This section of the FEDERAL REGISTER

contains notices to the public of the proposed

issuance of rules and regulations. The

purpose of these notices is to give interested

persons an opportunity to participate in the

rule making prior to the adoption of the final

rules.

Proposed Rules Federal Register

63892

Vol. 62, No. 232

Wednesday, December 3, 1997

NUCLEAR REGULATORY

COMMISSION

10 CFR Part 50

RIN 3150–AE26

Industry Codes and Standards;

Amended Requirements

AGENCY: Nuclear Regulatory

Commission.

ACTION: Proposed rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) regulations require

that nuclear power plant owners

construct Class 1, Class 2, and Class 3

components in accordance with the

rules provided in Section III, Division 1,

‘‘Requirements for Construction of

Nuclear Power Plant Components,’’ of

the American Society of Mechanical

Engineers (ASME) Boiler and Pressure

Vessel Code (BPV Code), inspect Class

1, Class 2, Class 3, Class MC (metal

containment) and Class CC (concrete

containment) components in accordance

with the rules provided in Section XI,

Division 1, ‘‘Requirements for Inservice

Inspection of Nuclear Power Plant

Components,’’ of the ASME BPV Code,

and test Class 1, Class 2, and Class 3

pumps and valves in accordance with

the rules provided in Section XI,

Division 1, of the ASME BPV Code.

The NRC proposes to amend 10 CFR

50.55a to revise the requirements for

construction, inservice inspection (ISI),

and inservice testing (IST) of nuclear

power plant components. For

construction, the proposed rule would

permit the use of Section III, Division 1,

of the ASME BPV Code, 1989 Addenda

through the 1996 Addenda, for Class 1,

Class 2, and Class 3 components with

six proposed limitations and a

modification.

For ISI, the proposed amendment

would require licensees to implement

Section XI, Division 1, of the ASME

BPV Code, 1995 Edition with the 1996

Addenda, for Class 1, Class 2, and Class

3 components with five proposed

limitations. Licensees would be

permitted to implement: Code Case N–

513 which addresses flaws in low and

moderate energy Class 3 piping; Code

Case N–523 which addresses the

temporary use of mechanical clamps in

Class 2 and 3 piping; and Subsection

IWE and Subsection IWL, 1995 Edition

with the 1996 Addenda.

The proposed rule would expedite

implementation of Appendix VIII,

‘‘Performance Demonstration for

Ultrasonic Examination Systems,’’ to

Section XI, Division 1, with three

proposed modifications. An expedited

implementation schedule would also be

required for a proposed modification to

Section XI which addresses volumetric

examination of the Class 1 high pressure

safety injection (HPSI) system in

pressurized water reactors (PWRs).

For IST, the proposed amendment

would require licensees to implement

the 1995 Edition with the 1996

Addenda of the ASME Code for

Operation and Maintenance of Nuclear

Power Plants (OM Code) for Class 1,

Class 2, and Class 3 pumps and valves

with one limitation and one

modification. 10 CFR 50.55a has been

clarified with respect to which pumps

and valves are to be included in a

licensee’s IST program. Licensees would

be permitted to implement: Code Case

OMN–1 with one modification in lieu of

stroke time testing; Appendix II (which

is an alternative to the check valve

condition monitoring program

provisions contained in Subsection

ISTC of the OM Code) with three

proposed modifications; and Subsection

ISTD for the IST of snubbers. Finally,

based upon supporting information

received since the last rulemaking, the

modification presently in § 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; email CAG@nrc.gov.

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

§ 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
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules 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 § 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 § 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 § 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 § 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
63894 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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 LightWater 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

§ 50.55a

The revisions to § 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 § 50.55a(g)(4)(i) and

§ 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

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

§ 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

§ 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 § 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 § 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

[§ 50.55a(b)(2)(xx)(A)] and the definition

of Construction Code

[§ 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,

§ 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
63896 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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 safetyrelated 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, § 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 § 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, LowPressure 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),

§ 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

§ 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, § 50.55a(b)(3), would

specify the editions and addenda of the

OM Code that have been incorporated

by reference into § 50.55a. Paragraph

50.55a(b)(3) together with § 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 § 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 § 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 § 50.55a.

Therefore, this proposed rule only

requires implementation of Subsections

ISTA, ISTB, and ISTC. However,

§ 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.
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2.3.2.3 Clarification of Safety-Related

Valves

The existing § 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 § 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, § 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 § 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 § 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

§ 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 § 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 § 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

§ 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 ondemand 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 MotorOperated 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 § 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

§ 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 § 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
63898 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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 § 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

§ 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 § 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

§ 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

§ 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 nonoptimum 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

§ 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-indepth 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
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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 plantspecific 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

§ 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, § 50.55a would

be modified to ensure consistency

between § 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 § 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 § 50.55a that are not

directly addressed by the BPV Code, the

licensee must receive NRC approval for

those activities or evaluations pursuant

to § 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 § 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
63900 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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

§ 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

§ 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,

§ 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

§ 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 § 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 § 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

§ 50.55a. In order to resolve this

possible discrepancy, the NRC proposes

to modify existing §§§ 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 § 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
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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 LightWater 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

§ 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 § 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 § 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 § 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, § 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 § 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

§ 50.55a(b)(2)(ix) would thus be

modified. According to

§ 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 posttensioning systems. The current

wording of § 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

§ 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 posttensioning 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
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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 § 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 § 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

§ 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 § 50.55a(b)(3)(ii) is

included in proposed § 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 motoroperated 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 § 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
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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 § 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 § 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 § 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 § 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

§ 50.55a(b)(3)(v).

