[Federal Register Volume 62, Number 232 (Wednesday, December 3, 1997)]
[Proposed Rules]
[Pages 63892-63911]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-31588]
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Proposed Rules
Federal Register
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This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 /
Proposed Rules
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AE26
Industry Codes and Standards; Amended Requirements
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) regulations require
that nuclear power plant owners construct Class 1, Class 2, and Class 3
components in accordance with the rules provided in Section III,
Division 1, ``Requirements for Construction of Nuclear Power Plant
Components,'' of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV Code), inspect Class 1, Class 2,
Class 3, Class MC (metal containment) and Class CC (concrete
containment) components in accordance with the rules provided in
Section XI, Division 1, ``Requirements for Inservice Inspection of
Nuclear Power Plant Components,'' of the ASME BPV Code, and test Class
1, Class 2, and Class 3 pumps and valves in accordance with the rules
provided in Section XI, Division 1, of the ASME BPV Code.
The NRC proposes to amend 10 CFR 50.55a to revise the requirements
for construction, inservice inspection (ISI), and inservice testing
(IST) of nuclear power plant components. For construction, the proposed
rule would permit the use of Section III, Division 1, of the ASME BPV
Code, 1989 Addenda through the 1996 Addenda, for Class 1, Class 2, and
Class 3 components with six proposed limitations and a modification.
For ISI, the proposed amendment would require licensees to
implement Section XI, Division 1, of the ASME BPV Code, 1995 Edition
with the 1996 Addenda, for Class 1, Class 2, and Class 3 components
with five proposed limitations. Licensees would be permitted to
implement: Code Case N-513 which addresses flaws in low and moderate
energy Class 3 piping; Code Case N-523 which addresses the temporary
use of mechanical clamps in Class 2 and 3 piping; and Subsection IWE
and Subsection IWL, 1995 Edition with the 1996 Addenda.
The proposed rule would expedite implementation of Appendix VIII,
``Performance Demonstration for Ultrasonic Examination Systems,'' to
Section XI, Division 1, with three proposed modifications. An expedited
implementation schedule would also be required for a proposed
modification to Section XI which addresses volumetric examination of
the Class 1 high pressure safety injection (HPSI) system in pressurized
water reactors (PWRs).
For IST, the proposed amendment would require licensees to
implement the 1995 Edition with the 1996 Addenda of the ASME Code for
Operation and Maintenance of Nuclear Power Plants (OM Code) for Class
1, Class 2, and Class 3 pumps and valves with one limitation and one
modification. 10 CFR 50.55a has been clarified with respect to which
pumps and valves are to be included in a licensee's IST program.
Licensees would be permitted to implement: Code Case OMN-1 with one
modification in lieu of stroke time testing; Appendix II (which is an
alternative to the check valve condition monitoring program provisions
contained in Subsection ISTC of the OM Code) with three proposed
modifications; and Subsection ISTD for the IST of snubbers. Finally,
based upon supporting information received since the last rulemaking,
the modification presently in Sec. 50.55a for containment isolation
valve inservice testing has been deleted.
The Statement of Considerations concludes by clarifying the NRC
position regarding ASME Code Interpretations, and discussing NRC
Direction Setting Issue Number 13 (DSI-13) with regard to NRC
endorsement of industry codes and standards.
DATES: Submit comments by March 3, 1998. Comments received after this
date will be considered if it is practical to do so, but the Commission
is able to ensure consideration only for comments received on or before
this date.
ADDRESSES: Comments may be sent to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. ATTN: Rulemaking and
Adjudications Staff. Hand deliver comments to 11545 Rockville Pike,
Rockville, Maryland, 20852, between 7:30 am and 4:15 pm on Federal
workdays.
You may also provide comments via the NRC's interactive rulemaking
website through the NRC home page (http://www.nrc.gov). This site
provides the availability to upload comments as files (any format), if
your web browser supports that function. For information about the
interactive website, contact Ms. Carol Gallagher, (301) 415-5905; e-
mail [email protected]
Single copies of this proposed rulemaking may be obtained by
written request or telefax to 301-415-2260 or from Frank C. Cherny,
Division of Engineering Technology, Office of Nuclear Regulatory
Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, Telephone: 301-415-6786, or Wallace E. Norris, Division of
Engineering Technology, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001, Telephone: 301-415-6796. Certain documents related to
this rulemaking, including comments received, may be examined at the
NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington,
DC. These same documents may also be viewed and downloaded via the
interactive rulemaking website as established by NRC for this
rulemaking.
FOR FURTHER INFORMATION CONTACT: Frank C. Cherny, Division of
Engineering Technology, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:
301-415-6786, or Wallace E. Norris, Division of Engineering Technology,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
Telephone: 301-415-6796.
SUPPLEMENTARY INFORMATION:
1. Background
2. Summary of Proposed Revisions to Sec. 50.55a
2.1 List of Each Revision and Implementation Schedule
2.2 Disscussion
2.3 120-Month Update
2.3.1 Section XI
2.3.1.1 Class 1, 2, and 3 Components, Including Supports
2.3.1.2 Limitations:
2.3.1.2.1 Engineering Judgment
2.3.1.2.2 Quality Assurance
2.3.1.2.3 Class 1 Piping
[[Page 63893]]
2.3.1.2.4 Class 2 Piping
2.3.1.2.5 Reconciliation of Quality Requirements
2.3.2 OM Code
2.3.2.1 Class 1, 2, and 3 Pumps and Valves
2.3.2.2 Background--OM Code
2.3.2.3 Clarification of Safety-Related Valves
2.3.2.4 Limitation:
2.3.2.4.1 Quality Assurance
2.3.2.5 Modification:
2.3.2.5.1 Stroke Time Testing
2.4 Expedited Implementation
2.4.1 Appendix VIII
2.4.1.1 Modifications:
2.4.1.1.1 Appendix VIII Personnel Qualification
2.4.1.1.2 Appendix VIII Specimen Set Cracks
2.4.1.1.3 Appendix VIII Specimen Set Microstructure
2.4.2 Generic Letter on Appendix VIII
2.4.3 Class 1 Piping Volumetric Examination
2.5 Voluntary Implementation
2.5.1 Section III
2.5.1.1 Limitations:
2.5.1.1.1 Engineering Judgement
2.5.1.1.2 Section III Materials
2.5.1.1.3 Weld Leg Dimensions
2.5.1.1.4 Seismic Design
2.5.1.1.5 Quality Assurance
2.5.1.1.6 Independence of Inspection
2.5.1.2 Modification:
2.5.1.2.1 Applicable Code Version for New Construction
2.5.2 Section XI
2.5.2.1 Subsection IWE and Subsection IWL
2.5.2.2 Flaws in Class 3 Piping; Mechanical Clamping Devices
2.5.3 OM Code
2.5.3.1 Code Case OMN-1
2.5.3.2 Appendix II
2.5.3.3 Subsection ISTD
2.5.3.4 Containment Isolation Valves
2.6 ASME Code Interpretations
2.7 DSI-13
2.8 Steam Generators
3. Finding of No Significant Environmental Impact
4. Paperwork Reduction Act Statement
5. Regulatory Analysis
6. Regulatory Flexibility Certification
7. Backfit Analysis
1. Background
The NRC is proposing to amend 10 CFR 50.55a, which defines the
requirements for applying industry codes and standards to nuclear power
plants. Section 50.55a presently requires that nuclear power plant
owners (1) construct Class 1, Class 2, and Class 3 components in
accordance with the rules provided in the 1989 Edition of Section III,
Division 1, ``Requirements for Construction of Nuclear Power Plant
Components'' of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code (BPV Code), (2) inspect Class 1, Class
2, and Class 3 components in accordance with the rules provided in the
1989 Edition of Section XI, Division 1, ``Requirements for Inservice
Inspection of Nuclear Power Plant Components,'' of the ASME BPV Code
with certain limitations and modifications, (3) inspect Class MC (metal
containment) and Class CC (concrete containment) components in
accordance with the rules provided in the 1992 Edition with the 1992
Addenda of Section XI, Division 1, with certain modifications, and (4)
test Class 1, Class 2, and Class 3 pumps and valves in accordance with
the rules provided in the 1989 Edition of Section XI, Division 1, of
the ASME BPV Code with certain limitations and modifications. Every 120
months licensees are required to update their ISI and IST programs to
meet the version of Section XI incorporated by reference into
Sec. 50.55a and in effect 12 months prior to the start of a new 120-
month interval.
The NRC proposes to amend 10 CFR 50.55a to revise the requirements
for construction, ISI, and IST of nuclear power plant components. For
construction, the proposed rule would permit the use of Section III,
Division 1, of the ASME BPV Code, 1989 Addenda through the 1996
Addenda, for Class 1, Class 2, and Class 3 components. Six proposed
limitations to the implementation of Section III are included which
address the issues of engineering judgement, Section III materials,
weld leg dimensions, seismic design, quality assurance, and
independence of inspection. A modification has been included addressing
the applicable Code version for new construction.
For ISI, the proposed amendment would require licensees to
implement Section XI, Division 1, of the ASME BPV Code, 1995 Edition
with the 1996 Addenda, for Class 1, Class 2, and Class 3. Five proposed
limitations to the implementation of Section XI are included which
address the issues of engineering judgement, quality assurance, Class 1
piping, Class 2 piping, and reconciliation of replacement items.
Licensees would be permitted to implement Code Case N-513 which
addresses flaws in low and moderate energy Class 3 piping, and Code
Case N-523 which addresses the temporary use of mechanical clamps in
Class 2 and 3 piping. Licensees would also be permitted to implement
Subsection IWE and Subsection IWL, 1995 Edition with the 1996 Addenda.
The proposed rule would expedite implementation of Appendix VIII,
``Performance Demonstration for Ultrasonic Examination Systems,'' to
Section XI, Division 1. Three proposed modifications to the
implementation of Appendix VIII are included to address the issues of
personnel qualification, specimen set cracks, and specimen set
microstructure. An expedited implementation schedule would also be
required for a proposed modification to Section XI which addresses
volumetric examination of the Class 1 high pressure safety injection
(HPSI) system in pressurized water reactors (PWRs).
For IST, the proposed amendment would require licensees to
implement the 1995 Edition with the 1996 Addenda of the ASME Code for
Operation and Maintenance of Nuclear Power Plants (OM Code) for Class
1, Class 2, and Class 3 pumps and valves. 10 CFR 50.55a has been
clarified with respect to which pumps and valves are to be included in
a licensee's IST program. A proposed limitation is included which
addresses the issue of quality assurance (QA). A proposed modification
to the implementation of the OM Code is included which addresses stroke
time testing. Licensees would be permitted to implement Code Case OMN-1
with one modification in lieu of stroke time testing. In addition,
Appendix II to the OM Code is an alternative to the check valve
condition monitoring program provisions contained in Subsection ISTC of
the OM Code. Three proposed modifications to the implementation of
Appendix II are included which supplement the appendix check valve
condition monitoring program. Licensees would be permitted to use
Subsection ISTD for the IST of snubbers. Finally, based upon supporting
information received since the last rulemaking, the modification
presently in Sec. 50.55a for containment isolation valve inservice
testing has been deleted.
The mechanism for endorsement of the ASME standards, which has been
used since the first endorsement in 1971, has been to incorporate by
reference the ASME BPV Code rules into Sec. 50.55a. The regulation
identifies which editions and addenda of the BPV Code have been
approved for use by the NRC. On August 6, 1992 (57 FR 34666), the NRC
published a final rule in the Federal Register to amend 10 CFR Part 50,
``Domestic Licensing of Production and Utilization Facilities.'' This
final rule amended Sec. 50.55a to incorporate by reference the 1986
Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III,
Division 1, and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989
Edition of Section XI, Division 1, of the BPV Code, with specified
modifications. The amendment imposed an augmented examination of
reactor vessel shell welds. The amendment also separated the
requirements for IST of pumps and valves from those for ISI of other
components by placing the requirements for inservice testing in a
[[Page 63894]]
separate paragraph. For IST of pumps and valves, the regulation,
through its incorporation by reference of the 1989 Edition of Section
XI, endorsed Part 1, ``Requirements for Inservice Performance Testing
of Nuclear Power Plant Pressure Relief Devices,'' Part 6, ``Inservice
Testing of Pumps in Light-Water Reactor Power Plants,'' and Part 10,
``Inservice Testing of Valves in Light-Water Reactor Power Plants,'' of
ASME/ANSI OMa-1988 to ASME/ANSI OM-1987.
On August 8, 1996 (61 FR 41303), the NRC published a final rule in
the Federal Register to amend 10 CFR 50.55a to incorporate by reference
for the first time ASME Section XI, Division 1, Subsection IWE,
``Requirements for Class MC and Metallic Liners of Class CC Components
of Light-Water Cooled Power Plants,'' and Subsection IWL,
``Requirements for Class CC Concrete Components of Light-Water Cooled
Power Plants.'' Subsection IWE provides criteria for visual inspection
of the surface of metal containments, the steel liners of concrete
containments, pressure-retaining bolts, and seals and gaskets.
Subsection IWL provides criteria for visual inspection of concrete
pressure-retaining shells and shell components and for the examination
of unbonded post-tensioning systems.
2. Summary of Proposed Revisions to Sec. 50.55a
The revisions to Sec. 50.55a which would result from adoption of
the 1989 Addenda through the 1996 Addenda have been divided into three
groups based on the proposed implementation schedule (i.e., 120-month
update, expedited, and voluntary). For each of these groups, it is
indicated in parentheses whether or not particular items are considered
a backfit under 10 CFR 50.109 as discussed in Section 8. Backfit
Analysis. This section provides a list of each revision and its
implementation schedule, followed by a discussion of the proposed
revisions.
2.1 List of Each Revision and Implementation Schedule
120-Month Update [in accordance with Sec. 50.55a(g)(4)(i) and
Sec. 50.55a(f)(4)(i)]
Section XI (Not A Backfit)
Class 1, 2, and 3 Components, Including Supports
Limitations
Engineering Judgement
Quality Assurance
Class 1 Piping
Class 2 Piping
Reconciliation of Quality Requirements
OM Code (Not A Backfit)
Class 1, 2, and 3 Pumps and Valves
Clarification of Safety-Related Valves
Limitation
Quality Assurance
Modification
Stroke Time Testing
Expedited Implementation [after 6 months from the date of the final
rule--Backfit]
Section XI
Appendix VIII (including three modifications)
Personnel Qualification
Specimen Set Cracks
Specimen Set Microstructure
Class 1 Piping Volumetric Examination
Voluntary Implementation [may be used when final rule published]
Section III (Not A Backfit)
Class 1, 2, and 3 Components
Limitations
Engineering Judgement
Section III Materials
Weld Leg Dimensions
Seismic Design
Quality Assurance
Independence of Inspection
Modification
Applicable Code Version for New Construction
Section XI (Not A Backfit)
Subsections IWE and IWL, 1995 Edition with the 1996 Addenda
Flaws in Class 3 Piping; Mechanical Clamping Devices
Limitation on Scope
OM Code (Not A Backfit)
Code Case OMN-1
Limitation on Length of Test Interval
Appendix II (including three modifications)
Valve Opening and Closing Functions
Limitation of Length of Initial Test Interval
Condition Monitoring Program
Subsection ISTD
Containment Isolation Valves
2.2 Discussion
2.3 120-Month Update
2.3.1 Section XI
2.3.1.1 Class 1, 2, and 3 Components, Including Supports
Section 50.55a(b)(2) together with Sec. 50.55a(g)(4) of the
proposed rule would require that licensees implement the 1995 Edition
with the 1996 Addenda of Section XI, Division 1, for Class 1, Class 2,
and Class 3 components and their supports. Five proposed limitations
would be included to address NRC positions on the use of Section XI.
