Industry Codes and Standards; Amended Requirements

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN 3150-AE26

Industry Codes and Standards; Amended Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) regulations require

that nuclear power plant owners construct Class 1, Class 2, and Class 3

components in accordance with the rules provided in Section III,

Division 1, ``Requirements for Construction of Nuclear Power Plant

Components,'' of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (BPV Code), inspect Class 1, Class 2,

Class 3, Class MC (metal containment) and Class CC (concrete

containment) components in accordance with the rules provided in

Section XI, Division 1, ``Requirements for Inservice Inspection of

Nuclear Power Plant Components,'' of the ASME BPV Code, and test Class

1, Class 2, and Class 3 pumps and valves in accordance with the rules

provided in Section XI, Division 1, of the ASME BPV Code.

The NRC proposes to amend 10 CFR 50.55a to revise the requirements

for construction, inservice inspection (ISI), and inservice testing

(IST) of nuclear power plant components. For construction, the proposed

rule would permit the use of Section III, Division 1, of the ASME BPV

Code, 1989 Addenda through the 1996 Addenda, for Class 1, Class 2, and

Class 3 components with six proposed limitations and a modification.

For ISI, the proposed amendment would require licensees to

implement Section XI, Division 1, of the ASME BPV Code, 1995 Edition

with the 1996 Addenda, for Class 1, Class 2, and Class 3 components

with five proposed limitations. Licensees would be permitted to

implement: Code Case N-513 which addresses flaws in low and moderate

energy Class 3 piping; Code Case N-523 which addresses the temporary

use of mechanical clamps in Class 2 and 3 piping; and Subsection IWE

and Subsection IWL, 1995 Edition with the 1996 Addenda.

The proposed rule would expedite implementation of Appendix VIII,

``Performance Demonstration for Ultrasonic Examination Systems,'' to

Section XI, Division 1, with three proposed modifications. An expedited

implementation schedule would also be required for a proposed

modification to Section XI which addresses volumetric examination of

the Class 1 high pressure safety injection (HPSI) system in pressurized

water reactors (PWRs).

For IST, the proposed amendment would require licensees to

implement the 1995 Edition with the 1996 Addenda of the ASME Code for

Operation and Maintenance of Nuclear Power Plants (OM Code) for Class

1, Class 2, and Class 3 pumps and valves with one limitation and one

modification. 10 CFR 50.55a has been clarified with respect to which

pumps and valves are to be included in a licensee's IST program.

Licensees would be permitted to implement: Code Case OMN-1 with one

modification in lieu of stroke time testing; Appendix II (which is an

alternative to the check valve condition monitoring program provisions

contained in Subsection ISTC of the OM Code) with three proposed

modifications; and Subsection ISTD for the IST of snubbers. Finally,

based upon supporting information received since the last rulemaking,

the modification presently in Sec. 50.55a for containment isolation

valve inservice testing has been deleted.

The Statement of Considerations concludes by clarifying the NRC

position regarding ASME Code Interpretations, and discussing NRC

Direction Setting Issue Number 13 (DSI-13) with regard to NRC

endorsement of industry codes and standards.

DATES: Submit comments by March 3, 1998. Comments received after this

date will be considered if it is practical to do so, but the Commission

is able to ensure consideration only for comments received on or before

this date.

ADDRESSES: Comments may be sent to: Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001. ATTN: Rulemaking and

Adjudications Staff. Hand deliver comments to 11545 Rockville Pike,

Rockville, Maryland, 20852, between 7:30 am and 4:15 pm on Federal

workdays.

You may also provide comments via the NRC's interactive rulemaking

website through the NRC home page (http://www.nrc.gov). This site

provides the availability to upload comments as files (any format), if

your web browser supports that function. For information about the

interactive website, contact Ms. Carol Gallagher, (301) 415-5905; e-

mail [email protected]

Single copies of this proposed rulemaking may be obtained by

written request or telefax to 301-415-2260 or from Frank C. Cherny,

Division of Engineering Technology, Office of Nuclear Regulatory

Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-

0001, Telephone: 301-415-6786, or Wallace E. Norris, Division of

Engineering Technology, U.S. Nuclear Regulatory Commission, Washington,

DC 20555-0001, Telephone: 301-415-6796. Certain documents related to

this rulemaking, including comments received, may be examined at the

NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington,

DC. These same documents may also be viewed and downloaded via the

interactive rulemaking website as established by NRC for this

rulemaking.

FOR FURTHER INFORMATION CONTACT: Frank C. Cherny, Division of

Engineering Technology, Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:

301-415-6786, or Wallace E. Norris, Division of Engineering Technology,

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

Telephone: 301-415-6796.

SUPPLEMENTARY INFORMATION:

1. Background

2. Summary of Proposed Revisions to Sec. 50.55a

2.1 List of Each Revision and Implementation Schedule

2.2 Disscussion

2.3 120-Month Update

2.3.1 Section XI

2.3.1.1 Class 1, 2, and 3 Components, Including Supports

2.3.1.2 Limitations:

2.3.1.2.1 Engineering Judgment

2.3.1.2.2 Quality Assurance

2.3.1.2.3 Class 1 Piping

[[Page 63893]]

2.3.1.2.4 Class 2 Piping

2.3.1.2.5 Reconciliation of Quality Requirements

2.3.2 OM Code

2.3.2.1 Class 1, 2, and 3 Pumps and Valves

2.3.2.2 Background--OM Code

2.3.2.3 Clarification of Safety-Related Valves

2.3.2.4 Limitation:

2.3.2.4.1 Quality Assurance

2.3.2.5 Modification:

2.3.2.5.1 Stroke Time Testing

2.4 Expedited Implementation

2.4.1 Appendix VIII

2.4.1.1 Modifications:

2.4.1.1.1 Appendix VIII Personnel Qualification

2.4.1.1.2 Appendix VIII Specimen Set Cracks

2.4.1.1.3 Appendix VIII Specimen Set Microstructure

2.4.2 Generic Letter on Appendix VIII

2.4.3 Class 1 Piping Volumetric Examination

2.5 Voluntary Implementation

2.5.1 Section III

2.5.1.1 Limitations:

2.5.1.1.1 Engineering Judgement

2.5.1.1.2 Section III Materials

2.5.1.1.3 Weld Leg Dimensions

2.5.1.1.4 Seismic Design

2.5.1.1.5 Quality Assurance

2.5.1.1.6 Independence of Inspection

2.5.1.2 Modification:

2.5.1.2.1 Applicable Code Version for New Construction

2.5.2 Section XI

2.5.2.1 Subsection IWE and Subsection IWL

2.5.2.2 Flaws in Class 3 Piping; Mechanical Clamping Devices

2.5.3 OM Code

2.5.3.1 Code Case OMN-1

2.5.3.2 Appendix II

2.5.3.3 Subsection ISTD

2.5.3.4 Containment Isolation Valves

2.6 ASME Code Interpretations

2.7 DSI-13

2.8 Steam Generators

3. Finding of No Significant Environmental Impact

4. Paperwork Reduction Act Statement

5. Regulatory Analysis

6. Regulatory Flexibility Certification

7. Backfit Analysis

1. Background

The NRC is proposing to amend 10 CFR 50.55a, which defines the

requirements for applying industry codes and standards to nuclear power

plants. Section 50.55a presently requires that nuclear power plant

owners (1) construct Class 1, Class 2, and Class 3 components in

accordance with the rules provided in the 1989 Edition of Section III,

Division 1, ``Requirements for Construction of Nuclear Power Plant

Components'' of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (BPV Code), (2) inspect Class 1, Class

2, and Class 3 components in accordance with the rules provided in the

1989 Edition of Section XI, Division 1, ``Requirements for Inservice

Inspection of Nuclear Power Plant Components,'' of the ASME BPV Code

with certain limitations and modifications, (3) inspect Class MC (metal

containment) and Class CC (concrete containment) components in

accordance with the rules provided in the 1992 Edition with the 1992

Addenda of Section XI, Division 1, with certain modifications, and (4)

test Class 1, Class 2, and Class 3 pumps and valves in accordance with

the rules provided in the 1989 Edition of Section XI, Division 1, of

the ASME BPV Code with certain limitations and modifications. Every 120

months licensees are required to update their ISI and IST programs to

meet the version of Section XI incorporated by reference into

Sec. 50.55a and in effect 12 months prior to the start of a new 120-

month interval.

The NRC proposes to amend 10 CFR 50.55a to revise the requirements

for construction, ISI, and IST of nuclear power plant components. For

construction, the proposed rule would permit the use of Section III,

Division 1, of the ASME BPV Code, 1989 Addenda through the 1996

Addenda, for Class 1, Class 2, and Class 3 components. Six proposed

limitations to the implementation of Section III are included which

address the issues of engineering judgement, Section III materials,

weld leg dimensions, seismic design, quality assurance, and

independence of inspection. A modification has been included addressing

the applicable Code version for new construction.

For ISI, the proposed amendment would require licensees to

implement Section XI, Division 1, of the ASME BPV Code, 1995 Edition

with the 1996 Addenda, for Class 1, Class 2, and Class 3. Five proposed

limitations to the implementation of Section XI are included which

address the issues of engineering judgement, quality assurance, Class 1

piping, Class 2 piping, and reconciliation of replacement items.

Licensees would be permitted to implement Code Case N-513 which

addresses flaws in low and moderate energy Class 3 piping, and Code

Case N-523 which addresses the temporary use of mechanical clamps in

Class 2 and 3 piping. Licensees would also be permitted to implement

Subsection IWE and Subsection IWL, 1995 Edition with the 1996 Addenda.

The proposed rule would expedite implementation of Appendix VIII,

``Performance Demonstration for Ultrasonic Examination Systems,'' to

Section XI, Division 1. Three proposed modifications to the

implementation of Appendix VIII are included to address the issues of

personnel qualification, specimen set cracks, and specimen set

microstructure. An expedited implementation schedule would also be

required for a proposed modification to Section XI which addresses

volumetric examination of the Class 1 high pressure safety injection

(HPSI) system in pressurized water reactors (PWRs).

For IST, the proposed amendment would require licensees to

implement the 1995 Edition with the 1996 Addenda of the ASME Code for

Operation and Maintenance of Nuclear Power Plants (OM Code) for Class

1, Class 2, and Class 3 pumps and valves. 10 CFR 50.55a has been

clarified with respect to which pumps and valves are to be included in

a licensee's IST program. A proposed limitation is included which

addresses the issue of quality assurance (QA). A proposed modification

to the implementation of the OM Code is included which addresses stroke

time testing. Licensees would be permitted to implement Code Case OMN-1

with one modification in lieu of stroke time testing. In addition,

Appendix II to the OM Code is an alternative to the check valve

condition monitoring program provisions contained in Subsection ISTC of

the OM Code. Three proposed modifications to the implementation of

Appendix II are included which supplement the appendix check valve

condition monitoring program. Licensees would be permitted to use

Subsection ISTD for the IST of snubbers. Finally, based upon supporting

information received since the last rulemaking, the modification

presently in Sec. 50.55a for containment isolation valve inservice

testing has been deleted.

The mechanism for endorsement of the ASME standards, which has been

used since the first endorsement in 1971, has been to incorporate by

reference the ASME BPV Code rules into Sec. 50.55a. The regulation

identifies which editions and addenda of the BPV Code have been

approved for use by the NRC. On August 6, 1992 (57 FR 34666), the NRC

published a final rule in the Federal Register to amend 10 CFR Part 50,

``Domestic Licensing of Production and Utilization Facilities.'' This

final rule amended Sec. 50.55a to incorporate by reference the 1986

Addenda, 1987 Addenda, 1988 Addenda, and 1989 Edition of Section III,

Division 1, and the 1986 Addenda, 1987 Addenda, 1988 Addenda, and 1989

Edition of Section XI, Division 1, of the BPV Code, with specified

modifications. The amendment imposed an augmented examination of

reactor vessel shell welds. The amendment also separated the

requirements for IST of pumps and valves from those for ISI of other

components by placing the requirements for inservice testing in a

[[Page 63894]]

separate paragraph. For IST of pumps and valves, the regulation,

through its incorporation by reference of the 1989 Edition of Section

XI, endorsed Part 1, ``Requirements for Inservice Performance Testing

of Nuclear Power Plant Pressure Relief Devices,'' Part 6, ``Inservice

Testing of Pumps in Light-Water Reactor Power Plants,'' and Part 10,

``Inservice Testing of Valves in Light-Water Reactor Power Plants,'' of

ASME/ANSI OMa-1988 to ASME/ANSI OM-1987.

On August 8, 1996 (61 FR 41303), the NRC published a final rule in

the Federal Register to amend 10 CFR 50.55a to incorporate by reference

for the first time ASME Section XI, Division 1, Subsection IWE,

``Requirements for Class MC and Metallic Liners of Class CC Components

of Light-Water Cooled Power Plants,'' and Subsection IWL,

``Requirements for Class CC Concrete Components of Light-Water Cooled

Power Plants.'' Subsection IWE provides criteria for visual inspection

of the surface of metal containments, the steel liners of concrete

containments, pressure-retaining bolts, and seals and gaskets.

Subsection IWL provides criteria for visual inspection of concrete

pressure-retaining shells and shell components and for the examination

of unbonded post-tensioning systems.

2. Summary of Proposed Revisions to Sec. 50.55a

The revisions to Sec. 50.55a which would result from adoption of

the 1989 Addenda through the 1996 Addenda have been divided into three

groups based on the proposed implementation schedule (i.e., 120-month

update, expedited, and voluntary). For each of these groups, it is

indicated in parentheses whether or not particular items are considered

a backfit under 10 CFR 50.109 as discussed in Section 8. Backfit

Analysis. This section provides a list of each revision and its

implementation schedule, followed by a discussion of the proposed

revisions.

2.1 List of Each Revision and Implementation Schedule

120-Month Update [in accordance with Sec. 50.55a(g)(4)(i) and

Sec. 50.55a(f)(4)(i)]

Section XI (Not A Backfit)

Class 1, 2, and 3 Components, Including Supports

Limitations

Engineering Judgement

Quality Assurance

Class 1 Piping

Class 2 Piping

Reconciliation of Quality Requirements

OM Code (Not A Backfit)

Class 1, 2, and 3 Pumps and Valves

Clarification of Safety-Related Valves

Limitation

Quality Assurance

Modification

Stroke Time Testing

Expedited Implementation [after 6 months from the date of the final

rule--Backfit]

Section XI

Appendix VIII (including three modifications)

Personnel Qualification

Specimen Set Cracks

Specimen Set Microstructure

Class 1 Piping Volumetric Examination

Voluntary Implementation [may be used when final rule published]

Section III (Not A Backfit)

Class 1, 2, and 3 Components

Limitations

Engineering Judgement

Section III Materials

Weld Leg Dimensions

Seismic Design

Quality Assurance

Independence of Inspection

Modification

Applicable Code Version for New Construction

Section XI (Not A Backfit)

Subsections IWE and IWL, 1995 Edition with the 1996 Addenda

Flaws in Class 3 Piping; Mechanical Clamping Devices

Limitation on Scope

OM Code (Not A Backfit)

Code Case OMN-1

Limitation on Length of Test Interval

Appendix II (including three modifications)

Valve Opening and Closing Functions

Limitation of Length of Initial Test Interval

Condition Monitoring Program

Subsection ISTD

Containment Isolation Valves

2.2 Discussion

2.3 120-Month Update

2.3.1 Section XI

2.3.1.1 Class 1, 2, and 3 Components, Including Supports

Section 50.55a(b)(2) together with Sec. 50.55a(g)(4) of the

proposed rule would require that licensees implement the 1995 Edition

with the 1996 Addenda of Section XI, Division 1, for Class 1, Class 2,

and Class 3 components and their supports. Five proposed limitations

would be included to address NRC positions on the use of Section XI.

