overlay.
Paragraph 5(a) of the Code Case requires ``visual and nondestructive
examination of the final structural overlay weld.'' Paragraph 5(a)(1)
of the Code Case specifically requires a VT-1 visual examination of the
completed weld overlay. Public commenters requested that the NRC remove
the second condition because it was redundant and unnecessary. The NRC
staff agrees and thus Condition (2) has been removed from the final
rule.
Code Case N-749 [Supplement 9, 2010 Edition]
Type: New.
Title: Alternative Acceptance Criteria for Flaws in Ferritic Steel
Components Operating in the Upper Shelf Temperature Range, Section XI,
Division 1.
The NRC has determined that instead of the upper shelf transition
temperature, Tc, as defined in the Code Case, the following
shall be used:
Tc = 154.8 [deg]F + 0.82 x RTNDT (in U.S
Customary Units), and
Tc = 82.8 [deg]C + 0.82 x RTNDT (in International
System (SI) Units).
Tc is the temperature above which the elastic plastic
fracture mechanics (EPFM) method must be applied. Additionally, the NRC
defines temperature Tc1 below, which the linear elastic
fracture mechanics (LEFM) method must be applied:
Tc1 = 95.36 [deg]F + 0.703 x RTNDT (in U.S
Customary Units), and
Tc1 = 47.7 [deg]C + 0.703 x RTNDT (in
International System (SI) Units).
Between Tc1 and Tc, while the fracture mode
is in transition from LEFM to EPFM, users should consider whether or
not it is appropriate to apply the EPFM method. Alternatively, the
licensee may use a different Tc value, if it can be
justified by plant[dash]specific Charpy curves.
Code Case N-749 provides acceptance criteria for flaws in ferritic
components for conditions when the material fracture resistance will be
controlled by upper-shelf toughness behavior. These procedures may be
used to accept a flaw in lieu of the requirements in Section XI,
paragraphs IWB-3610 and IWB-3620, which use LEFM to evaluate flaws that
exceed limits of Section XI, paragraph IWB-3500. Code Case N-749
employs EPFM methods (J[dash]integral) and is patterned after the
fracture methodology and acceptance criteria that currently exist in
Section XI, paragraph IWB-3730(b), and Section XI, Nonmandatory
Appendix K, ``Assessment of Reactor Vessels with Upper Shelf Charpy
Impact Energy Levels.'' The Code Case states that the proposed
methodology is applicable if the metal temperature of the component
exceeds the upper shelf transition temperature, Tc, which is
defined as nil-ductility reference temperature (RTNDT) plus
105 degrees F. The justification for this, as documented in the
underlying White Paper, PVP2012-78190, ``Alternative Acceptance
Criteria for Flaws in Ferritic Steel Components Operating in the Upper
Shelf Temperature Range,'' is that the ASME BPV Code, Section XI,
K1c curve will give a (T- RTNDT) value of 105
degrees F at K1c of 200 ksi[radic]inch.
Defining an upper shelf transition temperature purely based on LEFM
data is not convincing because it ignores EPFM data and Charpy data and
their relationship to the LEFM data. The NRC staff performed
calculations on several randomly selected reactor pressure vessel
surveillance materials with high upper-shelf energy values and low
RTNDT values from three plants and found that using
Tc, as defined in the Code Case, is nonconservative because
at the temperature of RTNDT + 105 degrees F, the Charpy
curves show that most of the materials will not reach their respective
upper-shelf energy levels. The NRC staff's condition is based on a 2015
ASME Pressure Vessels and Piping Conference paper (PVP2015-45307) by
Mark Kirk, Gary Stevens, Marjorie Erickson, William Server, and Hal
Gustin entitled, ``Options for Defining the Upper Shelf Transition
Temperature (Tc) for Ferritic Pressure Vessel Steels,'' where
Tc and Tc1 are defined as the intersections of
specific toughness curves of LEFM data and EPFM data, as shown in that
paper. Using the model in the 2015 PVP paper is justified because, in
addition to its theoretically motivated approach in applying the
temperature-dependent flow behavior of body-centered cubic materials,
the model is also supported by numerous LEFM data and 809 EPFM data in
the upper shelf region.
While the Tc proposed in Code Case N-749 is conservative
based on the intersection of the mean curves of the two sets of data,
the NRC determined that actual or bounding properties (on the
conservative side) should be used instead of mean material properties
for evaluating flaws detected in a ferritic component using the EPFM
approach. This will prevent inaccurate component failure predictions
using the EPFM approach, due to overestimated material properties.
Further, the NRC's approach considers the temperature range for
fracture mode transition between LEFM and EPFM. Based on the previous
discussion, the NRC is imposing a condition on the use of Code Case N-
749 that: (1) The two equations for Tc be used instead of
Tc, as proposed in the Code Case for requiring EPFM
application, when the temperature is above Tc, and (2) the
two equations for Tc1 be used for requiring LEFM application
when temperature is below Tc1. Between Tc1 and
Tc, while the fracture mode is in transition between LEFM
and EPFM, users should consider whether or not it is appropriate to
apply the EPFM method.
Alternatively, the licensee may use a different Tc
value, if it can be justified by plant-specific Charpy curves.
Code Case N-754 [Supplement 6, 2010 Edition]
Type: New.
Title: Optimized Structural Dissimilar Metal Weld Overlay for
Mitigation of PWR Class 1 Items, Section XI, Division 1.
The NRC approves Code Case N-754 with three conditions. Code Case
N-754 provides requirements for installing optimized structural weld
overlays (OWOL) on the outside surface of ASME Class 1 heavy-wall,
large-diameter piping composed of ferritic, austenitic stainless steel,
and nickel based alloy materials in pressurized water reactors
[[Page 2339]]
(PWRs) as a mitigation measure, where no known defect exists or the
defect depth is limited to 50 percent through wall. The upper 25
percent of the original pipe wall thickness is credited as a part of
the OWOL design in the analyses performed, in support of these repairs.
The technical basis supporting the use of OWOLs is provided in the
Electric Power Research Institute (EPRI) Materials Reliability Project
(MRP) Report MRP-169, Revision 1-A, entitled, ``Technical Basis for
Preemptive Weld Overlays for Alloy 82/182 Butt Welds in PWRs.'' By
letter dated August 9, 2010 (ADAMS Accession No. ML101620010), the NRC
informed the Nuclear Energy Institute (NEI) that the staff found that
MRP-169, Revision 1, as revised by letter dated February 3, 2010,
adequately described: (1) Methods for the weld overlay design; (2) the
supporting analyses of the design; (3) the experiments that verified
the analyses; and (4) the inspection requirements of the dissimilar
metal welds to be overlaid. However, the NRC identified the following
conditions.
The first condition requires that the conditions imposed on the use
of OWOLs contained in the NRC final safety evaluation for MRP-169,
Revision 1-A, must be satisfied. Eighteen limitations and conditions
are described in the final safety evaluation that address issues such
as fatigue crack growth rates, piping loads, design life of the weld
overlay, and reexamination frequencies. The imposition of the
conditions in the safety evaluation provide reasonable assurance that
the structural integrity of the pipes repaired through the use of weld
overlays will be maintained.
Code Case N-754 references Code Case N-770-2, ``Alternative
Examination Requirements and Acceptance Standards for Class 1 Pressure
Water Reactor (PWR) Piping and Vessel Nozzle Butt Welds Fabricated With
UNS N06082 or UNS W86182 Weld Filler Material With or Without
Application of Listed Mitigation Activities, Section XI, Division 1.''
The reference to Code Case N-770-2 provides the ASME requirements for
the performance of the preservice and ISI examinations of OWOLs, with
additional requirements if the ultrasonic examination is qualified for
axial flaws. The NRC approved Code Case N-770-2 with conditions in
Sec. 50.55a(g)(6)(ii)(F) on July 18, 2017 (82 FR 32934). Accordingly,
the second condition on the use of Code Case N-754 is that the
preservice and inservice inspections of OWOLs must satisfy Sec.
50.55a(g)(6)(ii)(F), i.e., meet the provisions of Code Case N-770-2.
The third condition addresses a potential implementation issue in
Code Case N-754 with respect to the deposition of the first layer of
weld metal. The second sentence in paragraph 1.2(f)(2) states that
``The first layer of weld metal deposited may not be credited toward
the required thickness, but the presence of this layer shall be
considered in the design analysis requirements in 2(b).'' The NRC found
that, among licensees, there can be various interpretations of the
words used in the ASME BPV Code and Code Cases. In this instance, the
NRC determined that the word ``may'' needed to be changed to ``shall''
in the second sentence in paragraph 1.2(f)(2), as a condition for use
of this Code Case. Accordingly, the NRC is adding a third condition to
clarify that the first layer shall not be credited toward the required
OWOL thickness unless the chromium content of the first layer is at
least 24 percent.
Code Case N-778 [Supplement 6, 2010 Edition]
Type: New.
Title: Alternative Requirements for Preparation and Submittal of
Inservice Inspection Plans, Schedules, and Preservice and Inservice
Summary Reports, Section XI, Division 1.
The NRC is approving Code Case N-778 with two conditions. Section
XI, paragraph IWA-1400(d), in the editions and addenda currently used
by the operating fleet, requires licensees to submit plans, schedules,
and preservice and ISI summary reports to the enforcement and
regulatory authorities having jurisdiction at the plant site. In the
licensees' pursuit to decrease burden, they have alluded to the
resources associated with the requirement to submit the items
previously listed. Code Case N-778 was developed to provide an
alternative to the requirements in the ASME BPV Code, in that the items
previously listed would only have to be submitted if specifically
required by the regulatory and enforcement authorities.
The NRC reviewed its needs with respect to the submittal of the
subject plans, schedules, and reports, and determined that it is not
necessary to require the submittal of plans and schedules. The NRC made
this determination because the latest up[dash]to[dash]date plans and
schedules are available at the plant site and can be requested by the
NRC at any time. However, the NRC determined that summary reports still
need to be submitted. Summary reports provide valuable information
regarding examinations that have been performed, conditions noted
during the examinations, the corrective actions performed, and the
status of the implementation of the ISI program. Accordingly, the NRC
is approving Code Case N-778 with conditions to require that licensees
continue to submit summary reports in accordance with paragraph IWA-
6240 of the 2009 Addenda of ASME Section XI, as addressed below.
The two conditions are modeled on the requirements currently in
paragraph IWA-6240 of the 2009 Addenda, Section XI. The requirements in
Section XI do not specify when the reports are to be submitted to the
regulatory authority; rather, the requirements only state that the
reports shall be completed. The first condition requires that the
preservice inspection summary report be submitted before the date of
placement of the unit into commercial service. The second condition
requires that the ISI summary report be submitted within 90 calendar
days of the completion of each refueling outage. The conditions rely on
the date of commercial service and the completion of a refueling outage
to determine when the reports are needed to be submitted to the
regulatory authority.
Code Case N-789 [Supplement 6, 2010 Edition]
Type: New.
Title: Alternative Requirements for Pad Reinforcement of Class 2
and 3 Moderate[dash]Energy Carbon Steel Piping for Raw Water Service,
Section XI, Division 1.
The NRC is approving Code Case N-789 with one condition. For
certain types of degradation, the Code Case provides requirements for
the temporary repair of degraded moderate energy Class 2 and Class 3
piping systems by external application of welded reinforcement pads.
The Code Case does not require inservice monitoring for the pressure
pad. However, the NRC determined that it is unacceptable to not monitor
the pressure pad because there may be instances where an unexpected
corrosion rate may cause the degraded area in the pipe to expand beyond
the area that is covered by the pressure pad. This could lead to the
pipe leaking and may challenge the structural integrity of the repaired
pipe. Therefore, the NRC is approving Code Case N-789 with a condition
to require a monthly visual examination of the installed pressure pad
for evidence of leakage.