2.3.1.2 Limitations
2.3.1.2.1 Engineering Judgement
The first proposed limitation to the implementation of Section XI
would address an NRC position with regard to the Foreword in the 1992
Addenda through the 1996 Addenda of the BPV Code. That Foreword
addresses the use of ``engineering judgement'' for ISI activities not
specifically considered by the Code. Proposed paragraph
50.55a(b)(2)(xi) would require that when a licensee relies on
engineering judgement for activities or evaluations of components or
systems within the scope of Sec. 50.55a that are not directly addressed
by the BPV Code, the licensee must receive NRC approval for those
activities or evaluations pursuant to 10 CFR 50.55a(a)(3).
2.3.1.2.2 Quality Assurance
The second proposed limitation to the implementation of Section XI
pertains to the use of NQA-1 with Section XI. Section XI references the
use of either NQA-1 or the Owner's Appendix B Quality Assurance Program
(10 CFR Part 50, Appendix B, ``Quality Assurance Criteria for Nuclear
Power Plants and Fuel Processing Plants'') as part of its individual
requirements for a QA program. At present, Sec. 50.55a endorses the
1989 Edition of the ASME Code which references NQA-1-1979 for Section
XI. The 1996 Addenda of the ASME Code references NQA-1-1992 for Section
XI.
The NRC has reviewed the requirements of NQA-1, 1986 Addenda
through the 1992 Addenda, that are part of the incorporation by
reference of Section XI, and has determined that by itself, NQA-1 would
not adequately describe how to satisfy the requirements of 10 CFR Part
50, Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants,'' since there are various aspects of
operational phase QA and administrative controls which are not
addressed by NQA-1.
10 CFR 50.34(b)(6)(ii) requires that ``The information on the
controls to be used for a nuclear power plant or a fuel reprocessing
plant shall include a discussion of how the applicable requirements of
Appendix B will be satisfied.'' This information is required to be
submitted to the NRC as part of the Final Safety Analysis Report
(FSAR). Standard Review Plan (SRP) 17.2, ``Quality Assurance During the
Operations Phase,'' states that ``The QA program description presented
in the FSAR must discuss how each criterion of Appendix B will be
met.'' Further, the SRP states ``The acceptance criteria include a
commitment to comply with the regulatory positions presented in the
appropriate issue of the Regulatory Guides including the requirements
of ANSI Standard N45.2.12 and the Branch
[[Page 63895]]
Technical Position listed in subsection V of SRP Section 17.1. Thus,
the commitment constitutes an integral part of the QA program
description and requirements.'' The NRC has determined that the
provisions of NQA-1, 1986 Addenda through the 1992 Addenda, would not
satisfy the criteria specified in SRP 17.2 for describing how the
requirements of Appendix B will be satisfied for operational
activities. There are numerous areas where American National Standards
Institute (ANSI) standards or NRC regulatory positions, which have been
long-standing cornerstones of an Owner's QA Program, are either
nonmandatory or missing altogether from the NQA-1 provisions. However,
the Owner's Section XI QA Program, which has been approved by the NRC,
is adequate. Thus, the Commission has determined that the requirements
of NQA-1, 1986 Addenda through the 1992 Addenda, are acceptable for use
in the context of Section XI, as permitted by IWA-1400, provided the
licensee utilizes its 10 CFR Part 50, Appendix B, QA program in
conjunction with Section XI. Changes to a licensee's QA program shall
be made in accordance with 10 CFR 50.54(a). Further, where NQA-1 and
Section XI do not address the commitments contained in the licensee's
Appendix B QA program description, such commitments shall be applied to
Section XI activities. Proposed Sec. 50.55a(b)(2)(xii) contains the
requirement addressing licensee's commitments related to Section XI.
2.3.1.2.3 Class 1 Piping
The third proposed limitation to the implementation of Section XI
would require licensees to use the rules for Section XI IWB-1220,
``Components Exempt from Examination,'' that are contained in the 1989
Edition in lieu of the rules in the 1989 Addenda through the 1996
Addenda. These later Code addenda contain provisions of Code Cases N-
198-1, ``Exemption from Examination for ASME Class 1 and Class 2 Piping
Located at Containment Penetrations;'' N-322, ``Examination
Requirements for Integrally Welded or Forged Attachments to Class 1
Piping at Containment Penetrations;'' and N-324, ``Examination
Requirements for Integrally Welded or Forged Attachments to Class 2
Piping at Containment Penetrations;'' which were found to be
unacceptable. Because the NRC had previously determined the Code cases
to be unacceptable, they were not endorsed in any revision of
Regulatory Guide 1.147, ``Inservice Inspection Code Case
Acceptability--ASME Section XI, Division 1.'' The provisions of Code
Case N-198-1 were determined by the NRC to be unacceptable because
industry experience has shown that welds in service-sensitive BWR
stainless steel piping, many of which are located in Containment
Penetrations, are subjected to an aggressive environment (BWR water at
reactor operating temperatures) and will experience Intergranular
Stress Corrosion Cracking. Exempting these welds from examination could
result in conditions which reduce the required margins to failure to
unacceptable levels. The provisions of Code Cases N-322 and N-324 were
determined to be unacceptable because some important piping was
exempted from inspection. Access difficulties was the basis in the Code
cases for exempting these areas from examination, but the NRC developed
the break exclusion zone design and examination criteria utilized for
most containment penetration piping expecting not only that Section XI
inspections would be performed but that augmented inspections would be
performed. These design and examination criteria are contained in
Branch Technical Position MEB 3-1, an attachment of NRC Standard Review
Plan 3.6.2, ``Determination of Rupture Locations and Dynamic Effects
Associated with the Postulated Rupture of Piping.'' Thus, proposed
Sec. 50.55a(b)(2)(xiii) would require licensees to use the rules for
IWB-1220 that are contained in the 1989 Edition in lieu of the rules in
the 1989 Addenda through the 1996 Addenda.
2.3.1.2.4 Class 2 Piping
The fourth proposed limitation to the implementation of Section XI,
contained in Sec. 50.55a(b)(2)(xiv), would confine implementation of
Section XI IWC-1220, ``Components Exempt from Examination,'' IWC-1221,
``Components Within RHR (Residual Heat Removal), ECC (Emergency Cool
Cooling), and CHR (Containment Heat Removal) Systems or Portions of
Systems,'' and IWC-1222, ``Components Within Systems or Portions of
Systems Other Than RHR, ECC, and CHR Systems,'' 1989 Addenda through
the 1996 Addenda. The provisions of Code Case N-408-3, ``Alternative
Rules for Examination of Class 2 Piping,'' were incorporated into
Subsection IWC in the 1989 Addenda. These provisions contain rules for
determining which Class 2 components are subject to volumetric and
surface examination. The NRC had previously determined that the
provisions of the Code Case were acceptable if the licensee defined the
Class 2 piping subject to volumetric and surface examination and
received approval prior to implementation. Approval was required to
ensure that safety significant components in the Residual Heat Removal,
Emergency Core Cooling, and Containment Heat Removal systems are not
exempted from appropriate examination requirements. Thus, the
requirements contained in IWC-1220, IWC-1221, and IWC-1222, 1989
Addenda through the 1996 Addenda, for determining the components
subject to examination and establishing examination requirements for
Class 2 piping may be used if the licensee defines the Class 2 piping
subject to volumetric and surface examination, and submits this
information to the NRC for approval pursuant to Sec. 50.55a(a)(3).
2.3.1.2.5 Reconciliation of Quality Requirements
The fifth proposed limitation to the implementation of Section XI
addresses reconciliation of replacement items
[Sec. 50.55a(b)(2)(xx)(A)] and the definition of Construction Code
[Sec. 50.55a(b)(2)(xx)(B)]. Changes to IWA-4222, ``Reconciliation of
Owner's Requirements,'' in the 1995 Addenda would permit a replacement
item produced at a facility not having a 10 CFR Part 50, Appendix B
qualified program to be used in safety-related applications. With
regard to the definition of Construction Code, a new definition of
Construction Code appeared in IWA-9000, ``Glossary,'' in the 1993
Addenda. Due to the changes made in IWA-4200 in the 1995 Addenda, the
change in definition could result in standards being utilized which do
not contain any QA requirements, or contain QA requirements that do not
fully comply with Appendix B. Thus, when implementing the 1995 Addenda
through the 1996 Addenda, Sec. 50.55a(b)(2)(xx)(A) would require
reconciliation of replacement items to the original QA requirements.
Section 50.55a(b)(2)(xx)(B) would require a licensee to reconcile
replacement items to the Construction Code and to the QA requirements
as described in the Owner's QA program.
Section XI Article IWA-4000 provides rules and requirements for the
repair and replacement of pressure retaining components and their
supports. Versions of IWA-4000 previous to the 1995 Addenda permitted a
licensee to purchase a replacement item to the standards of the
original Construction Code or a later version, provided that the
technical requirements of an item such as design and fabrication, as
well as the nontechnical requirements (identified as administrative
requirements in IWA-4222) such as QA
[[Page 63896]]
and Authorized Inspection of the later version were reconciled with
those of the original Construction Code and Owner's Requirements.
Reconciliation ensures that the replacement item meets certain
standards of quality so that it is satisfactory for the specified
design and operating conditions. In the 1995 Addenda, the provisions of
Code Case N-554, ``Alternative Requirements for Reconciliation of
Replacement Items,'' were incorporated into an extensive rewrite of
IWA-4200. As a result of these changes to IWA-4200, specifically IWA-
4222(a)(2), the nontechnical requirements for Class 1, 2, and 3 safety-
related replacement items would no longer need to be reconciled which
may result in noncompliance with 10 CFR Part 50, Appendix B. NRC
regulations require that any item which performs a safety-related
function must meet Appendix B. Appendix B invokes, among other things,
controls on suppliers of safety-related items. By not requiring
reconciliation of the administrative requirements, the provisions in
IWA-4222(a)(2) of the 1995 Addenda through the 1996 Addenda, would
allow vendors having a QA program which does not meet Appendix B to be
utilized, and may result in noncompliance with Appendix B. These
deficiencies could be resolved if the Code provided for commercial
grade item dedication in accordance with 10 CFR Part 21, ``Reporting of
Defects and Noncompliance.'' However, IWA-4222 does not address
commercial grade dedication. In addition, it should be pointed out that
a separate Code Case which provides an alternative for a specific
provision in IWA-4200, Code Case N-567, ``Alternative Requirements for
Class 1, 2, and 3 Replacement Components,'' was modified to require the
reconciliation of nontechnical requirements before the Code Case was
approved. Therefore, an inconsistency exists between the Code and a
Code Case. Thus, when implementing the 1995 Addenda through the 1996
Addenda, Sec. 50.55a(b)(2)(xx)(A) would require reconciliation of
replacement items to the original QA requirements.
The provisions of the Code in IWA-4222(a)(2) discussed above
address newly manufactured replacement parts. A further limitation on
the use of Article IWA-4200 in the 1995 Addenda through the 1996
Addenda is contained in Sec. 50.55a(b)(2)(xx)(B). IWA-4222(b) addresses
the use of items from a facility which was shutdown or for which
construction was halted. IWA-4222(b) permits the use of either the
administrative requirements of the Construction Code of the item being
replaced or the administrative requirements of the Construction Code of
the item being used for replacement. However, the definition of
``Construction Code'' was changed in the 1993 Addenda. In versions of
Section XI previous to the 1993 Addenda, Construction Code was defined
in IWA-9000, ``Glossary,'' as ``the body of technical requirements that
governed the construction of the item.'' Included in the body of
technical requirements that governed the construction of the item was a
requirement to reconcile the Owner's specification requirements, which
included NRC regulatory requirements, and applicable Owner design and
procurement specifications that invoke technical and nontechnical
requirements (e.g., 10 CFR Part 50, Appendix B). In the 1993 Addenda,
the definition became nationally recognized Codes such as ASME,
Specifications such as the American Society of Testing and Materials
(ASTM), and designated Code Cases. Either definition of Construction
Code would include the original Construction Codes for the design and
construction of piping, such as B31.1, ``Power Piping,'' and B31.7,
``Nuclear Piping,'' and those for the design and construction of
storage tanks, such as the American Petroleum Institute (API) 620,
``Design and Construction of Large, Welded, Low-Pressure Storage
Tanks,'' and API 650, ``Welded Steel Tanks for Oil Storage.'' However,
many of these standards utilized for construction do not contain any QA
requirements, or they contain QA requirements that do not fully comply
with Appendix B. Therefore, in order to satisfy Appendix B, QA
requirements similar to or meeting Appendix B were invoked in the
Owner's original procurement documents. Thus, when implementing IWA-
4200 (including subparagraphs IWA-4221, IWA-4222, IWA-4223, IWA-4224,
and IWA-5224), Sec. 50.55a(b)(2)(xx)(B) would require a licensee to
reconcile replacement items to the Construction Code and to the QA
requirements as described in the Owner's QA program.
2.3.2 OM Code (120-Month Update)
2.3.2.1 Class 1, 2, and 3 Pumps and Valves
The proposed amendment to Sec. 50.55a(f)(4) would require that IST
of pumps and valves be performed in accordance with the ASME ``Code for
Operation and Maintenance of Nuclear Power Plants'' (OM Code). A
proposed new section, Sec. 50.55a(b)(3), would specify the editions and
addenda of the OM Code that have been incorporated by reference into
Sec. 50.55a. Paragraph 50.55a(b)(3) together with Sec. 50.55a(f)(4) of
the proposed rule would require that licensees implement the 1995
Edition with the 1996 Addenda of the OM Code. Existing
Sec. 50.55a(f)(1) has been modified to clarify which pumps and valves
are to be included in the IST program. One proposed limitation to
implementation of the OM Code addressing QA, and one proposed
modification of the OM Code addressing stroke time testing have been
included.
2.3.2.2 Background--OM Code
Until 1990, the ASME Code requirements addressing IST of pumps and
valves were contained in Section XI Subsections IWP (pumps) and IWV
(valves). The provisions of IWP and IWV were last incorporated by
reference into Sec. 50.55a in a final rulemaking published on August 6,
1992 (57 FR 34666). In 1990, the ASME published the initial edition of
the OM Code which provides rules for IST of pumps and valves. The
requirements contained in the 1990 Edition are identical to the
requirements contained in the 1989 Edition of Section XI Subsections
IWP (pumps) and IWV (valves). The ASME Board on Nuclear Codes and
Standards has transferred responsibility for rules on IST from Section
XI to the OM Committee. As such, the Section XI rules for inservice
testing of pumps and valves that are presently incorporated by
reference into NRC regulations are no longer being updated by Section
XI.
The ASME 1990 Edition of the OM Code consists of one section
(Section IST) entitled ``Rules for Inservice Testing of Light-Water
Reactor Power Plants.'' This section is divided into four subsections,
ISTA, ``General Requirements,'' ISTB, ``Inservice Testing of Pumps in
Light-Water Reactor Power Plants,'' ISTC, ``Inservice Testing of Valves
in Light-Water Reactor Power Plants,'' and ISTD, ``Examination and
Performance Testing of Nuclear Power Plant Dynamic Restraints
(Snubbers).'' The IST of snubbers is governed by plant technical
specifications and, thus, has never been included in Sec. 50.55a.
Therefore, this proposed rule only requires implementation of
Subsections ISTA, ISTB, and ISTC. However, Sec. 50.55a(b)(3)(v) would
permit licensees to implement Subsection ISTD of the 1996 Addenda by
making a change to their technical specifications in accordance with
applicable NRC requirements.
[[Page 63897]]
2.3.2.3 Clarification of Safety-Related Valves
The existing Sec. 50.55a(f)(1) has been interpreted by some
licensees to mean that all safety-related pumps and valves regardless
of ASME Code Class (or equivalent) were to be included in the IST
program. The NRC proposes to modify this paragraph to clarify that the
provisions of Sec. 50.55a(f)(1) apply only to pumps and valves in
steam, water, air, and liquid radioactive waste systems that perform a
function to shut down the reactor, maintain the reactor in a safe
shutdown condition, mitigate the consequences of an accident, or
provide overpressure protection for such systems.
2.3.2.4 Limitation
2.3.2.4.1 Quality Assurance
The limitation to the implementation of the OM Code pertains to the
use of NQA-1, ``Quality Assurance Requirements for Nuclear
Facilities,'' with the OM Code. The OM Code references the use of
either NQA-1 or the Owner's Appendix B Quality Assurance Program as
part of its individual requirements for a QA program. At present,
Sec. 50.55a endorses NQA-1-1979 for the OM Code. The 1996 Addenda also
endorses NQA-1-1979. Thus, the 1996 OM Code has not endorsed a later
version of NQA-1. Because this rulemaking would incorporate the OM Code
by reference into Sec. 50.55a for the first time, a limitation is
included to address the same issues discussed previously in the Section
XI section on QA.
The NRC has determined that the provisions of NQA-1, 1979 Addenda,
would not adequately describe how to satisfy the requirements of
Appendix B as satisfied by Sec. 50.34(b)(6)(ii). Further, there are
various aspects of operational phase QA and administrative controls
which are not addressed by NQA-1. There are numerous areas where
American National Standards Institute (ANSI) standards or NRC
regulatory positions, which are specified in SRP 17.2, are either
nonmandatory or missing altogether from the NQA-1 provisions. However,
the Owner's QA Program, which has been approved by the NRC, is
adequate. Thus, the NRC has determined that the requirements of NQA-1-
1979, that are part of the incorporation by reference of the OM Code,
is acceptable for use in the context of the OM Code, as permitted by
ISTA 1.4, provided the licensee utilizes its 10 CFR Part 50, Appendix
B, QA program in conjunction with the OM Code. Changes to licensee's QA
program shall be made in accordance with 10 CFR 50.54. Further, where
NQA-1 and the OM Code do not address the commitments contained in the
licensee's Appendix B QA program description, such commitments shall be
applied to OM Code activities. Proposed Sec. 50.55a(b)(3)(i) addresses
licensee's commitments related to the OM Code.
2.3.2.5 Modification
2.3.2.5.1 Stroke Time Testing
Proposed Sec. 50.55a(b)(3)(ii) would require that the stroke time
testing requirement of Subsection ISTC of the OM Code applicable for
motor-operated valves (MOVs) be supplemented with programs that
licensees have previously committed to perform, prior to issuance of
this amendment to Sec. 50.55a, for demonstrating the design basis
capability of MOVs. Stroke time testing of MOVs has been specified in
ASME Section XI and is currently required by Sec. 50.55a(f). This same
testing is required by the OM Code. This testing is a useful tool and
complements other tests used to verify MOV function. Variation in
measured stroke times can indicate valve degradation. Additionally,
periodic stroking provides valve exercise and some measure of on-demand
reliability. However, as discussed in NRC Generic Letter (GL) 89-10
``Safety-Related Motor-Operated Valve Testing and Surveillance'' dated
June 28, 1989, it is now recognized that the stroke time testing alone
is not sufficient to provide assurance of MOV capability under design-
basis conditions.
Subsequent to licensees implementing programs pursuant to GL 89-10,
the NRC issued Generic Letter 96-05, ``Periodic Verification of Design-
Basis Capability of Safety-Related Motor-Operated Valves,'' on
September 18, 1996. This generic letter requested licensees to
establish a program, or to ensure the effectiveness of their current
program, to verify on a periodic basis that safety-related motor-
operated valves continue to be capable of performing their safety
functions within the current licensing bases of the facility. Prior to
issuance of this rule, licensees have made licensing commitments
pursuant to GL 96-05 that have been reviewed by the NRC staff. Most
licensees have committed to participate in the Joint Owners Group (JOG)
Program on MOV Periodic Verification. The JOG program includes three
phases: (1) licensees will establish an interim static diagnostic
testing program developed by JOG with a test frequency based on margin
and safety significance; (2) JOG will coordinate a dynamic testing
program over the next 5 years that includes approximately 150 MOVs with
participating licensees each testing a few MOVs three times over this
interval; and (3) based on the results of the dynamic testing program,
JOG will establish a long-term periodic test program. Proposed
Sec. 50.55a(b)(3)(ii) would require that licensees supplement the
stroke time testing requirements of the OM Code with these commitments.
2.4 Expedited Implementation
2.4.1 Appendix VIII
The proposed rule would require that licensees expedite
implementation of mandatory Appendix VIII, ``Performance Demonstration
for Ultrasonic Examination Systems,'' to Section XI, 1995 Edition with
the 1996 Addenda. Three proposed modifications would be included to
address NRC positions on the use of Appendix VIII. Licensees would be
required to implement Appendix VIII, including the modifications, for
all examinations of the pressure vessel, piping, nozzles, and bolts and
studs which occur after 6 months from the date of the final rule. The
proposed rule would not require any change to a licensee's ISI schedule
for examination of these components, but would require that the
provisions of Appendix VIII be used for all examinations after that
date rather than the ultrasonic testing (UT) procedures and personnel
requirements presently being utilized by licensees.
Appendix VIII provides the requirements for performance
demonstration for ultrasonic testing (UT) procedures, equipment, and
personnel used to detect flaws and size flaws. Its requirements are
applicable to all UT performed for Class 1, Class 2, and Class 3 items
(i.e., reactor vessel, nozzles, piping, and bolting and studs). These
requirements are also to be utilized when implementing the augmented
inservice inspection program for reactor vessel shell welds presently
required by Sec. 50.55a(g)(6)(ii)(A). The NRC has reviewed the 1995
Edition with the 1996 Addenda of Appendix VIII and has determined that
the provisions contained in this appendix should be used with three
modifications (addressed below). This mandatory appendix would normally
be adopted as part of the routine 120-month update specified in
Sec. 50.55a(g)(4), but because of the importance of the Appendix VIII
program, the NRC has determined that its requirements should be
implemented after 6 months from the date of the final rule. The
performance demonstration requirements in Appendix VIII would
[[Page 63898]]
substantially improve the ability of an examiner to detect and
characterize flaws in examined components. UT procedures and personnel
requirements are presently contained in Section XI but, as detailed in
the documented evaluation required by Sec. 50.109(a)(4), personnel
qualified to Appendix VIII are significantly better at detecting flaws.
The industry's Performance Demonstration Initiative (PDI) established a
process in accordance with Appendix VIII for reactor vessel, nozzle,
piping, and bolting examinations. PDI has received considerable support
from the industry, and every licensee has contributed financially. The
majority of the cost of PDI was in setting up the samples, which has
been completed. Proposed Sec. 50.55a(g)(6)(ii)(C)(1) would require
licensees to utilize the improved requirements in Appendix VIII for all
examinations of reactor vessels (including nozzles), piping, and
bolting performed after 6 months from the date of the final rule. To
date, the PDI program has qualified over 300 individuals for piping and
five teams for vessel examinations. Thus, the NRC does not believe that
a 6-month implementation period would result in hardship.
2.4.1.1 Modifications
2.4.1.1.1 Appendix VIII Personnel Qualification
The first proposed modification of Appendix VIII relates to its
requirement that ultrasonic examination personnel meet the requirements
of Appendix VII, ``Qualification of Nondestructive Examination
Personnel for Ultrasonic Examination,'' to Section XI. Appendix VII
first appeared in Section XI in the 1988 Addenda and was incorporated
by reference into Sec. 50.55a in a final rule published on August 6,
1992 (57 FR 34666). The NRC believes that the requirement in Appendix
VII-4240 for personnel to receive a minimum of 10 hours of training on
an annual basis is inadequate. Proposed Sec. 50.55a(b)(2)(xvii) would
require that all personnel qualified for performing ultrasonic
examinations in accordance with Appendix VIII receive 40 hours of
annual training which includes laboratory work and examination of
flawed specimens. Signals can be difficult to interpret, and as
detailed in the regulatory analysis for this rulemaking, experience and
studies indicate that the examiner must practice on a frequent basis to
maintain the capability for proper interpretation. In addition, these
studies have shown that this capability begins to diminish within
approximately 6 months if skills are not maintained. Thus, 10 hours of
annual training is not sufficient practice to maintain skills. The NRC
believes that a minimum of 40 hours of annual training, not 10 hours,
is required to maintain an examiner's abilities in this highly
specialized skill area. The NRC expects that licensees would distribute
the training over the course of the year to ensure that interpretation
skills do not diminish.
2.4.1.1.2 Appendix VIII Specimen Set Cracks
The second proposed modification of Appendix VIII would require
that all flaws in the specimen sets used for performance demonstration
for piping, vessels, and nozzles be cracks. For piping, Appendix VIII
requires that all of the flaws in a specimen set be cracks. However,
for vessels and nozzles, Appendix VIII would allow as many as 50% of
the flaws to be notches. For the purpose of demonstrating
nondestructive examination (NDE) capabilities, notches are not
realistic representations of service induced cracks. An inspector
cannot properly interpret service induced cracks by qualifying with
specimens containing notches. Notches are easier to detect than flaws
because notches have a higher amplitude and simpler signal
characteristics. Notches are easier to interpret and, in fact, the
probability of detecting notches can be much higher than the
probability of detecting cracks under similar conditions. In addition,
Appendix VIII provides a screening test that uses a relatively small
sample size containing few flaws. If some of the flaws are replaced by
notches that are unrealistic, the screening test becomes ineffective.
Because of these considerations, the flaws in the specimen sets
utilized for piping by EPRI for the PDI are all cracks. The regulatory
analysis for this rulemaking contains a detailed discussion of the
importance of using cracks in the specimens. Thus, proposed
Sec. 50.55a(b)(2)(xiii) would require that all flaws in the specimen
sets used for performance demonstration be cracks.
2.4.1.1.3 Appendix VIII Specimen Set Microstructure
The third proposed modification of Appendix VIII would require that
all specimens for single-side tests contain microstructures like the
components to be inspected and flaws with non-optimum characteristics
consistent with field experience that provide realistic challenges to
the UT technique. Appendix VIII does not distinguish specimens for two-
sided examinations from those used for single-sided examination.
Appendix VIII was originally developed using UT lessons learned
from two-sided examinations of welds. This UT experience provided the
input for designing specimens and selecting, locating, and
characterizing flaws. Studies have shown that defect characteristics
such as shape, size, depth, tilt angle, skew angle, roughness, and
crack tip affect the probability of detecting a particular flaw. For
example, it was demonstrated in one particular study (Reference 22 in
the documented evaluation) that a particular flaw was over three times
more reflective in one direction, thus easier to detect, than in the
opposite direction. Specimens designed for two-sided examination may
not have defects which are appropriate for single-sided performance
demonstration; i.e., the specimens may not adequately test an examiners
proficiency in detecting flaws. Therefore, in order to proceed with the
effort of qualifying UT systems (equipment, procedures, and personnel)
for single-sided examinations, proposed Sec. 50.55a(b)(2)(xx) would
require the industry to develop sets of specimens that contain
microstructures similar to the types found in the components to be
inspected and flaws with non-optimum characteristics, such as skew,
tilt, and roughness, consistent with field experience that provide
realistic challenges for single-sided performance demonstration.
2.4.2 Generic Letter on Appendix VIII
A draft generic letter was published in the Federal Register (61 FR
69120) for public comment on December 31, 1996, to alert the industry
to the importance of using equipment, procedures, and examiners capable
of reliably detecting and sizing flaws in the performance of
comprehensive examinations of reactor vessels and piping. The generic
letter stated that even though the need for improvement clearly
existed, the staff had reached the conclusion that immediate
backfitting of Appendix VIII in advance of this proposed rulemaking was
not warranted. This conclusion was based on consideration of defense-
in-depth measures, Code margins in component design, leakage monitoring
systems, and also that Appendix VIII was already being applied to
selected piping subject to intergranular stress corrosion cracking. The
NRC received 16 comment letters on the generic letter.
The comments generally were very similar and can be summarized in
the following five items: (1) it is inappropriate to request licensees
to voluntarily commit to a program in a
[[Page 63899]]
generic letter; (2) the urgency for licensee's to voluntarily commit to
implementing Appendix VIII is inconsistent with the statement in the
generic letter that a safety concern does not exist that would warrant
immediate backfitting in advance of the rulemaking; (3) the
performance-based qualification program of Appendix VIII should be
approved an alternative to the current ASME Code, and Appendix VIII as
implemented by PDI should be recognized as an acceptable alternative
for Appendix VIII; (4) the NRC should provide guidance on incorporating
Appendix VIII and/or PDI into plant-specific ISI programs; and (5) the
generic letter would request that licensees update their UT ISI and
augmented inspection commitments to a Code edition not yet referenced
in the regulations.
With regard to the first comment, the NRC disagrees that it is
inappropriate to request licensees to voluntarily commit to a program
in a generic letter. This is one mechanism available to the NRC for
alerting licensees, for example, to degraded conditions which may
unacceptably affect the function of safety-related components. The
second comment takes the generic letter statement out of context. What
the generic letter actually stated was that a safety concern did not
exist to warrant immediate backfitting in advance of the rulemaking
because of defense-in-depth measures, Code margins in design, and that
Appendix VIII was already being applied to selected piping subject to
intergranular stress corrosion cracking. The NRC strongly disagrees
that Appendix VIII and Appendix VIII as implemented by PDI should be
alternatives to the present Code rules. As detailed in the documented
evaluation for backfitting Appendix VIII, it has been demonstrated that
examiners previously considered qualified under Section XI generally
have marginal UT skills. This was evident from the discouragingly low
percentage of examiners initially satisfying the screening criteria for
detecting flaws under the PDI program. Comment four regarding guidance
on incorporating Appendix VIII into present ISI programs, and comment
five regarding Code edition are automatically resolved in a rulemaking
format.
At the time the generic letter was issued, this proposed rulemaking
was still under development. The purpose of the generic letter was to
alert the industry to the (1) generally poor performance in detecting
flaws and (2) the Commission's intent to endorse Appendix VIII via
rulemaking. Publication of a final rule would obviate the need for the
generic letter.
2.4.3 Class 1 Piping Volumetric Examination
A proposed modification of Section XI would require licensees of
pressurized water reactor plants to supplement the surface examination
of Class 1 High Pressure Safety Injection Systems (HPSI) piping as
required by Examination Category B-J of Table IWB-2500-1 for nominal
pipe sizes (NPS) between 4 (inches) and 1+ (inches), with a volumetric
(ultrasonic) examination. This requirement is proposed because (1)
inside diameter cracking of HPSI piping in the subject size range has
been previously discovered (as detailed in NRC Generic Letter 85-20,
``High Pressure Injection/Make-Up Nozzle Cracking in Babcock and Wilcox
Plants,'' and in NRC Information Notice 97-46, (``Unisolable Crack in
High-Pressure Injection Piping,''), (2) failure of this line could
result in a small break loss of coolant accident while directly
affecting the system designed to mitigate such an event, and (3)
volumetric examinations are already required by the Code for Class 2
portions of this system (Table IWC-2500-1, Examination Category C-F-1)
within the same NPS range. Thus, not only are the requirements between
Class 1 and Class 2 inconsistent (with the Class 1 portions being
subject to less stringent testing requirements as compared with Class 2
portions of the same type of piping), but operating experience has
shown that these reactor coolant pressure boundary (RCPB) pipe
examinations need to be more comprehensive. Proposed
Sec. p50.55a(b)(2)(xv) would require licensees to supplement the
Section XI required surface examination for the Class 1 portion of the
HPSI system with volumetric examination in order to ensure the
integrity of the reactor coolant pressure boundary as required by
General Design Criteria (GDC) 14, 10 CFR Part 50, Appendix A, or
similar provisions in the licensing basis for these facilities, and
Criteria II and XVI of 10 CFR Part 50, Appendix B. Licensees would be
required to perform the volumetric examination during any ISI program
inspection of the HPSI system performed after 6 months from the date of
the final rule. Utilization of licensee's existing ISI schedules will
result in the volumetric examinations being implemented in a reasonable
period of time while not impacting lengths of outages or requiring
facility shutdown solely for performance of these examinations.
2.5 Voluntary Implementation
2.5.1 Section III
The NRC has reviewed the 1989 Addenda, 1990 Addenda, 1991 Addenda,
1992 Edition, 1992 Addenda, 1993 Addenda, 1994 Addenda, 1995 Edition,
and 1996 Addenda of Section III, Division 1, for Class 1, Class 2, and
Class 3 components, and has determined that they are acceptable for
voluntary use with six proposed limitations. In addition, Sec. 50.55a
would be modified to ensure consistency between Sec. 50.55a and NCA-
1140.
The version of Section III utilized by licensees is chosen prior to
construction. Section 50.55a permits licensees to use the original
construction code during the operational phase or voluntarily update to
a later version which has been endorsed by Sec. 50.55a. Accordingly,
the proposed limitations to Section III become effective only when a
licensee voluntarily updates to a later version. The modification would
only apply to a applicant for a new construction permit.
2.5.1.1 Limitations
2.5.1.1.1 Engineering Judgement
The first proposed limitation to the implementation of Section III
would establish an NRC restriction with regard to the Foreword in the
1992 Addenda through the 1996 Addenda of the BPV Code. That Foreword
addresses the use of ``engineering judgement'' for construction
activities not specifically considered by the Code. Proposed paragraph
50.55a(b)(1)(i) would require that when a licensee relies on
engineering judgement for activities or evaluations of components or
systems within the scope of Sec. 50.55a that are not directly addressed
by the BPV Code, the licensee must receive NRC approval for those
activities or evaluations pursuant to Sec. 50.55a(a)(3).
2.5.1.1.2 Section III Materials
The second proposed limitation to the implementation of Section III
pertains to a reference to Section II, ``Materials,'' Part D,
``Properties.'' Section II, Part D, contained many printing errors in
the 1992 Edition. These errors were corrected in the 1992 Addenda.
Proposed Sec. 50.55a(b)(1)(ii) would require that Section II, 1992
Addenda, be applied when using the 1992 Edition of Section III. The
limitation is necessary to ensure that users of the Code use the design
stresses intended by the ASME Code.
2.5.1.1.3 Weld Leg Dimensions
The third proposed limitation to the implementation of Section III
would
[[Page 63900]]
correct a conflict in the design and construction requirements in
Subsection NB (Class 1 Components), Subsection NC (Class 2), and
Subsection ND (Class 3) of Section III, 1989 Addenda through the 1996
Addenda of the BPV Code. Two equations in NB-3683.4(c)(1), Footnote 11
to Figure NC-3673.2(b)-1, and Figure ND-3673.2(b)-1 were modified in
the 1989 Addenda and are no longer in agreement with Figures NB-4427-1,
NC-4427-1, and ND-4427-1. This change results in a different weld leg
dimension depending on whether the dimension is derived from the text
or calculated from the figures. Thus, to ensure consistency, proposed
Sec. 50.55a(b)(1)(iii) would require that licensees use the 1989
Edition for the above referenced paragraphs and figures in lieu of the
1989 Addenda through the 1996 Addenda.
2.5.1.1.4 Seismic Design
The fourth proposed limitation to the implementation of Section III
pertains to new requirements for piping design evaluation contained in
the 1994 Addenda through the 1996 Addenda of the BPV Code. The NRC has
determined that changes to subarticles NB-3200, ``Design by Analysis,''
NB-3600, ``Piping Design,'' NC-3600, ``Piping Design,'' and ND-3600,
``Piping Design,'' of Section III for Class 1, 2, and 3 piping design
evaluation for reversing dynamic loads (e.g., earthquake and other
similar type dynamic loads which cycle about a mean value) are
unacceptable. The new requirements are based on the premise that loads
such as earthquake loads are not capable of producing collapse or gross
distortion of a component. The requirements, in part, are based on
General Electric evaluations of the test data performed under
sponsorship of the Electric Power Research Institute (EPRI) and the
NRC. However, NRC evaluations of the data do not support the changes
and indicate lower margins than those estimated in earlier evaluations.
The ASME has established a special working group to reevaluate the
bases for the seismic design for piping. Thus, in proposed
Sec. 50.55a(b)(1)(iv), licensees would be permitted to use articles NB-
3200, NB-3600, NC-3600, and ND-3600, in the 1989 Addenda through the
1993 Addenda, but would be prohibited from using these requirements in
the 1994 Addenda through the 1996 Addenda.
2.5.1.1.5 Quality Assurance
The fifth proposed limitation to the implementation of Section III
pertains to the use of NQA-1, ``Quality Assurance Requirements for
Nuclear Facilities,'' with Section III. Section III references NQA-1 as
part of its individual requirements for a QA program by integrating
portions of NQA-1 into the QA program defined in NCA-4000, ``Quality
Assurance.'' At present, Sec. 50.55a endorses the 1989 Edition of the
ASME Code which references NQA-1-1986 for Section III. The 1996 Addenda
of the ASME Code references NQA-1-1992 for Section III.
The NRC has reviewed the requirements of NQA-1, 1986 Addenda
through the 1992 Addenda, that are part of the incorporation by
reference of Section III, and has determined that the provisions of
NQA-1 are acceptable for use in the context of Section III activities.
Portions of NQA-1 are integrated into Section III administrative,
quality, and technical provisions which provide a complete QA program
for design and construction. NQA-1 by itself would not adequately
describe how to satisfy the requirements of 10 CFR Part 50, Appendix B,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants.'' The additional criteria contained in Section
III, such as nuclear accreditation, audits, and third party inspection,
establishes a complete program and satisfies the requirements of
Appendix B (i.e., the provisions of Section III integrated with NQA-1).
Because licensees may voluntarily choose to apply later provisions of
Section III, proposed Sec. 50.55a(b)(1)(v) contains a limitation which
would require that the edition and addenda of NQA-1 specified by NCA-
4000 of Section III be used in conjunction with the administrative,
quality, and technical provisions contained in the edition of Section
III being utilized.
2.5.1.1.6 Independence of Inspection
The sixth proposed limitation to the implementation of Section III
would prohibit licensees from using subparagraph NCA-4134.10(a),
``Inspection,'' in the 1995 Edition through the 1996 Addenda. Prior to
this edition and addenda, NCA-4134.10(a) required that the provisions
of NQA-1, ``Quality Assurance Program Requirements for Nuclear
Facilities,'' Basic Requirement 10, ``Inspection,'' and Supplement 10S-
1, ``Supplementary Requirements for Inspection,'' be utilized without
exception. In the 1995 Edition, NCA-4134.10(a) was modified so that
paragraph 2 of Supplement 10S-1 and the requirements for independence
of inspection were no longer required. Supplement 10S-1, 2.1, states
that ``Inspection Personnel shall not report directly to the immediate
supervisors who are responsible for performing the work being
inspected.'' Subparagraph 2.2 states ``Each person who verifies
conformance of work activities for purposes of acceptance shall be
qualified to perform the assigned task.'' By exempting Supplement 10S-1
paragraph 2 from the requirements of NCA-4134.10, Section III could
promote noncompliance with 10 CFR 50, Appendix B, ``Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,''
Criterion 1, ``Organization.'' This criterion requires that persons
performing QA functions report to a management level such that
authority and organizational freedom, including sufficient independence
from cost and schedule when opposed to safety considerations, are
provided. Thus, in proposed Sec. 50.55a(b)(1)(vi), licensees would be
permitted to use the provisions contained in NCA-4134.10(a), in the
1989 Addenda through the 1994 Addenda, but would be prohibited from
using these provisions in the 1995 Edition through the 1996 Addenda.
2.5.1.2 Modification
2.5.1.2.1 Applicable Code Version for New Construction
The proposed modification of Section III addresses a possible
conflict between NCA-1140 and Sec. 50.55a for new construction. NCA-
1140 of Section III requires that the length of time between the date
of the edition and addenda used for new construction and the docket
date of the nuclear power plant be no greater than three years.
Paragraph 50.55a(b)(1) requires that the edition and addenda utilized
be incorporated by reference into the regulations. The possibility
exists that the edition and addenda required by the ASME Code to be
used for new construction would not be incorporated by reference into
Sec. 50.55a. In order to resolve this possible discrepancy, the NRC
proposes to modify existing Secs. Sec. 50.55a(c)(3)(i),
50.55a(d)(2)(i), and 50.55a(e)(2)(i), to permit an applicant for a
construction permit to use the latest edition and addenda which has
been incorporated by reference into Sec. 50.55a(b)(1) if the
requirements of the ASME Code and the regulations cannot simultaneously
be satisfied.
2.5.2 Section XI (Voluntary Implementation)
Licensees would be permitted to update from the 1992 Edition with
the 1992 Addenda of Subsection IWE and Subsection IWL to the 1995
Edition with the 1996 Addenda. In addition, licensees could implement
Code Case
[[Page 63901]]
N-513, ``Evaluation Criteria for Temporary Acceptance of Flaws in Class
3 Piping,'' and Code Case N-523-1, ``Mechanical Clamping Devices for
Class 2 and 3 Piping.''
2.5.2.1 Subsection IWE and Subsection IWL
Many of the provisions in Section XI Subsection IWL, ``Requirements
for Class CC Concrete Components of Light-Water Cooled Power Plants,''
pertaining to the inspection of the tendons of concrete containments
were based on guidance contained in Regulatory Guide 1.35, ``Inservice
Inspection of Ungrouted Tendons in Prestressed Concrete Containments.''
A final rule published on August 8, 1996 (61 FR 41303) incorporated by
reference the 1992 Edition with the 1992 Addenda of Subsection IWE,
``Requirements for Class MC and Metallic Liners of Class CC Components
of Light-Water Cooled Power Plants,'' and Subsection IWL. At that time,
there were several key positions in the regulatory guide addressing the
trending of prestress losses, unanticipated tendon elongation, grease
leakage, and excessive water in the sampled sheathing filler grease not
addressed in Subsection IWL because the ASME Code committees had not
yet completed consideration of these positions. Due to the importance
of these positions, the final rule addressed them in paragraphs
50.55a(b)(2)(ix)(A) through 50.55a(b)(2)(ix)(D)(3). In addition, the
final rule contained Sec. 50.55a(b)(2)(ix)(E) which addressed the
occurrence of degradation in inaccessible areas of containments.
Since publication of the 1992 Addenda, the ASME Code committees
have completed their consideration of those regulatory guide positions.
Most have been incorporated into subsequent edition and addenda, and
the 1995 Edition with the 1996 Addenda addresses all of the
modifications listed above except grease leakage and degradation in
inaccessible areas. Thus, licensees would be required to utilize the
modifications presently in Sec. 50.55a addressing grease leakage and
degradation in inaccessible areas. The NRC has determined that the
provisions contained in Subsection IWE and Subsection IWL, 1995 Edition
with the 1996 Addenda Code, in conjunction with the modifications,
would be acceptable.
The final rule published on August 8, 1996 (61 FR 41303)
incorporated Subsection IWE and Subsection IWL into Sec. 50.55a for the
first time. The final rule contained a requirement for licensees to
develop and implement a containment ISI program within five years. Each
plant had a pre-existing ISI program to address Class 1, Class 2, and
Class 3 components. The rule left it to the licensee's discretion
whether to have two separate ISI programs, or merge the containment ISI
program with the pre-existing program.
It has been over a year since the final rule was issued, and some
licensees have begun the development of a containment ISI program to
comply with the required 5-year implementation period. This containment
ISI program will be based on the 1992 Edition with the 1992 Addenda as
required by the final rule. However, other licensees have indicated
that they will request NRC approval pursuant to Sec. 50.55a(a)(3) to
use later editions and addenda of Subsection IWE and Subsection IWL
before this proposed rule becomes final. Thus, to provide flexibility,
Sec. 50.55a(b)(2)(vi) has been modified. Licensees would be permitted
to implement either the presently required 1992 Edition with the 1992
Addenda, or the latest containment examination provisions; i.e., 1995
Edition with the 1996 Addenda.
For those licensees implementing the 1992 Edition with the 1992
Addenda, all of the modifications contained in paragraphs
50.55a(b)(2)(ix)(A) through 50.55a(b)(2)(ix)(D)(3) must be applied as
presently required by Sec. 50.55a. Licensees wishing to implement the
1995 Edition with the 1996 Addenda would be required to apply
paragraphs 50.55a(b)(2)(ix)(A), 50.55a(b)(2)(ix)(D)(3), and
50.55a(b)(2)(ix)(E). Paragraph Sec. 50.55a(b)(2)(ix) would thus be
modified. According to Sec. 50.55a(g)(6)(ii)(B)(1), the containment
examinations performed during the 5-year implementation period are
those examinations which are required by Subsection IWE during the
first period of what will be the first containment inspection interval.
(Since Subsection IWL is based on a 5-year schedule, standard Section
XI periods do not apply for the examination of concrete containments
and their post-tensioning systems). With completion of the first period
examinations, the second period of the first containment ISI interval
would begin. The end of the third period completes the first
containment ISI interval, a containment ISI 120-month update has been
completed, and the second containment ISI interval would begin.
As licensees have begun developing their containment ISI programs,
the NRC has received requests to clarify the implementation schedule
for ISI of concrete containments and their post-tensioning systems. The
current wording of Sec. 50.55a(g)(6)(ii)(B)(2) requiring licensees to
implement ``the inservice examinations which correspond to the number
of years of operation which are specified in Subsection IWL'' has
created confusion regarding whether the first examination of concrete
is required to meet the examination schedule in Section XI, Subsection
IWL, IWL-2410, which is based on the date of the Structural Integrity
Test (SIT), or may be performed at any time between September 9, 1996
and September 9, 2001. According to Sec. 50.55a(g)(6)(ii)(B)(2) of the
final rulemaking, the first examination of concrete may be performed at
any time between September 9, 1996, and September 9, 2001. The date of
the first examination of concrete is not conditional upon compliance
with Subsection IWL-2410 or the SIT. The purpose of the italicized
words is to maintain the present 5-year schedule for examination of the
post-tensioning system as operating plants transition to Subsection
IWL. For operating reactors, there is no need to repeat the 1, 3, 5-
year implementation cycle.
Section 50.55a(g)(6)(ii)(B)(2) also stated that the first
examination performed shall serve the same purpose for operating plants
as the preservice examination specified for plants not yet in
operation. The affected plants are presently operating, but they will
be performing the examination of concrete under Subsection IWL for the
first time. Because the plants are operating, a Section XI preservice
examination cannot be performed. Therefore, the first concrete
examination is to be an inservice examination which will serve as the
baseline (the same purpose for operating plants as the preservice
examination specified for plants not yet in operation). With completion
of this first examination of concrete, the second five-year Subsection
IWL ISI period would begin. Likewise, examinations of the post-
tensioning system at the nth year (e.g., the 15th year post-tensioning
system examination), if performed to the requirements of Subsection
IWL, are to be performed to the ISI requirements, not the preservice
requirements.
The NRC has also been requested to clarify the schedule for future
examinations of concrete and their post-tensioning systems at both
operating and new plants. There is no requirement in Subsection IWL to
perform the examination of the concrete and the examination of the
post-tensioning system at the same time. The examination of the
concrete under Subsection IWL and the examination of
[[Page 63902]]
the liner plates of concrete containments under Subsection IWE may be
performed at any time during the 5-year expedited implementation. This
examination of the concrete and liner plate provides the baseline for
comparison with future containment ISI. Coordination of these schedules
in future examinations is left to each licensee. New plants would be
required to follow all of the provisions contained in Subsection IWL,
i.e., satisfy the preservice examination requirements and adopt the 1,
3, 5-year examination schedule ISI schedule.
2.5.2.2 Flaws in Class 3 Piping
Proposed Sec. 50.55a(b)(2)(xvi) would permit licensees to use Code
Case N-513, ``Evaluation Criteria for Temporary Acceptance of Flaws in
Class 3 Piping,'' and Code Case N-523-1, ``Mechanical Clamping Devices
for Class 2 and 3 Piping.'' Section XI contains repair methods for
pipes with a flaw exceeding acceptable limits. These repairs restore
the integrity of the flawed piping. There are certain cases, however,
where a Section XI Code repair may be impractical for a flaw detected
during plant operation (i.e., a plant shutdown would be required to
effect the Code repair). For many safety-related piping systems,
immediate repair is required regardless of plant status. However, it
has been determined that under certain conditions, temporary acceptance
of flaws, including through-wall leaking, of low and moderate energy
Class 3 piping is acceptable provided that the conditions are met, and
the repair is effected during the next outage. At present, licensees
must request NRC staff approval to defer Section XI Code repair for
these Class 3 moderate energy (200 xF, 275 psig) piping systems. The
NRC has reviewed Code Case N-513 and Code Case N-523-1 and has
determined that Code Case N-523-1 is acceptable. Code Case N-513 is
acceptable except for the scope and Section 4.0.
Section 1.0(a) of the Scope to Code Case N-513 limits the use of
the requirements to Class 3 piping. However, Section 1.0(c) would allow
the flaw evaluation criteria to be applied to all sizes of ferritic
steel and austenitic stainless steel pipe and tube. Without some
limitation on the scope of the Code Case, the flaw evaluation criteria
could be applied to components such as pumps and valves, original
construction deficiencies, and pressure boundary leakage; applications
for which the criteria should not be utilized. Thus, the NRC has
determined that the Code Case shall not be applied to: (1) components
other than pipe and tube, such as pumps, valves, expansion joints, and
heat exchangers; (2) the discovery and repair of flaws or deficiencies
remaining from original construction; (3) leakage through a flange
gasket; (4) threaded connections employing nonstructural seal welds for
leakage prevention (through seal weld leakage is not a structural flaw,
thread integrity must be maintained); and (5) degraded socket welds. A
proposed limitation would be added in Sec. 50.55a(b)(2)(xvi)(B) which
would preclude the use of Code Case N-513 for these applications.
The first paragraph of Section 4.0 of Code Case N-513 contains the
flaw acceptance criteria. The criteria provide a safety margin based on
service loading conditions. The second paragraph of Section 4.0,
however, would permit a reduction of the safety factors based on a
detailed engineering evaluation. No criteria or guidance is given for
justifying a reduction, or limiting the amount of reduction. The
acceptance criteria of the first paragraph are based on sound
principles. The second paragraph would allow ever finer calculation
until the available margins became unacceptably low. A limitation would
be added in proposed Sec. 50.55a(b)(2)(xvi)(A) requiring that when
implementing Code Case N-513, the specific safety factors in the first
paragraph of Section 4.0 be satisfied. The use of Code Case N-513, with
the limitations, and Code Case N-523-1 would obviate the need for
licensees to request approval for deferring repairs, thus saving NRC
and licensee resources.
2.5.3 OM Code (Voluntary Implementation)
Licensees would be permitted to implement Code Case OMN-1 in lieu
of stroke time testing as required in Subsection ISTC. Licensees would
also be permitted to implement Appendix II as an alternative to the
condition monitoring program provisions contained in Subsection ISTC.
However, licensees choosing to implement Appendix II would be required
to apply the three proposed modifications to Appendix II to supplement
check valve condition monitoring. In addition, licensees would be
permitted to use Subsection ISTD for the IST of snubbers.
2.5.3.1 Code Case OMN-1
An alternative to the provisions contained in Sec. 50.55a(b)(3)(ii)
is included in proposed Sec. 50.55a(b)(3)(iii) which would permit
licensees to voluntarily implement ASME Code Case OMN-1, ``Alternative
Rules for Preservice and Inservice Testing of Certain Electric Motor
Operated Valve Assemblies in LWR Power Plants.'' The NRC has determined
that for motor-operated valves, Code Case OMN-1 is acceptable in lieu
of Subsection ISTC, except for leakage rate testing (ISTC 4.3) which
must continue to be performed. As indicated in Attachment 1 to GL 96-
05, the Code case meets the intent of the generic letter, but with
certain limitations which were discussed in the generic letter. The NRC
supports the OMN-1 maximum motor-operated valve test interval of 10
years based on current knowledge and experience, but believes it
prudent to require that licensees evaluate the information obtained for
each motor-operated valve during the first five years of use of the
Code case, or three refueling outages (whichever is longer) to validate
assumptions made in justifying a longer test interval. These
limitations on the use of OMN-1 would be added to the rule as a
modification in Sec. 50.55a(b)(3)(iii)(A). Thus, Code Case OMN-1 is
acceptable in lieu of Subsection ISTC, other than leakage rate testing
requirements, with the modification that five years or three refueling
outages (whichever is longer) from initial implementation of Code Case
OMN-1, the adequacy of the test interval for each motor-operated valve
must be evaluated and adjusted as necessary.
In addition, as noted in GL 96-05, licensees are cautioned when
implementing Code Case OMN-1 that the benefits of performing a
particular test should be balanced against the potential adverse
effects placed on the valves or systems caused by this testing. Code
Case OMN-1 specifies that an IST program should consist of a mixture of
static and dynamic testing. While there may be benefits to performing
dynamic testing, there are also potential detriments to its use (i.e.,
valve damage). Licensees should be cognizant of this for each MOV when
selecting the appropriate method or combination of methods for the IST
program.
2.5.3.2 Appendix II
Paragraph ISTC 4.5.5 of Subsection ISTC permits the Owner to use
Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM
Code, as an alternative to the testing or examination provisions of
ISTC 4.5.1 through ISTC 4.5.4. If an Owner elects to use Appendix II,
the provisions of Appendix II become mandatory. However, upon reviewing
the appendix, the NRC has determined that the requirements in Appendix
II must be supplemented. The first area that the NRC believes requires
supplementation is the demonstration of acceptable valve performance.
Appendix II requires no testing or examination of the check
[[Page 63903]]
valve obturator movement to both the open and closed positions. Testing
or examination of the check valve obturator in one direction only
cannot assure the unambiguous detection of a functionally degraded
check valve. The valve obturator must be tested or examined in both the
opening and closing directions to assess its condition and confirm
acceptable performance. Proposed Sec. 50.55a(b)(3)(iv)(A) would require
bi-directional testing of check valves.
Length of test interval is the second area of Appendix II where the
NRC believes the rules must be supplemented. Appendix II was first
incorporated into the OM Code in the 1996 Addenda. Thus, the operating
experience database does not yet exist to support long term test
intervals for the condition monitoring concept. Under the current check
valve IST program, most valves are tested quarterly during plant
operation. The interval for certain valves has been extended to
refueling outages. Under the appendix, a licensee would be able to
extend the interval without limit. A policy of prudent and safe
interval extension dictates that any additional interval extension must
be limited to one fuel cycle, and this extension must be based on
sufficient experience to justify the additional time. Interval changes
or extensions must be justified and limited within the existing
performance and experience database. Condition monitoring and the
current experience data base may qualify some valves for an initial
extension to every other fuel cycle, while trending and evaluation of
the data may dictate that the testing interval for some valves be
reduced. Extensions of IST intervals must consider plant safety and be
supported by trending and evaluating both generic and plant-specific
performance data to ensure the component is capable of performing its
intended function over the entire IST interval. Proposed
Sec. 50.55a(b)(3)(iv)(B) would limit the time between the initial test
or examination and second test or examination to two fuel cycles or
three years (whichever is longer), with additional extensions limited
to one fuel cycle, and the total interval would be limited to a maximum
of 10 years. An extension or reduction in the interval between tests or
examinations would have to be supported by trending and evaluation of
performance data.
The final area in Appendix II which the Commission believes should
be supplemented is the requirement applicable to a licensee who
discontinues a condition monitoring program. A licensee who
discontinues use of Appendix II, under IST 4.5.5 is required to return
to the requirements of IST 4.5.4. However, the NRC believes the
requirements of IST 4.5.1 through IST 4.5.4 must be also met. Hence, if
the monitoring program is discontinued, proposed
Sec. 50.55a(b)(3)(iii)(C) would require a licensee to implement the
provisions of IST 4.5.1 through IST 4.5.4.
2.5.3.3 Subsection ISTD
The IST of dynamic restraints or snubbers is governed by plant
technical specification and, thus, has never been included in
Sec. 50.55a. However, the NRC has reviewed Subsection ISTD, 1995
Edition with the 1996 Addenda, and has determined that the provisions
for IST of snubbers are an acceptable alternative to the requirements
contained in the plant technical specifications. Subsection ISTD, 1996
Addenda, includes new provisions for service life monitoring of
snubbers. The new provisions require that the service lives of snubbers
be predicted and evaluated to ensure that the service life will not be
exceeded before the next scheduled refueling outage. These new
provisions simply formalize preventative maintenance practices
presently found in most plants. Because the IST of snubbers is governed
by plant technical specifications, Subsection ISTD is not included in
the proposed mandatory requirements of the rulemaking, but licensees
may choose to voluntarily implement Subsection ISTD, 1995 Edition with
the 1996 Addenda, by processing a change to their technical
specifications. This proposed modification is contained in
Sec. 50.55a(b)(3)(v).
2.5.3.4 Containment Isolation Valves
The proposed amendment would delete the existing modification in
Sec. 50.55a(b)(2)(vii) for IST of containment isolation valves (CIVs),
which was added to the regulations in a rulemaking effective on August
6, 1992 (57 FR 34666). That rulemaking incorporated by reference, among
other things, the 1989 Edition of ASME Section XI, Subsection IWV that
endorsed Part 10 of ASME/ANSI OMa1988 for valve inservice testing. A
modification to the testing requirements of Part 10 related to CIVs was
included in the rulemaking indicating that paragraphs 4.2.2.3(e) and
4.2.2.3(f) of Part 10 were to be applied to CIVs. As noted in the
``Supplementary Information'' for the August 6, 1992 rulemaking, the
ASME Operations and Maintenance (OM) Committee had initiated action to:
(1) perform a comprehensive review of OM Part 10 CIV testing
requirements and acceptance standards; and (2) develop a basis document
that would provide, as a minimum, a documented basis for not including
the requirements for analysis of leakage rates and corrective actions
in Part 10 for those CIVs that do not provide a reactor coolant system
pressure isolation function. The NRC made a commitment via the
Supplementary Information to reevaluate the need for the modification
to Section XI, Subsection IWV, following review of this OM Committee
basis document. This basis document was transmitted to the NRC in a
letter from Steve Weinman, Secretary, OM Committee, to Eric S.