2.3.1.2 Limitations

2.3.1.2.1 Engineering Judgement

The first proposed limitation to the implementation of Section XI

would address an NRC position with regard to the Foreword in the 1992

Addenda through the 1996 Addenda of the BPV Code. That Foreword

addresses the use of ``engineering judgement'' for ISI activities not

specifically considered by the Code. Proposed paragraph

50.55a(b)(2)(xi) would require that when a licensee relies on

engineering judgement for activities or evaluations of components or

systems within the scope of Sec. 50.55a that are not directly addressed

by the BPV Code, the licensee must receive NRC approval for those

activities or evaluations pursuant to 10 CFR 50.55a(a)(3).

2.3.1.2.2 Quality Assurance

The second proposed limitation to the implementation of Section XI

pertains to the use of NQA-1 with Section XI. Section XI references the

use of either NQA-1 or the Owner's Appendix B Quality Assurance Program

(10 CFR Part 50, Appendix B, ``Quality Assurance Criteria for Nuclear

Power Plants and Fuel Processing Plants'') as part of its individual

requirements for a QA program. At present, Sec. 50.55a endorses the

1989 Edition of the ASME Code which references NQA-1-1979 for Section

XI. The 1996 Addenda of the ASME Code references NQA-1-1992 for Section

XI.

The NRC has reviewed the requirements of NQA-1, 1986 Addenda

through the 1992 Addenda, that are part of the incorporation by

reference of Section XI, and has determined that by itself, NQA-1 would

not adequately describe how to satisfy the requirements of 10 CFR Part

50, Appendix B, ``Quality Assurance Criteria for Nuclear Power Plants

and Fuel Reprocessing Plants,'' since there are various aspects of

operational phase QA and administrative controls which are not

addressed by NQA-1.

10 CFR 50.34(b)(6)(ii) requires that ``The information on the

controls to be used for a nuclear power plant or a fuel reprocessing

plant shall include a discussion of how the applicable requirements of

Appendix B will be satisfied.'' This information is required to be

submitted to the NRC as part of the Final Safety Analysis Report

(FSAR). Standard Review Plan (SRP) 17.2, ``Quality Assurance During the

Operations Phase,'' states that ``The QA program description presented

in the FSAR must discuss how each criterion of Appendix B will be

met.'' Further, the SRP states ``The acceptance criteria include a

commitment to comply with the regulatory positions presented in the

appropriate issue of the Regulatory Guides including the requirements

of ANSI Standard N45.2.12 and the Branch

[[Page 63895]]

Technical Position listed in subsection V of SRP Section 17.1. Thus,

the commitment constitutes an integral part of the QA program

description and requirements.'' The NRC has determined that the

provisions of NQA-1, 1986 Addenda through the 1992 Addenda, would not

satisfy the criteria specified in SRP 17.2 for describing how the

requirements of Appendix B will be satisfied for operational

activities. There are numerous areas where American National Standards

Institute (ANSI) standards or NRC regulatory positions, which have been

long-standing cornerstones of an Owner's QA Program, are either

nonmandatory or missing altogether from the NQA-1 provisions. However,

the Owner's Section XI QA Program, which has been approved by the NRC,

is adequate. Thus, the Commission has determined that the requirements

of NQA-1, 1986 Addenda through the 1992 Addenda, are acceptable for use

in the context of Section XI, as permitted by IWA-1400, provided the

licensee utilizes its 10 CFR Part 50, Appendix B, QA program in

conjunction with Section XI. Changes to a licensee's QA program shall

be made in accordance with 10 CFR 50.54(a). Further, where NQA-1 and

Section XI do not address the commitments contained in the licensee's

Appendix B QA program description, such commitments shall be applied to

Section XI activities. Proposed Sec. 50.55a(b)(2)(xii) contains the

requirement addressing licensee's commitments related to Section XI.

2.3.1.2.3 Class 1 Piping

The third proposed limitation to the implementation of Section XI

would require licensees to use the rules for Section XI IWB-1220,

``Components Exempt from Examination,'' that are contained in the 1989

Edition in lieu of the rules in the 1989 Addenda through the 1996

Addenda. These later Code addenda contain provisions of Code Cases N-

198-1, ``Exemption from Examination for ASME Class 1 and Class 2 Piping

Located at Containment Penetrations;'' N-322, ``Examination

Requirements for Integrally Welded or Forged Attachments to Class 1

Piping at Containment Penetrations;'' and N-324, ``Examination

Requirements for Integrally Welded or Forged Attachments to Class 2

Piping at Containment Penetrations;'' which were found to be

unacceptable. Because the NRC had previously determined the Code cases

to be unacceptable, they were not endorsed in any revision of

Regulatory Guide 1.147, ``Inservice Inspection Code Case

Acceptability--ASME Section XI, Division 1.'' The provisions of Code

Case N-198-1 were determined by the NRC to be unacceptable because

industry experience has shown that welds in service-sensitive BWR

stainless steel piping, many of which are located in Containment

Penetrations, are subjected to an aggressive environment (BWR water at

reactor operating temperatures) and will experience Intergranular

Stress Corrosion Cracking. Exempting these welds from examination could

result in conditions which reduce the required margins to failure to

unacceptable levels. The provisions of Code Cases N-322 and N-324 were

determined to be unacceptable because some important piping was

exempted from inspection. Access difficulties was the basis in the Code

cases for exempting these areas from examination, but the NRC developed

the break exclusion zone design and examination criteria utilized for

most containment penetration piping expecting not only that Section XI

inspections would be performed but that augmented inspections would be

performed. These design and examination criteria are contained in

Branch Technical Position MEB 3-1, an attachment of NRC Standard Review

Plan 3.6.2, ``Determination of Rupture Locations and Dynamic Effects

Associated with the Postulated Rupture of Piping.'' Thus, proposed

Sec. 50.55a(b)(2)(xiii) would require licensees to use the rules for

IWB-1220 that are contained in the 1989 Edition in lieu of the rules in

the 1989 Addenda through the 1996 Addenda.

2.3.1.2.4 Class 2 Piping

The fourth proposed limitation to the implementation of Section XI,

contained in Sec. 50.55a(b)(2)(xiv), would confine implementation of

Section XI IWC-1220, ``Components Exempt from Examination,'' IWC-1221,

``Components Within RHR (Residual Heat Removal), ECC (Emergency Cool

Cooling), and CHR (Containment Heat Removal) Systems or Portions of

Systems,'' and IWC-1222, ``Components Within Systems or Portions of

Systems Other Than RHR, ECC, and CHR Systems,'' 1989 Addenda through

the 1996 Addenda. The provisions of Code Case N-408-3, ``Alternative

Rules for Examination of Class 2 Piping,'' were incorporated into

Subsection IWC in the 1989 Addenda. These provisions contain rules for

determining which Class 2 components are subject to volumetric and

surface examination. The NRC had previously determined that the

provisions of the Code Case were acceptable if the licensee defined the

Class 2 piping subject to volumetric and surface examination and

received approval prior to implementation. Approval was required to

ensure that safety significant components in the Residual Heat Removal,

Emergency Core Cooling, and Containment Heat Removal systems are not

exempted from appropriate examination requirements. Thus, the

requirements contained in IWC-1220, IWC-1221, and IWC-1222, 1989

Addenda through the 1996 Addenda, for determining the components

subject to examination and establishing examination requirements for

Class 2 piping may be used if the licensee defines the Class 2 piping

subject to volumetric and surface examination, and submits this

information to the NRC for approval pursuant to Sec. 50.55a(a)(3).

2.3.1.2.5 Reconciliation of Quality Requirements

The fifth proposed limitation to the implementation of Section XI

addresses reconciliation of replacement items

[Sec. 50.55a(b)(2)(xx)(A)] and the definition of Construction Code

[Sec. 50.55a(b)(2)(xx)(B)]. Changes to IWA-4222, ``Reconciliation of

Owner's Requirements,'' in the 1995 Addenda would permit a replacement

item produced at a facility not having a 10 CFR Part 50, Appendix B

qualified program to be used in safety-related applications. With

regard to the definition of Construction Code, a new definition of

Construction Code appeared in IWA-9000, ``Glossary,'' in the 1993

Addenda. Due to the changes made in IWA-4200 in the 1995 Addenda, the

change in definition could result in standards being utilized which do

not contain any QA requirements, or contain QA requirements that do not

fully comply with Appendix B. Thus, when implementing the 1995 Addenda

through the 1996 Addenda, Sec. 50.55a(b)(2)(xx)(A) would require

reconciliation of replacement items to the original QA requirements.

Section 50.55a(b)(2)(xx)(B) would require a licensee to reconcile

replacement items to the Construction Code and to the QA requirements

as described in the Owner's QA program.

Section XI Article IWA-4000 provides rules and requirements for the

repair and replacement of pressure retaining components and their

supports. Versions of IWA-4000 previous to the 1995 Addenda permitted a

licensee to purchase a replacement item to the standards of the

original Construction Code or a later version, provided that the

technical requirements of an item such as design and fabrication, as

well as the nontechnical requirements (identified as administrative

requirements in IWA-4222) such as QA

[[Page 63896]]

and Authorized Inspection of the later version were reconciled with

those of the original Construction Code and Owner's Requirements.

Reconciliation ensures that the replacement item meets certain

standards of quality so that it is satisfactory for the specified

design and operating conditions. In the 1995 Addenda, the provisions of

Code Case N-554, ``Alternative Requirements for Reconciliation of

Replacement Items,'' were incorporated into an extensive rewrite of

IWA-4200. As a result of these changes to IWA-4200, specifically IWA-

4222(a)(2), the nontechnical requirements for Class 1, 2, and 3 safety-

related replacement items would no longer need to be reconciled which

may result in noncompliance with 10 CFR Part 50, Appendix B. NRC

regulations require that any item which performs a safety-related

function must meet Appendix B. Appendix B invokes, among other things,

controls on suppliers of safety-related items. By not requiring

reconciliation of the administrative requirements, the provisions in

IWA-4222(a)(2) of the 1995 Addenda through the 1996 Addenda, would

allow vendors having a QA program which does not meet Appendix B to be

utilized, and may result in noncompliance with Appendix B. These

deficiencies could be resolved if the Code provided for commercial

grade item dedication in accordance with 10 CFR Part 21, ``Reporting of

Defects and Noncompliance.'' However, IWA-4222 does not address

commercial grade dedication. In addition, it should be pointed out that

a separate Code Case which provides an alternative for a specific

provision in IWA-4200, Code Case N-567, ``Alternative Requirements for

Class 1, 2, and 3 Replacement Components,'' was modified to require the

reconciliation of nontechnical requirements before the Code Case was

approved. Therefore, an inconsistency exists between the Code and a

Code Case. Thus, when implementing the 1995 Addenda through the 1996

Addenda, Sec. 50.55a(b)(2)(xx)(A) would require reconciliation of

replacement items to the original QA requirements.

The provisions of the Code in IWA-4222(a)(2) discussed above

address newly manufactured replacement parts. A further limitation on

the use of Article IWA-4200 in the 1995 Addenda through the 1996

Addenda is contained in Sec. 50.55a(b)(2)(xx)(B). IWA-4222(b) addresses

the use of items from a facility which was shutdown or for which

construction was halted. IWA-4222(b) permits the use of either the

administrative requirements of the Construction Code of the item being

replaced or the administrative requirements of the Construction Code of

the item being used for replacement. However, the definition of

``Construction Code'' was changed in the 1993 Addenda. In versions of

Section XI previous to the 1993 Addenda, Construction Code was defined

in IWA-9000, ``Glossary,'' as ``the body of technical requirements that

governed the construction of the item.'' Included in the body of

technical requirements that governed the construction of the item was a

requirement to reconcile the Owner's specification requirements, which

included NRC regulatory requirements, and applicable Owner design and

procurement specifications that invoke technical and nontechnical

requirements (e.g., 10 CFR Part 50, Appendix B). In the 1993 Addenda,

the definition became nationally recognized Codes such as ASME,

Specifications such as the American Society of Testing and Materials

(ASTM), and designated Code Cases. Either definition of Construction

Code would include the original Construction Codes for the design and

construction of piping, such as B31.1, ``Power Piping,'' and B31.7,

``Nuclear Piping,'' and those for the design and construction of

storage tanks, such as the American Petroleum Institute (API) 620,

``Design and Construction of Large, Welded, Low-Pressure Storage

Tanks,'' and API 650, ``Welded Steel Tanks for Oil Storage.'' However,

many of these standards utilized for construction do not contain any QA

requirements, or they contain QA requirements that do not fully comply

with Appendix B. Therefore, in order to satisfy Appendix B, QA

requirements similar to or meeting Appendix B were invoked in the

Owner's original procurement documents. Thus, when implementing IWA-

4200 (including subparagraphs IWA-4221, IWA-4222, IWA-4223, IWA-4224,

and IWA-5224), Sec. 50.55a(b)(2)(xx)(B) would require a licensee to

reconcile replacement items to the Construction Code and to the QA

requirements as described in the Owner's QA program.

2.3.2 OM Code (120-Month Update)

2.3.2.1 Class 1, 2, and 3 Pumps and Valves

The proposed amendment to Sec. 50.55a(f)(4) would require that IST

of pumps and valves be performed in accordance with the ASME ``Code for

Operation and Maintenance of Nuclear Power Plants'' (OM Code). A

proposed new section, Sec. 50.55a(b)(3), would specify the editions and

addenda of the OM Code that have been incorporated by reference into

Sec. 50.55a. Paragraph 50.55a(b)(3) together with Sec. 50.55a(f)(4) of

the proposed rule would require that licensees implement the 1995

Edition with the 1996 Addenda of the OM Code. Existing

Sec. 50.55a(f)(1) has been modified to clarify which pumps and valves

are to be included in the IST program. One proposed limitation to

implementation of the OM Code addressing QA, and one proposed

modification of the OM Code addressing stroke time testing have been

included.