In the proposed rule, the NRC expressed concern that the corrosion
rate specified in paragraph 3.1(1) of the Code Case may not address
certain scenarios. That paragraph would allow
[[Page 2340]]
either a corrosion rate of two times the actual measured corrosion rate
at the reinforcement pad installation location or four times the
estimated maximum corrosion rate for the system. To ensure that a
conservative corrosion rate is used to provide sufficient margin, the
NRC considered adding a second condition that requires that the design
of the pressure pad use the higher of the two corrosion rates
calculated, based on the same degradation mechanism as the degraded
location. However, as a result of a public comment, the NRC
reconsidered and determined that using a corrosion rate of either two
times the actual measured corrosion rate in that location, or four
times the estimated maximum corrosion rate for the system, already
provides a sufficiently conservative estimate of the corrosion rate;
therefore, a condition is not needed.
Code Case N-795 [Supplement 3, 2010 Edition]
Type: New.
Title: Alternative Requirements for BWR Class 1 System Leakage Test
Pressure Following Repair/Replacement Activities, Section XI, Division
1.
The NRC is approving Code Case N-795 with two conditions. The first
condition addresses a prohibition against the production of heat
through the use of a critical reactor core to raise the temperature of
the reactor coolant and pressurize the reactor coolant pressure
boundary (RCPB) (sometimes referred to as nuclear heat). The second
condition addresses the duration of the hold time when testing non-
insulated components to allow potential leakage to manifest itself
during the performance of system leakage tests.
Code Case N-795 was intended to address concerns that performing
the ASME-required pressure test for boiling water reactors (BWRs) under
shutdown conditions, (1) places the unit in a position of significantly
reduced margin, approaching the fracture toughness limits defined in
the Technical Specification Pressure[dash]Temperature (P-T) curves, and
(2) requires abnormal plant conditions/alignments, incurring additional
risks and delays, while providing little added benefit beyond tests,
which could be performed at slightly reduced pressures under normal
plant conditions. However, due to restrictions imposed by the pressure
control systems, most BWRs cannot obtain reactor pressure corresponding
to 100 percent rated power during normal startup operations at low
power levels that would be conducive to performing examinations for
leakage. The alternative test, provided by Code Case N-795, would be
performed at slightly reduced pressures and normal plant conditions,
which the NRC finds will constitute an adequate leak examination and
would reduce the risk associated with abnormal plant conditions and
alignments.
However, the NRC has had a long-standing prohibition against the
production of heat through the use of a critical reactor core to raise
the temperature of the reactor coolant and pressurize the RCPB. A
letter dated February 2, 1990, from James M. Taylor, Executive Director
for Operations, NRC, to Messrs. Nicholas S. Reynolds and Daniel F.
Stenger, Nuclear Utility Backfitting and Reform Group (ADAMS Accession
No. ML14273A002), established the NRC position with respect to use of a
critical reactor core to raise the temperature of the reactor coolant
and pressurize the RCPB. In summary, the NRC's position is that testing
under these conditions involves serious impediments to careful and
complete inspections, and therefore, inherent uncertainty with regard
to assuring the integrity of the RCPB. Further, the practice is not
consistent with basic defense-in-depth safety principles.
The NRC's position established in 1990, was reaffirmed in
Information Notice No. 98-13, ``Post-Refueling Outage Reactor Pressure
Vessel Leakage Testing Before Core Criticality,'' dated April 20, 1998.
The Information Notice was issued in response to a licensee that had
conducted an ASME BPV Code, Section XI, leakage test of the reactor
pressure vessel and subsequently discovered that it had violated 10 CFR
part 50, appendix G, IV.A.2.d. This regulation states that pressure
tests and leak tests of the reactor vessel that are required by Section
XI of the ASME Code must be completed before the core is critical. The
Information Notice references NRC Inspection Report 50-254/97-27 (ADAMS
Accession No. ML15216A276), which documents that licensee personnel
performing VT-2 examinations of the drywell at one BWR plant covered 50
examination areas in 12 minutes, calling into question the adequacy of
the VT-2 examinations.
The bases for the NRC's historical prohibition of pressure testing
with the core critical can be summarized as follows:
1. Nuclear operation of a plant should not commence before
completion of system hydrostatic and leakage testing to verify the
basic integrity of the RCPB, a principal defense-in-depth barrier to
the accidental release of fission products. In accordance with the
defense-in-depth safety precept, the nuclear power plant design
provides for multiple barriers to the accidental release of fission
products from the reactor.
2. Hydrotesting must be done essentially water solid (i.e., free of
pockets of air, steam or other gases) so that stored energy in the
reactor coolant is minimized during a hydrotest or leaktest.
3. The elevated reactor coolant temperatures, associated with
critical operation, result in a severely uncomfortable and difficult
working environment in plant spaces where the system leakage
inspections must be conducted. The greatly increased stored energy in
the reactor coolant, when the reactor is critical, increases the hazard
to personnel and equipment in the event of a leak. As a result, the
ability for plant workers to perform a comprehensive and careful
inspection becomes greatly diminished.
However, the NRC staff has determined that pressure testing with
the core critical is acceptable, if performed after repairs of a
limited scope, where only a few locations or a limited area needs to be
examined, and when ASME Code Section XI, Table IWB-2500-1, Category B-P
(the pressure test required once per cycle of the entire RCPB), has
been recently performed, thus verifying the integrity of the overall
RCPB. The NRC also notes that Code Case N-795 does not allow for the
use of the alternative test pressure following repairs/replacements on
the RPV, therefore it does not violate 10 CFR part 50, Appendix G. The
NRC determined that the risk associated with nuclear heat at low power
is comparable with the risk to the plant, when the test is performed
without nuclear heat (with the core subcritical) during mid-cycle
outages, when decay heat must be managed. Performing the pressure test
under shutdown conditions at full operating pressure without nuclear
heat requires securing certain key pressure control, heat removal, and
safety systems. Under such conditions, it is more difficult to control
temperature and pressure, when there is significant decay heat
production, such as after a mid-cycle outage, which may reduce the
margin available to prevent exceeding the plant pressure-temperature
limits.
The scope of repairs should be relatively small, when the pressure
test is conducted using nuclear heat, in order to minimize the
personnel safety risk and to avoid rushed examinations. Code Case N-795
does not place any restrictions on the size or scope of the repairs for
which the alternative may be used, other than the alternative test
pressure may not be used to satisfy
[[Page 2341]]
pressure test requirements following repair/replacement activities on
the reactor vessel. It is impractical to specify a particular number of
welded or mechanical repairs that would constitute a ``limited scope.''
However, if the plant is still in a refueling outage and has already
performed the ASME Section XI Category B-P pressure test of the entire
RCPB, it is likely that subsequent repairs would be performed only on
an emergent basis, and would generally be of a limited scope.
Additionally, the overall integrity of the RCPB will have been recently
confirmed via the Category B-P test. For mid-cycle maintenance outages,
the first condition allows the use of nuclear heat to perform the test,
if the outage duration is fourteen (14) days or less. This would tend
to limit the scope of repairs, and also limit use of the Code Case to
outages when decay heat was a significant problem. Therefore, the first
condition on Code Case N-795 states:
``The use of nuclear heat to conduct the BWR Class 1 system leakage
test is prohibited (i.e., the reactor must be in a non-critical
state), except during refueling outages in which the ASME Section XI
Category B-P pressure test has already been performed, or at the end
of mid-cycle maintenance outages fourteen (14) days or less in
duration.''
With respect to the second condition and adequate pressure test
hold time, the technical analysis supporting Code Case N-795 indicates
that the lower test pressure provides more than 90 percent of the flow,
which would result from the pressure corresponding to 100 percent
power. However, a reduced pressure means a lower leakage rate, so
additional time is required in order for there to be sufficient leakage
to be observed by inspection personnel. Section XI, paragraph IWA-5213,
``Test Condition Holding Time,'' does not require a holding time for
Class 1 components, once test pressure is obtained. To account for the
reduced pressure, Code Case N-795 would require a 15-minute hold time
for non-insulated components. The NRC has determined that 15 minutes
does not allow for an adequate examination, because it is not possible
to predict the entire range of scenarios or types of defects that could
result in leakage. While some types of defects could result in
immediate leakage, such as an improperly torqued bolted connection;
other types of defects, such as weld defects or tight cracks could
represent a more torturous path for leakage and may result in delayed
leakage. The staff determined that, due to the uncertainty in the time
required for leakage to occur to an extent, it would be readily
detectable by visual examination, hence, it is appropriate to
conservatively specify a longer hold time of 1 hour for non-insulated
components. Therefore, the final rule retains the one hour hold time
for non-insulated components.
Code Case N-799 [Supplement 4, 2010 Edition]
Type: New.
Title: Dissimilar Metal Welds Joining Vessel Nozzles to Components,
Section XI, Division 1.
The NRC approves Code Case N-799 with four conditions. Code Case N-
799 is a new Code Case developed to provide examination requirements
for the steam generator primary nozzle to pump casing attachment weld
for AP-1000 plants and dissimilar metal welds joining vessel nozzles to
pumps used in recent reactor designs (e.g., AP-1000, Advanced BWR).
Nuclear power plant pump casings are typically manufactured from cast
austenitic stainless steel (CASS) materials. The NRC is approving the
Code Case with conditions to address the shortcomings in the Code Case
with respect to requirements for ultrasonic examination.
The CASS is an anisotropic and inhomogeneous material. The
manufacturing process can result in varied and mixed structures. The
large size of the anisotropic grains affects the propagation of
ultrasound by causing severe attenuation, changes in velocity, and
scattering of ultrasonic energy. Refraction and reflection of the sound
beam occurs at the grain boundaries, which can result in specific
volumes of material not being examined, or defects being missed or
mischaracterized. The grain structure of the associated weldments also
impacts the effectiveness and reliability of the examinations.
Accordingly, it is paramount that robust examination techniques be
used.
Research has been conducted by several domestic and international
organizations attempting to address the shortcomings associated with
the use of conventional methods for the inspection of CASS materials.
The results of a study at Pacific Northwest National Laboratory (PNNL)
were published in NUREG/CR-6933, ``Assessment of Crack Detection in
Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-
Frequency Ultrasonic Methods'' (ADAMS Accession No. ML071020409). The
study demonstrated that additional measures were required to reliably
detect and characterize flaws in CASS materials and their associated
weldments.
Performance demonstration requirements for CASS components and
associated weldments have not yet been developed by the industry. To
ensure that effective and reliable examinations are performed, the NRC
is adopting the following four conditions on the Code Case.
The first condition addresses the gap between the probe and
component surface. Industry experience shows that effective ultrasonic
examinations depend, to a great extent, on limiting the gap between the
probe and component surface to less than 0.032[dash]inch. The BPV Code
does not have any requirements with respect to surface smoothness and
waviness. It has been demonstrated that reduced coupling and probe
lift-off on ``rough'' surfaces have the potential to present a
scattering effect at an interface where an acoustic beam impinges, to
redirect and mode convert some energy, which when returned to the probe
can be the source of spurious signals, or cause flaws to be mis-
characterized or missed altogether. Accordingly, the first condition
requires that the scanning surfaces have a gap less than 0.032-inch
beneath the ultrasonic testing probe. Gaps greater than 0.032[dash]inch
must be considered to be unexamined, unless it can be demonstrated, on
representative mockups, that a Section XI, Appendix VIII, Supplement
10, demonstration can be passed.
The second condition (No. 2a in DG-1296) is that the examination
requirements of Section XI, Mandatory Appendix I, paragraph I-3200(c)
must be applied. Code Case N-799 does not contain specific requirements
regarding examination techniques. Paragraph I-3200(c) contains specific
requirements that can be applied.
The third condition (No. 2c in DG-1296) is that ultrasonic depth
and sizing qualifications for CASS components must use the ASME BPV
Code requirements in Section XI, Appendix VIII, Supplement 10.
Supplement 10 contains qualification requirements for dissimilar metal
welds, and the use of these requirements will ensure that robust
techniques are applied.
The fourth condition (No. 2e in DG-1296) is that cracks that are
detected but cannot be depth-sized with performance-based procedures,
equipment, and personnel qualifications consistent with ASME Code
Section XI, Appendix VIII, shall be repaired or removed.
OM Code Cases (RG 1.192)
Code Case OMN-1, Revision 1 [2012 Edition]
Type: Revised.