Beckjord, Director, Office of Nuclear Regulatory Research, dated
February 16, 1994. The NRC has determined that the requirements of 10
CFR 50, Appendix J, ensure adequate identification analysis, and
corrective actions for leakage monitoring of CIVs, and that the
existing modification in Sec. 50.55a(b)(2)(vii) should be deleted. The
regulatory analysis for this proposed rule contains a detailed
discussion of the basis document findings and the NRC staff evaluation.
2.6 ASME Code Interpretations
The ASME issues Interpretations to clarify provisions of the BPV
and OM Codes. Requests for Interpretations are submitted by users, and
after appropriate committee deliberations and balloting, responses are
issued by the ASME. Generally, the NRC agrees with these
interpretations. When the NRC incorporates by reference specific
editions and addenda into its regulations, the NRC has a certain
understanding of those editions and addenda. Because an Interpretation
is issued subsequent to issuance of the provision to which it refers,
the Interpretation may affect that understanding. While the NRC
acknowledges that the ASME is the official interpreter of the Code, the
NRC will not accept ASME interpretations that, in NRC's opinion, are
contrary to NRC requirements or may adversely impact facility
operations. Interpretations have been issued which in some cases,
conflicted with or were inconsistent with NRC requirements. These
resulted in enforcement actions. Of particular concern are Code
Interpretations that may be implemented following initiation of
enforcement action by the NRC. ASME Code Interpretations were discussed
in Part 9900, Technical Guidance, of the NRC Inspection Manual. Part
9900 provides that licensees should exercise caution when applying
Interpretations as they are not specifically part of the
[[Page 63904]]
incorporation by reference into Sec. 50.55a and have not received NRC
approval.
2.7 DSI-13
Since 1992, when the Commission last revised Sec. 50.55a to endorse
new ASME Code Editions and addenda (57 FR 34666), several developments
have occurred which have raised some fundamental issues with respect to
the Commission's endorsement of ASME Codes. First, on October 21, 1993,
Entergy Operations, Inc. submitted a request that would relieve it from
updating its ISI and IST programs to the last ASME Code edition and
addenda incorporated by reference into Sec. 50.55a. The underlying
premise of the request was that a licensee should not be required to
upgrade its ISI and IST program without considering whether the costs
of the upgrade are warranted in light of the increased safety afforded
by the updated Code edition and addenda. Though the request was later
withdrawn, the underlying premise resulted in NRC reconsideration of
the 120-month update. Requiring Code updates every 120-months is still
under active consideration. However, the proposed rule has been
prepared under the traditional approach; i.e., licensees would be
required to update their ISI and IST programs every 120-months to the
latest edition and addenda incorporated by reference into Sec. 50.55a.
If a decision is reached subsequent to publication of the proposed rule
that is adverse to this approach, this position will be corrected prior
to publication of the final rule.
Second, the National Technology Transfer and Advancement Act of
1995, PL 104-113, was signed into law on March 7, 1996. The Act directs
federal agencies to achieve greater reliance on technical standards
developed by voluntary consensus standards development organizations.
Finally, the Commission commenced a Strategic Assessment and
Rebaselining Initiative. One of the issues addressed in this effort was
Direction Setting Issue (DSI) 13, which raised the question, ``In
performing its regulatory responsibilities, what consideration should
the NRC give to industry activities.'' A draft paper addressing DSI-13
was published for public comment on September 16, 1996, after which the
Commission held public meetings to facilitate understanding of the
issues and receive comments on the DSI-13 draft paper. Based on the
public comments, the Commission has directed the NRC Staff to address
how industry initiatives should be evaluated, and to evaluate several
issues related to NRC endorsement of industry codes and standards. As
part of this evaluation, the Staff is addressing issues relevant to the
NRC's endorsement of the ASME Code, including periodic updating, the
impact of 10 CFR 50.109 (the Backfit Rule), and streamlining the
process for NRC review and endorsement of the ASME Code.
2.8 Steam Generators
ASME Code requirements for repair of heat exchanger tubes by
sleeving were added to Section XI in the 1989 Addenda. Minimum Code
requirements for tube sleeving was added to the Code so that licensees
would not have to develop sleeving programs and have them approved by
the NRC on a case-by-case basis. The NRC has reviewed the Code
requirements for sleeving and determined that they are acceptable.
However, it should be recognized that there are other relevant
requirements, and that a considerable amount of effort is presently
being expended due to the number of occurrences of degraded steam
generator tubing. For example, licensees are required by either 10 CFR
50.55a(f) or by the plant technical specifications to perform periodic
inservice inspections and to repair (e.g., sleeving) or remove from
service (by installing plugs in the tube ends) all tubes found to
contain flaws exceeding the plugging limit (i.e., tube repair
criteria). In addition, current technical specifications contain
operational leakage limits. Licensee's have frequently found it
necessary to implement measures beyond minimum Code and technical
specification requirements to ensure adequate tube integrity when
significant degradation problems are encountered. Thus, the NRC
determination that the sleeving requirements are acceptable should be
kept in perspective.
3. Finding of No Significant Environmental Impact
Based upon an environmental assessment, 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 have a significant effect on
the quality of the human environment and therefore an environmental
impact statement is not required.
The proposed rule is one part of a regulatory framework directed to
ensuring pressure boundary integrity and the operational readiness of
pumps and valves. The proposed rule incorporates provisions contained
in the BPV Code and the OM Code for the construction, inservice
inspection, and inservice testing of components used in nuclear power
plants, has been updated to incorporate improved technology and
methodology. Therefore, in the general sense, the proposed rule would
have a positive impact on the environment.
The proposed rule would impose the Section XI 1995 Edition with the
1996 Addenda. As most of the technical changes to this edition/addenda
merely incorporate improved technology and methodology, imposition of
these requirements is not expected to either increase or decrease
occupational exposure. However, imposition of paragraphs IWF-2510,
Table IWF-2500-1, Examination Category F-A, and IWF-2430, would result
in fewer supports being examined which would decrease the occupational
exposure compared to present support inspection plans. It is estimated
that an examiner receives approximately 100 millirems for every 25
supports examined. Adoption of the new provisions is expected to
decrease the total number of supports to be examined by approximately
115 per unit per interval. Thus, the reduction in occupational exposure
is estimated to be 460 millirems per unit each inspection interval or
50.14 rems for 109 units.
The proposed rule would impose Appendix VIII to Section XI, 1995
Edition with the 1996 Addenda, BPV Code, for the first time and would
expedite its implementation. Appendix VIII provides rules for the
performance demonstration of ultrasonic examination systems,
procedures, and personnel. Implementation of this appendix should
result in a decrease in occupational exposure. Appendix VIII qualified
procedures and personnel should reduce repeat ultrasonic testing (UT),
which could reduce occupational exposure. In addition, flaws should be
detected at an earlier stage of growth resulting in less extensive
repair operations, which could further reduce occupational exposure.
The proposed rule would incorporate by reference into the
regulations the 1995 Edition with the 1996 Addenda of the OM Code.
Imposition of the OM Code is not expected to either increase or
decrease occupational exposure. The types of testing associated with
the 1995 Edition with the 1996 Addenda of the OM Code are essentially
the same as the OM standards contained in the 1989 Edition of Section
XI referenced in a final rule published on August 6, 1992 (57 FR
34666).
Actions required of applicants and licensees to implement the
proposed rule are of the same nature as those applicants and licensees
have been performing for many years. Therefore, this action should not
increase the
[[Page 63905]]
potential for a negative environmental impact.
The NRC has sent a copy of the Environmental Assessment and the
proposed rule to every State Liaison Officer and requested their
comments on the Environmental Assessment. The environmental assessment
is available for inspection at the NRC Public Document Room, 2120 L
Street NW (Lower Level), Washington, DC. Single copies of the
environmental assessment are available from Frank C. Cherny, Division
of Engineering Technology, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:
301-415-6786, or Wallace E. Norris, Division of Engineering Technology,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
Telephone: 301-415-6796.
4. Paperwork Reduction Act Statement
This proposed rule amends information collection requirements that
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). This rule has been submitted to the Office of Management and
Budget for review and approval of the paperwork requirements.
The public reporting burden for this information collection is
estimated to average 67 person-hours per response, including the time
for reviewing instructions, searching existing data sources, gathering
and maintaining the data needed, and completing and reviewing the
collection of information. The U.S. Nuclear Regulatory Commission is
seeking public comment on the potential impact of the information
collections contained in the proposed rule and on the following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
2. Is the estimate of burden accurate?
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
4. How can the burden of the information collection be minimized,
including the use of automated collection techniques?
Send comments on any aspect of this proposed collection of
information, including suggestions for further reducing the burden, to
the Information and Records Management Branch (T-6 F33), U.S. Nuclear
Regulatory Commission, Washington DC 20555-0001, or by Internet
electronic mail at [email protected]; and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of
Management and Budget, Washington DC 20503.
Comments to OMB on the information collections or on the above
issues should be submitted by January 2, 1998. Comments received after
this date will be considered if it is practical to do so, but assurance
of consideration cannot be given to comments received after this date.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.
5. Regulatory Analysis
The Commission has prepared a draft regulatory analysis on this
proposed regulation. The analysis examines the costs and benefits of
the alternatives considered by the Commission. The draft analysis is
available for inspection in the NRC Public Document Room, 2120 L Street
NW (Lower Level), Washington DC. The Commission requests public comment
on the draft analysis. Single copies of the analysis may be obtained
from Frank C. Cherny, Division of Engineering Technology, Office of
Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Telephone: 301-415-6786, Wallace E. Norris,
Division of Engineering Technology, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Telephone: 301-415-6796.
6. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rule will not, if
promulgated, have a significant economic impact on a substantial number
of small entities. This proposed 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 Small Business Size
Standards set out in regulations issued by the Small Business
Administration at 13 CFR Part 121.
7. Backfit Analysis
The Nuclear Regulatory Commission (NRC) regulations, 10 CFR 50.55a,
requires that nuclear power plant owners (1) construct Class 1, Class
2, and Class 3 components in accordance with the rules provided in
Section III, Division 1, ``Requirements for Construction of Nuclear
Power Plant Components,'' of the American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), (2)
inspect Class 1, Class 2, Class 3, Class MC (metal containment) and
Class CC (concrete containment) components in accordance with the rules
provided in Section XI, Division 1, ``Requirements for Inservice
Inspection of Nuclear Power Plant Components,'' of the BPV Code, and
(3) test Class 1, Class 2, and Class 3 pumps and valves in accordance
with the rules provided in Section XI, Division 1. Licensees are
required to update every 120 months to the version of Section XI
incorporated by reference into Sec. 50.55a 12 months prior to the start
of a new ten year interval.
The proposed amendment to Sec. 50.55a would require licensees to
update ISI in accordance with Section XI of the ASME BPV Code and IST
in accordance with the ASME OM Code. Licensees would be required to
implement the 1995 Edition with the 1996 Addenda of (1) Section XI,
Division 1 for Class 1, Class 2, Class 3, Class MC, and Class CC
components; (2) the ``Code for Operation and Maintenance of Nuclear
Power Plants'' (OM Code) for Class 1, Class 2, and Class 3 pumps and
valves; and (3) Appendix VIII, ``Performance Demonstration for
Ultrasonic Examination Systems,'' to Section XI, Division 1. As
permitted by Sec. 50.55a(a)(3), licensees may voluntarily update to the
1989 Addenda through the 1996 Addenda of Section III of the BPV Code,
with limitation. In addition, the modification for containment
isolation valve inservice testing that applied to the 1989 Edition of
the BPV Code has been deleted. Licensees will continue to be required
to update their ISI and IST programs every 120 months to the version of
Section XI and the OM Code incorporated by reference and in effect at
least 12 months prior to the start of a new 120-month interval.
The NRC position on the routine 120-month update to Sec. 50.55a has
consistently been that 10 CFR 50.109 does not require a backfit
analysis of the routine 120-month update to Sec. 50.55a. The basis for
the NRC position is that, (1) Section III, Division 1, update applies
only to new construction (i.e., the edition and addenda to be used in
the construction of a plant are selected based upon the date of the
construction permit and are not changed thereafter, except voluntarily
by the licensee), (2) licensees understand that Sec. 50.55a requires
that they update their inservice inspection program every 10 years to
the latest edition and addenda of Section XI that were incorporated by
reference in Sec. 50.55a and in effect 12 months before
[[Page 63906]]
the start of the next inspection interval, and (3) endorsing and
updating references to the ASME Code, a national consensus standard
developed by the participants (including the NRC) with broad and varied
interests, is consistent with both the intent and spirit of the backfit
rule (i.e., NRC provides for the protection of the public health and
safety, and does not unilaterally impose undue burden on applicants or
licensees). Finally, to ensure that any interested member of the public
that may not have had an opportunity to participate in the national
consensus standard process is able to communicate with the NRC,
proposed rules are published in the Federal Register.
The provisions for IST of pumps and valves were originally
contained in Section XI Subsections IWP and IWV. Section XI, 1989
Edition was incorporated by reference in the August 6, 1992 rulemaking
(57 FR 34666). The 1990 OM Code standards, Parts 1, 6, and 10 of ASME/
ANSI-OM-1987, are identical to Section XI, 1989 Edition. This proposed
amendment is an administrative change simply referencing the 1995
Edition with the 1996 Addenda of the OM Code. Therefore, imposition of
the 1995 Edition with the 1996 Addenda of the OM Code is not a backfit.
Appendix VIII, ``Performance Demonstration for Ultrasonic
Examination Systems,'' to Section XI would be used to demonstrate the
qualification of personnel and procedures for performing nondestructive
examination of welds in components of systems that include the reactor
coolant system and the emergency core cooling systems in nuclear power
facilities. Appendix VIII would greatly enhance the reliability of
detection and sizing of cracks and flaws, and it delineates a method
for qualification of the personnel and procedures. The appendix would
normally be imposed by the 120-month update requirement, but because of
its importance, implementation of Appendix VIII is being expedited by
the rulemaking. Because of the expedited implementation schedule, the
imposition of Appendix VIII is being considered a backfit. Licensees
would be required to implement Appendix VIII, including the
modifications, for all examinations of the pressure vessel, piping,
nozzles, and bolts and studs which occur after 6 months from the date
of the final rule. The proposed rule would not require any change to a
licensee's ISI schedule for examination of these components, but would
require that the provisions of Appendix VIII be used for all
examinations after that date rather than the UT procedures and
personnel requirements presently being utilized by licensees.
The NRC has concluded, on the basis of the documented evaluation
required by Sec. 50.109(a)(4), that imposition of Appendix VIII, which
would greatly enhance the overall level of assurance of the safety and
reliability of ultrasonic examination techniques in detecting and
sizing flaws, is necessary to bring the facilities described into
compliance with GDC 14, 10 CFR Part 50, Appendix A, or similar
provisions in the licensing basis for these facilities, and Criteria II
and XVI, of 10 CFR Part 50, Appendix B.
The modification to Section XI to require licensees to supplement
the surface examination of the Class 1 portion (RCPB) of the HPSI
system with volumetric examination would ensure the integrity of the
reactor coolant system pressure boundary and maintenance of emergency
core cooling system operability. The operability of this system is
necessary to ensure the protection of the public health and safety, and
the NRC has concluded, on the basis of the documented evaluation
required by Sec. 50.109(a)(4), that licensees must supplement the
Section XI required surface examination for the Class 1 portion of the
HPSI system with volumetric examination in order to ensure the
integrity of the reactor coolant pressure boundary as required by GDC
14, 10 CFR Part 50, Appendix A, or similar provisions in the licensing
basis for these facilities, and Criteria II and XVI, of 10 CFR Part 50,
Appendix B. Volumetric examination would be required during any ISI
program inspection of the HPSI system performed after 6 months from the
date of the final rule.
GDC 14, ``Reactor coolant pressure boundary,'' (RCPB) or similar
provisions in the licensing basis for these facilities, specify that
the RCPB be designed, fabricated, erected, and tested so as to have an
extremely low probability of abnormal leakage, or rapidly propagating
failure, and of gross rupture. There has recently been an occurrence of
gross rupture in the Class 1 portion of a HPSI system, and a number of
occurrences of abnormal leakage in the RCPB in other plants.
Imposition of Appendix VIII and the HPSI volumetric examination is
also necessary to bring the facilities described into compliance with
Criteria II, ``Quality Assurance Program,'' and Criteria XVI,
``Corrective Actions,'' of Appendix B to 10 CFR Part 50. Criteria II
requires, in part, that a QA program shall take into account the need
for special controls, processes, test equipment, tools, and skills to
attain the required quality and the need for verification of quality by
inspection and test. Evidence indicates that there are shortcomings in
the qualifications of personnel and procedures in ensuring the
reliability of the examinations. These safety significant revisions to
the Code include specific requirements for UT performance
demonstration, with statistically based acceptance criteria for blind
testing of UT systems (procedures, equipment, and personnel) used to
detect and size flaws. Criteria XVI requires that measures shall be
established to assure that conditions adverse to quality, such as
failures, malfunctions, deficiencies, deviations, defective material
and equipment, and nonconformances are promptly identified and
corrected. In analyzing the occurrences of pipe break and leakage, it
is apparent that the RCPB is subject to certain types of degradation.
Information gathered by the NRC staff indicates that many licensees
have not reacted to this serious safety concern by performing more
comprehensive examinations. The NRC believes that there is a basis for
reasonably concluding that such degradation could occur in virtually
all PWRs. Because of the serious degradation which has occurred, and
the belief that additional occurrences of noncompliance with GDC 14,
and Criteria II and XVI will be reported, the NRC has determined that
imposition of Appendix VIII and volumetric examination of the HPSI
system 6 months after the final rule has been published under the
compliance exception to Sec. 50.109(a)(4)(i) is appropriate, therefore,
a backfit analysis is not required and the cost-benefit standards of
Sec. 50.109(a)(3) do not apply. A complete discussion is contained in
the documented evaluation.
The rationale for application of the backfit rule and the backfit
justification for the various items contained in this proposed rule are
contained in the regulatory analysis and documented evaluation. The
regulatory analysis and documented evaluation are available for
inspection at the NRC Public Document Room, 2120 L Street NW (Lower
Level), Washington, DC. Single copies of the regulatory analysis and
documented evaluation are available from Frank C. Cherny, Division of
Engineering Technology, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:
301-415-6786, or Wallace E. Norris, Division of Engineering Technology,
Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory
Commission,
[[Page 63907]]
Washington, DC 20555-0001, Telephone: 301-415-6796.
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Fire prevention, Incorporation
by reference, Intergovernmental relations, Nuclear power plants and
reactors, Penalties, Radiation protection, Reactor siting criteria,
Reporting and recordkeeping requirements.
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. 553, the NRC is proposing to
adopt the following amendments to 10 CFR Part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for Part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 1444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub.
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
2. Section 50.55a is amended by removing and reserving paragraphs
(b)(2)(vii) and (g)(4)(iv), adding paragraphs (b)(2)(xi) through
(b)(2)(xx), (b)(3), (g)(6)(ii)(A)(6), and (g)(6)(ii)(C), and revising
the introductory text of paragraph (b), paragraph (b)(1), the
introductory text of paragraph (b)(2), paragraphs (b)(2)(iv),
(b)(2)(vi), (b)(2)(viii), the introductory text of paragraph
(b)(2)(ix), paragraphs (c)(3), (d)(2), (e)(2), the introductory text of
paragraph (f), paragraphs (f)(1), (f)(2), (f)(3)(iii), (f)(3)(iv), the
introductory text of paragraph (f)(4), paragraphs (g)(1), (g)(3)(i),
the introductory text of paragraph (g)(4), paragraphs (g)(6)(ii)(A)(1),
(g)(6)(ii)(A)(2), and Footnotes 5 and 7 to read as follows:
Sec. 50.55a Codes and standards.
* * * * *
(b) The ASME Boiler and Pressure Vessel Code, and the ASME Code for
Operation and Maintenance of Nuclear Power Plants, which are referenced
in the following paragraphs, were approved for incorporation by
reference by the Director of the Federal Register. A notice of any
changes made to the material incorporated by reference will be
published in the Federal Register. Copies of the ASME Boiler and
Pressure Vessel Code and the ASME Code for Operation and Maintenance of
Nuclear Power Plants may be purchased from the American Society of
Mechanical Engineers, United Engineering Center, 345 East 47th Street,
New York, NY 10017. They are also available for inspection at the NRC
Library, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland 20852-2738.
(1) As used in this section, references to Section III of the ASME
Boiler and Pressure Vessel Code refer to Section III, Division 1, and
include editions through the 1995 Edition and addenda through the 1996
Addenda, subject to the following limitations and modifications:
(i) Engineering judgement. When a licensee relies on engineering
judgment for activities or evaluations of components or systems within
the scope of 10 CFR 50.55a that are not directly addressed by the ASME
Boiler and Pressure Vessel Code, the NRC must approve the activities or
evaluations pursuant to 10 CFR 50.55a(a)(3).
(ii) Section III Materials. When applying the 1992 Edition of
Section III, licensees shall apply the 1992 Edition with the 1992
Addenda of Section II of the ASME Boiler and Pressure Vessel Code.
(iii) Weld leg dimensions. When applying the 1989 Addenda through
the 1996 Addenda of Section III, licensees shall not apply paragraph
NB-3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-
3673.2(b)-1, and shall continue to use the requirements in the 1989
Edition for this paragraph and figures.
(iv) Seismic design. Licensees may use Articles NB-3200, NB-3600,
NC-3600, and ND-3600 through the 1993 Addenda, subject to the
limitation specified in (b)(1)(iii) of this section. Licensees shall
not use the provisions in the 1994 Addenda through the 1996 Addenda for
these Articles.
(v) Quality assurance. When applying editions and addenda later
than the 1989 Edition of Section III, the requirements of NQA-1,
``Quality Assurance Requirements for Nuclear Facilities,'' 1986 Edition
through the 1992 Addenda are acceptable for use provided that both NQA-
1 and the quality assurance provisions specified in NCA-4000 are used
in conjunction with the administrative, quality, and technical
provisions contained in the edition and addenda of Section III being
utilized.
(vi) Independence of inspection. Licensees shall not apply NCA-
4134.10(a) of Section III, 1995 Edition with the 1996 Addenda, and
shall use NCA-4134.10(a), 1994 Addenda.
(2) As used in this section, references to Section XI of the ASME
Boiler and Pressure Vessel Code refer to Section XI, Division 1, and
include editions through the 1995 Edition and addenda through the 1996
Addenda, subject to the following limitations and modifications:
* * * * *
(iv) Pressure-retaining welds in ASME Code Class 2 piping (applies
to Tables IWC-2520 or IWC-2520-1, Category C-F).
(A) Appropriate Code Class 2 pipe welds in Residual Heat Removal
Systems, Emergency Core Cooling Systems, and Containment Heat Removal
Systems, must be examined. When applying editions and addenda up to the
1983 Edition through the Summer 1983 Addenda of Section XI of the ASME
Code, the extent of examination for these systems must be determined by
the requirements of paragraph IWC-1220, Table IWC-2520 Category C-F and
C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the
Summer 1975 Addenda.
(B) For a nuclear power plant whose application for a construction
permit was docketed prior to July 1, 1978, when applying editions and
addenda up to the 1983 Edition through the Summer 1983 Addenda of
Section XI of the ASME Code, the extent of examination for Code Class 2
pipe welds may be determined by the requirements of paragraph IWC-1220,
Table IWC-2520 Category C-F and C-G and paragraph IWC-2411 in the 1974
Edition and Addenda through the Summer 1975 Addenda of Section XI of
the ASME Code or other requirements the Commission may adopt.
* * * * *
[[Page 63908]]
(vi) Effective edition and addenda of Subsection IWE and Subsection
IWL, Section XI. Licensees shall use either the 1992 Edition with the
1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection
IWE and Subsection IWL as modified and supplemented by the requirements
in Sec. 50.55a(b)(2)(ix) and Sec. 50.55a(b)(2)(x).
(vii) [Reserved]
(viii) Section XI References to OM Part 4, OM Part 6 and OM Part 10
(Table IWA-1600-1). When using Table IWA-1600-1, ``Referenced Standards
and Specifications'' in the Section XI, Division 1, 1987 Addenda, 1988
Addenda, or 1989 Edition, the specified ``Revision Date or Indicator''
for ASME/ANSI OM Part 4, ASME/ANSI Part 6, and ASME/ANSI Part 10 shall
be the OMa-1988 Addenda to the OM-1987 Edition. These requirements have
been incorporated into the 1990 Edition of the OM Code which is
incorporated by reference in paragraph (b)(3) of this section.
(ix) Examination of concrete containments. Licensees applying
Subsection IWL, 1992 Edition with the 1992 Addenda, shall apply all of
the modifications in this paragraph. Licensees choosing to apply the
1995 Edition with the 1996 Addenda shall apply paragraphs
(b)(2)(ix)(A), (D)(3), and (E) of this section.
* * * * *
(xi) Engineering judgment. When a licensee relies on engineering
judgment for activities or evaluations of components or systems within
the scope of 10 CFR 50.55a that are not directly addressed by the ASME
Boiler and Pressure Vessel Code, the NRC must approve the activities or
evaluations pursuant to 10 CFR 50.55a(a)(3).
(xii) Quality Assurance. When applying Section XI editions and
addenda later than the 1989 Edition, the requirements of NQA-1,
``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda
through the 1989 Edition are acceptable as permitted by IWA-1400 of
Section XI, provided the licensee utilizes its 10 CFR Part 50, Appendix
B, quality assurance program, in conjunction with Section XI
requirements. Changes to licensee's quality assurance program shall be
made in accordance with 10 CFR 50.54(a). In addition, where NQA-1 and
Section XI do not address the commitments contained in the licensee's
Appendix B quality assurance program description, such commitments
shall be applied to Section XI activities.
(xiii) Class 1 piping. Licensees shall not apply IWB-1220,
``Components Exempt from Examination,'' of Section XI, 1989 Addenda
through the 1996 Addenda, and shall apply IWB-1220, 1989 Edition.
(xiv) Class 2 piping. Prior to applying the provisions of IWC-1220,
``Components Exempt from Examination,'' IWC-1221, ``Components Within
RHR, ECC, and CHR Systems or Portions of Systems,'' and IWC-1222,
``Components Within Systems or Portions of Systems Other Than RHR, ECC,
and CHR Systems,'' 1989 Addenda through the 1996 Addenda, licensees
shall define the Class 2 piping subject to volumetric and surface
examination, and submit this information for approval by the NRC staff
pursuant to Sec. 50.55a(a)(3) prior to implementation.
(xv) Class 1 piping volumetric examination. When performing weld
examinations of High Pressure Safety Injection Systems, as required by
Table IWB-2500-1, Examination Category B-J, Item Numbers B9.20, B9.21,
and B9.22, all licensees of pressurized water reactor facilities shall
perform volumetric examination of the Class 1 portion of the system
after [insert 6 months from the date of the final rule].
(xvi) Flaws in Class 3 piping moderate energy (200 xF, 275 psig)
piping. Licensees may use the provisions of Code Case N-513,
``Evaluation Criteria for Temporary Acceptance of Flaws in Class 3
Piping,'' Rev 0, and Code Case N-523-1, ``Mechanical Clamping Devices
for Class 2 and 3 Piping.'' Licensees choosing to apply Code Case N-
523-1 shall apply all of its provisions. Licensees choosing to apply
Code Case N-513 shall apply all of its provisions subject to the
following:
(A) When implementing Code Case N-513, the specific safety factors
in paragraph 4.0 must be satisfied.
(B) Code Case N-513 shall not be applied to:
(1) Components other than pipe and tube, such as pumps, valves,
expansion joints, and heat exchangers;
(2) The discovery and repair of flaws or deficiencies remaining
from original construction;
(3) Leakage through a flange gasket;
(4) Threaded connections employing nonstructural seal welds for
leakage prevention (through seal weld leakage is not a structural flaw,
thread integrity must be maintained); and
(5) Degraded socket welds.
(xvii) Appendix VIII personnel qualification. All personnel
qualified for performing ultrasonic examinations in accordance with
Appendix VIII shall receive 40 hours of annual training that includes
laboratory work and examination of flawed specimens.
(xviii) Appendix VIII specimen set cracks. All flaws in the
specimen sets used for performance demonstration for piping, vessels,
and nozzles shall be cracks.
(xix) Appendix VIII specimen set microstructure. All specimens for
single-side tests shall contain microstructures of the type found in
components to be inspected, and flaws with non-optimum characteristics
consistent with field experience that provide realistic challenges to
the UT techniques.
(xx) Reconciliation of Quality Requirements. The following
limitations apply when implementing Section XI, IWA-4200, 1995 Addenda
through the 1996 Addenda:
(A) Licensees shall not apply IWA-4200, of Section XI, 1995 Addenda
through the 1996 Addenda, for reconciliation of the administrative
requirements for replacement items, and shall reconcile the
administrative requirements with the original Construction Code and the
Owner's requirements as required by the 1995 Edition.
(B) Licensees shall not apply the definition of Construction Code
in IWA-9000, ``Glossary,'' 1993 Addenda through the 1996 Addenda, and
shall apply the definition of Construction Code in IWA-9000, 1992
Edition.
(3) As used in this section, references to the OM Code refer to the
ASME Code for Operation and Maintenance of Nuclear Power Plants, and
include addenda through the 1996 Addenda and editions through the 1995
Edition subject to the following limitations and modifications:
(i) Quality Assurance. When applying editions and addenda of the OM
Code, 1990 and later, the requirements of NQA-1, ``Quality Assurance
Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as
permitted by ISTA 1.4 of the OM Code, provided the licensee utilizes
its 10 CFR Part 50, Appendix B, quality assurance program, in
conjunction with the OM Code requirements. Changes to licensee's
quality assurance program shall be made in accordance with 10 CFR
50.54(a). In addition, where NQA-1 and the OM Code do not address the
commitments contained in the licensee's Appendix B quality assurance
program description, such commitments shall be applied to OM Code
activities.
(ii) Stroke time testing. Licensees shall comply with the
provisions on stroke time testing in OM Code ISTC 4.2, 1995 Edition
with the 1996 Addenda, and the programs developed under their licensing
commitments for demonstrating design basis capability of motor-operated
valves.
[[Page 63909]]
(iii) Code Case OMN-1. As an alternative to Sec. 50.55a(b)(3)(ii),
licensees may use Code Case OMN-1, ``Alternative Rules for Preservice
and Inservice Testing of Certain Electric Operated Valve Assemblies in
LWR Power Plants,'' Rev. 0, 1995 Edition with the 1996 Addenda, in
conjunction with ISTC 4.3, 1995 Edition with the 1996 Addenda.
Licensees choosing to apply the Code case shall apply all of its
provisions.
(A) The adequacy of the test interval for each valve shall be
evaluated and adjusted as necessary but not later than five years or
three refueling outages (whichever is longer) from initial
implementation of ASME Code Case OMN-1.
(B) [Reserved]
(iv) Appendix II. The following modifications apply when
implementing Appendix II, ``Check Valve Condition Monitoring Program,''
of the OM Code, 1995 Edition with the 1996 Addenda:
(A) Valve opening and closing functions must be demonstrated when
flow testing or examination methods (nonintrusive, or disassembly and
inspection) are used;
(B) The initial interval for tests and associated examinations
shall not exceed two fuel cycles or 3 years, whichever is longer; any
extension of this interval shall not exceed one fuel cycle per
extension with the maximum interval not to exceed 10 years; trending
and evaluation of existing data shall be used to reduce or extend time
the interval between tests.
(C) If the Appendix II condition monitoring program is
discontinued, then the requirements of ISTC 4.5.1 through 4.5.4 shall
be implemented.
(v) Subsection ISTD. Licensees may use Subsection ISTD, OM Code,
1995 Edition with the 1996 Addenda, by making a change to their
technical specifications in accordance with applicable NRC
requirements. Licensees choosing to apply the subsection shall apply
all of its provisions.
(c) * * *
(3) The Code Edition, Addenda, and optional Code Cases to be
applied to components of the reactor coolant pressure boundary must be
determined by the provisions of paragraph NCA-1140, Subsection NCA of
Section III of the ASME Boiler and Pressure Vessel Code, but:
(i) The edition and addenda applied to a component must be those
which are incorporated by reference in paragraph (b)(1) of this
section, and, in case of conflict between paragraph (b)(1) of this
section and paragraph NCA-1140, the latest edition and addenda
incorporated by reference in paragraph (b)(1) of this section shall be
applied,
(ii) The ASME Code provisions applied to the pressure vessel may be
dated no earlier than the Summer 1972 Addenda of the 1971 edition,
(iii) The ASME Code provisions applied to piping, pumps, and valves
may be dated no earlier than the Winter 1972 Addenda of the 1971
edition, and
* * * * *
(d) * * *
(2) The Code Edition, Addenda, and optional Code Cases6 to be
applied to the systems and components identified in paragraph (d)(1) of
this section must be determined by the rules of paragraph NCA-1140,
Subsection NCA of Section III of the ASME Boiler Vessel and Pressure
Code, but:
(i) The edition and addenda must be those which are incorporated by
reference in paragraph (b)(1) of this section, and, in case of conflict
between paragraph (b)(1) of this section and paragraph NCA-1140, the
latest edition and addenda incorporated by reference in paragraph
(b)(1) of this section shall be applied,
(ii) The ASME Code provisions applied to the systems and components
may be dated no earlier than the 1980 Edition, and
(iii) The ASME Code Cases6 must have been determined suitable for
use by the NRC.
(e) * * *
(2) The Code Edition, Addenda, and optional Code Cases6 to be
applied to the systems and components identified in paragraph (e)(1) of
this section must be determined by the rules of paragraph NCA-1140,
Subsection NCA of Section III of the ASME Boiler and Pressure Vessel
Code, but:
(i) The edition and addenda must be those which are incorporated by
reference in paragraph (b)(1) of this section, and, in case of conflict
between paragraph (b)(1) of this section and paragraph NCA-1140, the
latest edition and addenda incorporated by reference in paragraph
(b)(1) of this section shall be applied,
(ii) The ASME Code provisions applied to the systems and components
may be dated no earlier than the 1980 Edition, and
(iii) The ASME Code Cases must have been determined suitable for
use by the NRC.
(f) Inservice testing requirements. Requirements for inservice
inspection of Class 1, Class 2, Class 3, Class MC, and Class CC
components (including their supports) are located in Sec. 50.55a(g).
(1) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit was issued prior to January 1, 1971,
pumps and valves must meet the test requirements of paragraphs (f)(4)
and (f)(5) of this section to the extent practical. Pumps and valves
which are part of the reactor coolant pressure boundary must meet the
requirements applicable to components which are classified as ASME Code
Class 1. Other pumps and valves in steam, water, air, and liquid-
radioactive-waste systems that perform a function to shut down the
reactor or maintain the reactor in a safe shutdown condition, mitigate
the consequences of an accident, or provide overpressure protection for
such systems (in meeting the requirements of the 1986 Edition, or
later, of the Boiler and Pressure Vessel or OM Code), must meet the
test requirements applicable to components which are classified as ASME
Code Class 2 or Class 3.
(2) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit was issued on or after January 1,
1974, pumps and valves which are classified as ASME Code Class 1 and
Class 2 must be designed and be provided with access to enable the
performance of inservice tests for operational readiness set forth in
editions of Section XI of the ASME Boiler and Pressure Vessel Code and
Addenda6 in effect 6 months prior to the date of issuance of the
construction permit. The pumps and valves may meet the requirements set
forth in subsequent editions of this code and addenda which are
incorporated by reference in paragraph (b) of this section, subject to
limitations and modifications listed therein.
(3) * * *
(iii)(A) Pumps and valves, in facilities whose construction permit
was issued before [insert effective date of the final rule], which are
classified as ASME Code Class 1 must be designed and be provided with
access to enable the performance of inservice testing of the pumps and
valves for assessing operational readiness set forth in Section XI of
editions of the ASME Boiler and Pressure Vessel Code and Addenda6
applied to the construction of the particular pump or valve or the
Summer 1973 Addenda, whichever is later.
(B) Pumps and valves, in facilities whose construction permit is
issued on or after [insert effective date of the final rule], which are
classified as ASME Code Class 1 must be designed and be provided with
access to enable the performance of inservice testing of the pumps and
valves for assessing
[[Page 63910]]
operational readiness set forth in editions and addenda of the ASME OM
Code referenced in paragraph (b)(3) of this section at the time the
construction permit is issued.
(iv)(A) Pumps and valves, in facilities whose construction permit
was issued before [insert effective date of rule], which are classified
as ASME Code Class 2 and Class 3 must be designed and be provided with
access to enable the performance of inservice testing of the pumps and
valves for assessing operational readiness set forth in Section XI of
editions of the ASME Boiler and Pressure Vessel Code and Addenda6
applied to the construction of the particular pump or valve or the
Summer 1973 Addenda, whichever is later.
(B) Pumps and valves, in facilities whose construction permit is
issued on or after [insert effective date of the final rule], which are
classified as ASME Code Class 2 and 3 must be designed and be provided
with access to enable the performance of inservice testing of the pumps
and valves for assessing operational readiness set forth in editions
and addenda of the ASME OM Code referenced in paragraph (b)(3) of this
section at the time the construction permit is issued.
* * * * *
(4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, pumps and valves which are classified as
ASME Code Class 1, Class 2 and Class 3 must meet the inservice test
requirements, except design and access provisions, set forth in the
ASME OM Code and addenda that become effective subsequent to editions
and addenda specified in paragraphs (f)(2) and (f)(3) of this section
and that are incorporated by reference in paragraph (b) of this
section, to the extent practical within the limitations of design,
geometry and materials of construction of the components.
* * * * *
(g) * * *
(1) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit was issued before January 1, 1971,
components (including supports) must meet the requirements of
paragraphs (g)(4) and (g)(5) of this section to the extent practical.
Components which are part of the reactor coolant pressure boundary and
their supports must meet the requirements applicable to components
which are classified as ASME Code Class 1. Other pressure vessels,
piping, pumps and valves, and their supports in steam, water, air, and
liquid-radioactive-waste systems that provide pressure boundary
integrity for systems that perform a function to shut down the reactor
or maintain the reactor in a safe shutdown condition, or mitigate the
consequences of an accident, must meet the requirements applicable to
components which are classified as ASME Code Class 2 or Class 3.
* * * * *
(3) * * *
(i) Components (including supports) which are classified as ASME
Code Class 1 must be designed and be provided with access to enable the
performance of inservice examination of such components and must meet
the preservice examination requirements set forth in Section XI of
editions of the ASME Boiler and Pressure Vessel Code and Addenda6
applied to the construction of the particular component.
* * * * *
(4) Throughout the service life of a boiling or pressurized water-
cooled nuclear power facility, components (including supports) which
are classified as ASME Code Class 1, Class 2 and Class 3 must meet the
requirements, except design and access provisions and preservice
examination requirements, set forth in Section Xl of editions of the
ASME Boiler and Pressure Vessel Code and Addenda that become effective
subsequent to editions specified in paragraphs (g)(2) and (g)(3) of
this section and that are incorporated by reference in paragraph (b) of
this section, to the extent practical within the limitations of design,
geometry and materials of construction of the components. Components
which are classified as Class MC pressure retaining components and
their integral attachments, and components which are classified as
Class CC pressure retaining components and their integral attachments
must meet the requirements, except design and access provisions and
preservice examination requirements, set forth in Section XI of the
ASME Boiler and Pressure Vessel Code and Addenda that are incorporated
by reference in paragraph (b) of this section, subject to the
limitation listed in paragraph (b)(2)(vi) and the modifications listed
paragraph (b)(2)(ix) and (b)(2)(x) of this section, to the extent
practical within the limitation of design, geometry and materials of
construction of the components.
* * * * *
(iv) [Reserved]
(6) * * *
(ii) * * *
(A)(1) All previously granted reliefs under Sec. 50.55a to
licensees for the extent of volumetric examination of reactor vessel
shell welds specified in Item BI.10 of Examination Category B-A,
``Pressure Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1 of
Subsection IWB in applicable edition and addenda of Section XI,
Division 1, of the ASME Boiler and Pressure Vessel Code, during the
inservice inspection interval in effect on September 8, 1992 are hereby
revoked, subject to the specific modification in
Sec. 50.55a(g)(6)(ii)(A)(3)(iv) for licensees that defer the augmented
examination in accordance with Sec. 50.55a(g)(6)(ii)(A)(3).
(2) All licensees shall augment their reactor vessel examination by
implementing once, as part of the inservice inspection interval in
effect on September 8, 1992, the examination requirements for reactor
vessel shell welds specified in Item 81.10 of Examination Category B-A,
``Pressure Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1 of
Subsection IWB of the 1989 Edition of Section XI, Division 1, of the
ASME Boiler and Pressure Vessel Code, subject to the conditions
specified in Sec. 50.55a(g)(6)(ii)(A)(3) and (4). The augmented
examination, when not deferred in accordance with the provisions of
Sec. 50.55a(g)(6)(ii)(A)(3), shall be performed in accordance with the
related procedures specified in the Section XI edition and addenda
applicable to the inservice inspection interval in effect on September
8, 1992, and may be used as a substitute for the reactor vessel shell
weld examination scheduled for implementation during the inservice
inspection interval in effect on September 8, 1992. For the purpose of
this augmented examination, ``essentially 100%'' as used in Table IWB-
2500-1 means more than 90 percent of the examination volume of each
weld, where the reduction in coverage is due to interference by another
component, or part geometry.
* * * * *
(6) Augmented examinations of reactor vessel shell welds that are
performed in accordance with Sec. 50. 55a(g)(6)(ii)(A) after [insert 6
months from the date of the final rule] must be performed in accordance
with Sec. 50.55a(g)(6)(ii)(C).
* * * * *
(C) Application of Appendix VIII to Section Xl Examinations.
(1) All reactor vessel (including nozzles) ultrasonic examinations,
all piping ultrasonic examinations, and all bolting ultrasonic
examinations performed after insert 6 months from the date of the final
rule must be
[[Page 63911]]
performed in accordance with Appendix VIII of Section Xl, Division 1,
1995, Edition with the 1996 Addenda of the ASME Boiler and Pressure
Vessel Code.
(2) [Reserved]
* * * * *
\5\ For ASME Code Editions and Addenda issued prior to the
Winter 1977 Addenda, the Code Edition and Addenda applicable to the
component is governed by the order or contract date for the
component, not the contract date for the nuclear energy system. For
the Winter 1977 addenda and subsequent editions and addenda the
method for determining the applicable Code editions and addenda is
contained in Paragraph NCA-1140 of Section III of the ASME Code.
* * * * *
\7\ For purposes of this regulation the proposed IEEE-279 became
``in effect'' on August 30, 1968, and the revised issue IEEE-279-
1971 became ``in effect'' on June 3, 1971. Copies may be obtained
from the Institute of Electrical and Electronics Engineers, United
Engineering Center, 345 East 47th St., New York, NY 10017. Copies
are available for inspection at the NRC Library, Two White Flint
North, 11545, Rockville Pike, Rockville, Maryland 20852-2738.
* * * * *
Dated at Rockville, MD this 27th day of October 1997.
For the Nuclear Regulatory Commission.
L. Joseph Callan,
Executive Director for Operations.
[FR Doc. 97-31588 Filed 12-2-97; 8:45 am]
BILLING CODE 7590-01-P
