[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]


      
 ========================================================================
 Proposed Rules
                                                 Federal Register
 ________________________________________________________________________
 
 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.
 
 ========================================================================
 

  Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / 
Proposed Rules  

[[Page 63892]]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AE26


Industry Codes and Standards; Amended Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

-----------------------------------------------------------------------

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