Title: Alternative Rules for Preservice and Inservice Testing of
Active Electric
[[Page 2342]]
Motor[dash]Operated Valve Assemblies in Light-Water Reactor Power
Plants.
The conditions on Code Case OMN-1, Revision 1 [2012 Edition] are
identical to the conditions on OMN-1 [2006 Addenda] that were approved
by the NRC in Revision 1 of RG 1.192 in October 2014. The reasons for
imposing these conditions are not resolved by Code Case OMN-1, Revision
1 [2012 Edition] and, therefore, these conditions have been retained in
Revision 2 of RG 1.192.
Code Case OMN-3 [2012 Edition]
Type: Reaffirmed.
Title: Requirements for Safety Significance Categorization of
Components Using Risk Insights for Inservice Testing of LWR Power
Plants.
The conditions on Code Case OMN-3 [2012 Edition] are identical to
the conditions on OMN-3 [2004 Edition] that were approved by the NRC in
Revision 1 of RG 1.192 in October 2014. The reasons for imposing these
conditions are not resolved by Code Case OMN-3 [2012 Edition] and,
therefore, these conditions have been retained in Revision 2 of RG
1.192.
Code Case OMN-4 [2012 Edition]
Type: Reaffirmed.
Title: Requirements for Risk Insights for Inservice Testing of
Check Valves at LWR Power Plants.
The conditions on Code Case OMN-4 [2012 Edition] are identical to
the conditions on OMN-4 [2004 Edition] that were approved by the NRC in
Revision 1 of RG 1.192 in October 2014. The reasons for imposing these
conditions are not resolved by Code Case OMN-4 [2012 Edition] and,
therefore, these conditions have been retained in Revision 2 of RG
1.192.
Code Case OMN-9 [2012 Edition]
Type: Reaffirmed.
Title: Use of a Pump Curve for Testing.
The conditions on Code Case OMN-9 [2012 Edition] are identical to
the conditions on OMN-9 [2004 Edition] that were approved by the NRC in
Revision 1 of RG 1.192 in October 2014. The reasons for imposing these
conditions are not resolved by Code Case OMN-9 [2012 Edition] and,
therefore, these conditions have been retained in Revision 2 of RG
1.192.
Code Case OMN-12 [2012 Edition]
Type: Reaffirmed.
Title: Alternative Requirements for Inservice Testing Using Risk
Insights for Pneumatically and Hydraulically Operated Valve Assemblies
in Light-Water Reactor Power Plants (OM-Code 1998, Subsection ISTC).
The conditions on Code Case OMN-12 [2012 Edition] are identical to
the conditions on OMN-12 [2004 Edition] that were approved by the NRC
in Revision 1 of RG 1.192 in October 2014. The reasons for imposing
these conditions are not resolved by Code Case OMN-12 [2012 Edition]
and, therefore, these conditions have been retained in Revision 2 of RG
1.192.
Code Case OMN-16, Revision 1 [2012 Edition]
Type: Revised.
Title: Use of a Pump Curve for Testing.
Code Case OMN-16, 2006 Addenda, was approved by the NRC in
Regulatory Guide 1.192, Revision 1. With respect to Code Case OMN-16,
Revision 1, 2012 Edition, there was an editorial error in the
publishing of this Code Case in that Figure 1 from the original Code
Case (i.e., Rev. 0, 2006 Addenda) was omitted. Accordingly, the NRC
approves OMN-16, Revision 1, with a condition requiring that Figure 1
from the original Code Case be used when implementing OMN-16, Revision
1.
Code Case OMN-18 [2012 Edition]
Type: Reaffirmed.
Title: Alternate Testing Requirements for Pumps Tested Quarterly
Within 20% of Design Flow.
The ASME OM Code defines Group A pumps as those pumps that are
operated continuously or routinely during normal operation, cold
shutdown, or refueling operations. The OM Code specifies that each
Group A pump undergoes a Group A test quarterly and a comprehensive
test biennially. The OM Code requires that the reference value for a
comprehensive test to be within 20 percent of pump design flow, while
the reference value for a Group A test needs to be within 20 percent of
the pump design flow, if practicable. The biennial comprehensive test
was developed (first appeared in the 1995 Edition of the OM Code)
because pump performance concerns demonstrated that more stringent
periodic testing was needed at a flow rate within a more reasonable
range of the pump design flow rate, than typically performed during the
pump IST in the past.
Currently, when performing either the quarterly Group A test or the
biennial comprehensive pump test, licensees must comply with certain
limits for the flow Acceptable Range, the flow Required Action Range,
the differential pressure (or discharge pressure) Acceptable Range, and
the differential pressure (or discharge pressure) Required Action
Range. The limits for the quarterly Group A test are obtained by using
a factor of 1.10 times the flow reference value (Qr) or the
differential or discharge pressure reference value
([Delta]Pr or Pr), as applicable to the pump
type. The limits for the biennial comprehensive pump test
are obtained by using the factor of 1.03 times Qr or
[Delta]Pr (or Pr), as applicable to the pump
type, providing more restrictive test ranges and higher quality data.
Code Case OMN-18, 2012 Edition, would remove the Code requirement
to perform a biennial comprehensive pump test, where the quarterly
Group A pump test is performed within 20 percent of the
pump design flow rate, with instruments having the ability to obtain
the accuracies required for the comprehensive pump test. The NRC finds
the performance of a quarterly Group A pump test, at flow within 20 percent of the pump design flow rate, will be sufficient to
detect mechanical and hydraulic degradation of the tested pump. The NRC
finds that this will satisfy the intent of the biennial comprehensive
pump test, with the exception that the test acceptable ranges and
required action ranges are less precise than required for the
comprehensive test. Therefore, the NRC approves Code Case OMN-18, 2012
Edition, with a condition to specify the use of a factor of 1.06 for
the Group A test parameters, to be consistent with the test ranges for
the comprehensive test. The NRC concludes that the factor of 1.06 will
provide a reasonable test range, when applying Code Case OMN-18 to
Group A pumps tested quarterly, within 20 percent of the
pump design flow rate. The NRC finds that the quarterly Group A test
for pumps within 20 percent of the pump design flow rate,
combined with the provisions in the Code Case OMN-18 for the pump
instrumentation and the conditions in RG 1.192 for the test ranges,
will provide reasonable assurance of the operational readiness of these
pumps, as an acceptable alternative to the comprehensive pump test
provisions in the ASME OM Code.
Code Case OMN-19 [2012 Edition]
Type: Reaffirmed.
Title: Alternative Upper Limit for the Comprehensive Pump Test.
A requirement for a periodic pump verification test was added in
Mandatory Appendix V, ``Pump Periodic Verification Test Program,'' to
the 2012 Edition of the OM Code. The mandatory appendix is based on the
determination by the ASME that a pump periodic verification test is
needed to confirm that a pump can meet the required (differential or
discharge) pressure as applicable, at its highest
[[Page 2343]]
design basis accident flow rate. Code Case OMN-19, 2012 Edition, would
allow an applicant or licensee to use a multiplier of 1.06 times the
reference value in lieu of the 1.03 multiplier for the comprehensive
pump test's upper Acceptable Range criteria and Required Action Range,
High criteria reference in the ISTB test acceptance criteria tables.
The NRC considers Code Case OMN-19 to be acceptable where the
provisions of Appendix V for a pump periodic verification test as
referenced by ISTB-1400 are also satisfied to detect mechanical and
hydraulic degradation. Therefore, the NRC approves Code Case OMN-19,
2012 Edition, with the condition that the provisions in paragraph ISTB-
1400 and Mandatory Appendix V be applied when implementing the Code
Case.
Code Case OMN-20 [2012 Edition]
Type: New.
Title: Inservice Testing Frequency.
Surveillance Requirement (SR) 3.0.3 from Technical Specification
(TS) 5.5.6, ``Inservice Testing Program,'' allows licensees to apply a
delay period before declaring the SR for TS equipment ``not met,'' if a
licensee inadvertently exceeds or misses the time limit for performing
the TS surveillance. Licensees have been applying SR 3.0.3 to inservice
tests performed in accordance with the ASME Codes. The NRC has
determined that licensees cannot use TS 5.5.6 to apply SR 3.0.3 to
inservice tests under Sec. 50.55a(f) that are not associated with a TS
surveillance. To invoke SR 3.0.3, the licensee must first discover that
a TS surveillance was not performed at its specified frequency.
Therefore, the delay period that SR 3.0.3 provides does not apply to
non[dash]TS support components tested under Sec. 50.55a(f). The OM
Code does not provide for inservice test frequency reductions or
extensions. In order to provide inservice test frequency reductions or
extensions that cannot be provided by SR 3.0.3 from TS 5.5.6, ASME
developed OM Code Case OMN-20. The NRC has reviewed OM Code Case OMN-20
and has found it acceptable for use. The NRC determined that OM Code
Case OMN-20 may be applied to editions and addenda of the OM Code that
are listed in Sec. 50.55a(a)(1)(iv). Therefore, the NRC has included a
condition in RG 1.192, specifying that Code Case OMN-20 is applicable
to editions and addenda of the OM Code listed in Sec.
50.55a(a)(1)(iv).
C. ASME Code Cases Not Approved for Use (RG 1.193)
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 the 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[dash]temperature gas[dash]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; RG 1.193 confirms the
Code Cases that are not approved for use. The NRC is not adopting any
of the Code Cases listed in RG 1.193.
III. Opportunities for Public Participation
The proposed rule and draft RGs were published in the Federal
Register on March 2, 2016 (81 FR 10780), for a 75-day comment period.
The public comment period closed on May 16, 2016.
After the close of the public comment period, the NRC held a public
meeting on August 22, 2016, to discuss the status of this proposed
rule. The public meeting summary is available in ADAMS under Accession
No. ML16265A001.
IV. Public Comment Analysis
The NRC received a total of seven comment submissions on the
proposed rule and draft RGs. Table III lists the commenters, their
affiliation, and the ADAMS Accession Number for each submission.
Table III--Comment Submissions Received on the Proposed Rule and Draft RGs
----------------------------------------------------------------------------------------------------------------
ADAMS accession
Submission ID Commenter name Affiliation No.
----------------------------------------------------------------------------------------------------------------
1...................................... Paul Donavin............. Private Citizen.......... ML16063A509
2...................................... Gregory Frederick and Dan Electric Power Research ML16126A524
Patten. Institute.
3...................................... Anonymous................ Unknown.................. ML16133A422
4...................................... Charles Pierce........... Southern Nuclear ML16137A857
Operating Company.
5...................................... Ralph Hill III........... ASME..................... ML16138A835
6...................................... Mark Gowin............... Private Citizen.......... ML16139A798
7...................................... David Helker............. Exelon Generation ML16153A432
Company, LLC.
----------------------------------------------------------------------------------------------------------------
The NRC reviewed every comment submission and identified 32 unique
comments requiring the NRC's consideration and response. Comment
summaries and the NRC's responses are presented in this section. At the
end of each summary, the individual comments represented by the summary
are identified in the form [XX-YY] where XX represents the Submission
ID in Table III and YY represents the sequential comment within the
submission.
Public Comments on Draft Regulatory Guides
Regulatory Guide 1.84, Revision 37 (DG-1295)
No public comments were submitted regarding Regulatory Guide 1.84,
Revision 37 (Draft Guide (DG)-1295), therefore no NRC response is
needed.
Regulatory Guide 1.147, Revision 18 (DG-1296)
Code Case N-552-1
Comment: The proposed conditions on N-552-1 were incorporated into
the ASME BPV Code, Section XI, 2005 Addenda when Code Case N-552 was
incorporated into the code. However, these conditions have never been
incorporated into the Code Case itself. The proposed conditions are
identical to those imposed on Code Case N-552 in Revision 16 of RG
1.147. ASME does not object to these conditions. [ASME 5-2]
NRC Response: The NRC agrees with this comment.
No change was made to the final rule as a result of this comment.
Code Case N-576-2
Comment: Because the NRC has adopted the 2008 Addenda with no
conditions on IWA-4200, ASME recommends that the proposed
[[Page 2344]]
condition be revised to state ``. . . is to be performed in accordance
with IWA-4200 of the code of record for the current ISI Program.''
[ASME 5-3]
NRC Response: The NRC agrees, in part, with this comment. The NRC
staff has adopted the 2008 Addenda with no conditions on IWA-4200.
However, the staff does not agree that the proposed condition/note in
Regulatory Guide 1.147 should be revised to state ``. . . is to be
performed in accordance with IWA-4200 of the code of record for the
current ISI program'', because there may be licensees whose code of
record is prior to 2008 and such a condition is not necessary because
licensees would be required to follow IWA-4200 in their code of record,
if they were to adopt this Code Case. As a result, because use of the
repair method described in this Code Case (N-576-2) requires the NRC's
review and approval prior to implementation and licensees will be
required to follow IWA-4200 in their code of record, the NRC modified
the ``note'' on this Code Case to eliminate the portion of the ``note''
regarding reconciliation. The revised ``note'' now reads:
``Note: Steam generator tube repair methods require prior NRC
approval through the Technical Specifications. This Code Case does not
address certain aspects of this repair, e.g., the qualification of the
inspection and plugging criteria necessary for staff approval of the
repair method.''
Code Case N-638-6
Comment: Condition 1 was incorporated into IWA-4673(a)(2) of the
2013 Edition when N-638-6 was incorporated into the Code. This
condition has also been incorporated into N-638-8, which has been
published in the 2015 Code Case Book. Condition (2) was incorporated
into IWA-4671(b)(1) of the 2013 Edition when N-638-6 was incorporated
into the Code. Because there were no conditions imposed on the use of
IWA-4673(a)(2) or IWA-4671(b)(1) in the draft rule, to incorporate by
reference the 2013 Edition of the ASME BPV Code, Section XI, ASME
recommends that both of the proposed conditions be removed and Code
Case N-638-6 be moved to Table 1 of RG 1.147, Revision 18. [ASME 5-4]
NRC Response: The NRC agrees, in part, with this comment. Regarding
proposed Condition (1), the staff agrees that Condition (1) was
incorporated into IWA-4673(a)(2) of the 2013 Edition of ASME BPV Code,
Section XI, when ASME incorporated Code Case N-638-6 into the Section
XI. Proposed Condition (1) was also addressed in Code Case N-638-8.
However, Code Case N-638-6 does not address proposed Condition (1) and
this version of the Code Case will be available for use by licensees
who will not adopt the 2013 Edition of Section XI for several years.
Therefore, the NRC determined that it is appropriate to include
proposed Condition (1) in RG 1.147, Revision 18.
Regarding proposed Condition (2), Paragraph 1(b)(1) of Code Case N-
638-6 contains changes from the previous version of the Code Case,
which allows through-wall circumferential welds and includes additional
requirements when performing repairs that utilize through-wall
circumferential welds. At the time that this revision of the Code Case
was approved by the ASME, the staff had concerns related to through-
wall repairs. Subsequently, the NRC resolved its concerns. Therefore,
the NRC determined that proposed Condition (2) is unnecessary.
The NRC has removed proposed Condition (2) on Code Case N-638-6
from the final RG 1.147, Revision 18.
No change was made to the final rule as a result of this comment.
Code Cases N-666 and N-666-1
Comment: A new condition has been added to N-666, which is listed
as a Superseded Code Case: A surface (magnetic particle or liquid
penetrant) examination must be performed after installing the seal weld
and weld overlay on Class 1 and 2 piping socket welds. The fabrication
defects, if detected, must be dispositioned using the surface
examination acceptance criteria of the Construction Code identified in
the Repair/Replacement Plan.
As stated in our comment on N-666-1, the phrase ``seal weld and''
should be removed from the first sentence. Also, the addition of a new
condition to a Code Case that was previously unconditionally approved
in the Reg. Guide, and is now superseded, seems inappropriate. Several
plants would likely have this version of the Code Case in their Section
XI ``tool box'' until the end of their current Inspection Interval, and
would be apparently (but not obviously) bound by the new condition,
upon issuance of the new revision to Regulatory Guide. The third
paragraph under Section B. DISCUSSION, in the draft RG, includes the
statement ``If a Code Case is implemented by a licensee 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 ISI 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 the Code
Case that is necessary, for example, to enhance safety.'' Perhaps this
could be supplemented with another sentence such as, ``In this case,
the condition will be entered for the superseded Code Case under Table
5.'' [EPRI 2-4, Exelon 7-4]
NRC Response: The NRC agrees with this comment. The condition shown
in Table 5 of DG-1295 for Code Case N-666 was in error.
The condition on Code Case N-666 in Table 5 from the final RG
1.147, Revision 18 has been removed.
No change was made to the final rule as a result of this comment.
Comment: Condition 1--The construction code may not always require
a surface examination (depending on the construction code) on socket
welds. This condition is appropriate. However, the words ``and seal
weld'' in the first sentence should be removed from the condition
because it is inappropriate to require surface examination of non-
structural seal welds whose only function is to seal a leak. The ASME
recommends revising this condition to remove the words ``and seal
weld'' in the first sentence. Condition 2--This condition should be
removed as 5(a)(1) already required a Visual VT-1 examination of
completed weld overlays irrespective of the class of the joint. This
condition is redundant and only causes confusion. ASME recommends
removing this proposed condition. [EPRI 2-1, ASME 5-5]
NRC Response: The NRC agrees with this comment. The function of the
seal weld is to seal a leak so that sound weldment for the overlay can
be applied. The code case requires a visual examination of the seal
weld, remaining socket weld, and adjacent base material before the weld
overlay can be applied, which the NRC has determined is the appropriate
examination prior to the application of the weld overlay. Therefore,
Condition 1 has been revised to remove ``and seal weld.'' Regarding
Condition 2, the NRC agrees with the commenter. The code case requires
a visual examination of the seal pass and the completed weld overlay
and provides appropriate acceptance criteria. Therefore, the condition
is redundant and unnecessary. Condition 2 has been removed from Code
Case N-666 in Table 2 from the final RG 1.147, Revision 18.
No change was made to the final rule as a result of this comment.
[[Page 2345]]
Code Case N-711
Comment: ASME recommends that this Code Case N-711 be removed from
RG 1.193, Table 2 and added to Table 2 of RG 1.147 with appropriate
conditions to address NRC technical concerns with the use of this case.
[ASME 5-10]
NRC Response: The NRC disagrees with this comment. The NRC declines
at this time to adopt the recommended changes to the regulatory guides.
It would not be appropriate to include the Code Case in RG 1.147
without first having sought public comment on the adoption of the Code
Case. Nonetheless, the NRC has reviewed the information provided by
ASME and will consider approval of the Code Case in future rulemaking
activities.
No change was made to the final rule as a result of this comment.
Code Case N-722-2
Comment: ASME requests that the NRC identify any technical concerns
with N-722-2 and list these concerns in R.G. 1.193, Table 2. [ASME 5-
11]
NRC Response: The NRC disagrees with this comment. The NRC
disagrees with the comment because the NRC does not provide comments in
the Regulatory Guide 1.193 on ASME Code Cases, which the NRC mandates
for use as augmented inservice inspection programs under Sec.
50.55a(g)(6)(ii). Any conditions that the NRC finds necessary to
require are included under the particular section of Sec.
50.55a(g)(6)(ii)(D), (E) or (F), as applicable. This is to avoid
confusion such that a stakeholder does not use versions of these ASME
Code Cases in lieu of the mandated versions of the ASME Code Case in
Sec. 50.55a(g)(6)(ii). However, in order to be responsive to the
stakeholder comment, the NRC will provide the current concerns with the
implementation of ASME Code Case N-722-2, as a response to this comment
to be included in the Federal Register notice.
The NRC currently finds ASME Code Case N-722-2 unacceptable as
written due to the following main issues. First, the basis for the
removal of the Parts Examined from N-722-1 was found to be in error.
According to an ASME Code interpretation, XI-1-13-27, not all items
removed in N-722-2 were covered by the inspection requirements of ASME
Code Case N-770-1. The ASME Code Case N-722 will need to be revised
with a new basis for the removal of Parts Examined to be considered for
approval by the NRC. Second, Note 11 is not acceptable. The bases for
this concern is the same basis as Sec. 50.55a(g)(6)(ii)(F)(2), which
restricts the application of this material condition to exempt
volumetric and visual examination requirements in N-770-1. The NRC is
concerned that the wording of this exemption may allow insufficiently
mitigated items to be exempt from currently required visual inspection
requirements for components containing alloy 600/82/182 to maintain
structural and leak-tight integrity. Once again though, it is not the
intent of the NRC to include these items as conditions or limitations
in the regulatory guide. The current wording to redirect the user to
the applicable section of Sec. 50.55a(g)(6)(ii)(E) will remain,
because versions of this ASME Code Case, as well as N-729 and N-770,
are not alternatives to the Code requirements, but are mandated by
Sec. 50.55a as augmented ISI requirements. For these reasons the NRC
disagrees with the comment.
No change was made to the final rule as a result of this comment.
Code Case N-749
Comment: Public comment 5-6 raised two main points:
1. The comment takes issue with the temperature, Tc,
above which the staff suggests that EPFM techniques should be used. The
formula for Tc, given in the staff's condition, differs from
that proposed in Code Case N-749.
2. The comment takes issue with the part of the staff's condition
stating that ``Tc is the temperature above which elastic plastic
fracture mechanics (EPFM) must be applied.'' Item 4 of the public
comment suggests adopting a permissive rather than a perspective
condition by replacing the word ``must'' with the word ``may'' in the
preceding sentence. [ASME 5-6]
NRC Response: The NRC disagrees with this comment. The staff's
responses to these points are, as follows:
Concerning point 1, the technical bases for the staff's proposed
equation for Tc are well documented, as discussed
previously, and are well supported by data for RPV steels both before
and after neutron irradiation. This documentation appears in PVP 2015-
45307. Conversely, the Tc equation in the proposed Code Case
relates only to the intersection of the ASME KIc curve with
a fracture toughness (KIc) value of 220 MPa[radic]m, a value
that does not correspond well to any known materials data and,
moreover, does not account for the effects of irradiation
embrittlement. The NRC staff's proposal for Tc is thus
better supported by materials data than is the Code Case value.
Concerning point 2, in order for a permissive condition to be
acceptable (e.g., the use of ``may''), it would need to be demonstrated
that application of LEFM approaches to flaw assessment on the upper
shelf fracture behavior is always conservative relative to the more
technically correct EPFM approach. This has not been demonstrated in
either Code Case N-749 or in its supporting technical basis document.
As one example, an approach to using LEFM on the upper shelf fracture
behavior would be to continue to use the ASME KIc curve. At
upper shelf temperatures, the KIc curve over-estimates the
fracture toughness relative to the ductile fracture toughness (i.e.,
J0.1 or J-R), which is non-conservative.
No change was made to the final rule as a result of this comment.
Code Case N-754
Comment: The third condition proposed for this Code Case inversely
paraphrases existing statements in the Code Case, causing confusion to
the user as to what the condition actually adds to the existing
requirements. Further, by paraphrasing the requirements, essential
technical requirements, such as chrome content in the dilution zone,
are omitted which we do not believe is the intent of the condition. The
Federal Register states that the reason for this condition is that ``In
this instance, the NRC felt the word ``may'' needed to be changed to
``shall'' in the second sentence in paragraph 1.2(f)(2) as a condition
for use of this Code Case.'' In the English language, when the term
``may'' is followed by the word ``not'', the phrase means the same as
``shall not.'' However, if this phrase is truly a concern for some,
then the condition should be written exactly as the Code Case except
change the one word ``may'' to ``shall.'' [EPRI 2-2, ASME 5-7]
NRC Response: The NRC disagrees with this comment. Condition (3)
addresses the following two statements in Paragraph 1.2(f)(2) of Code
Case N-754 that reads: ``. . . The first layer of weld metal deposited
may not be credited toward the required thickness, but the presence of
this layer shall be considered in the design analysis requirements in
2(b). Alternatively, a first diluted layer may be credited toward the
required thickness, provided the layer and the associated dilution zone
contain at least 24% Cr [chromium] . . .'' The first sentence in
Paragraph 1.2(f)(2) could be interpreted so that the first weld layer
could be credited toward the required thickness because the word ``may
not'' does not absolutely prohibit such action. In addition, the first
sentence in the quoted statements does not have restriction on
[[Page 2346]]
the chromium contents for crediting the first weld layer toward the
required thickness.
The second sentence in the above quote limits the chromium content
of at least 24 percent; however, the second sentence began with the
word ``Alternatively.'' The word ``Alternatively'' implies that the
requirement in the second sentence is optional, i.e., a licensee may
choose to satisfy either the first sentence or the second sentence, but
the licensee does not need to satisfy both.
For example, a licensee deposits a first weld layer that contains
less than 24 percent chromium. The licensee could consider the first
layer, as part of the required weld overlay thickness, based on the
first sentence above because the first sentence does not identify a
specific chromium content. Therefore, it does not restrict the
consideration of the first layer for the required weld overlay
thickness. The second sentence in the above quote does require the
chromium content to be at least 24 percent. However, the licensee could
interpret that the second sentence does not apply to this case because
the second sentence is an alternate, optional requirement based on the
word ``Alternatively.''
The staff finds that Condition (3) does not omit the essential
technical requirements such as the chrome content in the dilution zone.
Condition (3) requires that if the first weld layer cannot achieve a
chromium content of at least 24 percent, it cannot be considered as
part of the weld overlay thickness. The staff recognizes that Condition
(3) provides the same requirements as in Paragraph 1.2(f)(2). However,
the purpose of Condition (3) is to clarify the requirements in
Paragraph 1.2(f)(2).
No change was made to the final rule as a result of this comment.
Code Case N-784
Comment: This Code Case enables personnel to receive credit for
experience hours for laboratory practice beyond the required number of
hours of laboratory training. For Level II certification, the total
experience hours may be reduced from 800 to 400 if the experience
consists of a combination of 80 hours of field experience and 320 hours
laboratory practice by scanning specimens containing flaws in materials
representative of those in actual power plant components. The field
experience will likely be in nuclear plants but there is no requirement
for UT examiners to obtain their experience in a nuclear plant. While
the experience credited would be on samples and mockups, those samples
would be required to contain actual flaws whereas over many hours of
field experience, fewer flaws may be encountered. Further, to ensure
the effectiveness of the laboratory practice, the Level II experience
time would be credited only after the individual passed an Appendix
VIII, Supplement 2 performance demonstration for length and depth
sizing. Since other performance demonstrations are required for
certification for vessels, ferritic piping and bolting, for example, it
is considered reasonable to only require the Supplement 2 performance
demonstration as a threshold for crediting the laboratory practice
hours. EPRI will provide reports (Nondestructive Evaluation: Fast-Track
NDE Work Force Enhancement, Volume 1; 1019119 and Nondestructive
Evaluation: Fast-Track NDE Work Force Enhancement, Volume 2, 1021150)
to the USNRC to support this Code Case and address the impact of the
reduced experience. This case does not reduce the training hours. [ASME
5-12]
NRC Response: The NRC disagrees with this comment. The ASME BPV
Code replaces field experience with training hours without a defined
technical basis. While the NRC is open to evidence related to a
technical basis for the substitution of laboratory experience as a
substitute for hours of work experience, the impact of the substitution
of laboratory hours for field experience and nuclear power plant
familiarization is unknown. The two documents cited in the comment
require 1,050 hours of hands-on practice with hundreds of hours of
additional classwork, not only 320 hours of laboratory training. If
future work showed that 320 hours would be sufficient or the Code Case
was modified to be in line with these documents, the NRC would consider
allowing the use of the Code Case.
No change was made to the final rule as a result of this comment.
Code Case N-789
Comment: The NRC Condition [2] does not allow the user to apply the
actual corrosion rate for the pressure pad design. This reflects the
staff position that the factors of 2 and 4 do not provide reasonable
assurance that actual corrosion rate is bounded. However, the
compensatory measures of inservice monitoring and the short acceptance
period of one operating cycle verify and provide assurance that both
structural and leak integrity will be maintained during the temporary
acceptance period. Condition (2) is contrary to several NRC SERs that
have evaluated and approved the Code Case for application at dozens of
domestic plants. Those SERs require that the reinforcing pad be
designed to accommodate twice the actual measured corrosion rate or if
unknown, then 4 times the maximum experienced in that or a similar
system at the same plant for the same degradation mechanism. Corrosion
rates are dependent upon many system variables--one primary factor
being the amount and frequency of fluid flow. To impose the rate that
may occur on a seldom-used dead-leg of a system to an area of active
flow, where the actual corrosion rate has been measured is technically
inappropriate. Since the monthly monitoring imposed by Condition (1)
was initiated for the same reason that this condition was proposed--
namely, the potential for an unexpected corrosion rate--this condition
should be removed. [EPRI 2-3, ASME 5-8]
NRC Response: The NRC agrees with this comment. The NRC determined
that the current language in the Code Case, which requires using a
corrosion rate of either two times the actual measured corrosion rate
in that location, or four times the estimated maximum corrosion rate
for the system, is reasonable and provides a conservative estimate of
the corrosion rate. This conservatively estimated corrosion rate,
coupled with proposed Condition (1) that requires enhanced inservice
monitoring, provides reasonable assurance that should corrosion rates
be more aggressive than originally predicted, there will be sufficient
time to initiate corrective actions prior to excessive leakage or loss
of structural integrity. Therefore, the NRC has determined that
proposed Condition (2) is not necessary.
The NRC has removed proposed Condition (2) on Code Case N-789 from
the final RG 1.147, Revision 18.
No change was made to the final rule as a result of this comment.
Comment: Paragraph 3.2(i) of Code Case N-789 has a typographic
error where it states ``. . . piping designed to NC-2650, ND-3650. . .
.'' NC-2650 should be NC-3650. Code Case N-789-2 corrected this
statement to read ``. . . piping designed to NC-3650 or ND-3650. . .
.'' The use of this Code Case N-789 should be conditioned to require
using the corrected language for paragraph 3.2(i) in N-789-2.
[Anonymous 3-1, Exelon 7-1]
NRC Response: The NRC agrees with the commenter. Code Case N-789
Paragraph 3.2(i) contains a typographical error. The code case
references NC-2650 and the correct reference is NC-3650. NC-2650 does
not exist in ASME Code Section III and
[[Page 2347]]
NC-3650 is the correct portion of the Code to use for the design of
reinforcing pads. The NRC does not believe that this typographical
error represents a safety concern. In order to prevent the delay of
issuance of the final rule by including a new condition on the code
case, the NRC will address this issue in a future rulemaking.
No change was made to the final rule as a result of this comment.
Code Case N-795
Comment: The commenters requested that one or both proposed
conditions on the use of this Code Case in DG-1296 be removed: (1)
Prohibition of use of nuclear heat to perform the leakage test; and (2)
Hold time for noninsulated components must be 1 hour versus 15 minutes
required by Code Case N-795. [Southern 4-1, ASME 5-9, and Exelon 7-2]
NRC Response: The NRC agrees, in part, with this comment. As
discussed in detail in the proposed rule in 81 FR 10780, dated March 2,
2016, the historical prohibition of the use of nuclear heat for
pressure testing is based on concerns about the quality of the VT-2
examinations performed with the core critical, due to the high
temperatures in containment, which limit stay times for inspectors, and
also concerns about personnel safety. However, the commenters
emphasized that Code Case N-795 is only intended for use in the case of
limited scope repairs, such as the replacement of a main steam relief
valve pilot valve (involving a single mechanical joint) when the relief
valve is found to be leaking during startup. Code Case N-795 states
that the alternative test pressure may not be used to satisfy the
requirements of Table IWB-2500-1, Category B-P (the pressure test
required once per cycle of the entire reactor coolant pressure
boundary). Code Case N-795 does not place any restrictions on the size
or scope of the repairs for which the alternative may be used, other
than the alternative test pressure may not be used to satisfy pressure
test requirements, following repair/replacement activities on the
reactor vessel.
However, upon review of the public comments, the staff has
determined that the risk associated with performing the pressure test
with nuclear heat at low power is comparable with the risk to the
plant, when the test is performed without nuclear heat (with the core
subcritical) during mid-cycle outages when decay heat must be managed.
Performing the pressure test under shutdown conditions at full
operating pressure without nuclear heat requires securing certain key
pressure control, heat removal, and safety systems. Under such
conditions, it is more difficult to control temperature and pressure,
when there is significant decay heat production, such as after a mid-
cycle outage, which may reduce the margin available to prevent
exceeding the plant pressure-temperature limits.
The NRC considers it desirable that the scope of repairs be
relatively small when the pressure test is conducted using nuclear
heat, in order to minimize the personnel safety risk and to avoid
rushed examinations. The staff considers it impractical to specify a
particular number of welded or mechanical repairs that would constitute
a ``limited scope.'' However, if the plant is still in a refueling
outage and has already performed the ASME Section XI Category B-P
pressure test of the entire RCPB, it is likely that subsequent repairs
would be performed only on an emergent basis and would generally be of
a limited scope. Additionally, the overall integrity of the RCPB will
have been recently confirmed via the Category B-P test. For mid-cycle
maintenance outages, the staff proposes to modify the condition to
incorporate a limit on the outage duration of fourteen (14) days. This
would tend to limit the scope of repairs, and also limit use of the
Code Case to outages when decay heat was a significant problem.
Therefore, the first condition on Code Case N-795 in Table 2 of DG-
1296, which currently reads:
1. The use of nuclear heat to conduct the BWR Class 1 system
leakage test is prohibited (i.e., the reactor must be in a non-
critical state).
a. This condition also applies to pressure testing of reactor
coolant pressure boundary components repaired or replaced in
accordance with Section XI, IWA-4000.
is modified to read:
1. The use of nuclear heat to conduct the BWR Class 1 system
leakage test is prohibited (i.e., the reactor must be in a non-
critical state), except during refueling outages in which the ASME
Section XI Category B-P pressure test has already been performed, or
at the end of mid-cycle maintenance outages fourteen (14) days or
less in duration.
With respect to the comment on the second condition, the NRC
disagrees with this comment. A one hour hold time is not
unreasonable for non-insulated components. Inspectors do not need to
be in containment during the hold time. Comment 5-9 (ASME) discussed
the technical basis for Code Case N-795, which stated that pressure
testing at 87 percent of full operating pressure would only result
in a 7 percent reduction in flow, while the hold time is being
increased by 50 percent from 10 minutes to 15 minutes. However, it
is not possible to predict the entire range of scenarios or types of
defects that could result in leakage. While some types of defects
could result in immediate leakage, such as an improperly torqued
bolted connection, other types of defects, such as weld defects or
tight cracks could represent a more torturous path for leakage and
may result in delayed leakage. Because the visual examination may be
conducted with the core critical, stay times for examiners in
containment may be limited; therefore, it is desirable that any
leakage be readily detectable. The staff determined that, due to the
uncertainty in the time required for leakage to occur, to an extent
that it would be readily detectable by visual examination, it is
appropriate to conservatively specify a longer hold time of 1 hour
for non-insulated components. Therefore, no changes are made to
Condition (2) requiring a 1-hour hold time for non-insulated
components.
No change was made to the final rule as a result of this comment.
Code Case N-799
Comment: This is a Code Case to define the examination volume/area
where older Section XI codes (up through 2010 Edition) do not recognize
the defined configuration. The conditions proposed in the Code Case are
not included in the proposed rule to accept the 2013 Edition of Section
XI and the Code Case configuration is defined in the 2013 Code Edition.
Commenters believe that this results in inconsistent requirements for
plants using older Code versions versus newer Code versions. The
examination conditions proposed for this Code Case use are not
appropriate for a volume of interest Code Case. If the NRC considers
the conditions appropriate, commenters believe that they should be
included in a revision to 10 CFR 50.55a to assure consistent
application, regardless of Code year and Addenda being applied.
Specifically Conditions (3) and (5) should be removed from the Code
Case. [Southern 4-2, Southern 4-3, and Exelon 7-3]
NRC Response: The NRC agrees, in part, with this comment.
Regarding the removal of proposed Condition (3) from N-799, the NRC
disagrees with the comment. The NRC doesn't find that the examination
of the inner \1/3\ of the component-to-component weld depicted in
Figure 1 of Code Case N-799 provides reasonable assurance that the
integrity of the component-to-component welds will be maintained
throughout the operating life of the plant. Code Case N-799 was written
to support new plant construction to provide examination requirements
for a weld configuration, which did not exist in Section XI (i.e.,
component-to-component welds). Specifically, the examination
requirements described in Code Case N-
[[Page 2348]]
799 would apply to the steam generator nozzle-to-reactor coolant pump
casing (SG-to-RCP) weld in the AP1000 design and the reactor vessel
nozzle-to-recirculation pump weld in the Advanced Boiling Water Reactor
(ABWR). The following discussion will focus on the AP1000 design, but
the staff's overall concern is also applicable to the reactor vessel-
to-reactor coolant pump connection for the ABWR design.
The AP1000 design is unique in that a reactor coolant pump is
welded directly to each of the two outlet nozzles on the steam
generator channel head. This SG-to-RCP weld is a dissimilar metal (low
alloy steel to cast austenitic stainless steel with Alloy 52/152 weld
metal) circumferential butt weld with a double sided weld joint
configuration, similar to that of a reactor vessel shell weld. Also,
this unique component-to-component weld is part of the reactor coolant
pressure boundary and is, therefore, subject to the examination
requirements of ASME Section XI, Subsection IWB.
ASME Section XI, IWB-2500 requires a full volume examination of all
component welds, except those welds found in piping and those found in
nozzles welded to piping. However, for the component-to-component welds
in question, Code Case N-799 only requires a licensee to perform a
volumetric examination of the inner \1/3\ of the weld and a surface
examination of the outer diameter. The staff notes that the
requirements of Code Case N-799 are identical to those in ASME Section
XI, Table IWB-2500-1, Examination Category B-F for welds between vessel
nozzles larger than NPS 4 and piping. As such, the staff does not
believe that examination requirements proposed in Code Case N-799 are
appropriate for the component-to-component welds because the service
conditions of the aforementioned welds are significantly different from
those that would be experienced by a traditional vessel nozzle-to-
piping/safe end butt weld. Specifically, in addition to the operating
environment (RCS pressure, temperature, and exposure to coolant) and
loads expected on a traditional nozzle-to-safe end weld, each SG-to-RCP
weld will support the full weight of a reactor coolant pump with no
other vertical or lateral supports. The SG-to-RCP welds will also be
subject to pump rotational forces and vibration loads from both the
steam generator and the reactor coolant pump during service. In the
absence of operating experience for the weld in question or a bounding
analysis, which demonstrates that a potential fabrication defect in the
outer \2/3\ of the weld will not experience subcritical crack growth,
the effects of these additional operating loads and stresses are
indeterminate. Absent either of the above, the staff finds that it is
inappropriate to limit the examination volume to the inner \1/3\ of the
weld as typical of a piping weld at this time. When the examination
volume that can be qualified by performance demonstration is less than
100 percent of the weld volume, a licensee should include an ultrasonic
examination to examine the qualified volume and perform a flaw
evaluation of the largest hypothetical crack that could exist in the
volume not qualified for ultrasonic examination. No change was made to
the rule as a result of this comment.
The NRC agrees that performing the examination in accordance with
Section XI, Appendix VIII, Supplement 10, for detection and sizing
would eliminate the need for the requirement to perform a flaw
evaluation, based on the largest hypothetical flaw in the unqualified
examination volume. However, the NRC determined a full volume
examination of the entire weld and heat affected zone is required to
provide reasonable assurance of structural integrity of the component-
to component welds addressed by Code Case N-799. The NRC also
determined that requiring the examination procedures to be qualified in
accordance with Section XI, Appendix VIII, Supplement 10, would
eliminate the need for several of the other conditions that were
proposed for N-799. Therefore, the final regulatory guide was modified
to specify only four conditions for Code Case N-799, as follows:
(i) Ultrasonic examination procedures, equipment, and personnel
shall be qualified by performance demonstration in accordance with
Section XI, Appendix VIII, Supplement 10. When applying the
examination requirements of Figure IWB-2500-8, the examination
volume shall be extended to include 100 percent of the weld.
(ii) Examination requirements of Section XI, Mandatory Appendix
I, paragraph I-3200(c) must be applied.
(iii) Ultrasonic depth and sizing qualifications for cast
austenitic stainless steel components must follow Appendix VIII,
Supplement 10, using representative cast austenitic stainless steel
mockups containing representative cracks and be independent of other
Supplement 10 qualifications.
(iv) Cracks detected and not depth sized to Appendix VIII type
performance-based procedures, equipment, and personnel
qualifications shall be repaired or removed.
The NRC agrees with the examination requirement regarding the
consistency between the Code Case and the codes, where the Code Case
that has been incorporated should be consistent. The NRC disagrees with
the statement that the proposed conditions are not appropriate for a
volume of interest Code Case. The NRC is planning to include this topic
in a future rulemaking.
Code Case N-806
Comment: ASME stated that it has taken action to address some of
these concerns and has published Code Case N-806-1, providing
additional requirements for determining wall thickness loss rates. The
ASME recommends that the NRC consider developing conditions on the use
of this case that would enable the endorsement of the case in Table 2
of RG 1.147. [ASME 5-13]
NRC Response: The NRC disagrees with this comment. The NRC
recognizes that ASME has addressed the NRC's concerns regarding the
derivation of the corrosion rate in predicting metal loss in piping and
has incorporated the corrosion rate derivation in the published Code
Case N-806-1. However, the current rulemaking is for Code Case N-806,
which does not contain sufficient information regarding the corrosion
rate. The ASME suggested that the NRC develop conditions on the use of
the Code Case such that the NRC could approve the Code Case for RG
1.147. The NRC has determined that approval of Code Case N-806 with
conditions would require too many conditions to address several open
issues regarding the relationship to the derivation of the corrosion
rate, which still need to be resolved. Therefore, the NRC cannot
approve Code Case N-806 in this rulemaking.
No change was made to the final rule as a result of this comment.
Code Case N-813
Comment: This Code Case should be removed from Table 2 of
Regulatory Guide 1.193 and added to Table 1 of Regulatory Guide 1.147
because of the following reasons.
1. The requirements of Code Case N-813 are identical to changes
made in the 2013 Edition of Section XI, which are being considered
under a separate draft 10 CFR 50.55a rule. The NRC has not proposed
any conditions on these requirements in the 2013 Edition. It is
inappropriate for the NRC to impose conditions on the same
requirements in Case N-813 as the requirements in the 2013 Edition.
2. This Case permits acceptance of subsurface flaws detected
during preservice examination using the same criteria applicable to
flaws detected during inservice examination. There is no greater
likelihood of subsurface flaws detected during preservice
examination to grow unacceptably than there is for the same flaws to
grow during inservice examination. Operating experience has
[[Page 2349]]
shown that the propensity for failure is increased by repairing such
flaws, whereas leaving them in place has never been shown to be a
precursor to failure. Without weld repair, there is no mechanism
expected to produce unacceptable flaw growth, whereas repair welding
itself has been repeatedly shown to cause flaws to grow to the point
of failure. The provisions of this Case, and the identical
provisions in the 2013 Edition, improve safety.
3. The technical basis for this Code Case and accompanying Code
revision states that the action is being sought to prevent the
unnecessary excavation and weld repair of subsurface indications,
which can be analytically shown to be benign over the expected
service lifetime of a component. Based on operating experience, it
is known that weld repairs and their associated stress fields often
serve as points of initiation for inservice degradation mechanisms
(e.g., intergranular stress corrosion cracking, primary water stress
corrosion cracking, etc.). Hence, it is in the best interest of the
long term safe operation of components being placed into service to
eliminate the need for weld repairs where they are not necessary to
correct fabrication problems, which will not challenge the
operability of the component over its service lifetime. This can be
achieved by permitting licensees to effectively utilize the flaw
evaluation rules of IWB-3600 and IWC-3600, which are already
accepted for the analysis of indications due to inservice
degradation.
4. It is important to note that any preservice flaw that has
been evaluated as acceptable is required to receive successive
examinations under IWB-2420(b) or IWC-2420(c) so if the flaw does
grow, it will be detected during these examinations. [ASME 5-14]
NRC Response: The NRC disagrees with this comment, in part. The NRC
has recognized that the provisions in Code Case N-813 are identical to
changes made in the 2013 Edition of the ASME BPV Code, Section XI. The
NRC addressed the contents of the 2013 Edition of the ASME BPV Code,
including the Code provisions identical to those allowed in Code Case
N-813, in a separate rulemaking.
The NRC recognizes that operating experience has shown that
repairing a weld that contains fabrication defects may cause the defect
to grow in the future. On the other hand, permitting a weld that
contains a known unacceptable fabrication defect prior to deployment is
not appropriate and is contrary to the fundamental engineering
principle of a good design. The fundamental engineering design is that
a component should not contain defects before placing it into service.
The staff has accepted the provision of ASME BPV Code, Section III that
permits acceptable flaws (i.e., small insignificant flaws) in a weld to
exist before deployment. The staff's objection to Code Case N-813 is
that the code case permits the existence of unacceptable flaws, which
do not meet the ASME Code preservice acceptance criteria, in welds
before their deployment. The code case allows these unacceptable flaws
to be accepted by analytical evaluation. The code case places no limits
on such flaws (i.e., a weld could have more than one unacceptable flaw
or numerous welds within a piping run could have flaws that did not
meet the preservice acceptance criteria), whereas the original fleet of
nuclear plants had no unacceptable preservice flaws. The staff
concludes that it cannot approve Code Case N-813 in this rulemaking.
The NRC will continue to evaluate operating experience relative to this
type of flaw to further inform decisions on possible approval of this
code case in future rulemakings.
No change was made to the final rule as a result of this comment.
Code Case N-818
Comment: Code Case N-818 should be removed from Regulatory Guide
1.193 and be allowed for use, as the reasons given in Regulatory Guide
1.193 to disallow Code Case N-818 have the following issues: (a) The
fact that the examination will be difficult should not be a reason to
prohibit it as Mandatory Appendix I requires that the technique(s) to
be applied for the volumetric procedure be demonstrated on specimens
simulating geometric, material and surface conditions to be encountered
during implementation. (b) The discussion that ultrasound may have
difficulties discerning between planar and volumetric flaws is not
relevant. There is no requirement in the Code Case to characterize the
flaw by type (i.e., planar or volumetric). (c) The suggestion that its
application should be limited to ferritic weldments defeats the purpose
of Code Case N-818. [EPRI 2-5, Southern 4-4]
NRC Response: The NRC disagrees with this comment, in part. At
present, the NRC has not received any supporting documents from the
industry to address the NRC's concern regarding this Code Case, such as
a demonstration of the adequacy of a full volume ultrasonic examination
for fabrication flaws in austenitic welds. Therefore, the wording of
the reasons given in RG 1.193 should not refer to the inspection being
difficult for austenitic materials and dissimilar metal welds, but
should instead refer to there not being an established technical basis
for the use of ultrasound to find fabrication flaws in these materials.
Additionally, the discussion of planar vs. volumetric flaws will be
removed from RG 1.193, as the Code Case does not require the examiner
to discriminate between these types of flaws. The revised wording for
RG 1.193 is:
The NRC has been conducting research at Pacific Northwest
National Laboratory on the examination of austenitic and ferritic
welds. The work has shown that performing a full volume ultrasonic
examination for fabrication flaws is significantly different from an
inservice examination. For example, examination from two directions
is necessary to detect certain circumferentially oriented
fabrication flaws such as lack of fusion. The work has also shown
that the second leg of a V-path can be applied to examine ferritic
materials on a limited basis but to date the technical basis has not
been established to show that these techniques will be effective on
austenitic materials and dissimilar metal welds. Another finding is
that surface conditions are critical with respect to detecting and
characterizing fabrication flaws. In summary, the NRC finds that an
analytical approach for the acceptance of certain fabrication flaws
could be acceptable if appropriately justified and the scope limited
to ferritic materials. The NRC finds that significant research will
be required to demonstrate that full-volume ultrasonic examination
for fabrication flaws is acceptable for austenitic and dissimilar
metal welds.
Regulatory Guide 1.192, Revision 2 (DG-1297)
Code Case OMN-20
Comment: Allow the use of Code Case OMN-20 for those plants that
implement ASME OM Code 2015 Edition and earlier editions and addenda.
[Gowin 6-1]
NRC Response: The NRC agrees, in part, with this comment. Code Case
OMN-20 cannot be implemented with the 2015 Edition of the ASME OM Code
because the 2015 Edition has not been incorporated by reference into
Sec. 50.55a. Code Case OMN-20 is currently applicable to the 2009
Edition through the OMa-2011 Addenda and all earlier editions and
addenda. Licensees who adopt the 2012 Edition of the ASME OM Code would
not be able to use Code Case OMN-20, without submitting a relief
request to the NRC for approval. For this reason, the NRC partially
agrees with the comment. The NRC believes that Code Case OMN-20 should
be allowed to be implemented with the 2012 Edition and earlier editions
and addenda of the ASME OM Code. The RG 1.192 was updated to add a
condition stating that Code Case OMN-20 is applicable to the editions
and addenda of the ASME OM Code listed in Sec. 50.55a(a)(1)(iv).
No change was made to the final rule as a result of this comment.
[[Page 2350]]
Public Comments on the Proposed Rule
Comment: The ASME Code is updated every year. Preparations are
underway to publish the 2017 edition. NRC is working on 2010 Edition.
It appears that NRC is not in compliance with National Technology
Transfer and Advancement Act of 1995 (NTTAA) by passive non-compliance.
Since NRC has many participants in the Code process, they should be
prepared to act as soon as final standards votes are counted. [Donavin
1-1]
NRC Response: The NRC disagrees with this comment. The NRC
appreciates the ASME's efforts to consider the NRC's concerns as
addressed in conditions to Sec. 50.55a. The NRC agrees that delays in
approving new ASME Code editions and Code Cases can be
counterproductive with respect to implementation of improvements in
ASME Code requirements. The NRC continues to assess ways to improve the
rulemaking process to find schedule efficiencies.
No change was made to the final rule as a result of this comment.
Comment: Many of the conditions are historical and are the result
of a single reviewer's opinion. An example is the rules for the 1994
edition where I watched an NRC reviewer living in Washington, DC
telling a PhD from Tokyo, Japan, that his seismic analysis defending
the edition was non conservative. If there are legitimate questions,
these should be separated from the ``not sufficiently conservative'' or
``insufficient information'' justifications. The Commission has set a
precedent in CVR for SECY-15-0106. ASME has endeavored to address
conditions with docketed letters and Code actions. [Donavin 1-2]
NRC Response: The NRC disagrees with this comment. Although a
single reviewer may state a contrary position, NRC reviews all Code
Cases and comments with appropriate staff and management. Code Cases
that the NRC finds to be conditionally acceptable are also listed in
RGs 1.84, 1.147, and 1.192, which are the subject of this rulemaking,
together with the conditions that must be used if the Code Case is
applied. The NRC determined 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.
No change was made to the final rule as a result of this comment.
Comment: ASME believes that it is not clear whether the word
``superseded'' applies to those Code Cases that are superseded by ASME
or those Code Cases that are listed as superseded in Table 5 of
Regulatory Guide 1.147.
ASME recommends revising the second sentence of this paragraph to
clarify that ``The older or superseded version of the Code Case, if
listed in Table 5, cannot be applied by the licensee or applicant for
the first time.'' [ASME 5-1]
NRC Response: The NRC agrees with this comment. The proposed
additional text will add clarity to the information presented in Table
5. The final RG 1.147 in the introductory paragraph to Table 5, has
been revised to include the statement, ``The older or superseded
version of the Code Case, if listed in Table 5, cannot be applied by
the licensee or applicant for the first time.'' at the end of the
explanatory text above Table 5.
No change was made to the final rule as a result of this comment.
Comment: The Code Case [N-711] would permit each licensee to
independently determine when achievement of a coverage requirement is
impractical, and when Code-required coverage is satisfied. As a result,
application of the Code Case for similar configurations at different
plants could result in potentially significant quantitative variations.
Furthermore, application of the Code Case is inconsistent with NRC's
responsibility for determining whether examinations are impractical,
and eliminates the NRC's ability to take exception to a licensee's
proposed action and impose additional measures where warranted in
accordance with 10 CFR 50.55a(g)(6)(i).
ASME recommends that this case be removed from RG 1.193, Table 2
and added to Table 2 of RG 1.147 with appropriate conditions to address
NRC technical concerns with the use of this case. [ASME 5-10]
NRC Response: The NRC agrees with this comment. However, this is a
new proposal and cannot be included in this rulemaking because it was
not provided for public comment. Rather than include the action in this
rulemaking, the NRC intends to include it within the scope of the
rulemaking that will incorporate by reference the 2015 edition of the
ASME BPV Code.
No change was made to the final rule as a result of this comment.
Comment: In Section IV, ``Section-by-Section Analysis'' of the
Proposed Rule dated March 2, 2016 (Federal Register Vol. 81, No. 41),
ASME believes that it is not clear whether the word ``superseded''
applies to those Code Cases that are superseded by ASME or those Code
Cases that are listed as superseded in Table 5 of Regulatory Guide
1.147 and in Table 5 of Regulatory Guide 1.84. [ASME 5-1 and ASME 5-15]
ASME provides the following recommendations:
i. ASME recommends that the NRC clarify the above concern in the
final rule.
ii. ASME recommends that the NRC review requirements for superseded
ASME Section III and OM Code Cases in RG 1.84 and RG 1.192 for similar
clarification.
NRC Response: The NRC agrees with this comment as noted in the
response to Comment 5-1. In addition to that clarifying text being
added in the introduction to Table 5 in RG 1.147, it will also be added
to the introduction of Table 5 in RG 1.84. The RG 1.192 does not
contain a table of superseded Code Cases, therefore, no change will be
made to the RG 1.192.
No change was made to the final rule as a result of this comment.
V. Section-by-Section Analysis
The following paragraphs in Sec. 50.55a, which list the three RGs
that are being incorporated by reference, are revised as follows:
Paragraphs (a)(3)(i): The reference to ``NRC Regulatory Guide 1.84,
Revision 36,'' is amended to remove ``Revision 36'' and add in its
place ``Revision 37.''
Paragraphs (a)(3)(ii): The reference to ``NRC Regulatory Guide
1.147, Revision 17,'' is amended to remove ``Revision 17'' and add in
its place ``Revision 18.''
Paragraphs (a)(3)(iii): The reference to ``NRC Regulatory Guide
1.192, Revision 1,'' is amended to remove ``Revision 1'' and add in its
place ``Revision 2.''
Overall Considerations on the Use of ASME Code Cases
This rulemaking amends Sec. 50.55a to incorporate by reference RG
1.84, Revision 37, which supersedes Revision 36; RG 1.147, Revision 18,
which supersedes Revision 17; and RG 1.192, Revision 2, which
supersedes Revision 1. The following general guidance applies to the
use of the ASME Code Cases approved in the latest versions of the RGs
that are incorporated by reference into Sec. 50.55a as part of this
rulemaking.
The approval of a Code Case in the NRC RGs constitutes acceptance
of its technical position for applications that are not precluded by
regulatory or other requirements or by the recommendations in these or
other RGs. The applicant and/or licensee are responsible for ensuring
that use of the Code Case does not conflict with regulatory
requirements or licensee
[[Page 2351]]
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
the Code Case, and does not supersede any condition(s) specified in the
Code Case, unless otherwise stated in the NRC condition.
The ASME Code Cases may be revised for many reasons (e.g., to
incorporate operational examination and testing experience and to
update material requirements based on research results). On occasion,
an inaccuracy in an equation is discovered or an examination, as
practiced, is found not to be adequate to detect a newly discovered
degradation mechanism. Hence, when an applicant or a licensee initially
implements a Code Case, Sec. 50.55a requires that the applicant or the
licensee implement the most recent version of that Code Case, as listed
in the RGs incorporated by reference. Code Cases superseded by revision
are no longer acceptable for new applications, unless otherwise
indicated.
Section III of the ASME BPV Code applies only to new construction
(i.e., the edition and addenda to be used in the construction of a
plant are selected based on the date of the construction permit and are
not changed thereafter, except voluntarily by the applicant or the
licensee). Hence, if a Section III Code Case is implemented by an
applicant or a licensee and a later version of the Code Case is
incorporated by reference into Sec. 50.55a and listed in the RGs, the
applicant or the licensee may use either version of the Code Case
(subject, however, to whatever change requirements apply to its
licensing basis (e.g., 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 Sec. 50.55a and
in effect 12 months prior to the start of the next inspection and
testing interval. Licensees who were using a Code Case prior to the
effective date of its revision may continue to use the previous version
for the remainder of the 120-month ISI or IST interval. This relieves
licensees of the burden of having to update their ISI or IST program
each time a Code Case is revised by the ASME and approved for use by
the NRC. Code Cases apply to specific editions and addenda, and Code
Cases may be revised if they are no longer accurate or adequate, so
licensees choosing to continue using a Code Case during the subsequent
ISI or IST interval must implement the latest version incorporated by
reference into Sec. 50.55a and listed in the RGs.
The ASME may annul Code Cases that are no longer required, are
determined to be inaccurate or inadequate, or have been incorporated
into the 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 Sec. 50.55a(z).
If a Code Case is incorporated by reference into Sec. 50.55a and later
annulled by the ASME because experience has shown that the design
analysis, construction method, examination method, or testing method is
inadequate, the NRC will amend Sec. 50.55a and the relevant RG to
remove the approval of the annulled Code Case. Applicants and licensees
should not begin to implement such annulled Code Cases in advance of
the rulemaking.
A Code Case may be revised, for example, to incorporate user
experience. The older or superseded version of the Code Case cannot be
applied by the licensee or applicant for the first time.
If an applicant or a licensee applied a Code Case before it was
listed as superseded, the applicant or the licensee may continue to use
the Code 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 Sec. 50.55a(z). If
a Code Case is incorporated by reference into Sec. 50.55a and later a
revised version is issued by the ASME because experience has shown that
the design analysis, construction method, examination method, or
testing method is inadequate; the NRC will amend Sec. 50.55a and the
relevant RG to remove the approval of the superseded Code Case.
Applicants and licensees should not begin to implement such superseded
Code Cases in advance of the rulemaking.
VI. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act, 5 U.S.C. 605(b), the NRC
certifies that this rule does not have a significant economic impact on
a substantial number of small entities. This final rule affects only
the licensing and operation of nuclear power plants. The companies that
own these plants do not fall within the scope of the definition of
``small entities'' set forth in the Regulatory Flexibility Act or the
size standards established by the NRC (Sec. 2.810).
VII. Regulatory Analysis
The NRC has prepared a final regulatory analysis on this
regulation. The analysis examines the costs and benefits of the
alternatives considered by the NRC. The total estimated net benefit of
this rule is $4.94 million (7% discount rate) and $5.68 million (3%
discount rate). The regulatory analysis is available as indicated in
the ``Availability of Documents'' section of this document.
VIII. Backfitting and Issue Finality
The provisions in this rule allow licensees and applicants to
voluntarily apply NRC-approved Code Cases, sometimes with NRC-specified
conditions. The approved Code Cases are listed in the three RGs that
are incorporated by reference into Sec. 50.55a.
An applicant's or a licensee's voluntary application of an approved
Code Case does not constitute backfitting, inasmuch as there is no
imposition of a new requirement or new position. Similarly, voluntary
application of an approved Code Case by a 10 CFR part 52 applicant or
licensee does not represent NRC imposition of a requirement or action
that is inconsistent with any issue finality provision in 10 CFR part
52. The NRC finds that this rule does not involve any provisions
requiring the preparation of a backfit analysis or documentation
demonstrating that one or more of the issue finality criteria in 10 CFR
part 52 are met.
IX. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31883).
X. Environmental Assessment and Final Finding of No Significant
Environmental Impact
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of 10 CFR part 51, that this rule, if adopted, would not be a
major Federal action significantly affecting the quality of the human
environment; therefore, an
[[Page 2352]]
environmental impact statement is not required.
The determination of this environmental assessment is that there
will be no significant effect on the quality of the human environment
from this action. As alternatives to the ASME Code, NRC-approved Code
Cases provide an equivalent level of safety. Therefore, the probability
or consequences of accidents is not changed. There are also no
significant, non-radiological impacts associated with this action
because no changes would be made affecting non-radiological plant
effluents and because no changes would be made in activities that would
adversely affect the environment. The determination of this
environmental assessment is that there will be no significant offsite
impact to the public from this action.
XI. Paperwork Reduction Act
This final rule contains new or amended collections of information
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). The collection of information was approved by the Office of
Management and Budget (approval number 3150-0011).
The burden to the public for these information collections is
estimated to average a reduction of 380 hours per response, including
the time for reviewing instructions, searching existing data sources,
gathering and maintaining the data needed, and completing and reviewing
the information collection.
The information collection is being conducted to document the plans
for and the results of inservice inspection and inservice testing
programs. The records are generally historical in nature and provide
data on which future activities can be based. The practical utility of
the information collection for the NRC is that appropriate records are
available for auditing by NRC personnel to determine if ASME BPV and OM
Code provisions for construction, inservice inspection, repairs, and
inservice testing are being properly implemented in accordance with
Sec. 50.55a of the NRC regulations, or whether specific enforcement
actions are necessary. Responses to this collection of information are
generally mandatory under Sec. 50.55a.
You may submit comments on any aspect of the information
collection(s), including suggestions for reducing the burden, by the
following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0059.
Mail comments to: Information Services Branch, Office of
the Chief Information Officer, Mail Stop: T-2 F43, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001 or to Aaron Szabo,
Desk Officer, Office of Information and Regulatory Affairs (3150-0011),
NEOB-10202, Office of Management and Budget, Washington, DC 20503;
telephone 202-395-3621, email: [email protected].
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement, unless the requesting document displays a currently valid
OMB control number.
XII. Congressional Review Act
This final rule is a rule as defined in the Congressional Review
Act (5 U.S.C. 801-808). However, the Office of Management and Budget
has not found it to be a major rule, as defined in the Congressional
Review Act.
XIII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies, unless using such a standard is inconsistent with
applicable law or is otherwise impractical. In this rule, the NRC is
continuing to use ASME BPV and OM Code Cases, which are ASME-approved
alternatives to compliance with various provisions of the ASME BPV and
OM Codes. The NRC's approval of the ASME Code Cases is accomplished by
amending the NRC's regulations to incorporate by reference the latest
revisions of the following, which are the subject of this rulemaking,
into Sec. 50.55a: RG 1.84, Revision 37; RG 1.147, Revision 18; and RG
1.192, Revision 2. 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 National Technology Transfer and Advancement Act of
1995 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.
XIV. Incorporation by Reference--Reasonable Availability to Interested
Parties
The NRC is incorporating by reference three NRC Regulatory Guides
that list new and revised ASME Code Cases, which the NRC has approved
as alternatives to certain provisions of NRC-required Editions and
Addenda of the ASME BPV Code and the ASME OM Code.
The NRC is required by law to obtain approval for incorporation by
reference from the Office of the Federal Register (OFR). The OFR's
requirements for incorporation by reference are set forth in 1 CFR part
51. On November 7, 2014, the OFR adopted changes to its regulations
governing incorporation by reference (79 FR 66267). The OFR regulations
require an agency to include, in a proposed rule, a discussion of the
ways that the materials the agency proposes to incorporate by reference
are reasonably available to interested parties or how it worked to make
those materials reasonably available to interested parties. The
discussion in this section complies with the requirement for final
rules, as set forth in 1 CFR 51.5(b).
The NRC considers ``interested parties'' to include all potential
NRC stakeholders, not only the individuals and entities regulated or
otherwise subject to the NRC's regulatory oversight. These NRC
stakeholders are not a homogenous group, so the considerations for
determining ``reasonable availability'' vary by class of interested
parties. The NRC identifies six classes of interested parties with
regard to the material to be incorporated by reference in an NRC rule:
Individuals and small entities regulated or otherwise
subject to the NRC's regulatory oversight. This class includes
applicants and potential applicants for licenses and other NRC
regulatory approvals, and who are subject to the material to be
incorporated by reference. In this context, ``small entities'' has the
same meaning as set out in Sec. 2.810.
Large entities otherwise subject to the NRC's regulatory
oversight. This class includes applicants and potential applicants for
licenses and other NRC regulatory approvals, and who are subject to the
material to be incorporated by reference. In this context, a ``large
entity'' is one which does not qualify as a ``small entity'' under
Sec. 2.810.
Non-governmental organizations with institutional
interests in the matters regulated by the NRC.
Other Federal agencies, states, local governmental bodies
(within the meaning of Sec. 2.315(c)).
Federally-recognized and State-recognized \7\ Indian
tribes.
---------------------------------------------------------------------------
\7\ State-recognized Indian tribes are not within the scope of
10 CFR 2.315(c). However, for purposes of the NRC's compliance with
1 CFR 51.5, ``interested parties'' includes a broad set of
stakeholders including State-recognized Indian tribes.
---------------------------------------------------------------------------
[[Page 2353]]
Members of the general public (i.e., individual,
unaffiliated members of the public who are not regulated or otherwise
subject to the NRC's regulatory oversight) and who need access to the
materials that the NRC proposes to incorporate by reference in order to
participate in the rulemaking.
The three regulatory guides being incorporated by reference in this
rule are available without cost and can be read online, downloaded, or
viewed, by appointment, at the NRC Technical Library, which is located
at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland
20852; telephone: 301-415-7000; e-mail: [email protected]; url:
www.nrc.gov/reading-rm/doc-collections/.
Because access to the three regulatory guides are available in
various forms and at no cost, the NRC determines that the three
regulatory guides, RG 1.84, Revision 37; RG 1.147, Revision 18; and RG
1.192, Revision 2, once approved by the OFR for incorporation by
reference, are reasonably available to all interested parties.
XV. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
Table IV--Rulemaking Related Documents
------------------------------------------------------------------------
ADAMS accession No./
Document title Federal Register citation/
web link
------------------------------------------------------------------------
Proposed Rule Documents:
Proposed Rule--Federal Register notice, 81 FR 10780.
March 2, 2016.
Draft Regulatory Analysis.............. ML15041A816.
Draft RG 1.84, Revision 37 (DG-1295)... ML15027A002.
Draft RG 1.147, Revision 18 (DG-1296).. ML15027A202.
Draft RG 1.192, Revision 2 (DG-1297)... ML15027A330.
Final Rule Documents:
Regulatory Analysis.................... ML16285A013.
RG 1.84, Revision 37................... ML16321A335.
RG 1.147, Revision 18.................. ML16321A336.
RG 1.192, Revision 2................... ML16321A337.
Related Documents:
Draft RG 1.193, ``ASME Code Cases Not ML15028A003.
Approved for Use,'' Revision 5. (DG-
1298).
Federal Register notice-- 82 FR 32934.
``Incorporation by Reference of
American Society of Mechanical
Engineers Codes and Code Cases,'' July
18, 2017.
Federal Register notice-- 80 FR 56820.
``Incorporation by Reference of
American Society of Mechanical
Engineers Codes and Code Cases,''
September 18, 2015.
Federal Register notice-- 68 FR 40469.
``Incorporation by Reference of ASME
BPV and OM Code Cases,'' July 8, 2003.
Federal Register notice--``Fracture 60 FR 65456.
Toughness Requirements for Light Water
Reactor Pressure Vessels,'' December
19, 1995.
Information Notice No. 98-13, ``Post- ML031050237.
Refueling Outage Reactor Pressure
Vessel Leakage Testing Before Core
Criticality,'' April 20, 1998.
Inspection Report 50-254/97-27......... ML15216A276.
Letter from James M. Taylor, Executive ML14273A002.
Director for Operations, NRC, to
Messrs. Nicholas S. Reynolds and
Daniel F. Stenger, Nuclear Utility
Backfitting and Reform Group, February
2, 1990.
Materials Reliability Project Report ML101660468.
MRP-169 Technical Basis for Preemptive
Weld Overlays for Alloy 82/182 Butt
Welds in PWRs (Revision 1), EPRI, Palo
Alto, CA: 2012. 1025295.
NUREG/CR-6933, ``Assessment of Crack ML071020409.
Detection in Heavy-Walled Cast
Stainless Steel Piping Welds Using
Advanced Low-Frequency Ultrasonic
Methods''.
White Paper, PVP2012-78190, http://
``Alternative Acceptance Criteria for proceedings.asmedigitalcol
Flaws in Ferritic Steel Components lection.asme.org/
Operating in the Upper Shelf proceeding.aspx?articleid=
Temperature Range,'' 2012. 1723450.
White Paper PVP2015-45307, ``Options http://
for Defining the Upper Shelf proceedings.asmedigitalcol
Transition Temperature (Tc) for lection.asme.org/
Ferritic Pressure Vessel Steels,'' proceeding.aspx?articleid=
2015. 2471884.
------------------------------------------------------------------------
List of Subjects in 10 CFR Part 50
Administrative practice and procedure, Antitrust, Classified
information, Criminal penalties, Education, Fire prevention, Fire
protection, Incorporation by reference, Intergovernmental relations,
Nuclear power plants and reactors, Penalties, Radiation protection,
Reactor siting criteria, Reporting and recordkeeping requirements,
Whistleblowing.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR part 50:
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. The authority citation for part 50 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103,
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135,
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236,
2237, 2239, 2273, 2282); Energy
[[Page 2354]]
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306
(42 U.S.C. 10226); National Environmental Policy Act of 1969 (42
U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-295, 94
Stat. 783.
0
2. In Sec. 50.55a, revise paragraphs (a)(3)(i) through (iii) to read
as follows:
Sec. 50.55a Codes and standards.
(a) * * *
(3) * * *
(i) NRC Regulatory Guide 1.84, Revision 37. NRC Regulatory Guide
1.84, Revision 37, ``Design, Fabrication, and Materials Code Case
Acceptability, ASME Section III,'' dated March 2017, with the
requirements in paragraph (b)(4) of this section.
(ii) NRC Regulatory Guide 1.147, Revision 18. NRC Regulatory Guide
1.147, Revision 18, ``Inservice Inspection Code Case Acceptability,
ASME Section XI, Division 1,'' dated March 2017, 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 2. NRC Regulatory Guide
1.192, Revision 2, ``Operation and Maintenance Code Case Acceptability,
ASME OM Code,'' dated March 2017, which lists ASME Code Cases that the
NRC has approved in accordance with the requirements in paragraph
(b)(6) of this section.
* * * * *
Dated at Rockville, Maryland, this 2nd day of August 2017.
For the Nuclear Regulatory Commission.
Brian E. Holian,
Acting Director, Office of Nuclear Reactor Regulation.
Editorial note: This document was received for publication by
the Office of the Federal Register on January 3, 2018.
[FR Doc. 2018-00112 Filed 1-16-18; 8:45 am]
BILLING CODE P
Office of the Federal Register, National Archives and Records Administration
Section
Rules and Regulations
Action
Final rule.
Dates
This final rule is effective on February 16, 2018. The incorporation by reference of certain publications listed in the regulation is approved by the Director of the Federal Register as of February 16, 2018.
Contact
Jennifer Tobin, Office of Nuclear Reactor Regulation, telephone: 301-415-2328, email: [email protected]; or Giovanni Facco, Office of Nuclear Regulatory Research, telephone: 301-415-6337; email: [email protected] Both are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
FR Citation
83
FR
2331
RIN Number
3150-AJ13
CFR Associated
Administrative Practice and Procedure; Antitrust; Classified Information; Criminal Penalties; Education; Fire Prevention; Fire Protection; Incorporation by Reference; Intergovernmental Relations; Nuclear Power Plants and Reactors; Penalties; Radiation Protection; Reactor Siting Criteria; Reporting and Recordkeeping Requirements
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