2.3.2.2 Background--OM Code

Until 1990, the ASME Code requirements addressing IST of pumps and

valves were contained in Section XI Subsections IWP (pumps) and IWV

(valves). The provisions of IWP and IWV were last incorporated by

reference into Sec. 50.55a in a final rulemaking published on August 6,

1992 (57 FR 34666). In 1990, the ASME published the initial edition of

the OM Code which provides rules for IST of pumps and valves. The

requirements contained in the 1990 Edition are identical to the

requirements contained in the 1989 Edition of Section XI Subsections

IWP (pumps) and IWV (valves). The ASME Board on Nuclear Codes and

Standards has transferred responsibility for rules on IST from Section

XI to the OM Committee. As such, the Section XI rules for inservice

testing of pumps and valves that are presently incorporated by

reference into NRC regulations are no longer being updated by Section

XI.

The ASME 1990 Edition of the OM Code consists of one section

(Section IST) entitled ``Rules for Inservice Testing of Light-Water

Reactor Power Plants.'' This section is divided into four subsections,

ISTA, ``General Requirements,'' ISTB, ``Inservice Testing of Pumps in

Light-Water Reactor Power Plants,'' ISTC, ``Inservice Testing of Valves

in Light-Water Reactor Power Plants,'' and ISTD, ``Examination and

Performance Testing of Nuclear Power Plant Dynamic Restraints

(Snubbers).'' The IST of snubbers is governed by plant technical

specifications and, thus, has never been included in Sec. 50.55a.

Therefore, this proposed rule only requires implementation of

Subsections ISTA, ISTB, and ISTC. However, Sec. 50.55a(b)(3)(v) would

permit licensees to implement Subsection ISTD of the 1996 Addenda by

making a change to their technical specifications in accordance with

applicable NRC requirements.

[[Page 63897]]

2.3.2.3 Clarification of Safety-Related Valves

The existing Sec. 50.55a(f)(1) has been interpreted by some

licensees to mean that all safety-related pumps and valves regardless

of ASME Code Class (or equivalent) were to be included in the IST

program. The NRC proposes to modify this paragraph to clarify that the

provisions of Sec. 50.55a(f)(1) apply only to pumps and valves in

steam, water, air, and liquid radioactive waste systems that perform a

function to shut down the reactor, maintain the reactor in a safe

shutdown condition, mitigate the consequences of an accident, or

provide overpressure protection for such systems.

2.3.2.4 Limitation

2.3.2.4.1 Quality Assurance

The limitation to the implementation of the OM Code pertains to the

use of NQA-1, ``Quality Assurance Requirements for Nuclear

Facilities,'' with the OM Code. The OM Code references the use of

either NQA-1 or the Owner's Appendix B Quality Assurance Program as

part of its individual requirements for a QA program. At present,

Sec. 50.55a endorses NQA-1-1979 for the OM Code. The 1996 Addenda also

endorses NQA-1-1979. Thus, the 1996 OM Code has not endorsed a later

version of NQA-1. Because this rulemaking would incorporate the OM Code

by reference into Sec. 50.55a for the first time, a limitation is

included to address the same issues discussed previously in the Section

XI section on QA.

The NRC has determined that the provisions of NQA-1, 1979 Addenda,

would not adequately describe how to satisfy the requirements of

Appendix B as satisfied by Sec. 50.34(b)(6)(ii). Further, there are

various aspects of operational phase QA and administrative controls

which are not addressed by NQA-1. There are numerous areas where

American National Standards Institute (ANSI) standards or NRC

regulatory positions, which are specified in SRP 17.2, are either

nonmandatory or missing altogether from the NQA-1 provisions. However,

the Owner's QA Program, which has been approved by the NRC, is

adequate. Thus, the NRC has determined that the requirements of NQA-1-

1979, that are part of the incorporation by reference of the OM Code,

is acceptable for use in the context of the OM Code, as permitted by

ISTA 1.4, provided the licensee utilizes its 10 CFR Part 50, Appendix

B, QA program in conjunction with the OM Code. Changes to licensee's QA

program shall be made in accordance with 10 CFR 50.54. Further, where

NQA-1 and the OM Code do not address the commitments contained in the

licensee's Appendix B QA program description, such commitments shall be

applied to OM Code activities. Proposed Sec. 50.55a(b)(3)(i) addresses

licensee's commitments related to the OM Code.

2.3.2.5 Modification

2.3.2.5.1 Stroke Time Testing

Proposed Sec. 50.55a(b)(3)(ii) would require that the stroke time

testing requirement of Subsection ISTC of the OM Code applicable for

motor-operated valves (MOVs) be supplemented with programs that

licensees have previously committed to perform, prior to issuance of

this amendment to Sec. 50.55a, for demonstrating the design basis

capability of MOVs. Stroke time testing of MOVs has been specified in

ASME Section XI and is currently required by Sec. 50.55a(f). This same

testing is required by the OM Code. This testing is a useful tool and

complements other tests used to verify MOV function. Variation in

measured stroke times can indicate valve degradation. Additionally,

periodic stroking provides valve exercise and some measure of on-demand

reliability. However, as discussed in NRC Generic Letter (GL) 89-10

``Safety-Related Motor-Operated Valve Testing and Surveillance'' dated

June 28, 1989, it is now recognized that the stroke time testing alone

is not sufficient to provide assurance of MOV capability under design-

basis conditions.

Subsequent to licensees implementing programs pursuant to GL 89-10,

the NRC issued Generic Letter 96-05, ``Periodic Verification of Design-

Basis Capability of Safety-Related Motor-Operated Valves,'' on

September 18, 1996. This generic letter requested licensees to

establish a program, or to ensure the effectiveness of their current

program, to verify on a periodic basis that safety-related motor-

operated valves continue to be capable of performing their safety

functions within the current licensing bases of the facility. Prior to

issuance of this rule, licensees have made licensing commitments

pursuant to GL 96-05 that have been reviewed by the NRC staff. Most

licensees have committed to participate in the Joint Owners Group (JOG)

Program on MOV Periodic Verification. The JOG program includes three

phases: (1) licensees will establish an interim static diagnostic

testing program developed by JOG with a test frequency based on margin

and safety significance; (2) JOG will coordinate a dynamic testing

program over the next 5 years that includes approximately 150 MOVs with

participating licensees each testing a few MOVs three times over this

interval; and (3) based on the results of the dynamic testing program,

JOG will establish a long-term periodic test program. Proposed

Sec. 50.55a(b)(3)(ii) would require that licensees supplement the

stroke time testing requirements of the OM Code with these commitments.

2.4 Expedited Implementation

2.4.1 Appendix VIII

The proposed rule would require that licensees expedite

implementation of mandatory Appendix VIII, ``Performance Demonstration

for Ultrasonic Examination Systems,'' to Section XI, 1995 Edition with

the 1996 Addenda. Three proposed modifications would be included to

address NRC positions on the use of Appendix VIII. Licensees would be

required to implement Appendix VIII, including the modifications, for

all examinations of the pressure vessel, piping, nozzles, and bolts and

studs which occur after 6 months from the date of the final rule. The

proposed rule would not require any change to a licensee's ISI schedule

for examination of these components, but would require that the

provisions of Appendix VIII be used for all examinations after that

date rather than the ultrasonic testing (UT) procedures and personnel

requirements presently being utilized by licensees.

Appendix VIII provides the requirements for performance

demonstration for ultrasonic testing (UT) procedures, equipment, and

personnel used to detect flaws and size flaws. Its requirements are

applicable to all UT performed for Class 1, Class 2, and Class 3 items

(i.e., reactor vessel, nozzles, piping, and bolting and studs). These

requirements are also to be utilized when implementing the augmented

inservice inspection program for reactor vessel shell welds presently

required by Sec. 50.55a(g)(6)(ii)(A). The NRC has reviewed the 1995

Edition with the 1996 Addenda of Appendix VIII and has determined that

the provisions contained in this appendix should be used with three

modifications (addressed below). This mandatory appendix would normally

be adopted as part of the routine 120-month update specified in

Sec. 50.55a(g)(4), but because of the importance of the Appendix VIII

program, the NRC has determined that its requirements should be

implemented after 6 months from the date of the final rule. The

performance demonstration requirements in Appendix VIII would

[[Page 63898]]

substantially improve the ability of an examiner to detect and

characterize flaws in examined components. UT procedures and personnel

requirements are presently contained in Section XI but, as detailed in

the documented evaluation required by Sec. 50.109(a)(4), personnel

qualified to Appendix VIII are significantly better at detecting flaws.

The industry's Performance Demonstration Initiative (PDI) established a

process in accordance with Appendix VIII for reactor vessel, nozzle,

piping, and bolting examinations. PDI has received considerable support

from the industry, and every licensee has contributed financially. The

majority of the cost of PDI was in setting up the samples, which has

been completed. Proposed Sec. 50.55a(g)(6)(ii)(C)(1) would require

licensees to utilize the improved requirements in Appendix VIII for all

examinations of reactor vessels (including nozzles), piping, and

bolting performed after 6 months from the date of the final rule. To

date, the PDI program has qualified over 300 individuals for piping and

five teams for vessel examinations. Thus, the NRC does not believe that

a 6-month implementation period would result in hardship.

2.4.1.1 Modifications

2.4.1.1.1 Appendix VIII Personnel Qualification

The first proposed modification of Appendix VIII relates to its

requirement that ultrasonic examination personnel meet the requirements

of Appendix VII, ``Qualification of Nondestructive Examination

Personnel for Ultrasonic Examination,'' to Section XI. Appendix VII

first appeared in Section XI in the 1988 Addenda and was incorporated

by reference into Sec. 50.55a in a final rule published on August 6,

1992 (57 FR 34666). The NRC believes that the requirement in Appendix

VII-4240 for personnel to receive a minimum of 10 hours of training on

an annual basis is inadequate. Proposed Sec. 50.55a(b)(2)(xvii) would

require that all personnel qualified for performing ultrasonic

examinations in accordance with Appendix VIII receive 40 hours of

annual training which includes laboratory work and examination of

flawed specimens. Signals can be difficult to interpret, and as

detailed in the regulatory analysis for this rulemaking, experience and

studies indicate that the examiner must practice on a frequent basis to

maintain the capability for proper interpretation. In addition, these

studies have shown that this capability begins to diminish within

approximately 6 months if skills are not maintained. Thus, 10 hours of

annual training is not sufficient practice to maintain skills. The NRC

believes that a minimum of 40 hours of annual training, not 10 hours,

is required to maintain an examiner's abilities in this highly

specialized skill area. The NRC expects that licensees would distribute

the training over the course of the year to ensure that interpretation

skills do not diminish.

2.4.1.1.2 Appendix VIII Specimen Set Cracks

The second proposed modification of Appendix VIII would require

that all flaws in the specimen sets used for performance demonstration

for piping, vessels, and nozzles be cracks. For piping, Appendix VIII

requires that all of the flaws in a specimen set be cracks. However,

for vessels and nozzles, Appendix VIII would allow as many as 50% of

the flaws to be notches. For the purpose of demonstrating

nondestructive examination (NDE) capabilities, notches are not

realistic representations of service induced cracks. An inspector

cannot properly interpret service induced cracks by qualifying with

specimens containing notches. Notches are easier to detect than flaws

because notches have a higher amplitude and simpler signal

characteristics. Notches are easier to interpret and, in fact, the

probability of detecting notches can be much higher than the

probability of detecting cracks under similar conditions. In addition,

Appendix VIII provides a screening test that uses a relatively small

sample size containing few flaws. If some of the flaws are replaced by

notches that are unrealistic, the screening test becomes ineffective.

Because of these considerations, the flaws in the specimen sets

utilized for piping by EPRI for the PDI are all cracks. The regulatory

analysis for this rulemaking contains a detailed discussion of the

importance of using cracks in the specimens. Thus, proposed

Sec. 50.55a(b)(2)(xiii) would require that all flaws in the specimen

sets used for performance demonstration be cracks.

2.4.1.1.3 Appendix VIII Specimen Set Microstructure

The third proposed modification of Appendix VIII would require that

all specimens for single-side tests contain microstructures like the

components to be inspected and flaws with non-optimum characteristics

consistent with field experience that provide realistic challenges to

the UT technique. Appendix VIII does not distinguish specimens for two-

sided examinations from those used for single-sided examination.

Appendix VIII was originally developed using UT lessons learned

from two-sided examinations of welds. This UT experience provided the

input for designing specimens and selecting, locating, and

characterizing flaws. Studies have shown that defect characteristics

such as shape, size, depth, tilt angle, skew angle, roughness, and

crack tip affect the probability of detecting a particular flaw. For

example, it was demonstrated in one particular study (Reference 22 in

the documented evaluation) that a particular flaw was over three times

more reflective in one direction, thus easier to detect, than in the

opposite direction. Specimens designed for two-sided examination may

not have defects which are appropriate for single-sided performance

demonstration; i.e., the specimens may not adequately test an examiners

proficiency in detecting flaws. Therefore, in order to proceed with the

effort of qualifying UT systems (equipment, procedures, and personnel)

for single-sided examinations, proposed Sec. 50.55a(b)(2)(xx) would

require the industry to develop sets of specimens that contain

microstructures similar to the types found in the components to be

inspected and flaws with non-optimum characteristics, such as skew,

tilt, and roughness, consistent with field experience that provide

realistic challenges for single-sided performance demonstration.

2.4.2 Generic Letter on Appendix VIII

A draft generic letter was published in the Federal Register (61 FR

69120) for public comment on December 31, 1996, to alert the industry

to the importance of using equipment, procedures, and examiners capable

of reliably detecting and sizing flaws in the performance of

comprehensive examinations of reactor vessels and piping. The generic

letter stated that even though the need for improvement clearly

existed, the staff had reached the conclusion that immediate

backfitting of Appendix VIII in advance of this proposed rulemaking was

not warranted. This conclusion was based on consideration of defense-

in-depth measures, Code margins in component design, leakage monitoring

systems, and also that Appendix VIII was already being applied to

selected piping subject to intergranular stress corrosion cracking. The

NRC received 16 comment letters on the generic letter.

The comments generally were very similar and can be summarized in

the following five items: (1) it is inappropriate to request licensees

to voluntarily commit to a program in a

[[Page 63899]]

generic letter; (2) the urgency for licensee's to voluntarily commit to

implementing Appendix VIII is inconsistent with the statement in the

generic letter that a safety concern does not exist that would warrant

immediate backfitting in advance of the rulemaking; (3) the

performance-based qualification program of Appendix VIII should be

approved an alternative to the current ASME Code, and Appendix VIII as

implemented by PDI should be recognized as an acceptable alternative

for Appendix VIII; (4) the NRC should provide guidance on incorporating

Appendix VIII and/or PDI into plant-specific ISI programs; and (5) the

generic letter would request that licensees update their UT ISI and

augmented inspection commitments to a Code edition not yet referenced

in the regulations.

With regard to the first comment, the NRC disagrees that it is

inappropriate to request licensees to voluntarily commit to a program

in a generic letter. This is one mechanism available to the NRC for

alerting licensees, for example, to degraded conditions which may

unacceptably affect the function of safety-related components. The

second comment takes the generic letter statement out of context. What

the generic letter actually stated was that a safety concern did not

exist to warrant immediate backfitting in advance of the rulemaking

because of defense-in-depth measures, Code margins in design, and that

Appendix VIII was already being applied to selected piping subject to

intergranular stress corrosion cracking. The NRC strongly disagrees

that Appendix VIII and Appendix VIII as implemented by PDI should be

alternatives to the present Code rules. As detailed in the documented

evaluation for backfitting Appendix VIII, it has been demonstrated that

examiners previously considered qualified under Section XI generally

have marginal UT skills. This was evident from the discouragingly low

percentage of examiners initially satisfying the screening criteria for

detecting flaws under the PDI program. Comment four regarding guidance

on incorporating Appendix VIII into present ISI programs, and comment

five regarding Code edition are automatically resolved in a rulemaking

format.

At the time the generic letter was issued, this proposed rulemaking

was still under development. The purpose of the generic letter was to

alert the industry to the (1) generally poor performance in detecting

flaws and (2) the Commission's intent to endorse Appendix VIII via

rulemaking. Publication of a final rule would obviate the need for the

generic letter.

2.4.3 Class 1 Piping Volumetric Examination

A proposed modification of Section XI would require licensees of

pressurized water reactor plants to supplement the surface examination

of Class 1 High Pressure Safety Injection Systems (HPSI) piping as

required by Examination Category B-J of Table IWB-2500-1 for nominal

pipe sizes (NPS) between 4 (inches) and 1+ (inches), with a volumetric

(ultrasonic) examination. This requirement is proposed because (1)

inside diameter cracking of HPSI piping in the subject size range has

been previously discovered (as detailed in NRC Generic Letter 85-20,

``High Pressure Injection/Make-Up Nozzle Cracking in Babcock and Wilcox

Plants,'' and in NRC Information Notice 97-46, (``Unisolable Crack in

High-Pressure Injection Piping,''), (2) failure of this line could

result in a small break loss of coolant accident while directly

affecting the system designed to mitigate such an event, and (3)

volumetric examinations are already required by the Code for Class 2

portions of this system (Table IWC-2500-1, Examination Category C-F-1)

within the same NPS range. Thus, not only are the requirements between

Class 1 and Class 2 inconsistent (with the Class 1 portions being

subject to less stringent testing requirements as compared with Class 2

portions of the same type of piping), but operating experience has

shown that these reactor coolant pressure boundary (RCPB) pipe

examinations need to be more comprehensive. Proposed

Sec. p50.55a(b)(2)(xv) would require licensees to supplement the

Section XI required surface examination for the Class 1 portion of the

HPSI system with volumetric examination in order to ensure the

integrity of the reactor coolant pressure boundary as required by

General Design Criteria (GDC) 14, 10 CFR Part 50, Appendix A, or

similar provisions in the licensing basis for these facilities, and

Criteria II and XVI of 10 CFR Part 50, Appendix B. Licensees would be

required to perform the volumetric examination during any ISI program

inspection of the HPSI system performed after 6 months from the date of

the final rule. Utilization of licensee's existing ISI schedules will

result in the volumetric examinations being implemented in a reasonable

period of time while not impacting lengths of outages or requiring

facility shutdown solely for performance of these examinations.

2.5 Voluntary Implementation

2.5.1 Section III

The NRC has reviewed the 1989 Addenda, 1990 Addenda, 1991 Addenda,

1992 Edition, 1992 Addenda, 1993 Addenda, 1994 Addenda, 1995 Edition,

and 1996 Addenda of Section III, Division 1, for Class 1, Class 2, and

Class 3 components, and has determined that they are acceptable for

voluntary use with six proposed limitations. In addition, Sec. 50.55a

would be modified to ensure consistency between Sec. 50.55a and NCA-

1140.

The version of Section III utilized by licensees is chosen prior to

construction. Section 50.55a permits licensees to use the original

construction code during the operational phase or voluntarily update to

a later version which has been endorsed by Sec. 50.55a. Accordingly,

the proposed limitations to Section III become effective only when a

licensee voluntarily updates to a later version. The modification would

only apply to a applicant for a new construction permit.

2.5.1.1 Limitations

2.5.1.1.1 Engineering Judgement

The first proposed limitation to the implementation of Section III

would establish an NRC restriction with regard to the Foreword in the

1992 Addenda through the 1996 Addenda of the BPV Code. That Foreword

addresses the use of ``engineering judgement'' for construction

activities not specifically considered by the Code. Proposed paragraph

50.55a(b)(1)(i) would require that when a licensee relies on

engineering judgement for activities or evaluations of components or

systems within the scope of Sec. 50.55a that are not directly addressed

by the BPV Code, the licensee must receive NRC approval for those

activities or evaluations pursuant to Sec. 50.55a(a)(3).

2.5.1.1.2 Section III Materials

The second proposed limitation to the implementation of Section III

pertains to a reference to Section II, ``Materials,'' Part D,

``Properties.'' Section II, Part D, contained many printing errors in

the 1992 Edition. These errors were corrected in the 1992 Addenda.

Proposed Sec. 50.55a(b)(1)(ii) would require that Section II, 1992

Addenda, be applied when using the 1992 Edition of Section III. The

limitation is necessary to ensure that users of the Code use the design

stresses intended by the ASME Code.

2.5.1.1.3 Weld Leg Dimensions

The third proposed limitation to the implementation of Section III

would

[[Page 63900]]

correct a conflict in the design and construction requirements in

Subsection NB (Class 1 Components), Subsection NC (Class 2), and

Subsection ND (Class 3) of Section III, 1989 Addenda through the 1996

Addenda of the BPV Code. Two equations in NB-3683.4(c)(1), Footnote 11

to Figure NC-3673.2(b)-1, and Figure ND-3673.2(b)-1 were modified in

the 1989 Addenda and are no longer in agreement with Figures NB-4427-1,

NC-4427-1, and ND-4427-1. This change results in a different weld leg

dimension depending on whether the dimension is derived from the text

or calculated from the figures. Thus, to ensure consistency, proposed

Sec. 50.55a(b)(1)(iii) would require that licensees use the 1989

Edition for the above referenced paragraphs and figures in lieu of the

1989 Addenda through the 1996 Addenda.

2.5.1.1.4 Seismic Design

The fourth proposed limitation to the implementation of Section III

pertains to new requirements for piping design evaluation contained in

the 1994 Addenda through the 1996 Addenda of the BPV Code. The NRC has

determined that changes to subarticles NB-3200, ``Design by Analysis,''

NB-3600, ``Piping Design,'' NC-3600, ``Piping Design,'' and ND-3600,

``Piping Design,'' of Section III for Class 1, 2, and 3 piping design

evaluation for reversing dynamic loads (e.g., earthquake and other

similar type dynamic loads which cycle about a mean value) are

unacceptable. The new requirements are based on the premise that loads

such as earthquake loads are not capable of producing collapse or gross

distortion of a component. The requirements, in part, are based on

General Electric evaluations of the test data performed under

sponsorship of the Electric Power Research Institute (EPRI) and the

NRC. However, NRC evaluations of the data do not support the changes

and indicate lower margins than those estimated in earlier evaluations.

The ASME has established a special working group to reevaluate the

bases for the seismic design for piping. Thus, in proposed

Sec. 50.55a(b)(1)(iv), licensees would be permitted to use articles NB-

3200, NB-3600, NC-3600, and ND-3600, in the 1989 Addenda through the

1993 Addenda, but would be prohibited from using these requirements in

the 1994 Addenda through the 1996 Addenda.

2.5.1.1.5 Quality Assurance

The fifth proposed limitation to the implementation of Section III

pertains to the use of NQA-1, ``Quality Assurance Requirements for

Nuclear Facilities,'' with Section III. Section III references NQA-1 as

part of its individual requirements for a QA program by integrating

portions of NQA-1 into the QA program defined in NCA-4000, ``Quality

Assurance.'' At present, Sec. 50.55a endorses the 1989 Edition of the

ASME Code which references NQA-1-1986 for Section III. The 1996 Addenda

of the ASME Code references NQA-1-1992 for Section III.

The NRC has reviewed the requirements of NQA-1, 1986 Addenda

through the 1992 Addenda, that are part of the incorporation by

reference of Section III, and has determined that the provisions of

NQA-1 are acceptable for use in the context of Section III activities.

Portions of NQA-1 are integrated into Section III administrative,

quality, and technical provisions which provide a complete QA program

for design and construction. NQA-1 by itself would not adequately

describe how to satisfy the requirements of 10 CFR Part 50, Appendix B,

``Quality Assurance Criteria for Nuclear Power Plants and Fuel

Reprocessing Plants.'' The additional criteria contained in Section

III, such as nuclear accreditation, audits, and third party inspection,

establishes a complete program and satisfies the requirements of

Appendix B (i.e., the provisions of Section III integrated with NQA-1).

Because licensees may voluntarily choose to apply later provisions of

Section III, proposed Sec. 50.55a(b)(1)(v) contains a limitation which

would require that the edition and addenda of NQA-1 specified by NCA-

4000 of Section III be used in conjunction with the administrative,

quality, and technical provisions contained in the edition of Section

III being utilized.

2.5.1.1.6 Independence of Inspection

The sixth proposed limitation to the implementation of Section III

would prohibit licensees from using subparagraph NCA-4134.10(a),

``Inspection,'' in the 1995 Edition through the 1996 Addenda. Prior to

this edition and addenda, NCA-4134.10(a) required that the provisions

of NQA-1, ``Quality Assurance Program Requirements for Nuclear

Facilities,'' Basic Requirement 10, ``Inspection,'' and Supplement 10S-

1, ``Supplementary Requirements for Inspection,'' be utilized without

exception. In the 1995 Edition, NCA-4134.10(a) was modified so that

paragraph 2 of Supplement 10S-1 and the requirements for independence

of inspection were no longer required. Supplement 10S-1, 2.1, states

that ``Inspection Personnel shall not report directly to the immediate

supervisors who are responsible for performing the work being

inspected.'' Subparagraph 2.2 states ``Each person who verifies

conformance of work activities for purposes of acceptance shall be

qualified to perform the assigned task.'' By exempting Supplement 10S-1

paragraph 2 from the requirements of NCA-4134.10, Section III could

promote noncompliance with 10 CFR 50, Appendix B, ``Quality Assurance

Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,''

Criterion 1, ``Organization.'' This criterion requires that persons

performing QA functions report to a management level such that

authority and organizational freedom, including sufficient independence

from cost and schedule when opposed to safety considerations, are

provided. Thus, in proposed Sec. 50.55a(b)(1)(vi), licensees would be

permitted to use the provisions contained in NCA-4134.10(a), in the

1989 Addenda through the 1994 Addenda, but would be prohibited from

using these provisions in the 1995 Edition through the 1996 Addenda.

2.5.1.2 Modification

2.5.1.2.1 Applicable Code Version for New Construction

The proposed modification of Section III addresses a possible

conflict between NCA-1140 and Sec. 50.55a for new construction. NCA-

1140 of Section III requires that the length of time between the date

of the edition and addenda used for new construction and the docket

date of the nuclear power plant be no greater than three years.

Paragraph 50.55a(b)(1) requires that the edition and addenda utilized

be incorporated by reference into the regulations. The possibility

exists that the edition and addenda required by the ASME Code to be

used for new construction would not be incorporated by reference into

Sec. 50.55a. In order to resolve this possible discrepancy, the NRC

proposes to modify existing Secs. Sec. 50.55a(c)(3)(i),

50.55a(d)(2)(i), and 50.55a(e)(2)(i), to permit an applicant for a

construction permit to use the latest edition and addenda which has

been incorporated by reference into Sec. 50.55a(b)(1) if the

requirements of the ASME Code and the regulations cannot simultaneously

be satisfied.

2.5.2 Section XI (Voluntary Implementation)

Licensees would be permitted to update from the 1992 Edition with

the 1992 Addenda of Subsection IWE and Subsection IWL to the 1995

Edition with the 1996 Addenda. In addition, licensees could implement

Code Case

[[Page 63901]]

N-513, ``Evaluation Criteria for Temporary Acceptance of Flaws in Class

3 Piping,'' and Code Case N-523-1, ``Mechanical Clamping Devices for

Class 2 and 3 Piping.''

2.5.2.1 Subsection IWE and Subsection IWL

Many of the provisions in Section XI Subsection IWL, ``Requirements

for Class CC Concrete Components of Light-Water Cooled Power Plants,''

pertaining to the inspection of the tendons of concrete containments

were based on guidance contained in Regulatory Guide 1.35, ``Inservice

Inspection of Ungrouted Tendons in Prestressed Concrete Containments.''

A final rule published on August 8, 1996 (61 FR 41303) incorporated by

reference the 1992 Edition with the 1992 Addenda of Subsection IWE,

``Requirements for Class MC and Metallic Liners of Class CC Components

of Light-Water Cooled Power Plants,'' and Subsection IWL. At that time,

there were several key positions in the regulatory guide addressing the

trending of prestress losses, unanticipated tendon elongation, grease

leakage, and excessive water in the sampled sheathing filler grease not

addressed in Subsection IWL because the ASME Code committees had not

yet completed consideration of these positions. Due to the importance

of these positions, the final rule addressed them in paragraphs

50.55a(b)(2)(ix)(A) through 50.55a(b)(2)(ix)(D)(3). In addition, the

final rule contained Sec. 50.55a(b)(2)(ix)(E) which addressed the

occurrence of degradation in inaccessible areas of containments.

Since publication of the 1992 Addenda, the ASME Code committees

have completed their consideration of those regulatory guide positions.

Most have been incorporated into subsequent edition and addenda, and

the 1995 Edition with the 1996 Addenda addresses all of the

modifications listed above except grease leakage and degradation in

inaccessible areas. Thus, licensees would be required to utilize the

modifications presently in Sec. 50.55a addressing grease leakage and

degradation in inaccessible areas. The NRC has determined that the

provisions contained in Subsection IWE and Subsection IWL, 1995 Edition

with the 1996 Addenda Code, in conjunction with the modifications,

would be acceptable.

The final rule published on August 8, 1996 (61 FR 41303)

incorporated Subsection IWE and Subsection IWL into Sec. 50.55a for the

first time. The final rule contained a requirement for licensees to

develop and implement a containment ISI program within five years. Each

plant had a pre-existing ISI program to address Class 1, Class 2, and

Class 3 components. The rule left it to the licensee's discretion

whether to have two separate ISI programs, or merge the containment ISI

program with the pre-existing program.

It has been over a year since the final rule was issued, and some

licensees have begun the development of a containment ISI program to

comply with the required 5-year implementation period. This containment

ISI program will be based on the 1992 Edition with the 1992 Addenda as

required by the final rule. However, other licensees have indicated

that they will request NRC approval pursuant to Sec. 50.55a(a)(3) to

use later editions and addenda of Subsection IWE and Subsection IWL

before this proposed rule becomes final. Thus, to provide flexibility,

Sec. 50.55a(b)(2)(vi) has been modified. Licensees would be permitted

to implement either the presently required 1992 Edition with the 1992

Addenda, or the latest containment examination provisions; i.e., 1995

Edition with the 1996 Addenda.

For those licensees implementing the 1992 Edition with the 1992

Addenda, all of the modifications contained in paragraphs

50.55a(b)(2)(ix)(A) through 50.55a(b)(2)(ix)(D)(3) must be applied as

presently required by Sec. 50.55a. Licensees wishing to implement the

1995 Edition with the 1996 Addenda would be required to apply

paragraphs 50.55a(b)(2)(ix)(A), 50.55a(b)(2)(ix)(D)(3), and

50.55a(b)(2)(ix)(E). Paragraph Sec. 50.55a(b)(2)(ix) would thus be

modified. According to Sec. 50.55a(g)(6)(ii)(B)(1), the containment

examinations performed during the 5-year implementation period are

those examinations which are required by Subsection IWE during the

first period of what will be the first containment inspection interval.

(Since Subsection IWL is based on a 5-year schedule, standard Section

XI periods do not apply for the examination of concrete containments

and their post-tensioning systems). With completion of the first period

examinations, the second period of the first containment ISI interval

would begin. The end of the third period completes the first

containment ISI interval, a containment ISI 120-month update has been

completed, and the second containment ISI interval would begin.

As licensees have begun developing their containment ISI programs,

the NRC has received requests to clarify the implementation schedule

for ISI of concrete containments and their post-tensioning systems. The

current wording of Sec. 50.55a(g)(6)(ii)(B)(2) requiring licensees to

implement ``the inservice examinations which correspond to the number

of years of operation which are specified in Subsection IWL'' has

created confusion regarding whether the first examination of concrete

is required to meet the examination schedule in Section XI, Subsection

IWL, IWL-2410, which is based on the date of the Structural Integrity

Test (SIT), or may be performed at any time between September 9, 1996

and September 9, 2001. According to Sec. 50.55a(g)(6)(ii)(B)(2) of the

final rulemaking, the first examination of concrete may be performed at

any time between September 9, 1996, and September 9, 2001. The date of

the first examination of concrete is not conditional upon compliance

with Subsection IWL-2410 or the SIT. The purpose of the italicized

words is to maintain the present 5-year schedule for examination of the

post-tensioning system as operating plants transition to Subsection

IWL. For operating reactors, there is no need to repeat the 1, 3, 5-

year implementation cycle.

Section 50.55a(g)(6)(ii)(B)(2) also stated that the first

examination performed shall serve the same purpose for operating plants

as the preservice examination specified for plants not yet in

operation. The affected plants are presently operating, but they will

be performing the examination of concrete under Subsection IWL for the

first time. Because the plants are operating, a Section XI preservice

examination cannot be performed. Therefore, the first concrete

examination is to be an inservice examination which will serve as the

baseline (the same purpose for operating plants as the preservice

examination specified for plants not yet in operation). With completion

of this first examination of concrete, the second five-year Subsection

IWL ISI period would begin. Likewise, examinations of the post-

tensioning system at the nth year (e.g., the 15th year post-tensioning

system examination), if performed to the requirements of Subsection

IWL, are to be performed to the ISI requirements, not the preservice

requirements.

The NRC has also been requested to clarify the schedule for future

examinations of concrete and their post-tensioning systems at both

operating and new plants. There is no requirement in Subsection IWL to

perform the examination of the concrete and the examination of the

post-tensioning system at the same time. The examination of the

concrete under Subsection IWL and the examination of

[[Page 63902]]

the liner plates of concrete containments under Subsection IWE may be

performed at any time during the 5-year expedited implementation. This

examination of the concrete and liner plate provides the baseline for

comparison with future containment ISI. Coordination of these schedules

in future examinations is left to each licensee. New plants would be

required to follow all of the provisions contained in Subsection IWL,

i.e., satisfy the preservice examination requirements and adopt the 1,

3, 5-year examination schedule ISI schedule.

2.5.2.2 Flaws in Class 3 Piping

Proposed Sec. 50.55a(b)(2)(xvi) would permit licensees to use Code

Case N-513, ``Evaluation Criteria for Temporary Acceptance of Flaws in

Class 3 Piping,'' and Code Case N-523-1, ``Mechanical Clamping Devices

for Class 2 and 3 Piping.'' Section XI contains repair methods for

pipes with a flaw exceeding acceptable limits. These repairs restore

the integrity of the flawed piping. There are certain cases, however,

where a Section XI Code repair may be impractical for a flaw detected

during plant operation (i.e., a plant shutdown would be required to

effect the Code repair). For many safety-related piping systems,

immediate repair is required regardless of plant status. However, it

has been determined that under certain conditions, temporary acceptance

of flaws, including through-wall leaking, of low and moderate energy

Class 3 piping is acceptable provided that the conditions are met, and

the repair is effected during the next outage. At present, licensees

must request NRC staff approval to defer Section XI Code repair for

these Class 3 moderate energy (200 xF, 275 psig) piping systems. The

NRC has reviewed Code Case N-513 and Code Case N-523-1 and has

determined that Code Case N-523-1 is acceptable. Code Case N-513 is

acceptable except for the scope and Section 4.0.

Section 1.0(a) of the Scope to Code Case N-513 limits the use of

the requirements to Class 3 piping. However, Section 1.0(c) would allow

the flaw evaluation criteria to be applied to all sizes of ferritic

steel and austenitic stainless steel pipe and tube. Without some

limitation on the scope of the Code Case, the flaw evaluation criteria

could be applied to components such as pumps and valves, original

construction deficiencies, and pressure boundary leakage; applications

for which the criteria should not be utilized. Thus, the NRC has

determined that the Code Case shall not be applied to: (1) components

other than pipe and tube, such as pumps, valves, expansion joints, and

heat exchangers; (2) the discovery and repair of flaws or deficiencies

remaining from original construction; (3) leakage through a flange

gasket; (4) threaded connections employing nonstructural seal welds for

leakage prevention (through seal weld leakage is not a structural flaw,

thread integrity must be maintained); and (5) degraded socket welds. A

proposed limitation would be added in Sec. 50.55a(b)(2)(xvi)(B) which

would preclude the use of Code Case N-513 for these applications.

The first paragraph of Section 4.0 of Code Case N-513 contains the

flaw acceptance criteria. The criteria provide a safety margin based on

service loading conditions. The second paragraph of Section 4.0,

however, would permit a reduction of the safety factors based on a

detailed engineering evaluation. No criteria or guidance is given for

justifying a reduction, or limiting the amount of reduction. The

acceptance criteria of the first paragraph are based on sound

principles. The second paragraph would allow ever finer calculation

until the available margins became unacceptably low. A limitation would

be added in proposed Sec. 50.55a(b)(2)(xvi)(A) requiring that when

implementing Code Case N-513, the specific safety factors in the first

paragraph of Section 4.0 be satisfied. The use of Code Case N-513, with

the limitations, and Code Case N-523-1 would obviate the need for

licensees to request approval for deferring repairs, thus saving NRC

and licensee resources.

2.5.3 OM Code (Voluntary Implementation)

Licensees would be permitted to implement Code Case OMN-1 in lieu

of stroke time testing as required in Subsection ISTC. Licensees would

also be permitted to implement Appendix II as an alternative to the

condition monitoring program provisions contained in Subsection ISTC.

However, licensees choosing to implement Appendix II would be required

to apply the three proposed modifications to Appendix II to supplement

check valve condition monitoring. In addition, licensees would be

permitted to use Subsection ISTD for the IST of snubbers.

2.5.3.1 Code Case OMN-1

An alternative to the provisions contained in Sec. 50.55a(b)(3)(ii)

is included in proposed Sec. 50.55a(b)(3)(iii) which would permit

licensees to voluntarily implement ASME Code Case OMN-1, ``Alternative

Rules for Preservice and Inservice Testing of Certain Electric Motor

Operated Valve Assemblies in LWR Power Plants.'' The NRC has determined

that for motor-operated valves, Code Case OMN-1 is acceptable in lieu

of Subsection ISTC, except for leakage rate testing (ISTC 4.3) which

must continue to be performed. As indicated in Attachment 1 to GL 96-

05, the Code case meets the intent of the generic letter, but with

certain limitations which were discussed in the generic letter. The NRC

supports the OMN-1 maximum motor-operated valve test interval of 10

years based on current knowledge and experience, but believes it

prudent to require that licensees evaluate the information obtained for

each motor-operated valve during the first five years of use of the

Code case, or three refueling outages (whichever is longer) to validate

assumptions made in justifying a longer test interval. These

limitations on the use of OMN-1 would be added to the rule as a

modification in Sec. 50.55a(b)(3)(iii)(A). Thus, Code Case OMN-1 is

acceptable in lieu of Subsection ISTC, other than leakage rate testing

requirements, with the modification that five years or three refueling

outages (whichever is longer) from initial implementation of Code Case

OMN-1, the adequacy of the test interval for each motor-operated valve

must be evaluated and adjusted as necessary.

In addition, as noted in GL 96-05, licensees are cautioned when

implementing Code Case OMN-1 that the benefits of performing a

particular test should be balanced against the potential adverse

effects placed on the valves or systems caused by this testing. Code

Case OMN-1 specifies that an IST program should consist of a mixture of

static and dynamic testing. While there may be benefits to performing

dynamic testing, there are also potential detriments to its use (i.e.,

valve damage). Licensees should be cognizant of this for each MOV when

selecting the appropriate method or combination of methods for the IST

program.

2.5.3.2 Appendix II

Paragraph ISTC 4.5.5 of Subsection ISTC permits the Owner to use

Appendix II, ``Check Valve Condition Monitoring Program,'' of the OM

Code, as an alternative to the testing or examination provisions of

ISTC 4.5.1 through ISTC 4.5.4. If an Owner elects to use Appendix II,

the provisions of Appendix II become mandatory. However, upon reviewing

the appendix, the NRC has determined that the requirements in Appendix

II must be supplemented. The first area that the NRC believes requires

supplementation is the demonstration of acceptable valve performance.

Appendix II requires no testing or examination of the check

[[Page 63903]]

valve obturator movement to both the open and closed positions. Testing

or examination of the check valve obturator in one direction only

cannot assure the unambiguous detection of a functionally degraded

check valve. The valve obturator must be tested or examined in both the

opening and closing directions to assess its condition and confirm

acceptable performance. Proposed Sec. 50.55a(b)(3)(iv)(A) would require

bi-directional testing of check valves.

Length of test interval is the second area of Appendix II where the

NRC believes the rules must be supplemented. Appendix II was first

incorporated into the OM Code in the 1996 Addenda. Thus, the operating

experience database does not yet exist to support long term test

intervals for the condition monitoring concept. Under the current check

valve IST program, most valves are tested quarterly during plant

operation. The interval for certain valves has been extended to

refueling outages. Under the appendix, a licensee would be able to

extend the interval without limit. A policy of prudent and safe

interval extension dictates that any additional interval extension must

be limited to one fuel cycle, and this extension must be based on

sufficient experience to justify the additional time. Interval changes

or extensions must be justified and limited within the existing

performance and experience database. Condition monitoring and the

current experience data base may qualify some valves for an initial

extension to every other fuel cycle, while trending and evaluation of

the data may dictate that the testing interval for some valves be

reduced. Extensions of IST intervals must consider plant safety and be

supported by trending and evaluating both generic and plant-specific

performance data to ensure the component is capable of performing its

intended function over the entire IST interval. Proposed

Sec. 50.55a(b)(3)(iv)(B) would limit the time between the initial test

or examination and second test or examination to two fuel cycles or

three years (whichever is longer), with additional extensions limited

to one fuel cycle, and the total interval would be limited to a maximum

of 10 years. An extension or reduction in the interval between tests or

examinations would have to be supported by trending and evaluation of

performance data.

The final area in Appendix II which the Commission believes should

be supplemented is the requirement applicable to a licensee who

discontinues a condition monitoring program. A licensee who

discontinues use of Appendix II, under IST 4.5.5 is required to return

to the requirements of IST 4.5.4. However, the NRC believes the

requirements of IST 4.5.1 through IST 4.5.4 must be also met. Hence, if

the monitoring program is discontinued, proposed

Sec. 50.55a(b)(3)(iii)(C) would require a licensee to implement the

provisions of IST 4.5.1 through IST 4.5.4.

2.5.3.3 Subsection ISTD

The IST of dynamic restraints or snubbers is governed by plant

technical specification and, thus, has never been included in

Sec. 50.55a. However, the NRC has reviewed Subsection ISTD, 1995

Edition with the 1996 Addenda, and has determined that the provisions

for IST of snubbers are an acceptable alternative to the requirements

contained in the plant technical specifications. Subsection ISTD, 1996

Addenda, includes new provisions for service life monitoring of

snubbers. The new provisions require that the service lives of snubbers

be predicted and evaluated to ensure that the service life will not be

exceeded before the next scheduled refueling outage. These new

provisions simply formalize preventative maintenance practices

presently found in most plants. Because the IST of snubbers is governed

by plant technical specifications, Subsection ISTD is not included in

the proposed mandatory requirements of the rulemaking, but licensees

may choose to voluntarily implement Subsection ISTD, 1995 Edition with

the 1996 Addenda, by processing a change to their technical

specifications. This proposed modification is contained in

Sec. 50.55a(b)(3)(v).

2.5.3.4 Containment Isolation Valves

The proposed amendment would delete the existing modification in

Sec. 50.55a(b)(2)(vii) for IST of containment isolation valves (CIVs),

which was added to the regulations in a rulemaking effective on August

6, 1992 (57 FR 34666). That rulemaking incorporated by reference, among

other things, the 1989 Edition of ASME Section XI, Subsection IWV that

endorsed Part 10 of ASME/ANSI OMa1988 for valve inservice testing. A

modification to the testing requirements of Part 10 related to CIVs was

included in the rulemaking indicating that paragraphs 4.2.2.3(e) and

4.2.2.3(f) of Part 10 were to be applied to CIVs. As noted in the

``Supplementary Information'' for the August 6, 1992 rulemaking, the

ASME Operations and Maintenance (OM) Committee had initiated action to:

(1) perform a comprehensive review of OM Part 10 CIV testing

requirements and acceptance standards; and (2) develop a basis document

that would provide, as a minimum, a documented basis for not including

the requirements for analysis of leakage rates and corrective actions

in Part 10 for those CIVs that do not provide a reactor coolant system

pressure isolation function. The NRC made a commitment via the

Supplementary Information to reevaluate the need for the modification

to Section XI, Subsection IWV, following review of this OM Committee

basis document. This basis document was transmitted to the NRC in a

letter from Steve Weinman, Secretary, OM Committee, to Eric S.

Beckjord, Director, Office of Nuclear Regulatory Research, dated

February 16, 1994. The NRC has determined that the requirements of 10

CFR 50, Appendix J, ensure adequate identification analysis, and

corrective actions for leakage monitoring of CIVs, and that the

existing modification in Sec. 50.55a(b)(2)(vii) should be deleted. The

regulatory analysis for this proposed rule contains a detailed

discussion of the basis document findings and the NRC staff evaluation.

2.6 ASME Code Interpretations

The ASME issues Interpretations to clarify provisions of the BPV

and OM Codes. Requests for Interpretations are submitted by users, and

after appropriate committee deliberations and balloting, responses are

issued by the ASME. Generally, the NRC agrees with these

interpretations. When the NRC incorporates by reference specific

editions and addenda into its regulations, the NRC has a certain

understanding of those editions and addenda. Because an Interpretation

is issued subsequent to issuance of the provision to which it refers,

the Interpretation may affect that understanding. While the NRC

acknowledges that the ASME is the official interpreter of the Code, the

NRC will not accept ASME interpretations that, in NRC's opinion, are

contrary to NRC requirements or may adversely impact facility

operations. Interpretations have been issued which in some cases,

conflicted with or were inconsistent with NRC requirements. These

resulted in enforcement actions. Of particular concern are Code

Interpretations that may be implemented following initiation of

enforcement action by the NRC. ASME Code Interpretations were discussed

in Part 9900, Technical Guidance, of the NRC Inspection Manual. Part

9900 provides that licensees should exercise caution when applying

Interpretations as they are not specifically part of the

[[Page 63904]]

incorporation by reference into Sec. 50.55a and have not received NRC

approval.

2.7 DSI-13

Since 1992, when the Commission last revised Sec. 50.55a to endorse

new ASME Code Editions and addenda (57 FR 34666), several developments

have occurred which have raised some fundamental issues with respect to

the Commission's endorsement of ASME Codes. First, on October 21, 1993,

Entergy Operations, Inc. submitted a request that would relieve it from

updating its ISI and IST programs to the last ASME Code edition and

addenda incorporated by reference into Sec. 50.55a. The underlying

premise of the request was that a licensee should not be required to

upgrade its ISI and IST program without considering whether the costs

of the upgrade are warranted in light of the increased safety afforded

by the updated Code edition and addenda. Though the request was later

withdrawn, the underlying premise resulted in NRC reconsideration of

the 120-month update. Requiring Code updates every 120-months is still

under active consideration. However, the proposed rule has been

prepared under the traditional approach; i.e., licensees would be

required to update their ISI and IST programs every 120-months to the

latest edition and addenda incorporated by reference into Sec. 50.55a.

If a decision is reached subsequent to publication of the proposed rule

that is adverse to this approach, this position will be corrected prior

to publication of the final rule.

Second, the National Technology Transfer and Advancement Act of

1995, PL 104-113, was signed into law on March 7, 1996. The Act directs

federal agencies to achieve greater reliance on technical standards

developed by voluntary consensus standards development organizations.

Finally, the Commission commenced a Strategic Assessment and

Rebaselining Initiative. One of the issues addressed in this effort was

Direction Setting Issue (DSI) 13, which raised the question, ``In

performing its regulatory responsibilities, what consideration should

the NRC give to industry activities.'' A draft paper addressing DSI-13

was published for public comment on September 16, 1996, after which the

Commission held public meetings to facilitate understanding of the

issues and receive comments on the DSI-13 draft paper. Based on the

public comments, the Commission has directed the NRC Staff to address

how industry initiatives should be evaluated, and to evaluate several

issues related to NRC endorsement of industry codes and standards. As

part of this evaluation, the Staff is addressing issues relevant to the

NRC's endorsement of the ASME Code, including periodic updating, the

impact of 10 CFR 50.109 (the Backfit Rule), and streamlining the

process for NRC review and endorsement of the ASME Code.

2.8 Steam Generators

ASME Code requirements for repair of heat exchanger tubes by

sleeving were added to Section XI in the 1989 Addenda. Minimum Code

requirements for tube sleeving was added to the Code so that licensees

would not have to develop sleeving programs and have them approved by

the NRC on a case-by-case basis. The NRC has reviewed the Code

requirements for sleeving and determined that they are acceptable.

However, it should be recognized that there are other relevant

requirements, and that a considerable amount of effort is presently

being expended due to the number of occurrences of degraded steam

generator tubing. For example, licensees are required by either 10 CFR

50.55a(f) or by the plant technical specifications to perform periodic

inservice inspections and to repair (e.g., sleeving) or remove from

service (by installing plugs in the tube ends) all tubes found to

contain flaws exceeding the plugging limit (i.e., tube repair

criteria). In addition, current technical specifications contain

operational leakage limits. Licensee's have frequently found it

necessary to implement measures beyond minimum Code and technical

specification requirements to ensure adequate tube integrity when

significant degradation problems are encountered. Thus, the NRC

determination that the sleeving requirements are acceptable should be

kept in perspective.

3. Finding of No Significant Environmental Impact

Based upon an environmental assessment, the Commission has

determined, under the National Environmental Policy Act of 1969, as

amended, and the Commission's regulations in Subpart A of 10 CFR Part

51, that this rule, if adopted, would not have a significant effect on

the quality of the human environment and therefore an environmental

impact statement is not required.

The proposed rule is one part of a regulatory framework directed to

ensuring pressure boundary integrity and the operational readiness of

pumps and valves. The proposed rule incorporates provisions contained

in the BPV Code and the OM Code for the construction, inservice

inspection, and inservice testing of components used in nuclear power

plants, has been updated to incorporate improved technology and

methodology. Therefore, in the general sense, the proposed rule would

have a positive impact on the environment.

The proposed rule would impose the Section XI 1995 Edition with the

1996 Addenda. As most of the technical changes to this edition/addenda

merely incorporate improved technology and methodology, imposition of

these requirements is not expected to either increase or decrease

occupational exposure. However, imposition of paragraphs IWF-2510,

Table IWF-2500-1, Examination Category F-A, and IWF-2430, would result

in fewer supports being examined which would decrease the occupational

exposure compared to present support inspection plans. It is estimated

that an examiner receives approximately 100 millirems for every 25

supports examined. Adoption of the new provisions is expected to

decrease the total number of supports to be examined by approximately

115 per unit per interval. Thus, the reduction in occupational exposure

is estimated to be 460 millirems per unit each inspection interval or

50.14 rems for 109 units.

The proposed rule would impose Appendix VIII to Section XI, 1995

Edition with the 1996 Addenda, BPV Code, for the first time and would

expedite its implementation. Appendix VIII provides rules for the

performance demonstration of ultrasonic examination systems,

procedures, and personnel. Implementation of this appendix should

result in a decrease in occupational exposure. Appendix VIII qualified

procedures and personnel should reduce repeat ultrasonic testing (UT),

which could reduce occupational exposure. In addition, flaws should be

detected at an earlier stage of growth resulting in less extensive

repair operations, which could further reduce occupational exposure.

The proposed rule would incorporate by reference into the

regulations the 1995 Edition with the 1996 Addenda of the OM Code.

Imposition of the OM Code is not expected to either increase or

decrease occupational exposure. The types of testing associated with

the 1995 Edition with the 1996 Addenda of the OM Code are essentially

the same as the OM standards contained in the 1989 Edition of Section

XI referenced in a final rule published on August 6, 1992 (57 FR

34666).

Actions required of applicants and licensees to implement the

proposed rule are of the same nature as those applicants and licensees

have been performing for many years. Therefore, this action should not

increase the

[[Page 63905]]

potential for a negative environmental impact.

The NRC has sent a copy of the Environmental Assessment and the

proposed rule to every State Liaison Officer and requested their

comments on the Environmental Assessment. The environmental assessment

is available for inspection at the NRC Public Document Room, 2120 L

Street NW (Lower Level), Washington, DC. Single copies of the

environmental assessment are available from Frank C. Cherny, Division

of Engineering Technology, Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:

301-415-6786, or Wallace E. Norris, Division of Engineering Technology,

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

Telephone: 301-415-6796.

4. Paperwork Reduction Act Statement

This proposed rule amends information collection requirements that

are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et

seq.). This rule has been submitted to the Office of Management and

Budget for review and approval of the paperwork requirements.

The public reporting burden for this information collection is

estimated to average 67 person-hours per response, including the time

for reviewing instructions, searching existing data sources, gathering

and maintaining the data needed, and completing and reviewing the

collection of information. The U.S. Nuclear Regulatory Commission is

seeking public comment on the potential impact of the information

collections contained in the proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper

performance of the functions of the NRC, including whether the

information will have practical utility?

2. Is the estimate of burden accurate?

3. Is there a way to enhance the quality, utility, and clarity of

the information to be collected?

4. How can the burden of the information collection be minimized,

including the use of automated collection techniques?

Send comments on any aspect of this proposed collection of

information, including suggestions for further reducing the burden, to

the Information and Records Management Branch (T-6 F33), U.S. Nuclear

Regulatory Commission, Washington DC 20555-0001, or by Internet

electronic mail at [email protected]; and to the Desk Officer, Office of

Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of

Management and Budget, Washington DC 20503.

Comments to OMB on the information collections or on the above

issues should be submitted by January 2, 1998. Comments received after

this date will be considered if it is practical to do so, but assurance

of consideration cannot be given to comments received after this date.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to

respond to, a collection of information unless it displays a currently

valid OMB control number.

5. Regulatory Analysis

The Commission has prepared a draft regulatory analysis on this

proposed regulation. The analysis examines the costs and benefits of

the alternatives considered by the Commission. The draft analysis is

available for inspection in the NRC Public Document Room, 2120 L Street

NW (Lower Level), Washington DC. The Commission requests public comment

on the draft analysis. Single copies of the analysis may be obtained

from Frank C. Cherny, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, Telephone: 301-415-6786, Wallace E. Norris,

Division of Engineering Technology, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, Telephone: 301-415-6796.

6. Regulatory Flexibility Certification

In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.

605(b), the Commission certifies that this rule will not, if

promulgated, have a significant economic impact on a substantial number

of small entities. This proposed rule affects only the licensing and

operation of nuclear power plants. The companies that own these plants

do not fall within the scope of the definition of ``small entities''

set forth in the Regulatory Flexibility Act or the Small Business Size

Standards set out in regulations issued by the Small Business

Administration at 13 CFR Part 121.

7. Backfit Analysis

The Nuclear Regulatory Commission (NRC) regulations, 10 CFR 50.55a,

requires that nuclear power plant owners (1) construct Class 1, Class

2, and Class 3 components in accordance with the rules provided in

Section III, Division 1, ``Requirements for Construction of Nuclear

Power Plant Components,'' of the American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), (2)

inspect Class 1, Class 2, Class 3, Class MC (metal containment) and

Class CC (concrete containment) components in accordance with the rules

provided in Section XI, Division 1, ``Requirements for Inservice

Inspection of Nuclear Power Plant Components,'' of the BPV Code, and

(3) test Class 1, Class 2, and Class 3 pumps and valves in accordance

with the rules provided in Section XI, Division 1. Licensees are

required to update every 120 months to the version of Section XI

incorporated by reference into Sec. 50.55a 12 months prior to the start

of a new ten year interval.

The proposed amendment to Sec. 50.55a would require licensees to

update ISI in accordance with Section XI of the ASME BPV Code and IST

in accordance with the ASME OM Code. Licensees would be required to

implement the 1995 Edition with the 1996 Addenda of (1) Section XI,

Division 1 for Class 1, Class 2, Class 3, Class MC, and Class CC

components; (2) the ``Code for Operation and Maintenance of Nuclear

Power Plants'' (OM Code) for Class 1, Class 2, and Class 3 pumps and

valves; and (3) Appendix VIII, ``Performance Demonstration for

Ultrasonic Examination Systems,'' to Section XI, Division 1. As

permitted by Sec. 50.55a(a)(3), licensees may voluntarily update to the

1989 Addenda through the 1996 Addenda of Section III of the BPV Code,

with limitation. In addition, the modification for containment

isolation valve inservice testing that applied to the 1989 Edition of

the BPV Code has been deleted. Licensees will continue to be required

to update their ISI and IST programs every 120 months to the version of

Section XI and the OM Code incorporated by reference and in effect at

least 12 months prior to the start of a new 120-month interval.

The NRC position on the routine 120-month update to Sec. 50.55a has

consistently been that 10 CFR 50.109 does not require a backfit

analysis of the routine 120-month update to Sec. 50.55a. The basis for

the NRC position is that, (1) Section III, Division 1, update applies

only to new construction (i.e., the edition and addenda to be used in

the construction of a plant are selected based upon the date of the

construction permit and are not changed thereafter, except voluntarily

by the licensee), (2) licensees understand that Sec. 50.55a requires

that they update their inservice inspection program every 10 years to

the latest edition and addenda of Section XI that were incorporated by

reference in Sec. 50.55a and in effect 12 months before

[[Page 63906]]

the start of the next inspection interval, and (3) endorsing and

updating references to the ASME Code, a national consensus standard

developed by the participants (including the NRC) with broad and varied

interests, is consistent with both the intent and spirit of the backfit

rule (i.e., NRC provides for the protection of the public health and

safety, and does not unilaterally impose undue burden on applicants or

licensees). Finally, to ensure that any interested member of the public

that may not have had an opportunity to participate in the national

consensus standard process is able to communicate with the NRC,

proposed rules are published in the Federal Register.

The provisions for IST of pumps and valves were originally

contained in Section XI Subsections IWP and IWV. Section XI, 1989

Edition was incorporated by reference in the August 6, 1992 rulemaking

(57 FR 34666). The 1990 OM Code standards, Parts 1, 6, and 10 of ASME/

ANSI-OM-1987, are identical to Section XI, 1989 Edition. This proposed

amendment is an administrative change simply referencing the 1995

Edition with the 1996 Addenda of the OM Code. Therefore, imposition of

the 1995 Edition with the 1996 Addenda of the OM Code is not a backfit.

Appendix VIII, ``Performance Demonstration for Ultrasonic

Examination Systems,'' to Section XI would be used to demonstrate the

qualification of personnel and procedures for performing nondestructive

examination of welds in components of systems that include the reactor

coolant system and the emergency core cooling systems in nuclear power

facilities. Appendix VIII would greatly enhance the reliability of

detection and sizing of cracks and flaws, and it delineates a method

for qualification of the personnel and procedures. The appendix would

normally be imposed by the 120-month update requirement, but because of

its importance, implementation of Appendix VIII is being expedited by

the rulemaking. Because of the expedited implementation schedule, the

imposition of Appendix VIII is being considered a backfit. Licensees

would be required to implement Appendix VIII, including the

modifications, for all examinations of the pressure vessel, piping,

nozzles, and bolts and studs which occur after 6 months from the date

of the final rule. The proposed rule would not require any change to a

licensee's ISI schedule for examination of these components, but would

require that the provisions of Appendix VIII be used for all

examinations after that date rather than the UT procedures and

personnel requirements presently being utilized by licensees.

The NRC has concluded, on the basis of the documented evaluation

required by Sec. 50.109(a)(4), that imposition of Appendix VIII, which

would greatly enhance the overall level of assurance of the safety and

reliability of ultrasonic examination techniques in detecting and

sizing flaws, is necessary to bring the facilities described into

compliance with GDC 14, 10 CFR Part 50, Appendix A, or similar

provisions in the licensing basis for these facilities, and Criteria II

and XVI, of 10 CFR Part 50, Appendix B.

The modification to Section XI to require licensees to supplement

the surface examination of the Class 1 portion (RCPB) of the HPSI

system with volumetric examination would ensure the integrity of the

reactor coolant system pressure boundary and maintenance of emergency

core cooling system operability. The operability of this system is

necessary to ensure the protection of the public health and safety, and

the NRC has concluded, on the basis of the documented evaluation

required by Sec. 50.109(a)(4), that licensees must supplement the

Section XI required surface examination for the Class 1 portion of the

HPSI system with volumetric examination in order to ensure the

integrity of the reactor coolant pressure boundary as required by GDC

14, 10 CFR Part 50, Appendix A, or similar provisions in the licensing

basis for these facilities, and Criteria II and XVI, of 10 CFR Part 50,

Appendix B. Volumetric examination would be required during any ISI

program inspection of the HPSI system performed after 6 months from the

date of the final rule.

GDC 14, ``Reactor coolant pressure boundary,'' (RCPB) or similar

provisions in the licensing basis for these facilities, specify that

the RCPB be designed, fabricated, erected, and tested so as to have an

extremely low probability of abnormal leakage, or rapidly propagating

failure, and of gross rupture. There has recently been an occurrence of

gross rupture in the Class 1 portion of a HPSI system, and a number of

occurrences of abnormal leakage in the RCPB in other plants.

Imposition of Appendix VIII and the HPSI volumetric examination is

also necessary to bring the facilities described into compliance with

Criteria II, ``Quality Assurance Program,'' and Criteria XVI,

``Corrective Actions,'' of Appendix B to 10 CFR Part 50. Criteria II

requires, in part, that a QA program shall take into account the need

for special controls, processes, test equipment, tools, and skills to

attain the required quality and the need for verification of quality by

inspection and test. Evidence indicates that there are shortcomings in

the qualifications of personnel and procedures in ensuring the

reliability of the examinations. These safety significant revisions to

the Code include specific requirements for UT performance

demonstration, with statistically based acceptance criteria for blind

testing of UT systems (procedures, equipment, and personnel) used to

detect and size flaws. Criteria XVI requires that measures shall be

established to assure that conditions adverse to quality, such as

failures, malfunctions, deficiencies, deviations, defective material

and equipment, and nonconformances are promptly identified and

corrected. In analyzing the occurrences of pipe break and leakage, it

is apparent that the RCPB is subject to certain types of degradation.

Information gathered by the NRC staff indicates that many licensees

have not reacted to this serious safety concern by performing more

comprehensive examinations. The NRC believes that there is a basis for

reasonably concluding that such degradation could occur in virtually

all PWRs. Because of the serious degradation which has occurred, and

the belief that additional occurrences of noncompliance with GDC 14,

and Criteria II and XVI will be reported, the NRC has determined that

imposition of Appendix VIII and volumetric examination of the HPSI

system 6 months after the final rule has been published under the

compliance exception to Sec. 50.109(a)(4)(i) is appropriate, therefore,

a backfit analysis is not required and the cost-benefit standards of

Sec. 50.109(a)(3) do not apply. A complete discussion is contained in

the documented evaluation.

The rationale for application of the backfit rule and the backfit

justification for the various items contained in this proposed rule are

contained in the regulatory analysis and documented evaluation. The

regulatory analysis and documented evaluation are available for

inspection at the NRC Public Document Room, 2120 L Street NW (Lower

Level), Washington, DC. Single copies of the regulatory analysis and

documented evaluation are available from Frank C. Cherny, Division of

Engineering Technology, Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:

301-415-6786, or Wallace E. Norris, Division of Engineering Technology,

Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory

Commission,

[[Page 63907]]

Washington, DC 20555-0001, Telephone: 301-415-6796.

List of Subjects in 10 CFR Part 50

Antitrust, Classified information, Fire prevention, Incorporation

by reference, Intergovernmental relations, Nuclear power plants and

reactors, Penalties, Radiation protection, Reactor siting criteria,

Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of

the Atomic Energy Act of 1954, as amended, the Energy Reorganization

Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to

adopt the following amendments to 10 CFR Part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for Part 50 continues to read as follows:

Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68

Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,

83 Stat. 1444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,

2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88

Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.

2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,

185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub.

L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),

and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42

U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued

under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,

50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83

Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued

under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,

50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42

U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939

(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,

68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued

under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

2. Section 50.55a is amended by removing and reserving paragraphs

(b)(2)(vii) and (g)(4)(iv), adding paragraphs (b)(2)(xi) through

(b)(2)(xx), (b)(3), (g)(6)(ii)(A)(6), and (g)(6)(ii)(C), and revising

the introductory text of paragraph (b), paragraph (b)(1), the

introductory text of paragraph (b)(2), paragraphs (b)(2)(iv),

(b)(2)(vi), (b)(2)(viii), the introductory text of paragraph

(b)(2)(ix), paragraphs (c)(3), (d)(2), (e)(2), the introductory text of

paragraph (f), paragraphs (f)(1), (f)(2), (f)(3)(iii), (f)(3)(iv), the

introductory text of paragraph (f)(4), paragraphs (g)(1), (g)(3)(i),

the introductory text of paragraph (g)(4), paragraphs (g)(6)(ii)(A)(1),

(g)(6)(ii)(A)(2), and Footnotes 5 and 7 to read as follows:

Sec. 50.55a Codes and standards.

* * * * *

(b) The ASME Boiler and Pressure Vessel Code, and the ASME Code for

Operation and Maintenance of Nuclear Power Plants, which are referenced

in the following paragraphs, were approved for incorporation by

reference by the Director of the Federal Register. A notice of any

changes made to the material incorporated by reference will be

published in the Federal Register. Copies of the ASME Boiler and

Pressure Vessel Code and the ASME Code for Operation and Maintenance of

Nuclear Power Plants may be purchased from the American Society of

Mechanical Engineers, United Engineering Center, 345 East 47th Street,

New York, NY 10017. They are also available for inspection at the NRC

Library, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland 20852-2738.

(1) As used in this section, references to Section III of the ASME

Boiler and Pressure Vessel Code refer to Section III, Division 1, and

include editions through the 1995 Edition and addenda through the 1996

Addenda, subject to the following limitations and modifications:

(i) Engineering judgement. When a licensee relies on engineering

judgment for activities or evaluations of components or systems within

the scope of 10 CFR 50.55a that are not directly addressed by the ASME

Boiler and Pressure Vessel Code, the NRC must approve the activities or

evaluations pursuant to 10 CFR 50.55a(a)(3).

(ii) Section III Materials. When applying the 1992 Edition of

Section III, licensees shall apply the 1992 Edition with the 1992

Addenda of Section II of the ASME Boiler and Pressure Vessel Code.

(iii) Weld leg dimensions. When applying the 1989 Addenda through

the 1996 Addenda of Section III, licensees shall not apply paragraph

NB-3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-

3673.2(b)-1, and shall continue to use the requirements in the 1989

Edition for this paragraph and figures.

(iv) Seismic design. Licensees may use Articles NB-3200, NB-3600,

NC-3600, and ND-3600 through the 1993 Addenda, subject to the

limitation specified in (b)(1)(iii) of this section. Licensees shall

not use the provisions in the 1994 Addenda through the 1996 Addenda for

these Articles.

(v) Quality assurance. When applying editions and addenda later

than the 1989 Edition of Section III, the requirements of NQA-1,

``Quality Assurance Requirements for Nuclear Facilities,'' 1986 Edition

through the 1992 Addenda are acceptable for use provided that both NQA-

1 and the quality assurance provisions specified in NCA-4000 are used

in conjunction with the administrative, quality, and technical

provisions contained in the edition and addenda of Section III being

utilized.

(vi) Independence of inspection. Licensees shall not apply NCA-

4134.10(a) of Section III, 1995 Edition with the 1996 Addenda, and

shall use NCA-4134.10(a), 1994 Addenda.

(2) As used in this section, references to Section XI of the ASME

Boiler and Pressure Vessel Code refer to Section XI, Division 1, and

include editions through the 1995 Edition and addenda through the 1996

Addenda, subject to the following limitations and modifications:

* * * * *

(iv) Pressure-retaining welds in ASME Code Class 2 piping (applies

to Tables IWC-2520 or IWC-2520-1, Category C-F).

(A) Appropriate Code Class 2 pipe welds in Residual Heat Removal

Systems, Emergency Core Cooling Systems, and Containment Heat Removal

Systems, must be examined. When applying editions and addenda up to the

1983 Edition through the Summer 1983 Addenda of Section XI of the ASME

Code, the extent of examination for these systems must be determined by

the requirements of paragraph IWC-1220, Table IWC-2520 Category C-F and

C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the

Summer 1975 Addenda.

(B) For a nuclear power plant whose application for a construction

permit was docketed prior to July 1, 1978, when applying editions and

addenda up to the 1983 Edition through the Summer 1983 Addenda of

Section XI of the ASME Code, the extent of examination for Code Class 2

pipe welds may be determined by the requirements of paragraph IWC-1220,

Table IWC-2520 Category C-F and C-G and paragraph IWC-2411 in the 1974

Edition and Addenda through the Summer 1975 Addenda of Section XI of

the ASME Code or other requirements the Commission may adopt.

* * * * *

[[Page 63908]]

(vi) Effective edition and addenda of Subsection IWE and Subsection

IWL, Section XI. Licensees shall use either the 1992 Edition with the

1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection

IWE and Subsection IWL as modified and supplemented by the requirements

in Sec. 50.55a(b)(2)(ix) and Sec. 50.55a(b)(2)(x).

(vii) [Reserved]

(viii) Section XI References to OM Part 4, OM Part 6 and OM Part 10

(Table IWA-1600-1). When using Table IWA-1600-1, ``Referenced Standards

and Specifications'' in the Section XI, Division 1, 1987 Addenda, 1988

Addenda, or 1989 Edition, the specified ``Revision Date or Indicator''

for ASME/ANSI OM Part 4, ASME/ANSI Part 6, and ASME/ANSI Part 10 shall

be the OMa-1988 Addenda to the OM-1987 Edition. These requirements have

been incorporated into the 1990 Edition of the OM Code which is

incorporated by reference in paragraph (b)(3) of this section.

(ix) Examination of concrete containments. Licensees applying

Subsection IWL, 1992 Edition with the 1992 Addenda, shall apply all of

the modifications in this paragraph. Licensees choosing to apply the

1995 Edition with the 1996 Addenda shall apply paragraphs

(b)(2)(ix)(A), (D)(3), and (E) of this section.

* * * * *

(xi) Engineering judgment. When a licensee relies on engineering

judgment for activities or evaluations of components or systems within

the scope of 10 CFR 50.55a that are not directly addressed by the ASME

Boiler and Pressure Vessel Code, the NRC must approve the activities or

evaluations pursuant to 10 CFR 50.55a(a)(3).

(xii) Quality Assurance. When applying Section XI editions and

addenda later than the 1989 Edition, the requirements of NQA-1,

``Quality Assurance Requirements for Nuclear Facilities,'' 1979 Addenda

through the 1989 Edition are acceptable as permitted by IWA-1400 of

Section XI, provided the licensee utilizes its 10 CFR Part 50, Appendix

B, quality assurance program, in conjunction with Section XI

requirements. Changes to licensee's quality assurance program shall be

made in accordance with 10 CFR 50.54(a). In addition, where NQA-1 and

Section XI do not address the commitments contained in the licensee's

Appendix B quality assurance program description, such commitments

shall be applied to Section XI activities.

(xiii) Class 1 piping. Licensees shall not apply IWB-1220,

``Components Exempt from Examination,'' of Section XI, 1989 Addenda

through the 1996 Addenda, and shall apply IWB-1220, 1989 Edition.

(xiv) Class 2 piping. Prior to applying the provisions of IWC-1220,

``Components Exempt from Examination,'' IWC-1221, ``Components Within

RHR, ECC, and CHR Systems or Portions of Systems,'' and IWC-1222,

``Components Within Systems or Portions of Systems Other Than RHR, ECC,

and CHR Systems,'' 1989 Addenda through the 1996 Addenda, licensees

shall define the Class 2 piping subject to volumetric and surface

examination, and submit this information for approval by the NRC staff

pursuant to Sec. 50.55a(a)(3) prior to implementation.

(xv) Class 1 piping volumetric examination. When performing weld

examinations of High Pressure Safety Injection Systems, as required by

Table IWB-2500-1, Examination Category B-J, Item Numbers B9.20, B9.21,

and B9.22, all licensees of pressurized water reactor facilities shall

perform volumetric examination of the Class 1 portion of the system

after [insert 6 months from the date of the final rule].

(xvi) Flaws in Class 3 piping moderate energy (200 xF, 275 psig)

piping. Licensees may use the provisions of Code Case N-513,

``Evaluation Criteria for Temporary Acceptance of Flaws in Class 3

Piping,'' Rev 0, and Code Case N-523-1, ``Mechanical Clamping Devices

for Class 2 and 3 Piping.'' Licensees choosing to apply Code Case N-

523-1 shall apply all of its provisions. Licensees choosing to apply

Code Case N-513 shall apply all of its provisions subject to the

following:

(A) When implementing Code Case N-513, the specific safety factors

in paragraph 4.0 must be satisfied.

(B) Code Case N-513 shall not be applied to:

(1) Components other than pipe and tube, such as pumps, valves,

expansion joints, and heat exchangers;

(2) The discovery and repair of flaws or deficiencies remaining

from original construction;

(3) Leakage through a flange gasket;

(4) Threaded connections employing nonstructural seal welds for

leakage prevention (through seal weld leakage is not a structural flaw,

thread integrity must be maintained); and

(5) Degraded socket welds.

(xvii) Appendix VIII personnel qualification. All personnel

qualified for performing ultrasonic examinations in accordance with

Appendix VIII shall receive 40 hours of annual training that includes

laboratory work and examination of flawed specimens.

(xviii) Appendix VIII specimen set cracks. All flaws in the

specimen sets used for performance demonstration for piping, vessels,

and nozzles shall be cracks.

(xix) Appendix VIII specimen set microstructure. All specimens for

single-side tests shall contain microstructures of the type found in

components to be inspected, and flaws with non-optimum characteristics

consistent with field experience that provide realistic challenges to

the UT techniques.

(xx) Reconciliation of Quality Requirements. The following

limitations apply when implementing Section XI, IWA-4200, 1995 Addenda

through the 1996 Addenda:

(A) Licensees shall not apply IWA-4200, of Section XI, 1995 Addenda

through the 1996 Addenda, for reconciliation of the administrative

requirements for replacement items, and shall reconcile the

administrative requirements with the original Construction Code and the

Owner's requirements as required by the 1995 Edition.

(B) Licensees shall not apply the definition of Construction Code

in IWA-9000, ``Glossary,'' 1993 Addenda through the 1996 Addenda, and

shall apply the definition of Construction Code in IWA-9000, 1992

Edition.

(3) As used in this section, references to the OM Code refer to the

ASME Code for Operation and Maintenance of Nuclear Power Plants, and

include addenda through the 1996 Addenda and editions through the 1995

Edition subject to the following limitations and modifications:

(i) Quality Assurance. When applying editions and addenda of the OM

Code, 1990 and later, the requirements of NQA-1, ``Quality Assurance

Requirements for Nuclear Facilities,'' 1979 Addenda, are acceptable as

permitted by ISTA 1.4 of the OM Code, provided the licensee utilizes

its 10 CFR Part 50, Appendix B, quality assurance program, in

conjunction with the OM Code requirements. Changes to licensee's

quality assurance program shall be made in accordance with 10 CFR

50.54(a). In addition, where NQA-1 and the OM Code do not address the

commitments contained in the licensee's Appendix B quality assurance

program description, such commitments shall be applied to OM Code

activities.

(ii) Stroke time testing. Licensees shall comply with the

provisions on stroke time testing in OM Code ISTC 4.2, 1995 Edition

with the 1996 Addenda, and the programs developed under their licensing

commitments for demonstrating design basis capability of motor-operated

valves.

[[Page 63909]]

(iii) Code Case OMN-1. As an alternative to Sec. 50.55a(b)(3)(ii),

licensees may use Code Case OMN-1, ``Alternative Rules for Preservice

and Inservice Testing of Certain Electric Operated Valve Assemblies in

LWR Power Plants,'' Rev. 0, 1995 Edition with the 1996 Addenda, in

conjunction with ISTC 4.3, 1995 Edition with the 1996 Addenda.

Licensees choosing to apply the Code case shall apply all of its

provisions.

(A) The adequacy of the test interval for each valve shall be

evaluated and adjusted as necessary but not later than five years or

three refueling outages (whichever is longer) from initial

implementation of ASME Code Case OMN-1.

(B) [Reserved]

(iv) Appendix II. The following modifications apply when

implementing Appendix II, ``Check Valve Condition Monitoring Program,''

of the OM Code, 1995 Edition with the 1996 Addenda:

(A) Valve opening and closing functions must be demonstrated when

flow testing or examination methods (nonintrusive, or disassembly and

inspection) are used;

(B) The initial interval for tests and associated examinations

shall not exceed two fuel cycles or 3 years, whichever is longer; any

extension of this interval shall not exceed one fuel cycle per

extension with the maximum interval not to exceed 10 years; trending

and evaluation of existing data shall be used to reduce or extend time

the interval between tests.

(C) If the Appendix II condition monitoring program is

discontinued, then the requirements of ISTC 4.5.1 through 4.5.4 shall

be implemented.

(v) Subsection ISTD. Licensees may use Subsection ISTD, OM Code,

1995 Edition with the 1996 Addenda, by making a change to their

technical specifications in accordance with applicable NRC

requirements. Licensees choosing to apply the subsection shall apply

all of its provisions.

(c) * * *

(3) The Code Edition, Addenda, and optional Code Cases to be

applied to components of the reactor coolant pressure boundary must be

determined by the provisions of paragraph NCA-1140, Subsection NCA of

Section III of the ASME Boiler and Pressure Vessel Code, but:

(i) The edition and addenda applied to a component must be those

which are incorporated by reference in paragraph (b)(1) of this

section, and, in case of conflict between paragraph (b)(1) of this

section and paragraph NCA-1140, the latest edition and addenda

incorporated by reference in paragraph (b)(1) of this section shall be

applied,

(ii) The ASME Code provisions applied to the pressure vessel may be

dated no earlier than the Summer 1972 Addenda of the 1971 edition,

(iii) The ASME Code provisions applied to piping, pumps, and valves

may be dated no earlier than the Winter 1972 Addenda of the 1971

edition, and

* * * * *

(d) * * *

(2) The Code Edition, Addenda, and optional Code Cases6 to be

applied to the systems and components identified in paragraph (d)(1) of

this section must be determined by the rules of paragraph NCA-1140,

Subsection NCA of Section III of the ASME Boiler Vessel and Pressure

Code, but:

(i) The edition and addenda must be those which are incorporated by

reference in paragraph (b)(1) of this section, and, in case of conflict

between paragraph (b)(1) of this section and paragraph NCA-1140, the

latest edition and addenda incorporated by reference in paragraph

(b)(1) of this section shall be applied,

(ii) The ASME Code provisions applied to the systems and components

may be dated no earlier than the 1980 Edition, and

(iii) The ASME Code Cases6 must have been determined suitable for

use by the NRC.

(e) * * *

(2) The Code Edition, Addenda, and optional Code Cases6 to be

applied to the systems and components identified in paragraph (e)(1) of

this section must be determined by the rules of paragraph NCA-1140,

Subsection NCA of Section III of the ASME Boiler and Pressure Vessel

Code, but:

(i) The edition and addenda must be those which are incorporated by

reference in paragraph (b)(1) of this section, and, in case of conflict

between paragraph (b)(1) of this section and paragraph NCA-1140, the

latest edition and addenda incorporated by reference in paragraph

(b)(1) of this section shall be applied,

(ii) The ASME Code provisions applied to the systems and components

may be dated no earlier than the 1980 Edition, and

(iii) The ASME Code Cases must have been determined suitable for

use by the NRC.

(f) Inservice testing requirements. Requirements for inservice

inspection of Class 1, Class 2, Class 3, Class MC, and Class CC

components (including their supports) are located in Sec. 50.55a(g).

(1) For a boiling or pressurized water-cooled nuclear power

facility whose construction permit was issued prior to January 1, 1971,

pumps and valves must meet the test requirements of paragraphs (f)(4)

and (f)(5) of this section to the extent practical. Pumps and valves

which are part of the reactor coolant pressure boundary must meet the

requirements applicable to components which are classified as ASME Code

Class 1. Other pumps and valves in steam, water, air, and liquid-

radioactive-waste systems that perform a function to shut down the

reactor or maintain the reactor in a safe shutdown condition, mitigate

the consequences of an accident, or provide overpressure protection for

such systems (in meeting the requirements of the 1986 Edition, or

later, of the Boiler and Pressure Vessel or OM Code), must meet the

test requirements applicable to components which are classified as ASME

Code Class 2 or Class 3.

(2) For a boiling or pressurized water-cooled nuclear power

facility whose construction permit was issued on or after January 1,

1974, pumps and valves which are classified as ASME Code Class 1 and

Class 2 must be designed and be provided with access to enable the

performance of inservice tests for operational readiness set forth in

editions of Section XI of the ASME Boiler and Pressure Vessel Code and

Addenda6 in effect 6 months prior to the date of issuance of the

construction permit. The pumps and valves may meet the requirements set

forth in subsequent editions of this code and addenda which are

incorporated by reference in paragraph (b) of this section, subject to

limitations and modifications listed therein.

(3) * * *

(iii)(A) Pumps and valves, in facilities whose construction permit

was issued before [insert effective date of the final rule], which are

classified as ASME Code Class 1 must be designed and be provided with

access to enable the performance of inservice testing of the pumps and

valves for assessing operational readiness set forth in Section XI of

editions of the ASME Boiler and Pressure Vessel Code and Addenda6

applied to the construction of the particular pump or valve or the

Summer 1973 Addenda, whichever is later.

(B) Pumps and valves, in facilities whose construction permit is

issued on or after [insert effective date of the final rule], which are

classified as ASME Code Class 1 must be designed and be provided with

access to enable the performance of inservice testing of the pumps and

valves for assessing

[[Page 63910]]

operational readiness set forth in editions and addenda of the ASME OM

Code referenced in paragraph (b)(3) of this section at the time the

construction permit is issued.

(iv)(A) Pumps and valves, in facilities whose construction permit

was issued before [insert effective date of rule], which are classified

as ASME Code Class 2 and Class 3 must be designed and be provided with

access to enable the performance of inservice testing of the pumps and

valves for assessing operational readiness set forth in Section XI of

editions of the ASME Boiler and Pressure Vessel Code and Addenda6

applied to the construction of the particular pump or valve or the

Summer 1973 Addenda, whichever is later.

(B) Pumps and valves, in facilities whose construction permit is

issued on or after [insert effective date of the final rule], which are

classified as ASME Code Class 2 and 3 must be designed and be provided

with access to enable the performance of inservice testing of the pumps

and valves for assessing operational readiness set forth in editions

and addenda of the ASME OM Code referenced in paragraph (b)(3) of this

section at the time the construction permit is issued.

* * * * *

(4) Throughout the service life of a boiling or pressurized water-

cooled nuclear power facility, pumps and valves which are classified as

ASME Code Class 1, Class 2 and Class 3 must meet the inservice test

requirements, except design and access provisions, set forth in the

ASME OM Code and addenda that become effective subsequent to editions

and addenda specified in paragraphs (f)(2) and (f)(3) of this section

and that are incorporated by reference in paragraph (b) of this

section, to the extent practical within the limitations of design,

geometry and materials of construction of the components.

* * * * *

(g) * * *

(1) For a boiling or pressurized water-cooled nuclear power

facility whose construction permit was issued before January 1, 1971,

components (including supports) must meet the requirements of

paragraphs (g)(4) and (g)(5) of this section to the extent practical.

Components which are part of the reactor coolant pressure boundary and

their supports must meet the requirements applicable to components

which are classified as ASME Code Class 1. Other pressure vessels,

piping, pumps and valves, and their supports in steam, water, air, and

liquid-radioactive-waste systems that provide pressure boundary

integrity for systems that perform a function to shut down the reactor

or maintain the reactor in a safe shutdown condition, or mitigate the

consequences of an accident, must meet the requirements applicable to

components which are classified as ASME Code Class 2 or Class 3.

* * * * *

(3) * * *

(i) Components (including supports) which are classified as ASME

Code Class 1 must be designed and be provided with access to enable the

performance of inservice examination of such components and must meet

the preservice examination requirements set forth in Section XI of

editions of the ASME Boiler and Pressure Vessel Code and Addenda6

applied to the construction of the particular component.

* * * * *

(4) Throughout the service life of a boiling or pressurized water-

cooled nuclear power facility, components (including supports) which

are classified as ASME Code Class 1, Class 2 and Class 3 must meet the

requirements, except design and access provisions and preservice

examination requirements, set forth in Section Xl of editions of the

ASME Boiler and Pressure Vessel Code and Addenda that become effective

subsequent to editions specified in paragraphs (g)(2) and (g)(3) of

this section and that are incorporated by reference in paragraph (b) of

this section, to the extent practical within the limitations of design,

geometry and materials of construction of the components. Components

which are classified as Class MC pressure retaining components and

their integral attachments, and components which are classified as

Class CC pressure retaining components and their integral attachments

must meet the requirements, except design and access provisions and

preservice examination requirements, set forth in Section XI of the

ASME Boiler and Pressure Vessel Code and Addenda that are incorporated

by reference in paragraph (b) of this section, subject to the

limitation listed in paragraph (b)(2)(vi) and the modifications listed

paragraph (b)(2)(ix) and (b)(2)(x) of this section, to the extent

practical within the limitation of design, geometry and materials of

construction of the components.

* * * * *

(iv) [Reserved]

(6) * * *

(ii) * * *

(A)(1) All previously granted reliefs under Sec. 50.55a to

licensees for the extent of volumetric examination of reactor vessel

shell welds specified in Item BI.10 of Examination Category B-A,

``Pressure Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1 of

Subsection IWB in applicable edition and addenda of Section XI,

Division 1, of the ASME Boiler and Pressure Vessel Code, during the

inservice inspection interval in effect on September 8, 1992 are hereby

revoked, subject to the specific modification in

Sec. 50.55a(g)(6)(ii)(A)(3)(iv) for licensees that defer the augmented

examination in accordance with Sec. 50.55a(g)(6)(ii)(A)(3).

(2) All licensees shall augment their reactor vessel examination by

implementing once, as part of the inservice inspection interval in

effect on September 8, 1992, the examination requirements for reactor

vessel shell welds specified in Item 81.10 of Examination Category B-A,

``Pressure Retaining Welds in Reactor Vessel,'' in Table IWB-2500-1 of

Subsection IWB of the 1989 Edition of Section XI, Division 1, of the

ASME Boiler and Pressure Vessel Code, subject to the conditions

specified in Sec. 50.55a(g)(6)(ii)(A)(3) and (4). The augmented

examination, when not deferred in accordance with the provisions of

Sec. 50.55a(g)(6)(ii)(A)(3), shall be performed in accordance with the

related procedures specified in the Section XI edition and addenda

applicable to the inservice inspection interval in effect on September

8, 1992, and may be used as a substitute for the reactor vessel shell

weld examination scheduled for implementation during the inservice

inspection interval in effect on September 8, 1992. For the purpose of

this augmented examination, ``essentially 100%'' as used in Table IWB-

2500-1 means more than 90 percent of the examination volume of each

weld, where the reduction in coverage is due to interference by another

component, or part geometry.

* * * * *

(6) Augmented examinations of reactor vessel shell welds that are

performed in accordance with Sec. 50. 55a(g)(6)(ii)(A) after [insert 6

months from the date of the final rule] must be performed in accordance

with Sec. 50.55a(g)(6)(ii)(C).

* * * * *

(C) Application of Appendix VIII to Section Xl Examinations.

(1) All reactor vessel (including nozzles) ultrasonic examinations,

all piping ultrasonic examinations, and all bolting ultrasonic

examinations performed after insert 6 months from the date of the final

rule must be

[[Page 63911]]

performed in accordance with Appendix VIII of Section Xl, Division 1,

1995, Edition with the 1996 Addenda of the ASME Boiler and Pressure

Vessel Code.

(2) [Reserved]

* * * * *

\5\ For ASME Code Editions and Addenda issued prior to the

Winter 1977 Addenda, the Code Edition and Addenda applicable to the

component is governed by the order or contract date for the

component, not the contract date for the nuclear energy system. For

the Winter 1977 addenda and subsequent editions and addenda the

method for determining the applicable Code editions and addenda is

contained in Paragraph NCA-1140 of Section III of the ASME Code.

* * * * *

\7\ For purposes of this regulation the proposed IEEE-279 became

``in effect'' on August 30, 1968, and the revised issue IEEE-279-

1971 became ``in effect'' on June 3, 1971. Copies may be obtained

from the Institute of Electrical and Electronics Engineers, United

Engineering Center, 345 East 47th St., New York, NY 10017. Copies

are available for inspection at the NRC Library, Two White Flint

North, 11545, Rockville Pike, Rockville, Maryland 20852-2738.

* * * * *

Dated at Rockville, MD this 27th day of October 1997.

For the Nuclear Regulatory Commission.

L. Joseph Callan,

Executive Director for Operations.

[FR Doc. 97-31588 Filed 12-2-97; 8:45 am]

BILLING CODE 7590-01-P