2.5.3.4 Containment Isolation Valves

The proposed amendment would

delete the existing modification in

§ 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

§ 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
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incorporation by reference into § 50.55a

and have not received NRC approval.

2.7 DSI–13

Since 1992, when the Commission

last revised § 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 § 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 § 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 caseby-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
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules 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

BJS1@NRC.Gov; 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 § 50.55a 12 months

prior to the start of a new ten year

interval.

The proposed amendment to § 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

§ 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 § 50.55a has

consistently been that 10 CFR 50.109

does not require a backfit analysis of the

routine 120-month update to § 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 § 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

§ 50.55a and in effect 12 months before
63906 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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 § 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 § 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 § 50.109(a)(4)(i) is

appropriate, therefore, a backfit analysis

is not required and the cost-benefit

standards of § 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,
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules 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:

§ 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.

* * * * *
63908 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

(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

§ 50.55a(b)(2)(ix) and § 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

§ 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.
Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules 63909

(iii) Code Case OMN–1. As an

alternative to § 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 § 50.55a(g).

(1) For a boiling or pressurized watercooled 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-radioactivewaste 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 watercooled 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
63910 Federal Register / Vol. 62, No. 232 / Wednesday, December 3, 1997 / Proposed Rules

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 watercooled 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-radioactivewaste 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 § 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

§ 50.55a(g)(6)(ii)(A)(3)(iv) for licensees

that defer the augmented examination in

accordance with § 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 § 50.55a(g)(6)(ii)(A)(3) and

(4). The augmented examination, when

not deferred in accordance with the

provisions of § 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 § 50.

55a(g)(6)(ii)(A) after [insert 6 months

from the date of the final rule] must be

performed in accordance with

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

* * * * * 5For 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.

* * * * * 7For 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

NUCLEAR REGULATORY

COMMISSION

10 CFR Parts 50 and 70

RIN 3150–AF87

Criticality Accident Requirements

AGENCY: Nuclear Regulatory

Commission.

ACTION: Proposed rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is amending its

regulations to provide light-water

nuclear power reactor licensees with

greater flexibility in meeting the

requirement that licensees authorized to

possess more than a small amount of

special nuclear material (SNM) maintain

a criticality monitoring system in each

area where the material is handled,

used, or stored. This action is taken as

a result of the experience gained in

processing and evaluating a number of

exemption requests from power reactor

licensees and NRC’s safety assessments

in response to these requests that

concluded that the likelihood of

criticality was negligible.

DATES: Comments on the proposed rule

must be received on or before January 2,

1998.

ADDRESSES: Mail comments to:

Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and

Adjudication Staff. Hand deliver

comments to 11555 Rockville Pike,

Maryland, between 7:45 am and 4:15

pm on Federal workdays.

Copies of any comments received may

be examined at the NRC Public

Document Room, 2120 L Street NW.

(Lower Level), Washington, DC.

For information on submitting

comments electronically, see the

discussion under Electronic Access in

the Supplementary Information section.

FOR FURTHER INFORMATION CONTACT: Stan

Turel, Office of Nuclear Regulatory

Research, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, telephone (301) 415–6234, e-mail

spt@nrc.gov.

SUPPLEMENTARY INFORMATION: For

additional information see the Direct

Final Rule published in the rules

section of this Federal Register.

Procedural Background

Because NRC considers this action

noncontroversial and routine, we are

publishing this proposed rule

concurrently as a direct final rule. The

direct final rule will become effective on

February 17, 1998. However, if the NRC

receives significant adverse comments

on the direct final rule by January 2,

1998, then the NRC will publish a

document that withdraws the direct

final rule. If the direct final rule is

withdrawn, the NRC will address in a

Final Rule the comments received in

response to the proposed revisions in a

subsequent final rule. Absent significant

modifications to the proposed revisions

requiring republication, the NRC will

not initiate a second comment period

for this action in the event the direct

final rule is withdrawn.

Electronic Access

You may also provide comments via

the NRC’s interactive rulemaking web

site 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 rulemaking site,

contact Ms. Carol Gallagher, (301) 415–

6215; e-mail CAG@nrc.gov.

List of Subjects

10 CFR Part 50

Antitrust, Classified information,

Criminal penalties, Fire prevention,

Intergovernmental relations, Nuclear

power plants and reactors, Radiation

protection, Reactor siting criteria,

Reporting and recordkeeping

requirements.

10 CFR Part 70

Criminal penalties, Hazardous

materials transportation, Material

control and accounting, Nuclear

materials, Packaging and containers,

Radiation protection, Reporting and

recordkeeping requirements, Scientific

equipment, Security measures, Special

nuclear material.

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, the National

Environmental Policy Act of 1969, as

amended, and 5 U.S.C. 553, the NRC is

considering adopting the following

amendments to 10 CFR Parts 50 and 70.

PART 50—DOMESTIC LICENSING OF

PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for 10 CFR

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. 444, 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, as amended by

Pub. L. 102–486, sec. 2902, 106 Stat. 3123,

(42 U.S.C. 5851). Sections 50.10 also issued

under secs. 101, 185, 68 Stat. 936, 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 50.68 is added under the

center heading ‘‘Issuance, Limitations,

and Conditions of Licenses and

Construction Permits’’ to read as

follows: