NUREG-1144

Rev. 2

Nuclear Plant -Aging Research

(NPAR) Program Plan, - II

Status and Accomplishments

t

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

i

_
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NUREG-1144

Rev. 2

RD, RG, RM, RV,

R9, 9A, 9B

Nuclear Plant Aging Research

Nuclear Plant Aging Research

(NPAR) Program Plan

Status .and Accomplishments

Manuscript Completed: March 1991

Date Published: June 1991

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555
ABSTRACT

A comprehensive Nuclear Plant Aging Research (NPAR) Program was implemented by the U.S. NRC Office of Nuclear Regulatory Research in 1985 to identify and resolve technical safety issues related to the aging of systems,

structures, and components in operating nuclear power plants. This is Revision 2 to the Nuclear Plant Aging Research Program Plan. This plan defines

the goals of the program, the current status of research, and summarizes

utilization of the research results in the regulatory process. The plan also

describes major milestones and schedules for coordinating research within the

agency and with organizations and institutions outside the agency, both

domestic and foreign.

Currently, the NPAR Program comprises seven major areas: 1) hardwareoriented engineering research involving components and structures; 2) systemoriented aging interaction studies; 3) development of technical bases for

license renewal rulemaking; 4) determining risk significance of aging phenomena; 5) development of technical bases for resolving generic safety issues;

6) recommendations for field inspection and maintenance addressing aging concerns;,7) and residual lifetime evaluations of major LWR components and structures. The NPAR technical database comprises approximately 100 NUREG/CR

reports by June 1991, plus numerous published papers and proceedings that

offer regulators and industry important insights to aging characteristics and

aging management of safety-related equipment. Regulatory applications include

revisions to and development of regulatory guides and technical specifications; support to resolve generic safety issues; development of codes and

standards; evaluation of diagnostic techniques, (e.g., for cables and valves);

and technical support for development of the license renewal rule.

ii

-p
- CONTENTS

ABSTRACT . . . . . . . . . . . .' . . . . . . . . . . . . . .

FOREWORD . . . . . . . . . . .'. .'. .-. . . . . . . . . .

. . . iii

xiii

SUMMARY . . . . . . . . ... . ..

ACKNOWLEDGMENTS . . . . . . . . . . . . . . . . . .

ACRONYMS AND INITIALISMS . . . . . . . . . . . . . .

WORKING DEFINITIONS FOR NPAR . . . . . . . . . . .

1.O. INTRODUCTION . . . . . . . . . . . . . . . . .

1.1 GOALS AND OBJECTIVES OF THE NPAR PROGRAM : . . .

-1.2 BACKGROUND AND NEED .'. . . . ... .'. . .

1.3 ACCOMPLISHMENTS . . .,. . . . ... . . . .

1.4 ORGANIZATION OF NPAR PROGRAM'PLAN DOCUMENT

xv

xvii

xix

xxii

....... ........ 1.1

1.4

1.4

1.7

1.8

1.5 NUCLEAR PLANT AGING RESEARCH PLANNING INSTRUMENTS

2.0 BASIS FOR THE NUCLEAR PLANT AGING'RESEARCH PROGRAM

2.1 BASIS FOR IDENTIFYING AND RESOLVING TECHNICAL

SAFETY ISSUES .. . . . . . . . . . . . . . . . . .

2.2 TECHNICAL SAFETY ISSUES ... . . . . . ... . .

2.3 IMPLEMENTATION . . . . . . . ..... . . . . . . . . .

3.0 AGING ASSESSMENTS OF SPECIFIC COMPONENTS AND SYSTEMS,

UTILIZING THE NPAR PROGRAM'STRATEGY .. . . . .. . . . .

3.1 PRE-PHASE I - PRIORITIZATION OF, SYSTEMS, STRUCTURES,

AND COMPONENTS . . . .. '. . ... ''. . . .. '. . . .

3.2 PHASE I - INTERIM AGING ASSESSMENT . . . . . . .. . .

. . . 1.9

. . . 2.1

. . . 2.1'

2.1

. . . 2.2

. . . 3.1

. . . 3.1

. .-. . 3.6

3.2.1 Cons'ultat'ions with Nuclear Regulatory Commission

Staff . . . . . . . . . . . . . . ... . . . . . . .

3.2.2 Review of Design Information andApplications .

3.2.3 Survey of Operating Experience and Failure

Evaluation . . . .'. . . . . . . . .'... . . .'. .

3.9

3.9

'3.10

v
3.2.4 Review and Evaluation of Inspection, Surveillance,

and Monitoring . . . . . . . . . . . . . . . . . . 3.11

3.2.5 Interim Aging Assessment and Recommendations . . . 3.11

3.3 PHASE II - COMPREHENSIVE AGING ASSESSMENT . . . . . . . . 3.11

3.3.1 Review and Verification of Improved Inspection,

Surveillance, and Monitoring Methods . . . . . . . 3.12

3.3.2 Testing of Naturally-Aged Components . . . . . . . 3.13

3.4 UTILIZATION OF RESEARCH RESULTS . . . . . . . . . . . . . 3.13

3.4.1 Resolution of Generic Safety Issues . . . . . . . . 3.13

3.4.2 Considerations for License Renewal . . . . . . . . . 3.14

3.4.3 Guidelines for Inspection, Surveillance,

Monitoring Methods, and Maintenance . . . . . . . . 3.14

3.4.4 Guidelines for Service-Life Predictions . . . . . . 3.15

3.4.5 Recommendations for Standards and Guides 3.15

3.4.6 Dissemination of Technical Results . . . . . . . . 3.16

3.4.7 Innovative Materials and Design . . . . . . . . . . 3.16

4.0 SPECIAL TOPICS . . . . . . . . . . . . . . . . . . . . . . . . . 4.1

4.1 RISK EVALUATIONS OF SIGNIFICANT AGING EFFECTS . . . . . . 4.2

4.2 RISK-BASED COMPONENT PRIORITIZATION AND SELECTION . . . . 4.3

4.3 SHIPPINGPORT AGING ASSESSMENT . . ... . . . . . . . . . . 4.4

4.4 AGING ASSESSMENT AND MITIGATION OF MAJOR LWR

COMPONENTS AND STRUCTURES . . . . . . . . . . . . . . . . 4.5

4.5 DEVELOPMENT OF THE TECHNICAL BASIS FOR GUIDANCE TO

SUPPORT REGULATORY ACTIONS INVOLVING NUCLEAR

PLANT AGING ..... . . . . . . . . . . . . . . . . . . 4.6

4.6 TECHNICAL SPECIFICATIONS - FROM AGING PERSPECTIVE . . . . 4.7

4.7 GUIDANCE FOR RECORDKEEPING . . . . . . . . . . . . . . . . 4.8

4.8 THE USE OF NPAR RESULTS IN INSPECTION ACTIVITIES . . . . . 4.9

4.9 AGING RELIABILITY MODELING OF ACTIVE COMPONENTS 4.9

vi
4.10

4.11

4.12

4.13

4.14

4.15

4.16

REGULATORY INSTRUMENT REVIEW . . . . . . ..

REVIEW OF INDUSTRY TECHNICAL REPORTS FOR:,

LICENSE RENEWAL .................

AGING EFFECTS IN BALANCE-OF-PLANT SYSTEMS . . . . .

MAINTENANCE TO MANAGE AGING ... . . . . . . . . .

AGING DEGRADATION OF CABLES . . . . . . . . . .

FIRE VULNERABILITY OF AGED ELECTRICAL COMPONENTSOTHER SPECIFIC TOPICS .... . .... . . . . . . .

4.16.1 Aging/Seismic Shock Interaction

4.16.2 Role of Safety-Centered'AgingManagement -.

4.16.3 Status of the Maintenance Rule.,.-..

; .. 4.10

. .. . 4.11

* * 4.12

. . 4.13

. . 4.13

. . 4.14

. . 4.15

. . 4.15

. . 4.16

. . "I 4.17

*

. .

. .

5.0 LICENSE 5.0 ICESE RENWALNEWL . . ..... . . . . . . . . . . . . . . . ... . . . . .5.11

5.1 REGULATORY GUIDANCE FOR LICENSE RENEWAL . . . . . . . . . 5.2

5.2 NPAR PARTICIPATION IN THE LICENSE RENEWAL PROCESS . . . . 5.2

5.3 INDUSTRY PERSPECTIVE . . . . . . . . . . . . . . . . . . . 5.3

5.4 REGULATORY RESEARCH PERSPECTIVE . . . . . . . . . . . . . 5.5

5.4.1 Technical Issues . . . . . . . . . . . . ... . . 5.5

5.4.2 Plans and Schedule 5.6

5.5 INDUSTRY REPORT REVIEWS . . ... . . . . . . . . ... :5t6

5.6 LEAD PLANT IMPLEMENTATION . . . . . . . . ... 5.7

5.7 SUPPORT FROM CODES AND STANDARDS AND INDUSTRY GROUPS . .. 5.7

6.0 UTILIZATION OF RESEARCH RESULTS .. 6.1

6.1 STATUS OF NUCLEAR PLANT AGING RESEARCH-RESULTS . . . . . 6.1

6.2 STATUS OF NUCLEAR PLANT AGING RESEARCH

DOCUMENTATION . . .. . . 6.1

6.3 TECHNICAL APPLICATIONS OF NPAR RESULTS . . . . . . . . . . 6.2

6.3.1 General Technical Applications . . . . . . . . . . 6.2

vi i
6.3.2 Specific Technical Applications . . . . . . . . . . 6.3

6.4 REGULATORY APPLICATIONS OF NPAR RESULTS . . . . . . . . . 6.5

6.4.1 General Regulatory Applications . . . . . . . . . . 6.5

6.4.2 Specific Regulatory Applications . . . . . . . . . 6.5

6.5 APPLICATION OF NPAR RESULTS TO UNDERSTAND AND

MANAGE AGING . . . . . . . . . . . . . . . . . . . . . . . 6.11

6.5.1 Results from Aging Investigations on

Components and Systems . . . . . . . . . . . . . . 6.11

6.5.2 Results from Nuclear Plant Aging Research

Special Topics . . . . . . . . . . . . . . . . . . . 6.21

6.6 NPAR SUPPORT TO LICENSE RENEWAL . . . . . . . . . . . . . 6.35

7.0 NPAR REVIEW AND INTEGRATION ACTIVITIES . . ... . . . . . . . . 7.1

7.1 TECHNICAL INTEGRATION REVIEW GROUP FQR AGING AND

LIFE EXTENSION . . . . . . . . . . . . . . . . . . . . . . 7.1

7.2 AGING AND LIFE EXTENSION COORDINATING COMMITTEE . . . . . 7.1

7.3 NPAR RESEARCH REVIEW GROUP .... . . . . . . . . . . . . 7.1

7.4 OTHER NPAR REVIEW ACTIVITIES . . . . . . . . . . . . . . . 7.2

8.0 COORDINATION WITH OTHER PROGRAMS, INSTITUTIONS, AND

ORGANIZATIONS . . . . . . . . . . . . . . . . . . . . . . . . . 8.1

9.0 SCHEDULES AND RESOURCE REQUIREMENTS . . . . . . . . . . . . . . 9.1

10.0 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . 10.1

APPENDIX A - NPAR PROGRAM PARTICIPANTS . . . . . . . . . . . . . . . A.1

APPENDIX B - UNDERSTANDING AND MANAGING AGING . . . . . . . . . . . B.1

APPENDIX C - ONGOING PROGRAMS RELATED TO NPAR . . . . . . . . . . . C.1

APPENDIX D - ALTERNATIVE APPROACHES FOR CONDUCTING NPAR

PHASE-II COMPREHENSIVE ASSESSMENTS . . . . . . . . . . D.1

APPENDIX E - SHIPPINGPORT STATION AGING EVALUATION

NRC CONTRACTOR PUBLICATIONS . . . . . . . . . . . . . . E.1

Viii
FIGURES

1.1 Aging Research Structure . . . .'. . . . . . . . . . . .

1.2 NPAR Program Elements . ... ... . . . . . . . . . . . . .

3.1 NPAR Structure .....................

3.2 NPAR Program Strategy . . . . ... . . . . . . . . . . . .

5.1 Nuclear Power Plant Licensing Regimes and Aging

Importance . ... . . . . . . .- . . . . . . ... . . . .

5.2 Schedule of License Renewal Rule Development and

Implementation . . . . . . . . . . . .'. . . . . . . . .

6.1 Emergency Diesel Generator Subsystem and Component

Failure Rates ......................

6.2 Functional Indicators for Dielectric Integrity

6.3 Functional Indicators for Rotational Integrity . . . . .

6.4 Functional Indicators for Mechanical Integrity . . . . .

6.5 Understanding and Managing Aging of Emergency Diesel

Generators . . . . . . . . . . . . . . . . . . . . . . . .

6.6 Understanding and Managing Aging of PWR Pressure Vessels

6.7 Understanding and Managing Aging of BWR Pressure Vessels

6.8 Understanding and Managing Aging of BWR Mark I

Containments . . . . . . . . . . . . . . . . . . . . . .

6.9 Understanding and Managing Aging of PWR Reactor Internals

6.10 Understanding and Managing Aging of BWR Reactor Internals

6.11 Understanding and Managing Aging of PWR RCS Piping

and Nozzles . . . . . . . . . . . . . . . . . . . . . . .

6.12 Understanding and Managing Aging of PWR Steam

Generator Tubes .....................

6.13 Understanding and Managing Aging of Pressurizer, Surge,

and Spray Lines .....................

. . . 1.2

. . . 1.3

. . .I 3.2

.. . 3.2 . . .. 3.3

. . . 5.1

. . . 5.4

. . . 6.14

. . . 6.17

. . . 6.18

. . . 6.19

. . . 6.23

. . . 6.24

. . . 6.25

. . . 6.26

. . . 6.27

. . . 6.28

. . . 6.29

. . . 6.30

. . . 6.31

ix
6.14 Understanding and Managing Aging of BWR Recirculation

Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.32

6.15 Understanding and Managing Aging of LWR Coolant Pumps 6.33

9.1 NPAR Milestones and Schedules--Systems . . . . . . . . . . . . 9.2

9.2 NPAR Milestones and Schedules--Components . . . . . . . . . . . 9.6

9.3 NPAR Milestones and Schedules--Special Topics . . . . . . . . . 9.10

9.4 NPAR Milestones and Schedules--Aging Program and

Issues Coordination ..... . . . . . . . . . . . . . . . . . 9.11

x
TABLES

1.1 Schedule of License Expirations for the 30 Oldest

Operating Plants ....................

3.1 Components of Current Interest in the NPAR Program

and Their Completion Schedule . . . . . . . . . . . . . .

3.2 Systems of Current Interest in the NPAR Program

and Their Completion Schedule . . . . . . . . . ... . . .

3.3 Generic Safety Issues, with Elements of Aging . . . . . .

4.1 Special Topics of Current Interest in the NPAR Program

and Their Completion Schedule . . . . . . . . . . . . .

6.1 Potential Use of NPAR Results, Involving Aging Consideration, for Systems, Structures, and Components . . . .

6.2 Selected ASME Standards for Operation and Maintenance of

Mechanical Equipment . . . . . . . . . . . . . . . . . .

6.3 IEEE Standards for Electrical Equipment for Nuclear Power

Plants that Provide Information Relevant to Aging and LR

8.1 Selected Programs Relevant to NPAR Aging Studies

and Programs ......................

9.1 Planned NPAR Accomplishments for FY 1991 to 93 . . . . .

. . . 1.6

. . . 3.7

. . . 3.8

. . . 3.14

. . . 4.1

. . . 6.7

. . . 6.12

6.13

. . . 8.2

9.12

xi
FOREWORD

The U.S. Nuclear Regulatory Commission's (NRC's) hardware-oriented engineering research program for plant'aging-and degradation'monitoring of components and systems was first discussed in the initial version of the program

plan, NUREG-1144, issued in July 1985. The program is identified as the

Nuclear Plant Aging Research (NPAR) Program.. It was stated in the plan that

NUREG-1144 would be a living document to be'revised periodically. The revisions would reflect-the experience' gained-from-implementing the plan. To be

incorporated were 1) comments received from within the NRC, industrial codes

and standards committees, and-from domestic and foreign organizations and

institutions; 2) research results and experience gained from the utilization

of the plan in the regulatory process.. *The first revision to the plan,

NUREG-1144, Rev. 1, was issued-in September 1987.

The Office of Nuclear Regulatory'Research' (RES) staff, which oversees the

development and implementation ofNUREG-1144, has since received numerous

additional comments and updates'on'this'document from various offices within

the NRC and from individuals, organizations, and institutions outside NRC,

both domestic and foreign. This revision reflects those comments.

The NRC provided planning guidance for needed safety research on plant

aging and license renewal in its 1987 Policy 'and Planning Guidance document

(NUREG-0885, Issue 5). -The Executive Director-for Operations (EDO) provided

specific program guidance to the staff for FY 1986 to 1988 planning and program development. The NRC staff provided their comments on the current

research program and needs for additional'research and prioritization by

"user-need" letters to RES and thrdugh'the Technical Integration Review Group

for Aging and Life Extension (TIRGALEX) review of the NPAR Program. More

recent review and guidance have come from the NPAR Review Group, represented

by staff from the NRC offices and from the Regions; from the Nuclear Safety

Research Review Committee (NSRRC); and from the Advisory Committee on Reactor

Safeguards (ACRS).

This document presents a revised research plan that addresses the identification and resolution of technical safety issues relevant to plant aging

and license renewal. This document also describes the utilization of results

of aging assessments in the regulatory process and presents the current status

of major program activities and significant accomplishments. Additionally,

this plan describes methods to mitigate the effects of age-related degradation

on electrical and mechanical components and nuclear plant safety systems.

These methods include inspection, diagnostics, condition monitoring, maintenance, trending, and recordkeeping.

In conjunction with related program plans for primary system pressure

boundary components and civil structures, this plan forms the overall framework for NPAR within the Division of Engineering, Office of Nuclear Regulatory

Research of the NRC.

xiii
We welcome comments on this plan and will consider them in developing

subsequent editions of this document. Comments need not be restricted to the

research activities described herein; comments identifying omissions and/or

recommending additional research are also welcome.

ndra P. -, NPAR Program Coordinator

Mil&n Vagins, 6hief '

Electrical and Mechanical

Engineering Branch

X 7\ ,

Approved by: L. C. Shao, Director

Division of Engineering

Office of Nuclear Regulatory Research

xiv
SUMMARY I

The charter of the Nuclear Plant Aging Research (NPAR) Program described

in this plan is to identify and address technical safety issues related to the

aging of electrical-and mechanical components and safety systems and.support

systems in commercial nuclear powerplants (NPPs). The aging term of inter-''

est includes the'pe'riod of normal'licensed plant operation as well'as the period of extended plant"operation that may.be requested-in utility applications

for license renewals."The principal'goals of the program are to-understand

the effects of age-related degradation inmNPPs.and how to manage and mitigate

them effectively.

The NPAR Program comprises seven major'areas: 1) hardware-oriented engineering research involving components and structures; 2)..system-oriented aging

interaction studies;'3) development of technical.bases'for license renewal

rulemaking; 4) risk significance'of aging phenomena; 5) development of technical bases to resolve generic safety 'issues;: 6)yrecommendations.for field

inspection and maintenance addressing aging concerns; 7)'and.residual..lifetime

evaluations of major light-water reactor (LWR) components'and structures.

Significant progress has been.made incompleting the aging assessments of

key components and systems since th6e'prior' revision of NUREG-1144.' Nineteen

safety-related components are under NPAR investigation, including five for

which Phase-Il assessments are completed.' Seventeen'safety-related systems

are under NPAR investigation, including'two foriwhich'Phase-II assessments are

completed. .Progress has also been made.in developing models and-approaches to

evaluate the effects of 'age-related'degradatib'n on plant'risk and'effectiveness of maintenance to alleviate aging concerns. Studies to identify degradation sites and life-limiting processes for each major component have produced

useful data. The Shippingport Aging'Evalu'ation provided over 200 naturallyaged components and specimens of mater'ials for'aging evaluations. Operating

experience and expert opinion have been applied to these aging assessments.

Initiatives to develop the-te'chnica -bases for.license renewal of'NPPs

have advanced since the last program plan update. On the regulatory side,

plant license'renewal has been ident'ified as an important cur'rent NRC

initiative.' The NRC is'developing''a License.Renewal Rule, and the'NPAR'Program has several initiatives underway to support the Rule, including development of a regulatory guide on Standard Format and Content of Technical Information for ADplications'to'Renew Nuclear Power Plant Operating Licenses'

(DG-1009); technical support to the Office of Nuclear Reactor Regulations

(NRR) in the following activities: development-of Safety.Evaluation Reports

(SERs); development of a Standard Review Plan for License Renewal (SRP-LR);

and reviews of lead plant license renewal applications; also,.participation in

reviews of Industry Reports. On the"'industry side, two lead plants are

developing license renewal-requests, 'planned for:submission to NRC.in.1991.

The two plants are the Monticello boiling-water reactor (BWR) and the.Yankee

Rowe pressurized-water reactor (PWR). The license renewal initiatives are

xv
sponsored by the U.S. Department of Energy, Electric Power Research Institute,

and utilities, coordinated by the Nuclear Utility Management and Resources

Council (NUMARC).

The NPAR staff has been prominent in international initiatives on NPP

aging. A 1987 International Atomic Energy Agency (IAEA) symposium on NPP

Aging and Life Extension in Vienna, Austria, was attended by delegates from

30 countries. The program included 14 presentations by NPAR speakers from a

total of 40 papers. A major IAEA initiative, now underway on NPP aging, has

been guided to a major extent by methods and results developed under the NPAR

program. In August 1988, the NPAR Program hosted an international symposium

on NPP aging in Bethesda, Maryland. Speakers from 10 nations participated,

and a proceedings has been published (NUREG/CP-0100).

Through its research, the NPAR Program is providing information for

timely and sound regulatory decisions on the operation of NPPs of all ages,

including the possibility of extended operation. The research results are

providing specific inputs to the regulations, including revisions to regulatory guides, technical specifications, generic safety issues, and codes and

standards.

Future NPAR plans include the following:

* continuation of research on safety-related systems, structures, and

components to address technical safety issues related to aging

* continuing studies on special topics, including aging effects on

plant risk, and residual life assessments of major LWR components

* providing guidance for effective field inspections and maintenance

that address understanding and managing aging in safety-related components and structures

* providing input to the license renewal process, including support to

the License Renewal Rule, assistance with reviews of industry positions, and further resolution of aging issues that are specific to

extended operation; support to NRR in developing SERs and the SRP-LR

for license renewal

* application of NPAR strategy and database to assessment of aging

impacts on mothballed plants

* support to develop the Maintenance R;ule and a Regulatory Guide

* motivation to implement well-structured aging management programs

* application of lessons learned from NPAR aging assessments to

advanced reactor technology.

xvi
ACKNOWLEDGMENTS

Staff members(a) from the following laboratories have contributed to this

revision to the NPAR Program Plan, Status, and Accomplishments:

BNL

INEL

ORNL

PNL

SAIC

SNL

J. H. Taylor

H. L. Magleby

D. M. Eissenberg

A. B. Johnson, Jr. (with major

W. E. Vesely

M. Bohn

M. J. Jacobus

contributions from W. C. Morgan)

The following NRC/EMEB Piroject Managers also contributed to the

document:

S. Aggarwal

J. Burns

W. Farmer

G. Weidenhamer

John Nageley and Robert Pedersen at PNL have provided editorial

assistance.

(a) Represented by the corresponding NPAR project managers.

xvii
ACRONYMS AND INITIALISMS

ACRS Advisory Committee on Reactor Safeguards

AEOD Office of Analysis and Evaluation of Operational Data of NRC

AFW auxiliary feedwater

ALEXCC Aging and Life Extension Coordinating Committee

ANL Argonne National Laboratory

ANS American Nuclear Society

ANSI American Nuclear Standards Institute

ASME American Society for Mechanical Engineers

BNL Brookhaven National Laboratory

BOP balance of plant

BWR boiling-water reactor

DOE U.S. Department of Energy

ECCAD electrical circuit characterization and diagnostic

ECCS emergency core cooling system

EDG emergency diesel generator

EDO Executive Director for Operations

EIS environmental impact statement

EMEB Electrical and Mechanical Engineering Branch of the Division of

Engineering

EPRI Electric Power Research Institute

EQAG Equipment Qualification Advisory Group of EPRI

GSI generic safety issues

HPCI high pressure coolant injection system (PWRs)

IAEA International Atomic Energy Agency

xix
I&C instrumentation and control

IEEE Institute of Electrical and Electronic Engineers

INEL Idaho National Engineering Laboratory

IPRDS In-Plant Reliability Data System

ISMM inspection, surveillance, and monitoring methods

IR industry report

LER licensee event report

LR license renewal

LWR light-water reactor

MCC motor control center

MEB Materials Engineering Branch of the Division of Engineering

MOV motor operated valve

NDE nondestructive examination

NIST National Institute of Standards and Technology

NOAC Nuclear Operations Analysis Center at ORNL

NPAR Nuclear Plant Aging Research

NPP nuclear power plant

NPRDS Nuclear Plant Reliability Data System

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation of NRC

NSAC Nuclear Safety Analysis Center, operated by EPRI

NSSS nuclear steam supply system

NUMARC Nuclear Utility Management and Resources Council

NUPLEX Nuclear Utility Plant Life Extension working group of NUMARC

ORNL Oak Ridge National Laboratory

xx
PLEX plant life extension

PNL Pacific Northwest Laboratory

PORV power-operatedrelief valve.

PRA. probabilistic risk assessment

PSA probabilistic safety analysis

PWR pressurized-water reactor

RCIC reactor core isolation-cooling

RES Office of Nuclear Regulatory Research of NRC

RHR residual heat removal

RPS reactor.protection system,,,-

SSCs systems, structures, and components

SNL Sandia NationalLaboratory ,

SNUG snubber utility group

SRP-LR Standard Review Plan for License Renewal

SSEB Str uctural, and Seismic Engineering Branch of the Division of

Engineering

SWS service water system

SWWG Service.Water,Working.Group of EPRI

TIRGALEX Technical Integration Review Group for Aging and Life Extension

xxi
WORKING DEFINITIONS FOR NPAR(a)

* accelerated aging* - synonym for artificial aging (when the simulation of natural aging is performed by application of stressors

beyond those found in service in order to obtain observable aging

effects in a reasonable period of test time).

* agina* - showing the effects of time or use in the physical characteristics of a system, structure, dr component.

* acing degradation* - gradual deterioration in the physical characteristics of a system, structure, or component that is due to aging

mechanisms, occurs with time or use under pre-service or service

conditions, and could impair its ability to perform any of its

design functions.

* aging management* - engineering, operations, and maintenance

activities to control aging degradation and failures due to aging

of systems, structures, or components to within acceptable limits.

* aging mechanism* - process that gradually changes the physical

characteristics of a system, structure, or component with time or

use.

• common cause failure* - two or more redundant component failures

due to a single cause (IEEE Standard 100).

• common mode failure* - two or more redundant component failures in

the same manner or mode.

* condition monitoring* - continuous or periodic measurement and

trending of the performance or physical characteristics of a system, structure, or component to indicate its current or future

performance.

• extended operation - operation of a nuclear power plant beyond the

term authorized by its current operating license.

* important to safety - defined by 10 CFR 50, Appendix A, as "Those

structures, systems and components that provide reasonable assurance that the facility can be operated without undue risk to the

health and safety of the public." For details, see 10 CFR 50.49.

(a) Starred entries were developed by a Committee on Common Aging Terminology with representatives from NRC, research laboratories, electric utilities, EPRI, and NUMARC. The initiative was sponsored by EPRI and was

coordinated by MPR Associates. A report is expected to be issued in

1991.

xxi 1
* inspection* - observation or.measurement to verify that the physical characteristics'of a system,;structure,;or component conform to

acceptance criteria. t o c o conform'to

* license extension - authorized extension of the operating license

to allow for the time elapsed between-the dates of the construction

permit (CP)'and operating.license:.(OL).

* license renewal -`2the issuance of an operating'license which supersedes an existing operating license for a NPP.

maintenance. corrective* actions.that restore, by repair, overhaul, or replacement, the capability of a failed system, structure,

or component.to perform its.design-.function within-acceptance

criteria.

maintenance, predictive* - a'form of preventive maintenance performed periodically or continuously to monitor, inspect, test,

diagnose, or trend a system'.s, structure's, or component's-performance or condition.indicators;'results indicate or.forecast

functional ability or the-:nature.and schedule of planned

maintenance prior to failure.

maintenance. preventive* -' periodic, predictive, or planned maintenance performed prior to failure' of a system,' structure, or component to extend its service.life'by.controlling.degradation or'

failure.

* natural aging* - aging of a system, structure, or component which

occurs under preservice and service conditions.

* Dlant life assessment - the process of ensuring that NPP components, systems, and structures can be maintained in a condition to

safely and reliably perform their design functions throughout

extended operation.

* safety-related items - defined by 10 CFR 100, Appendix A, as a subset of components "important to-safety" (see definition above).

This definition includes those systems, structures, or components

designed to remain functional in the event that a Safe Shutdown

Earthquake occurs. . The required safety functions include

maintaining

1. the integrity of the reactor coolant pressure boundary

2. the capability to shut down the reactor and maintain it in a

safe shutdown condition

xxiii
3. the capability to prevent or mitigate the consequences of

accidents that could result in potential offsite exposure

comparable to the guideline exposures of this part

(10 CFR 100).

Note: Regulatory Guide 1.29, "Seismic Design Classification,"

lists light-water reactor SSCs that are to be designed to

Seismic Category I and, therefore, are considered safetyrelated according to 10 CFR 100, Appendix A.

Note: The ASME Boiler and Pressure Vessel Code, in Section III, (ASME 1983), requires the classification of components into 3 classes. Class 1 is safety-related pressure

retaining equipment; Class 2 is the pressure retaining equipment that is essential for safe operation of the plant, but

does not, under normal operation contain nuclear materials;

and Class 3 is for other equipment.

surveillance* - observation or measurement of the performance or

physical characteristics of a system, structure, or component to

verify that it conforms to acceptance criteria.

trigger event - an operational transient or minor accident that can

lead to a more serious event when followed by failures in safetyrelated systems. Emphasis is placed on the relationship between

failures (causes and modes) expected to be experienced during operation and those that would potentially occur under the stresses

associated with design basis or trigger events.

Xxi V
1.0 INTRODUCTION

The United States has more than 110 nuclear reactors in commercial operation and several of these reactors have operated more than 20 years. As the

population of light-water reactors (LWRs) advances in age; the need for a

research program that systematically assesses the effects of age-related

degradation on the safety of these plants has been recognized. The Director

of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission (NRC), in his comments on the Long-Range Research Plan, identified a need

for a research program to investigate the safety aspects of aging processes in

commercial nuclear power plants (NPPs). The Advisory Committee on Reactor

Safeguards (ACRS), in their 1983 report to Congress, also recommended initiating an aging research program.

The NRC provided guidance for needed safety research on plant aging and

license renewal in its Policy and Planning Guidance document (NUREG-0885).

Also, the Executive Director for Operations (EDO) provided specific program

guidance to the staff for FY 1986 to 1988 planning and program development.

In response to this need, the NRC Office of Nuclear Regulatory Research

(RES) developed and implemented a hardware-oriented engineering research program for understanding and managing aging of safety-related components and

systems. This program is identified'as the Nuclear Plant Aging Research

(NPAR) Program. As shown in Figures 1.1 and 1.2,:the program is conducted by

the Electrical and Mechanical Engineering Branch of the Division of Engineering of the Office of Nuclear Regulatory Research. Similar programs .are conducted on other aspects of aging in NPPs (Vagins and Taboada 1985; Muscara and

Serpan 1985; Vagins and Strosnider 1985; Muscara 1985). One of these programs focuses on aging in NPP vessels, piping, and steam generators and nondestructive examination techniques, and is being conducted by the Materials

Engineering Branch of the Division of Engineering (MEB). The program plan

developed by MEB is a companion to the NPAR program plan. Another related

program involves an assessment of the impact of age-related degradation on

plant civil structures and is being conducted by the Structural and Seismic

Engineering Branch (Naus et al. 1989). These three programs form the

foundation for research on aging and are implemented within the Division of

Engineering,'Office of Nuclear Regulatory Research (NRR) of the NRC.

The NPAR Program Plan was first described in the July 1985 issue of

NUREG-1144-(Morris and Vora .1985).' In September 1987, the Program Plan was

revised and reissued as NUREG-1144, Rev. 1 (Vora 1987). This report, Revision 2, describes the current NPAR Program Plan, discusses progress, status,

and accomplishments in the program since 'issuance of Revision 1, and includes

future plans. Revision 2 also emphasizes the research currently being conducted to resolve the technical safety issues relevant to the aging of NPPs,

both in the current licensing period and in extended operation. The utilization of the NPAR technical data in the regulatory process is also included.

1.1
| Components Systems Vessels Piping | [ L - I r ~essels Sale Ends] LŽ~rturieL

Breakers HP-ECCS

Relays LP-ECCS Steam _

SOVSWS Generator

Bistables/Switches CCWS

s Transformers RPS PressurizerRPV

PORV IE Dist Sys. internals

Heat Exchangers AFS

ors Reactor Internals CRD

ions Batteries SLCS Supports Others

Motors IAS

ssors Chargerslinverters ADS

Motor Control ESF

CCS

I&C

Turb/Cntrl 39103085.1

FIGURE 1.1. Aging Research Structure
Electrical and Mechanical

Engineering Branch

(Aging and Components Section)

.__ . __ __ __ __ __ __ __ __ __ _I 

'__ _ _ _ _._ _ _ l __I

Components

l

I

I.

I

I

*I

Special Studies/

Aging Program and

Issues Coordination

I

Systems

II

I

I

I I

I

W.

Risk Significance of Aging

Component Prioritization

Residual Lifetime Assessment

Shippingport Aging Evaluation

Review of Technical Specifications

Recordkeeping Needs

Preparation of Regulatory Guide(s)

SRP Review from Aging Perspective

Review of Applicable Reg. Instruments

Review of Bulletins and Notices

Maintenance to-Manage Aging

Aging of Mothballed Nuclear Power Plants

39103085.2

FIGURE 1.2. NPAR Program Elements
1.1 GOALS AND OBJECTIVES OF THE NPAR PROGRAM

The main goals of the NPAR Program are to understand aging and to identify ways to manage aging of safety-related systems, structures, and components (SSCs) in NPPs. The following are the technical objectives of the

Program:

* identify and characterize aging effects which, if unmitigated, could

cause degradation of SSCs and thereby impair plant safety

* develop supporting data and information to facilitate management of

age-related degradation

* identify methods of inspection, surveillance, and monitoring, or of

evaluating residual-life of SSCs, which will ensure timely detection

of significant aging effects before loss of safety function

* evaluate the effectiveness of storage, maintenance, repair, and

replacement practices in mitigating the effects of aging and diminishing the rate and extent of degradation caused by aging

* provide technical bases and support for the License Renewal Rule and

the license renewal process and develop a regulatory guide on the

format and technical information content for license renewal

applications.

The NPAR Program has been developed to meet these goals and objectives.

Currently, the NPAR Program comprises seven major areas: 1) hardware-oriented

engineering research on components and structures; 2) systems-oriented aging

interaction studies; 3) development of technical bases for license renewal

rulemaking; 4) risk significance of aging phenomena; 5) development of technical bases for resolving generic safety issues; 6) recommendations for field

inspection and maintenance to address aging concerns; and 7) residual lifetime

evaluations of major LWR components and structures.

Other ongoing NRC programs, industry-sponsored research, and programs

being conducted in foreign countries have been considered in developing the

NPAR Program Plan. In cases where relevant results from aging studies are

available or are being developed, the NPAR Program has been planned to avoid

duplication of effort.

1.2 BACKGROUND AND NEED

The NPAR Program has identified aging as the cumulative, time-dependent

degradation of a system, structure, or component in a NPP that, if unmitigated, could compromise continuing safe operation of the plant. Necessary

measures must, therefore, be taken to ensure that aging does not reduce the

operational readiness of a plant's safety systems and does not result in

common-mode failures of redundant, safety-related equipment, thus reducing

1.4
defense-in-depth. It is also' necessary to'ensure that aging does not lead to

equipment failure which could cause an accident.

In December 1990, there were 112 licensed commercial LWR power plants in

operation in the U.S. -The following summarizes the age distribution of these

plants:

Operating Lifetime Number

(Years Since Onerating License) of Plants

More than 20: 12

Between 15 and 20 37

lBetween 10 and 15 16

Between 5 and 10 26

Less than 5 21

The 30 oldest operating plants are listed in Table 1.1; the two oldest,

Yankee Rowe and Big Rock Point, have been in operation for 30 and 28 years,

respectively. However, they are demonstration plants with design power of

<200 MWe. The next oldest plant, San Onofre-1, has a net capacity of 436 MWe'

and has been licensed for 24 years. In addition to the plants in operation,

there are 6 more plants' under construction. Most of these plants. are expected

to be in operation within the next decade.

As the population of U.S. LWRs has aged,'problems have occurred'as a

result of time-dependent degradation mechanisms such as stress'corrosion,

thermal aging, radiation embrittlement, fatigue, and erosion. 'These problems

have included failures in pumps, valves, and relays, embrittlement of cable

insulation, and cracking of the heat-'treated anchor heads .for post-te'nsioning

systems in containment. Although progress is being made to mitigate-the

degradation that has already been identified, significant questions concerning

age-related degradation of SSCs remain because of the variety of components in

a commercial power reactor, the complexity of the aging process, and the

limited experience with prolonged operation of these power plants.

The NPAR Program also provides key information'to enable the NRC-to

resolve technical safety issues and define its policy and regulatory position

on license renewal (LR). The License Renewal Rule (55 FR 29043) is being

developed by the NRC to provide a clearily defined policy and regulatory position regarding the safe operation of aged plants for extended 'operation.

Clearly defined policies and criteria-are needed'to'ensure'thatrequests for

LR address the primary regulatory concernsC'and relevant technicalsafety

issues. Reactors are licensed for u'p to 40;years' of' operation'under current

regulations. Currentregulations also permit'rehnewi1of an 'operating license.

The NRC Technical Integration Review'Group for"Agi'ng a'nd Life Extension

(TIRGALEX), established in '1986,'deVeloped a working definition for LR.

*1.5
TABLE 1.1. Schedule of License Expirations for the 30 Oldest Operating Plants

Design

Rating Date Operating License

No. Name of Plant PWR/BWR (Net MWe) License Issued Expiration

I Yankee Rowe PWR 175 07/60 2000

2 Big Rock Point BWR 72 08/62 2002

3 San Onofre-1 PWR 436 03/67 2007

4 Haddam Neck PWR 582 06/67 2007

5 Nine Mile Point-1 BWR 620 08/68 2008

6 Oyster Creek-1 BWR 650 08/69 2009

7 Ginna PWR 470 09/69 2009

8 Dresden-2 BWR 794 12/69 2009

9 Monticello-1 BWR 545 09/70 2010

10 Robinson-2 PWR 700 09/70 2010

11 Millstone-1 BWR 660 10/70 2010

12 Point Beach-1 PWR 497 10/70 2010

13 Dresden-3 BWR 794 03/71 2011

14 Surry-1 PWR 788 05/72 2012

15 Point Beach-2 PWR 497 05/72 2012

16 Turkey Point-3 PWR 693 07/72 2012

17 Pilgrim-1 BWR 655 09/72 2012

18 Palisades PWR 805 10/72 2012

19 Quad Cities-2 BWR 789 12/72 2012

20 Quad Cities-1 BWR 789 12/72 2012

21 Surry-2 PWR 788 01/73 2013

22 Vermont Yankee BWR 540 02/73 2013

23 Oconee-1 PWR 887 02/73 2013

24 Turkey Point-4 PWR 693 04/73 2013

25 Maine Yankee PWR 825 06/73 2013

26 Fort Calhoun-1 PWR 478 08/73 2013

27 Prairie-1 PWR 530 08/73 2013

28 Indian Point-2 PWR 873 09/73 2013

29 Oconee-2 PWR 887 10/73 2013

30 Zion-I PWR 1040 10/73 2013

License renewal is defined to include operation beyond the original license

term of 40 years, requiring a program for understanding and managing of plant

SSCs.

The first license for a large plant (>400 MWe) will not expire until the

year 2007. However, to allow for the long lead times required for planning

and construction, the utilities need to decide approximately 10 to 15 years

before the end of the license period whether to request a plant LR or to construct replacement generating capacity. Utilities have defined a tentative

schedule for several important steps in the LR process (see Section 5.0). Two

representative LWRs were the subject of EPRI and DOE utility-sponsored pilot

studies on plant life extension. These were the Surry PWR plant, Virginia

Electric Power Company, (EPRI NP-6232M) and the Monticello BWR plant, Northern

1.6
States Power Company, (EPRI NP-5836M)'. Subsequently, a bidding process was

used to select two lead plants to develop formal requests for LR. The two

plants that were selected are the' 175-MWe Yankee Rowe:PWR'and Monticello, a

545-MWe BWR. At a-technical level, this project is'to provide an initial

evaluation of the-effects of aging'onrcommercial NPPs' and-establish the scope

of the effort to extend the operating lifetime of-these plants beyond their

initial 40 years of licensed operation. The first submittal to the NRC is

expected in 1991. If the NRC accepts-the LR-concept, many additional submittals for LR can be expected over the next decade (up to 45 by the year

2002).

1.3 ACCOMPLISHMENTS '

From key results of NPAR aging studies, the NRC recognizes that agerelated degradation, a process that occurs in all technologies, needs systematic inspection andcasse's'sment for NPP~safety;''Communication of the 'need to

understand and manage aging has been'.armajortcontribution of'the NPAR Program.

For example, it has been acknowledged that "the focus of license renewal is

mitigation and 'management of age-related degradation'to'ensure an'adequate

level of-safety" (NRC November 1989,'Session'-1).

While applying a systematic, comprehensive research strategy adaptable

to a broad range of SSCs, the NPAR laboratories have developed technical bases

for identifying age-related degradation of SSCs'and developed effective measures for understanding and managing aging in the SSCs.- The current scope of

the technical -bases is defined in Sections 3.0 through 6.0 of this plan for

potential users, both regulatory and industrial; '

The NPAR program also has made progress on major topics, including the

Shippingport 'Aging Evaluation', determining the 'isk significance of aging

phenomena, recommendations for field inspections,-maintenance, and residuallife investigations of major LWR components-and structures.

Cited in 'Section 6.0 are numerous specific ca'ses where NPAR Program

results have been used in the regulatory process; including input to regulatory guides, 'resolution of 'generic safety 'issues', revisions to the technical

specifications, and input to development of codes and 'standards. Technical

results have been used to recommend improved maintenance techniques and surveillance methods. ' -- ' '

Finally, while focusing on the need to'address aging in the current

licensing period, the NPAR Program has compiled'te'chnical "information that

contributes to the technical bases for the LR Rule, scheduled to be issued in

April 1992. ! ', I I .

1.7
1.4 ORGANIZATION OF NPAR PROGRAM PLAN DOCUMENT

This revision to the program plan, NUREG 1144, Rev. 2, emphasizes the

major NPAR Program elements: the hardware studies, the special topics, and

the recent emergence of LR. Included for each element is status, accomplishments, and future plans. Organization of the plan is summarized below.

Section 2.0 describes the basis from which the NPAR program conducts

aging research.

Section 3.0 contains an outline of the NPAR strategy, the systematic

approach used in the program for assessing the effects of aging on plant

safety-related SSCs. The criteria used to identify safety-related SSCs are

discussed, as is the phased approach developed to study the effects of agerelated degradation.

Section 4.0 reviews special topics relevant to the aging of NPPs that

have been and are being investigated in the program.

Section 5.0 summarizes the issues involved in LR, from both the industry

and regulatory perspectives. Plans and schedules for actions required to

resolve the issues and complete the LR process for the first NPPs are also

discussed.

Section 6.0 contains an overview of accomplishments of the NPAR Program

in the regulatory process. A synopsis of results then follows. The discussion of utilization is divided into five categories: rulemaking, generic

safety issues; maintenance and surveillance; inspection of safety systems and

components; and codes and standards.

Section 7.0 contains an overview of the NPAR review and integration

activities within NRC, with other government agencies, and with external

institutions and organizations, domestic and foreign.

Section 8.0 describes the program coordination with other government

agencies and external institutions and organizations, both domestic and

foreign. It also contains a discussion of the purpose and activities of the

NPAR Research Review Groups.

Section 9.0 contains the schedules and research requirements developed

for the various NPAR activities and a discussion of them. The schedules

include research activities supporting LR initiatives and the continuation of

age-related confirmatory research.

Appendix A contains a brief overview of NPAR participants.

Appendix B discusses bases for understanding and managing aging in NPPs.

Appendix C contains further details, beyond the overview in Section 8.0,

of ongoing programs related to aging and extended operation. The coordination

1.8
required between the NPAR investigations and other ongoing activities is discussed, with emphasis on the need to optimize the use of available resources.

Appendix D cites examples of alternative approaches for conducting NPAR

Phase-II comprehensive assessments.

Appendix E lists publications of NRC contractors on various aspects of

the Shippingport Station aging evaluation.

1.5 NUCLEAR PLANT AGING RESEARCH PLANNING INSTRUMENTS

The following are major planning instruments used in the NPAR Program:

Multi-year - The NPAR Program Plan, NUREG-1144. This plan was

first issued in (Morris and Vora) 1985. Revision 1 was issued in

(Vora) 1987. This version, Revision 2, was issued in 1991.

Annual/Monthly/Periodic

- Program assumptions/program briefs

- Form 189s

- Program reviews

- Programmatic discussions at the individual laboratories and/or

at NRC Headquarters

- Monthly reports

- Users' needs

- Direction from Advisory Committee on Reactor Safeguards

- Recommendations of Research Review Groups

- Interactions with EPRI and DOE

- Interactions with outside institutions and organizations,

domestic and foreign.

1.9
. 2.0 BASIS FOR THE NUCLEAR PLANT AGINGRESEARCH PROGRAM

. The U.S. Nuclear Regulatory Commission (NRC) is responsible for ensuring

that nuclear-power plants are operated safely during-their initial licensed

period and continue to operate safely during extended operation. *To help fulfill this responsibility, the NRC instituted the Nuclear Plan Aging Research

(NPAR) program to understand how aging may affect safety-related systems,

structures, and components (SSCs),.to identify!the measures available to

manage age-related degradation,-'and to anticipate problems that may result

from plant aging. A set of technical safety issues has been developed for the

NPAR program to provide focus and direction for aging assessments of SSCs and

to maintain safe plant operation.'JThe'basis for developing'the technical

safety issues is described,' followed by 'anhoverviewcof their implementation.

2.1 BASIS FOR IDENTIFYING AND RESOLVING TECHNICAL SAFETY ISSUES

-The Technical Integration. Review:Group for-Agi.ng and Life Extension

(TIRGALEX) was established in.-1986 by the Executive Director for Operations

(EDO) to facilitate the planning and integration of'NRC plant aging and

license renewal/life extension activities. The initial objectives of TIRGALEX

were to clearly define the technical safety and regulatory policy issues associated with plant aging and extended operation and develop'a plan for'resolving the issues in a timely, well-integrated, and effective manner.

The framework recommended by TIRGALEX for planning and integrating agency

activities related to:plant'aging and license renewal (LR) was adopted in the

NPAR-Program. There has been an.ongoing effort'to'coordinate the various .NRC

initiatives that are relevant tobplant aging:' first through TIRGALEX, subsequently through-the Aging and Life Extension Coordinating Committee (ALEXCC),

and currently through the NPAR Review Group.

2.2 TECHNICAL SAFETY ISSUES

A set of technical safety'issues was-developed by the TIRGALEX to provide

focus and 'direction for'the NPAR Program. -These issues are based 'on :operating

experience,' expert judgment,' and risk significance. ;These technical 'safety

issues-must be addressed to.-ensure that safety levels are maintained as the;

service lives of the present',generation of reactors increase. ''-]

* What SSCs are susceptible.to aging-effects that could adversely.! - affect public health and-safety? Which of these SSCs are main- '

s tained and are replaceable? ' -

What are the degradation processesiof-materials, components, and

structures that could, if improperly maintained.and/or.not replaced,

affect safety during normal design life and during extended life?

2.1
* How can operational readiness of aged SSCs be ensured during the

40-year design life and during extended life of a reactor?

* Are currently available nondestructive examination and test methods

adequate to identify all relevant aging mechanisms before safety is

affected? If not, what efforts are under way to improve the

methods.

* What criteria are required to evaluate residual life of SSCs? What

supporting evidence (data, analyses, inspections, etc.) will be

needed?

* How should SSCs be selected for comprehensive aging assessments and

residual-life evaluations? Which SSC(s) should be selected?

* How effective are current programs for mitigating aging (e.g.,

control of environment, maintenance, replacement, and repair)?

* What kinds of reliability assurance and maintenance programs will be

needed to ensure operational readiness of aged safety systems and

components?

* What additional changes will be needed in codes and standards to

address aging? What schedule should be followed?

The technical safety issues were developed by first examining the nature

of the aging process and then examining the potential role aging plays in

plant safety, and the NRC's mission to address plant aging and extended operation and LR. The technical objectives of the NPAR program have been developed

to address the technical safety issues. Together, the technical objectives

and the technical safety issues provide the framework required for developing

and guiding the individual aging assessments in the NPAR program.

2.3 IMPLEMENTATION

In accord with the TIRGALEX Integration Plan, the technical data currently developed in related projects and the regulatory needs identified by

the Office of Nuclear Reactor Regulation (NRR) establish the priority of the

NPAR Program activities. The Office of Nuclear Regulatory Research (RES) has

the responsibility for carrying out the necessary research programs.

Hardware-oriented engineering research needed to resolve the issues

related to aging is being conducted in the Division of Engineering under three

programs. The NPAR Program for aging investigations on components and systems

is being performed by the Electrical and Mechanical Engineering Branch of the

Division of Engineering. The aging research program on the vessels, piping,

steam generator, and nondestructive examination techniques is being performed

by the Materials Engineering Branch of the Division of Engineering. The aging

research program and civil structures is conducted by the Structural and

Seismic Engineering Branch of the Division of Engineering.

2.2
As the principal activities in these research programs are completed, the

research results are made available for use in the regulatory process. RES

also makes use of research findings as they contribute to the developing

regulatory criteria, guides and standards, and review procedures.

2.3
3.0 AGING ASSESSMENTS OF SPECIFIC COMPONENTS AND SYSTEMS

UTILIZING THE NPAR PROGRAM STRATEGY

Aging assessments have been conducted under the'Nuclear Plant Aging

Research (NPAR) Program on approximately 40 categories of components and systems considered risk significant, including a cross section of representative

electrical and mechanical components'and- systems. The selection of the components and systems has been based on criteria outlined in Section 3.1. After

the selection process was completed, participating laboratories and contractors were assigned specific components and systems for aging assessments (see

Figure 3.1). In all, five national laboratories and several private institutions and organizations are conducting aging assessments, which are listed in

Appendix A.

The aging assessments are based on a phased approach identified as the

NPAR Program Strategy, which is illustrated in Figure 3.2. The NPAR Program

Strategy involves the following phases:,-

* Pre-Phase I - Prioritization of systems, structures, and components

* Phase I - Interim Aging Assessment

* Phase II - Comprehensive Aging Assessment

* Utilization of Research Results.

The processes for selecting components and systems and applying the NPAR

Strategy to the aging assessments are'discussed in the following sections.

Approaches to understand and manage aging are discussed in Appendix B, including descriptions of key degradation mechanisms.

3.1 PRE-PHASE I --PRIORITIZATION OF SYSTEMS, STRUCTURES, AND COMPONENTS

Before Phase I of the NPAR Program was implemented, the specific structures, systems, and components (SSCs) to be studied were selected for in-depth

engineering studies. The selection criteria were based on the potential risk

contributed from failures of SSCs; experience obtained from operating plants;

expert judgment on the tendency of age-related degradation; and user a needs.

Studies by Vesely et al. (1983) and Davis et al. (1985) provided risk-based

inputs to the selection process. User needs include resolving generic safety

issues, revising regulatory documents, upgrading plant maintenance and surveillance, and developing technical bases for license renewal (LR)

consideration.

(a) In the U.S. Nuclear Regulatory Commission (NRC), the user is the' Office

of Nuclear Reactor Regulation (NRR).

3.1
Farmer

ORNL

Valves-MOV. CV, AOV, SOV. PORV

Instrument and Control

Reactor core Internals

Turbine Drives, Governors for Pumps

Main Steam Isolation Valves

AFW System

CRD (BWR)

AFW Pumps

Compressors

U-#ott<:nr

Bums

PNL

Snubbers AD System

Diesels SLC System

Shippingport Chillers

SWS System ESF

Aggarwal

8NI

Motor Control Centers

RCICS

Reliability Modeling

Inspection Integration

Chargers / Inverters

CCW System

CRD (PWR)

Instrument Air

Circuit Breakers I Relays

Bistables / Switches

RHR System / Low Pressure ECCS

Containment Cooling System

BOP Systems

Motors

CW Cable 

Farmer NIST

Diagnotic

Farmer

SNL

Aging - FIRE

Cables

Connectors

I Penetrations

Ri k- a edPVora

NPAR

Coordinator

NPAR Integration

/~~~~G s

/ / R~~~eg. G~uides

/ / ~~~Tech. Specs.

/ / ~~~SERS 8 SRP

/ | ~~~IAEA NASA

/ | ~~~NUMARO FMAA

/ ~~~Good Practices for Aging Managemne

/ ~~~Licenso Renewal and Mainlenanca Ru

/ \ R~~~~~isk-Based Priorilization

/ \ ~~~~~Regulatory Inst. Review

t \ ~~~~~Rocord Keeping

N

Aggarwal

K ~~WYLEBreakers

Relavs

Weidenharner

INFI

Transformers, Batteries

High Pressure ECCS

RPS System

Class I E System

Residual Life Assessment

Risk Gvaluation of Aoing

Weidenhamer

SAco

Risk Sianiticance of Aaina

Fe

les

FIGURE 3.I. NPAR Structure
NPAR Program Strategy

PRE-PHASE I -

PRIORITIZATION

OF CS

PHASE I - INTERIM AGING

ASSESSMENT OF CS

Select Candidate

Components

and Systems

(Deterministic)

Develop . - .

Aging-risk .

Methodology

(Probabilistic).I

Prioritize

Components -

and Systems -

Hybrid

Approach.

Deterministic and

Probabilistic

. Evaluate operating experience, available

data for CS (e.g., IPRDS, NPRDS, and LERs.

Plant-Specific Data)

* Review and analyze available

Information from

* Designs

. Specifications

* Operational parameters

* Ongoing research

* Utilize knowledge of experts and

Industry practices

* Characterize aging mechanism for CS

Peer Review

Continued

.

('3

; .

L;

.. ..

* Revlew ISMM for CS

v Current methods

; Current technology

* Interact with key NRC staff

CS = Components Systems

39102078.2

FIGURE 3.2. NPAR Program Strategy
NPAR Program Strategy

PHASE II - COMPREHENSIVE AGING

ASSESSMENT OF CS - OPTIONAL

APPROACHES

UTILIZATION OF

RESEARCH RESULTS

1st Option - Testing and Evaluation

* Conduct In situ assessments

* Test naturally-aged components,

component models, and samples

with simulated degradation

* Verify Improved ISMM

L Develop recommendations

for ISMM

Per

Conduct

detailed

analysis and

evaluation

Conduct

detailed

analysis and

evaluation

V Develop application guidelines

* Interact with codes and standards

committees

* Interact with Industry

* Interact with NRC staff

Ner Revlew

2nd Option - Expert Approach and

Field Evaluation

Systematic Interactions with:

* Experts/consultants

* Vendors and manufacturers

* Utilities and field evaluations

* Other equipment users

* Codes and standards groups

Apply results of research

. Resolve GSis

. Develop technical bases for license renewal

* Develop guidelines for ISMM

. Develop guidelines for service-life predictions

. Formulate recommendations for standards

and guides

* Dissemination of technical research

* Mothbalilng, reactivation of plants

* Advanced LWR

L . Develop recommendations

for ISMM

ISMM = Inspection, Surveillance, and Monitoring Methods

CS = Components Systems 39102078.3

FIGURE 3.2. (contd)
NPAR Program Strategy. . q ".

PHASE I ACTIVITIES

1. Preliminary Identification of susceptibility

of materials to age-related degradation.

2. IdentificatIon of stressors and related

environmental factors that could cause

age-related degradation.-

3. identlficatlon 'of degradation sites.

4. Identification'of falfure modes and causes.

.5. Development bf'functlonal parameters to

- Indicate present status and aging

'progression. ; '

6. Evaluation' of current'Inispection, surveillance,

- and monitoring methods.

7. Evaluation of current maintenance practices.

8. Recommendations for Phase 11 studies.

PHASE 11 -ACTIVITIES

1. In situ assessment of naturally-aged CS.

2. Tests of naturally-aged components or

tests of components.with simulated aging.

3. Laboratory or In-plant verification of

- 'methods of Inspection, monitoring, and

surveillance.- " " -

4. Verification'of methods of residual'> -

service-llfe evaluation.'

5. Development of recommendations for

Inspection or 'monitoring methods and

periods. -

6. Recommendations for effective maintenance

practices. ' :

7. Recommendations for acceptable methods

to determine residual life of equipment.

39102078.4

en

FIGURE 3.2. (contd)
The major constituents of the SSCs were selected on the basis of the

safety criterion that the release of fission products that could occur during

an accident should be contained within the plant (Davis et al. 1985; Vesely

1987; Muscara and Serpan 1985; Vagins and Taboada 1985); these constituents

include the pressure boundary components, containment, and supporting structures. Also included are components related to reactor control systems and

reactor safety systems. The reactor internals are included because their

failure may prevent control rod insertion or may cause fuel failure. Detailed

evaluations of reactor pressure vessels, reactor coolant piping and safe ends,

and steam generators are being performed in the program sponsored by the Materials Engineering Branch of the Division of Engineering.

The scope of NPAR aging assessments encompasses 1) pressurized-water

reactors-(PWRs) from the three U.S. nuclear steam supply system vendors,

2) boiling-water reactors (BWRs), 3) plants with numerous variations in

design, applications, and suppliers, and 4) operation and maintenance with

differing practices and philosophies. The NPAR Program includes studies of

selected electrical and mechanical components and representative safety

systems and support systems, special topics studies, and the utilization of

results from aging assessments.

NOTE: It is not the intent of the NPAR Program to conduct in-depth

engineering evaluations of aging and defect characterization, and methods for inspection, surveillance, and monitoring methods and mitigation practices of all significant

plant elements. It is the industry's responsibility to

characterize and evaluate their own plant SSCs to ensure

their operational safety as the plants advance in age.

Table 3.1 contains a listing of components that have been identified to

have aging-related impacts on plant safety and support systems and their

availability and safety margins. Current status of work on the study of the

aging of these components is also noted.

Table 3.2 provides the research status of systems in nuclear power plants

(NPPs) that are of current interest. These systems are considered important

for accident prevention or mitigation.

3.2 PHASE I - INTERIM AGING ASSESSMENT

After the components and systems are selected, the first step is to

establish a boundary to define what is to be included in the component or system under consideration and also to identify important interfaces between the

component or system to be investigated and other adjacent components or systems. The Phase-I assessment is based on information from public and private

databases, vendor information, open literature, utility sources, operating

experience, and expert opinion. The information includes the identification

of failure modes, preliminary identification of failures caused by age-related

degradation, and reviews of inspection, surveillance, and monitoring methods

3.6
TABLE 3.1. Components of Current Interest in-the NPAR Program

and Their Completion Schedule :-

Topic

Motor operated valves

Check valves

Solenoid valves

Air.operated valves

Auxiliary feedwater pumps

Small electric motors*

Large electric motors

Chargers/inverters

Batteries

Power operated relief valves

Snubbers

Circuit breakers/relays

Electrical penetrations

Connectors, terminal blocks

Chillers

Cables

Diesel generators

Transformers

Heat exchangers

Compressors

Bistables/switches

Main steam isolation valves

Accumulators

Surge arrestors

Isolation condensers (BWR)

Purge and vent valves

Safety relief valves

Service water and component

cooling water pumps I '

Laboratorv

ORNL

ORNL

ORNL

ORNL

ORNL

ORNL

BNL

BNL

INEL

ORNL

PNL

BNL, Wyle

SNL

SNL

PNL

SNL

PNL

INEL

ORNL

ORNL

BNL

ORNL

Schedule

Complete in FY-91

Complete in FY-91

Complete in FY-91

Initiate Phase 1 in FY-91

Complete-in FY-91

Completed in FY-88

Initiate Phase I in FY-92

Completed in FY-90

Completed in FY-90

Completed in FY-89

Complete Phase 2 in FY-91

Complete Phase 2 in FY-91

Complete Phase 1 in FY-91

Initiate Phase 1 in FY-91

Initiate Phase 1 in FY-91

Complete Phase 2 in FY-91

Phase 2 completed in FY-89

Complete Phase I in FY-91

Complete Phase 1 in FY-91

Phase 1 completed in FY-90

Initiate Phase 1 in FY-91

Initiate Phase 1 in FY-91

No initiative

No initiative'

No initiative

No initiative

No initiative

No initiative

3.7
TABLE 3.2. Systems of Current Interest in the NPAR Program

and Their Completion Schedule

Topic

High pressure emergency core

cooling system

RHR/Low pressure emergency

core cooling system

Service water

Component cooling water

Reactor protection

Class lE electric distribution

Auxiliary feed water

Control rod drive, PWR (W)

Control rod drive, PWR (B&W, CE)

Control rod drive, BWR

Motor control centers

Instrument air

Containment cooling

Engineered Safety Features

Instrument and control

Automatic depressurization

(BWR)

Standby liquid control (BWR)

Core internals

Turbine main generator

and controls

Containment isolation

Recirculation pump trip

actuation instrumentation

(BWR)

Reactor core isolation cooling

Laboratory

INEL

Schedule

Complete Phase 1 in FY-91

BNL Complete Phase 2 in FY-91

PNL

BNL

INEL

INEL

ORNL

BNL

BNL

ORNL

BNL

BNL

BNL

PNL

ORNL

PNL

Phase 2 completed in FY-90

Complete Phase 2 in FY-92

Complete Phase 2 in FY-91

Complete Phase 2 in FY-91

Initiate Phase 1 in FY-91

Phase 1 completed in FY-90

Complete Phase 1 in FY-92

Complete Phase 1 in FY-91

Completed in FY-89

Complete Phase 2 in FY-92

Complete Phase 1 in FY-91

Initiate Phase 1 in FY-91

Complete Phase 1 in FY-92

Complete Pre-Phase 1 in

FY-91

Complete Phase 1 in FY-91

Initiate Phase 1 in FY-91

Initiate Phase 1 in FY-91

PNL

ORNL

ORNL

No initiative

No initiative

No initiative

(ISMM), including manufacturer-recommended surveillance and maintenance practices. Performance parameters or condition indicators potentially useful in

detecting degradation are identified, and preliminary recommendations are made

regarding improved ISMM and maintenance practices. Recommendations are also

developed to identify the detailed engineering tests and analyses conducted in

Phase II.

3.8
A key outcome of.the Phase-I investigation is a decision regarding the

need for. a Phase'-II assessment on' specific' SSCs. The following sections summarize the'activities that comprise the Phase-I in'ter'im investigation.

3.2.1 Consultations with Nuclear-Reaulatorv Commission Staff

An important Phase-I consideration is to identify where similar aging

research or-related regulatory initiatives.are active in the NRC while avoiding duplication..'--Aging'-research'sho'ul'd potentially-augment the value of the

related research and.regulatory'in'itia'tives. .For'example, an NRR study on

service water system.(SWS).biological,.fouling provided important input to an

NPAR SWS'aging investigation. Identifying related regulatory initiatives,

such as the.resolution of.generic'-"sa'fety.is'sues or'needed revisions to regulatory'guides or technical specific'ations',;;'can determine how NPAR aging

research may, contribute: to resolving regulatory needs..

3.2.2 Review of Design Information and Applications

During th'e'preliminary.research,'design information and-applications are-..

reviewed as follows: ' ; -

1. Design and Specifications. ;The first-ofthe three elements of the

Phase-I assessment begins with-a review of the'design -data and specifications for the hardware being:studied. This includes.such items

as nonproprietary design documents,;final.safety analysis reports,

operating and maintenance manuals,..and product literature. Addi-.

tional sources of information are also reviewed. These include . vendor surveys, utility contacts,-published- reportts,.and expert,-

: opinion. .. . - - .::--- :J

-.2.. Materials. An important aspect of Phase Ilis to identify all susceptibilities of.significant:materials that comprise.the hardware

under review. The susceptibilities of specific materials and parts

to age-related degradation are.evaluated.- - -

3. Operating and Environmental Stressors. The degradation of the SSCs

involves time-dependent:phenomena any, among other things, depends

on operating environment' and operating-history. The environmental

stressors considered include temperature,-radiation,';chemicals,-.:

contaminants, atmospheric conditions, and humidity! The-environ-.

mental conditions considered include operational parameters and

environmental conditions that prevail during other periods, such as

during testing,-shutdowns,;storage jperiods,. accident conditions, and

post-accident situation..

The operating history a'ssessment'includes 'thee thermal-, mechanicall,

and electrical stressors that SSCs experience during their operating

lifetimes.:.'-Normal operating'conditionsarnticipated transients -.

off-normal-condit'ions,-and accident:and post-'accident conditions:.are

studied to determine the influence'and effects of stressors'and

13.9
environment on degradation processes. Typical examples of electrical stressors include slow-switching transients, fast transients

of the lightning variety, and low frequency 50- to 60-Hz signals,

which can occur singly or in various combinations. Examples of

mechanical stressors are static loading stresses, dynamic loading

stresses, and seismic and vibrational stresses.

4. Performance Requirements and Condition or Functional Indicators.

Performance requirements are reviewed to assess whether aging

degrades the ability of SSCs to perform required safety functions

during normal, abnormal, and accident conditions. Here it may be

possible to identify condition or functional indicators. These

consist of indicators that are practical to monitor and that provide

cost-effective means to identify and manage degradation. Finally,

ongoing research is reviewed, and applicable results are included in

the assessment of hardware under study.

This review and analysis of materials, designs and specifications, stressors and environment, and operational parameters are performed on all components and systems selected for the comprehensive aging assessment.

3.2.3 Survey of Operating Experience and Failure Evaluation

Another activity of the Phase-I assessment is a critical survey of the

documented operating experience obtained on SSCs being evaluated. This review

provides information on the failure rates and reliability that can be

expected, and the aging-related failure modes and causes that have been

experienced. The sources of this information include data from the in-plant

reliability data system (IPRDS) sponsored by the NRC and the nuclear plant

reliability data system (NPRDS) managed by the Institute of Nuclear Power

Operation. Other sources include Licensee Event Reports (LERs), Nuclear Plant

Experience, Plant Maintenance Records, aqd Inservice Inspection Reports.

From the critical survey, hardware failures are evaluated to identify the

following:

* Failure Mechanisms. These are established through the process of

identifying dominant stressors; studying materials and designs of

components and parts; reviewing service environments and applications; and then evaluating the nature of, and the factors contributing to, aging and failures.

* Failure Modes. The indicators of failures (for example, leaks,

fatigue failures, radiation, thermal embrittlement of materials, and

voltage collapse or disturbance in current signature) are assessed

and critical age-related failure modes are identified.

* Failure Causes. The conditions of design, manufacture, service

environments, and applications that may lead to failures are

determined.

3.10
3.2.4 Review and Evaluation of Inspection, Surveillance, and Monitoring

The third element in the. Phase-I investigation is the review of the

inspection, surveillance, and monitoring methods (ISMM). This element also

involves a review of artificial or accelerated aging techniques used to

qualify hardware for end-of-life performance.

Existing methods for ISMM are evaluated to detbrmine those methods likely

to be effective in detecting aging at an incipient stage before loss of safety

function. Also it-is important that the methods not be unduly'expensive to

implement and not result in unacceptable levels of occupational exposure.

Surveillance and monitoring methods being evaluated include periodic inspections, both visual and instrument-aided, and on-line 'instrumented techniques.

The evaluation seeks to-identify performance parameters and functional indicators that represent the capability'of equipment and are useful for managing

aging. Key considerations are to monitor the selected parameters'and'indicators at operating plants for reasonable costs.

Artificial or accelerated aging techniques may be reviewed and compared'

to data available from naturally-aged hardware to demonstrate the applicability of current practices.

3.2.5 Interim Aging Assessment and Recommendations

The result of a Phase-I evaluation-of SSCs is an interim aging assessment

and an evaluation of the safety significance of the-probable failure'modes

(Wu 1989). An interim evaluation-*of current ISMM technology is given, and

potential performance parameters and .'condition indicators are developed.

Interim recommendations are made-for'Phase-Il studies based on the results and

reviews of research activities-completed in Phase I. 'The results of the

Phase-I study are issued-in a technical progress report or a NUREG/CR-report.

Generally, NUREG/CR reports are subject-to peer review before publication.

Continuation to Phase-II activities,Jfor a given component or system, is

deemed unnecessary if 1) an adequate database and experience -exists'within the

industry;.2) industry-sponsored programs'adequately address the research

needs; and 3) resources need to-be'directed-to other'research activities.

3.3 PHASE II - COMPREHENSIVE!AGING ASSESSMENT

Phase-II assessments of SSCs usually involve-some combination -of one or

more of the following: 1) tests of naturally-aged equipment or equipment with

simulated degradation; 2) laboratory or in-plant verification of methods for

ISMM; 3) aging assessments by expert:panels; 4) development of recommendations for inspection or monitoring techniques, 5) verification-of methods for

evaluating:residual service lifetime; 6) identification of-effective'maintenance practices; 7) in-situ examination and data gathering for operating

equipment; 8) verification of failure causes using results from in-situ and

postservice examinations.

3.11
When fully conducted, the Phase-II assessment involves several in-depth

assessments, which include validating advanced ISMM through laboratory and

field testing of samples and validating accelerated aging techniques. It also

may include developing models to simulate degradation, in-situ aging assessment, and testing of naturally-aged equipment from operating NPPs.

In-depth laboratory investigation of selected naturally-aged components

has been an element of NPAR Phase II for certain categories of equipment such

as valves, motors, and battery chargers/inverters. However, detailed laboratory investigations on large components such as emergency diesel generators

(EDGs) are not practical at current NPAR funding levels. Therefore, the

Phase-Il assessment of such components has involved and will continue to

involve extensive use of experts and coordination with utilities and other

industry sources to identify aging characteristics. Budget limitations have

required developing innovative approaches to investigations of aging mechanisms without reliance on in-depth laboratory studies and studies on naturallyaged components. Alternative approaches to comprehensive aging assessments

that have been developed and applied in NPAR studies are described in

Appendix D.

3.3.1 Review and Verification of Improved Inspection. Surveillance, and

Monitoring Methods

The Phase-II research of ISMM involves reviewing advanced methods and

technology for each category of components and systems under study. In this

phase, advanced techniques and technologies, either in use or under development, are investigated. When available, the sources of information technology

both within and outside the nuclear industry are used. The sources outside

the nuclear industry include fossil plants, the petrochemical industry, the

aerospace industry, various branches of the Department of Defense, and other

government agencies. The practical feasibility of applying these technologies

to nuclear plant components is also explored.

Laboratory and field application and verification tests of ISMM candidate technologies are carried out. The objective of the tests is to demonstrate that methods are appropriate to follow the dynamics of the performance

parameters and functional indicators of interest; methods have adequate selectivity (will not give false indications) and sensitivity (will detect in the

incipient stage); and suitable acceptance/rejection criteria are available so

that maintenance needs can be correctly identified.

Some laboratory tests could involve simulating defects of varying degrees

of propagation in prototype hardware to determine sensitivity and detection

criteria. Various defect and environment combinations could be used to determine selectivity. These laboratory tests could be conducted to verify that

the methods are applicable for in-situ use at power plants. Field tests are

recommended at cooperating utilities to confirm the laboratory results, to

provide information about the frequency and method of data collection and

analysis, and to estimate cost effectiveness and practicality of application.

3.12
3.3.2 Testing of Naturally-Aged Components

.A second element..in the!Phase-II assessment is examining.and.te'sting

naturally-aged components.that have .had extended.service in NPPs. Obtaining

and testing naturally-aged components-is sometimes difficult or costly'; however, the information obtained from 1the testingjis valuable to quantify the

effects of aging and to determine'whether adequate''safety margins exist 'to

ensure the operational readiness of.naturally-aged components.and systems that

remain in service. -

Equipment that has experienced significantoperating and environmental

stressors was obtained from the decommissioned ShippingportPWR. Some examinations of.the naturally-aged Shippingport 'equipment:were conducted in-situ.

Other Shippingport'components and'specimen's were'shiipped to participating

laboratories for aging evaluations. Results from.the Shippingport aging

studies that have been reported are referr6d to'in'Section 4'.'3 Appendix E.

In-situ monitoring of operating'equipment at LWRs is recommended to gain

an understanding of the interaction between age-related defect characteri.zation and inspection, surveillance,' and maintenance.' Also, aging investigations are sometimes undertaken on equipment that has failed during operation.

Determining the root cause of'the-failure 'is the key 'considerati.on.

3.4 UTILIZATION OF RESEARCH RESULTS

With the completion of the aging assessment of.SSCs, ,a.technical ,basis

becomes available for use in developing'and refining the regulatory process.

Examples of possible uses include implementing improved ISMM; modifying present codes and standards; resolving generic'safety'issues;' and developing

guidelines and review procedures for LR.

In the NPAR Program Strategy (Figure 3.2), the research performed for the

Phase-I and Phase-II assessments.leads to developing application guidelines.

It may also provide a systematic collection ofthistorical baseline data and

trending information for evaluating component 'nd system aging effects. The

specific areas of application,'shown -in 'Figure 3.2, provide highlights of the

end uses of the NPAR Program'. 'An-important'part of 'the application guideline

phase 'is' the'technical integration of the'results`obtained.by NPAR :and other

major programs'.' In 'developing guidelines, -the NPAR staff work with all the

NRC offices involved in'programs'relevant to nuclear'plant aging and LR, with

codes and standards committees, and with industry groups.' --

3.4.1 Resolution of Generic'Safety Issues

The NRC report, NUREG-0933 (Emrit et al. 1991) contains a recommended

priority list to assist in the timely and efficient resolution of generic

safety issues.-' The NPAR Program generates -guidelines, -develops' criteria, and

supports resolution of thelgeneric'saffty -issues.- The.'generic'safety issues

that would directly benefit' from the'NPAR Program'results are listed in

Table 3.3.-

3.13
TABLE 3.3. Generic Safety Issues,(a) with Elements of Aging

Issue Number Title

15 Radiation Effects on Reactor Vessel Supports (Revised)

23 Reactor Coolant Pump Seal Failures

51 Proposed Requirements for Improving the Reliability of Open

Cycle Service Water Systems

87 Failure of HPCI Steam Line Without Isolation

113 Dynamic Qualification Testing of Large Bore Hydraulic Snubbers

130 Essential Service Water Pump Failures

142 Leakage Through Electrical Isolators in Instrument Circuits

143 Availability of Chilled Water System and Room Cooling

153 Loss of Essential Service Water in LWR's

B-55 Improve Reliability of Target Rack Safety Relief Valves

B-56 Diesel Reliability

SlIA License Renewal Rulemaking

S17 Electric Power Reliability

(a) From NUREG-0933 (Emrit, et al. 1991).

3.4.2 Considerations for License Renewal

An important objective of the NPAR Program is to identify and resolve the

technical safety issues involved in requests for LR of nuclear power plants.

An end product of NPAR initiatives will be guidance or recommendations to NRC

users on subjects such as revisions to ISMM methods, residual lifetimes of

major components, key technical information required in applications for LR,

ensuring continued safe operation of plants, and providing support to the LR

rulemaking process. This topic is discussed in Section 5.0.

3.4.3 Guidelines for Inspection. Surveillance, Monitoring Methods, and

Maintenance

A principal element in the engineering evaluation of age-related degradation is evaluating methods used in ISMM of NPP systems and components. A

review is made of methods currently in use in Phase I. Also, an evaluation is

provided of their effectiveness in detecting age-related degradation at an

3.14
incipient stage before a loss of safety function. Advanced techniques are

also reviewed, and may be tested in Phase II.

The results of the NPAR aging assessments are'also utilized to evaluate

the role of maintenance in mitigating age-related degradation and developing

guidelines for revised or preferred maintenance practices. This work consists

of the following activities: reviewing current-practices and procedures;

reviewing vendors' recommendations; evaluating the merits of performing

preventive, corrective, or predictive maintenance; identifying failures caused

by maintenance procedures; and developing recommendations for a preferred

maintenance approach.

3.4.4 Guidelines for Service-Life Predictions

Aging and residual life of major LWRcomponents is evaluated in the NPAR

Program. The current methods for predicting the service life of major mechanical components and structures are reviewed to develop an approach for life-,

time evaluation. The resources required to inspect and monitor components and

structures are considered to determine whether technically acceptable methods

for predicting service life-could be'substituted. The information generated

in this research has two principal objectives. One is to assist in developing

criteria that ensure that the degradation of major components does not impair

safe plant operation. The second is to generate a technical basis for establishing criteria and developing guidelines to be used in license review

procedures for LR. Although these objectives complement the objectives of

ongoing industry-sponsored pilot projects on LR, they focus only on safety

aspects.

The approach to the residual-life assessment involves identifying and

prioritizing major components with respect to safe plant operation. This is

followed by an initial effort to determine the life-limiting processes for

each of the major components, whereby degradation sites and failure modes

during normal operation and accident conditions are identified.

The residual-life assessment also includes current and potential methods

for ISMM. In the utilization phase, the work is focused on integrating currently available technical information relevant to aging during current operations and during LR. Further description of residual-life assessments of

major components is addressed in Section 4.4.

3.4.5 Recommendations for Standards and Guides

The NPAR Program develops recommendations for revising relevant industry

codes and standards for continuing the operation of aged plants. The NPAR

Program also provides input to technical bases for preparing NRC regulatory'

guides, review procedures for the continued operation of NPPs, and for LR

considerations.

3.15
3.4.6 Dissemination of Technical Results

The research information developed in NPAR is disseminated by preparing

technical papers (for example, journal articles) and reports (for example,

NUREG/CRs) and by sponsoring workshops, training seminars, society presentations, symposia, information exchange programs, and international (IAEA)

initiatives. To investigate approaches to make the NPAR technical bases more

accessible, two EDG reports (one sponsored by NPAR, one sponsored by DOE) were

compiled on computer disks and were cross-referenced using a Hypertext

approach. Wider use of the approach to improve accessibility of this database

is being considered.

3.4.7 Innovative Materials and Design

Another element in the NPAR Program Strategy (Figure 3.2) is the identification of needs for innovative materials and designs. Here recommendations

may be provided (and it is up to the industry to implement them) to evaluate

design changes to existing SSCs that would make them less susceptible to

aging. Innovative materials and designs could also find end uses in the other

NRC programs (such as equipment qualification and advanced LWR designs).

3.16
i4.o0 SPECIAL 'TOPICS -

The special topics, listed in Table 4.1 and reviewed in this' section,

constitute a major element of the Nuclear Plant Aging Research (NPAR) Program...

Most of the'research'on'these topics'-interacts''with'the'research 'on'the,' r

hardware-oriented initiatives discussed in'Section 3'.0.: For exam'plel; the

Shippingport aging evaluation has provided naturally-aged components to several of the systems and components studies., Most.of. the remaining topics

derive inputs from the hardware'studies', for example,-specific needs for

TABLE 4.1. Special'Topics 'obf'Current'Inte'rest in thetNPAR Program

and Their Completion Schedule i'''

* : Toi c ^ * - Laboratorv Scheduie

Risk evaluations of significant aging effects .

Risk-based component prioritization and selection ;

- ~~~~~~ **. ..,.

Shippingport aging assessment '_

Aging assessment and mitigation of major LWR components and

structures

Development of technical bases for:

* Regulatory guides ;

* Standard review plans -

* Criteria for license renewal

* Maintenance rulemaking

* Manual for aging management * :

Technical specifications from aging perspective

Guidance for recordkeeping

Inspection integration

Reliability modeling of component aging

Review of regulatory instruments .

Review of industry' technical reports for-license renewal -- i

Aging Effects in Balance-of-Plant Systems

Maintenance to manage aging

.4

SAIC . . Complete.Phase lin FY-90

PNL Coml s 1 in FY-90 ... . I ....., ,,.-.

PNL 'Complete'in-FY-90

-NEL Complete Phase I in FY-90

PNL. Complete Phase .inFY-91

PNL ; Complete Phase 1 in FY-91

PNL ' Complete Phase 1 in FY-910

PNL, Continuing

PNL Complete Phase 1 in FY-91

PNL Complete Phase .in FY-9i

PNL' Complete Phase 1 in FY-91

PNL Complete Phase 1 in FY-91

BNL Complete Phase. in FY-91

.PNL Complete Phase 2 in.FY-91

P(Al) Complete in FY-9 1

;I . .~ - , . .

BNL - On Hold

- PNL r Complete Phase I in FY-90

'SNC, Complete in FY-p9 F I , .Aging degradation of cables. - .

* . .. ...

Fire vulnerability of aged electrical components:'

Aging effects in mothballed PNL plants

.SNLPNL

I -

Complete Phase I in

Planned for FY-92

FY-90

4.1
changes to the technical specifications and nuclear power plant (NPP) recordkeeping are identified in part from the NPAR aging studies identified in

Section 3.0. The special topics also are providing important input to NRC

license renewal (LR) initiatives.

The scope and status of the special topics identified in Table 4.1 are

summarized in the following sections.

4.1 RISK EVALUATIONS OF SIGNIFICANT AGING EFFECTS

Objective: To develop and demonstrate an age-dependent PRA-based methodology for prioritizing the risk contributions to and maintenance importances of aged components.

Contractor: Science Applications International Corporation (SAIC)

Methodology: A methodology has been developed that takes into consideration the aging contributions to core melt frequency and to other risk

results by a Taylor expansion of the basic risk equations. This approach

allows current probabilistic risk assessments (PRAs) to be used to determine

the risk importance that can then be combined with aging effects on components

and structures, which are determined using separate aging models. Using this

approach, PRAs do not have to be recalculated to incorporate the aging

effects; instead, current PRA results can be combined with aging effects of a

component and structural level to obtain the core melt frequency increases and

other risk result increases caused by aging. In this methodology, multiple

aging effects are incorporated, aging contributors are prioritized in detail,

and maintenance programs can be quantified for their risk effectiveness in

controlling aging effects. Uncertainty evaluations are also included.

Status: The methodology has been tested and demonstrated. A NUREG/CR

has been issued (Vesely, Kurth, and Scalzo 1990). Software programs have been

initially developed for applications of the methodology. Data analysis

programs have also been developed to analyze data for aging effects that apply

to the risk evaluations to determine the risk implications of age-related

degradation.

Results: Three recent PRAs have been used to evaluate the risk effectiveness of test and maintenance programs using the linear aging model.

Analyses have also been conducted to support regulatory analyses for LR

rulemaking.

Plans: Plans are to extend the methodology to include more comprehensive models of aging effects, test procedures, and maintenance programs!

Further plant-specific data are to be analyzed, and production schemes are to

be developed. Possible applications to assist regulatory decision-making will

also be investigated.

4.2
4.2 RISK-BASED COMPONENT PRIORITIZATION'-AND'SELECTION

Objective: To develop and demonstrate methodologies and procedures for

implementing a hybrid (PRA-based and deterministic) prioritization of aged

components and structures.

Contractor: Pacific Northwes't-Laboratory (PNL)

Methodologiv: 'The PNL hybrid approach seldcts and prioritizes all

important-to-safety components'based on-risk contribution resulting from

component risk importance and aging effects. The hybrid approach accomplishes

this by combining the active-components's'elected by an-age-dependent PRAbased methodology with passive components and structures (SC) not included in

the PRA. PNL's approach will seek to incorporate'the SC not included in the

PRA by using the results from the PRA as input'to an expert-panel deterministic method, the Analytical Hierarchy'Process (AHP). 'In this approach, the

risk importance for'active'components'-from-the PRA are used to provide benchmarks for the AHP values for thdese'components. The'relative risk importance

of the passive components and structures would the'n be 'derived from'their

hierarchy relative to theactive components'by'thle expert panel. 'The effects

of aging on the'failure probability of each passive SC will then be calculated

using inputs on failure rate, effective test-interval, and effective renewal/

replacement interval, as was previously done for the'PRA-based active

components. The resultant aging failure probability of each SC will be combined with its risk importance to produce the resultant'risk contribution'from

age-related degradation. -

Status: The PRA-based component prioritization methodology incorporating the effects -of-multiple component aging was developed. This PRA-based

methodology was' used -to prioritize active 'co'mponeh'rtsiusing two NUREG-1150

PRAs, one each for a PWR and BWR.'-- From'the prioritization, high-priority components were identified and selected 'for improved maintenance, which was

implemented through'redyctions-in the inspection and renewal intervals. A

report on this'effo'rtaJ is undergoing peer review at'NRC before its anticipated publication by the end of 1991.

Results: The PRA-based studies provided two important perspectives on

aging effects. 'The first is 'that aging'Iof plant components may result'in a

significant-increase to plant risk;.-the amount of increase may be plantspecific. The results of incorporating the'effects of multiple component''

aging demonstrated that aging may be viewed-as an important common cause of

component failure. Integrating single and multiple components in a prioritized list of components also demonstrated the importance of developing more

comprehensive PRAs to incorporate all important contributors to plant risk.

(a) Levy, I. S., W. E. Vesely. Dr'aft September 28, '1990.' 'PRA-Based

Prioritization of Risk Contribultions''and Maintenance Importance for Aged

Active Components. -NUREG/CR-5587, Pacific Northwest Laboratory,

Richland, Washington.

4.3
Plans: Completion of the hybrid component selection methodology is

expected by the end of 1992.

4.3 SHIPPINGPORT AGING ASSESSMENT

Objective: To use data and naturally-aged components and specimens from

the decommissioned Shippingport pressurized-water reactor (PWR) to evaluate

aging effects, resulting from reactor service, on components and systems being

investigated in the NPAR Program and by other NRC organizations.

Contractor: Pacific Northwest Laboratory (PNL)

Methodology: As the Shippingport PWR was decommissioned, the aging

assessment was conducted by 1) evaluating the relevance of Shippingpovt to

operating reactors; 2) in-situ measurements on circuits and materials;

3) removing samples from selected components; 4) acquiring components and

systems for use in studies at participating laboratories; a1 and 5) acquiring

operating data, manuals, etc., to accompany equipment items that were removed

for investigation by NPAR and other NRC organizations.

Status: Over 200 components (e.g., battery chargers, valves, a motor

control center) and specimens (e.g., from neutron shield tank, concrete, cast

stainless steel components, piping) were removed and sent to participating

laboratories; all buildings and equipment have been removed and the decommissioning is complete; PNL continues to provide assistance with data and information on the components and specimens that have been acquired.

Results: Data from examining the Shippingport components and specimens

are providing characteristics of thermal and radiation embrittlement (by ANL),

aging characteristics of inverters/battery chargers (by BNL), check valves and

motor operated valves (by ORNL) and radiological assessments from coolant system samples (by PNL). A cable sizing problem on a motor operated valve was

identified (by INEL), resulting in issuance of NRC Information Notice

No. 89-11.

Plans: Investigations are continuing on components and specimens from

Shippingport; PNL will continue to provide records support while the investigations are in progress. Attention will be given to proper disposal of

Shippingport equipment used in the investigations.

(a) Elements and participants in the Shippingport aging evaluation: - Coordination PNL

- In situ assessments ANL, INEL, PNL - Components BNL, INEL, NIST, ORNL, PNL - Samples of materials ANL.

4.4
4.4 AGING ASSESSMENT AND MITIGATION OF MAJOR LWR COMPONENTS AND STRUCTURES

Objective: To develop an understanding of the aging effects of major

light-water reactor (LWR) components and structures so that the-effects can be

effectively managed. The major'emphasis of this project is on integrating,

evaluating, -and updating the research' results from 'related programs, including

those sponsored by the NRC and industry. -.The five specific objectives of this

project are to 1) identify and evalu'ate technical safety issues related to"

plant aging and LR; 2) evaluate and define surveillance and mitigation

methods; 3) recommend revisions of 'codes and standards to support aging

management and LR; 4) identify additional research projects'needed to resolve

the technical safety issues related to plant aging and LR; and 5) develop And

evaluate life-assessment'models and procedures for the major LWR components

and structures. - - - - - - E - -

Contractor: Idaho National Engineering Laboratory (INEL) -

Methodoloqv: '!The project consists o'f integrating, evaluating, 'and'-'

updating all the relevant technical information on degradation mechanisms,

nondestructive inspection techniques, and monitoring methods relevant to major

LWR components and-structures."':The 'major'sources of information include

reports from completed and ongoing NRC-and-industry.research programs; NRC

Information Notices, Bulletins, and Generic Letters; and the Nuclear Power

Experience database.. Contributions from several technical experts from-industry, universities, and laboratories are' sought to prepare'the project reports.

Several other experts are sought to critically review the reports. ''The proj-'-

ect results are being used to develop and modify regulatory guides, technical

specifications, Standard Review Plans, and'ASME Section XI Codes'and Standards

related to aging and LR. A five-step approach is used to accomplish the

project objectives: -- ' '

1. Identify and prioritize major components and structures according

to relevance to plant-safety. ;-

2. Identify for each component, degradation sites, mechanisms and

stressors;. identify potential 'failure modes,; and then evaluate

current inservice inspection methods. -

3. Assess advanced inspection,' surve'illance, and monitoring methods

that can be used to detect, 'size, 'and trend-age-related degradation. Evaluate current and emerging methods to mitigate aging ';

damage.

4. Develop life-assessment procedur'es;for LWR components and

structures. ' -

5. Support the development of technical criteria for' LR. ' -

Status: Qualitative aging assessments of'22 major components including

primary pressure boundary components, containments, emergency diesel~generators, and "cables :and, conne'ctors- have been'. completed.': This assessment marks'

the completion of the first two steps. The remaining three steps are

-4.'5
partially completed. For the third step, evaluation methods for materials

properties, monitoring techniques for acoustic emissions, and methods for monitoring fatigue are being evaluated. As part of the fourth step, draft

reports on the life-assessment procedures for PWR reactor pressure vessels,

LWR metal containments, PWR steam generator tubes, and LWR cast stainless

steel components are published. In the fifth step, the project personnel

provide input to the reports, which are being prepared by PNL for the

technical support of the regulatory activities related to aging and LR.

Results: Qualitative assessment has identified the technical safety

issues related to aging of major components and structures. The assessment

results show that the conventional inservice inspection techniques and procedures sometimes have questionable capabilities in detecting age-related

degradation and have inadequate capabilities in sizing aging damage. Emerging

inspection and monitoring methods have been identified for detection, sizing,

and trending of aging damage. Modifications of ASME Section XI codes and.

standards have been identified. The project has also identified several

methods to mitigate aging damage in the major components and structures.

4.5 DEVELOPMENT OF THE TECHNICAL BASIS FOR GUIDANCE TO SUPPORT REGULATORY

ACTIONS INVOLVING NUCLEAR PLANT AGING

Objective: To develop technical guidance needed to implement the

License Renewal Rule and to support future regulatory activities that address

aging in nuclear power plants.

Contractor: Pacific Northwest Laboratory (PNL)

Methodology: In developing the technical basis for regulatory guidance

and thiesupporting technical information, the approach has been 1) to collect

and organize all available information related to age-related degradation in

NPPs, and 2) to perform the integration and analysis required to apply this

information to regulatory needs. The NPAR Program has been a major source of

basic technical information for this topic. Additional sources include

reports from other NRC research programs, industry technical reports, and a

wide variety of reports that document results of domestic and foreign aging

research programs. The collected body of information is summarized to a

degree suitable for the required regulatory guidance. This information is

then organized in an appropriate format to address the requirements in the

proposed regulatory action, such as the LR Rule.

Status: Several products, some of which represent multiple revisions,

have been generated. These products include 1) a detailed criteria and

screening methodology for selecting SSCs important to LR; 2) criteria and

requirements for programs for understanding and managing aging in support of

LR; 3) the detailed technical basis for a regulatory guide on technical information requirements for LR applications; 4) a draft handbook that recommends

practices to manage aging in NPPs; 5) organizing the current body of NPP aging

technology in a format that corresponds directly with the Standard Review Plan

(NUREG-0800); 6) a survey of the major regulatory instruments, including codes

and standards, to assess adequacy with which they address age-related degra4.6
dation; and 7) a quantitative, deterministic ranking of safety-related PWR

systems and their constituent components in terms of safety significance and

aging effects on safety margins. ;Work continues on each of these activities.

Results: Guidance developed under this'task has contributed to a draft

regulatory guide (DG-1009) and.NUREG/CR 5562, which supplements the regulatory guide and updated versions of regulatory-instruments such as the Standard Review Plan. These documents represent the primary channels through

which results of the NPAR program and other research on aging phenomena will

be applied in the regulatory process. Anticipated products include the

following:

* the technical basis for a regulatory.guide that identifies the

technical information to be supplied as part of an application for

an operating LR; this regulatory guide will be needed to supplement

the License Renewal Rule (55 FR 29043).

* a comprehensive handbook reviewing information useful for managing . age-related degradation in NPPs; this handbook will supplement the

regulatory guide.with.detailed, technical.information presented in

a systems format that mirrors the Standard Review Plan

(NUREG-0800).

Plans: Development of the regulatory guide on technical content of LR

applications will proceed in parallel with LR rulemaking. The goal is to

issue a final regulatory guide.in-April 1992. 'The handbook will be complete

in December 1991 and will be updated annually or biannually.

4.6 TECHNICAL SPECIFICATIONS - FROM AGING PERSPECTIVE

Objective: To evaluate the lessons learned from NPAR aging studies that

have potential application to technical specification (TS) requirements.

Contractor: Pacific Northwest Laboratory (PNL)

Methodology:. -The TS.requirements will be evaluated from three perspectives: 1) their adequacy in consideringiage-related degradation mechanisms,'

2) as potential contributors to aging'(e.g., TS requirements for frequent fast

starts of diesel generators of.the loads imposed on auxiliary feedwater pumps

while testing in the pumping mode), and 3) as a means for detecting, 'trending,

and managing aging through surveillance testing. These evaluations will be

performed by reviewing, from an agingperspective, the-products of the NRC

Technical Specification Improvement program and identifying potential modifications that would account for aging. . This task will make recommendations-for

better testing and monitoring of the condition of SSCs and better detection of

incipient failures. The emphasis will be placed on "quality over quantity"

and frequency of testing.

4.7
Status: The revised TS have been subject to a recent comment period.

Some NPAR studies are at a point where recommendations for TS improvement can

be made. Other NPAR studies are in progress and identification of TS inputs

is premature.

Results: The NPAR program provides for component- and system-specific

identification of condition indicators and age-related degradation sites. The

results of the TS assessment will be useful to focus on the quality of tests,

such as detecting defects in the incipient or degraded state.

Plans: Key considerations are determining whether the TS effectively

address detection and management of age-related degradation, and determining

the extent that TS requirements contribute to age-related degradation as a

result of test procedures.

4.7 GUIDANCE FOR RECORDKEEPING

Objective: To determine if changes are needed in nuclear plant recordkeeping practices to support aging management and potential extended plant

operation. Such changes could affect the types of information collected, how

it is stored, and how it is used.

Contractor: Pacific Northwest Laboratory (PNL)

Methodologv: This task addressed the current status of nuclear plant

records systems and the recordkeeping implications of NPAR research and similar programs in the United States and abroad. Also considered were the capabilities of recordkeeping techniques and technologies to make changes when

needed. This research was conducted by reviewing the nuclear plant aging and

records management literature and through discussions with nuclear plant

records personnel.

Status: A draft NUREG/CR report(a) has been prepared for review; decisions concerning additional regulation needed in recordkeeping will be part of

the overall LR rulemaking process. Although not formally representing the

NPAR Program or the NRC, PNL staff participated in an IAEA Consultant's

Meeting in July 1989, held for the purpose of drafting preliminary guidelines

for recordkeeping needed to support nuclear plant life extension and LR.

Results: Plant records cannot resolve all the aging management issues.

Although nuclear plant recordkeeping is extensive, implementation at different

plants is highly variable and does not always fully support engineering, maintenance, and operational activities. Research on this task suggests that regulatory guidance and standards need to focus more on how recordkeeping can

(a) Dukelow, J. S. 1990. "Recordkeeping Needs to Mitigate the Impact of

Aging Degradation, Including During a License Renewal Period." (Draft),

Pacific Northwest Laboratories, Richland, Washington.

4.8
support safe plant operation. Details 'about information that shouldbe'collected to support aging management and potential life extension require careful plant-specific analysis.! Existing re'cordkeeping :techniques and technology

are adequate; but surveillance,-'monitoring, and testing techniques may not be

adequate to determine residual life for all SSCs.

Plans: Future work on this task'will include l)continuing review'of

the recordkeeping-implications of other NPAR aging assessments, 2) review of'

the treatment of recordkeeping needs in 'industry topicalrreports (prepared

under the auspices of NUMARC), and 3) inputfor the technical' basis for such

regulatory guidance, if:the NRC decides-to issue formal regulatory guidance on

recordkeeping 'needs ,for LR.'

4.8 THE USE OF NPAR RESULTS IN INSPECTION ACTIVITIES

ObJective: To explore the areas'in'which NPAR Program results could

enhance the NRC inspector's'understanding of age-related'degradation and contribute to a more effective inspection. Also, to identify the NPAR informal

tion that can focus inspections on those'components'and systems vulnerable to

age-related degradation, and to develop a method for integrating research

results into the inspection process.

Contractor: Brookhaven National Laboratory (BNL)

'Methodologv: -Several tasks have been-accomplished toward:integrating'

NPAR information into the NRC In'spection Program. These tasks-determined"

which NPAR information is relevant to the inspector, the'way in which the

NRC's Inspection Program addresses the performance of components and 'systems,

and the format in which the NPAR information can be presented so that it can

be readily accessed-and updated. ; ;-

;Status: A report has been issued describing the NPAR results'that can

enhance NRC inspection activities. Recommendations-are provided for communicating pertinent information to NRC inspectors.: These recommendations are

based on a detailed 'assessment of the 'NRC's 'Inspect'ion Program 'and feedback - from resident-and 'regional inspectors. -'NPAR report summaries andtinspection ' guides for aging have been completed for components and systems studied at'

BNL. - :. - " e -

Plans: BNL will continue td'work with the other NPAR-contractors and

with Office of Nuclear Reactor' Regulation/Inspection and Licensing Program.-

Branch totestablish a complete manual 'of summaries'and guides available to the

NRC inspectors. ' ' -" ' -

4.9 AGING RELIABILITY MODELING OF ACTIVE'COMPONENTS

Objective: - To''develop'mathemat'ical:models using'operatir'g experience'

data to characterize'the degradation rate' of a component with its 'age; rthese'

models will estimate 'the component failure rate from' this degradation rate 'and

calculate the effects of the component aging on the plant risks.

4.9
Contractor: Brookhaven National Laboratory (BNL)

Methodologv: The aging reliability modeling focuses on analyzing the

times of component degradation to model how the degradation rates change with

the age of the component. Similar trends are also developed for the failure

rates as they change with the age of the component. The methodology also

discusses the effectiveness of maintenance the applies to aging evaluations.

Status: The specific applications of this technique that are completed

include residual heat removal pumps, service water system pumps, and compressors. Plant-specific data on a particular component are preferred when compared to that from the nuclear plant reliability data system. The study also

discussed several limitations on the model and statistical tests for combining

data for similar components from other plants.

Results: Models for other active components will continue to be developed. The failure rates developed will be used in the PRA models for specific

plants to calculate risks caused by aging in these components. Finally, a

screening methodology for selecting and prioritizing components will be developed to improve the present aging mitigation programs.

4.10 REGULATORY INSTRUMENT REVIEW

Objective: To investigate how effectively regulatory instruments (documentation), such as regulatory guides, standards, specifications, and codes,

provide guidance for aging of major safety-related LWR components.

Contractor: Pacific Northwest Laboratory (PNL)

Methodology: Eight regulatory instruments were chosen for review: the

Code of Federal Regulations, the NRC Regulatory Guides, ASME Boiler and Pressure Vessel Code, generic safety issues, standard technical specifications,

American Nuclear Society standards, the Standard Review Plan and Institute of

Electrical and Electronic Engineers (IEEE) standards. Components for the

review were selected from the NPAR program's major LWR safety-related component BWR and PWR plants. Typical components chosen for review were the

reactor pressure vessel, primary piping, cables, emergency diesel generators,

containment and basemat, and selected pumps and valves; these components are

representative of mechanical, electrical, vessels/piping, and structural

facilities. The review technique consists of determining the principal aging

issues for each component, e.g., corrosion and fatigue, and then investigating the instruments to determine aging guidance with reference to the aging

issue chosen. Results of the reviews have been reported in tables that list

each component versus the instruments and aging issues.

Status: The review began in FY 1988. In March of 1989, a PNL report

was published entitled "Guide to Regulatory Instruments for LWR Reactor Pressure Vessels: Aging and License Renewal Considerations" (Werry 1989). In

FY 1990 a draft report, "Regulatory Instrument Review: Management of Aging of

LWR Major Safety-Related Components, Volume I" (Werry 1990), was issued for

the following components: the reactor pressure vessel, the primary piping,

4.10
the steam generator, the-pressurizer vessel, and the emergency diesel generator. During FY 1990, the instruments were'evaluated for-containment and basemat and cables.

Results: The results of the review have shown,4with some exceptionsfound in the IEEE standards, that the instruments included in this review do

not explicitly address aging or LR. However, aging management does exist

because safety-related design,'construction, and'operation are consistent with

principles needed to provide aging management and extended operation. The

results indicate that-irtevisions are-needed in the instruments.- -Principal

suggested revisions are-functional'criteria-that clearly define the'requirements'of aging management... Otherirevisions -are needed to enhance and encourage improvements in nondestructive examination (NDE) methods-and tools.

Plans: The instrument review will continue through-FY 1991.with the

investigation of selected pumps and valves. The reviews of these components,

together with the reviews of.-containment-and-basemat, will be published as

Volume II of NUREG/CR-5490.' Reviewstof NRC notices and bulletins will be

updated in-FY 1991-. ';! -:

It is suggested that revisions-to the regulatory instruments are needed

to fully address the aging aspects of nuclear power plants. Theseinstrument

revisions will require additional information and research. More conclusive

information is-needed in the following areas: -rate of time-dependent material

degradation'for SSCs,-e.g., rate of consumption of fatigue design life; ' evaluation-of-aging initiatives in industrial/codes and-standards for-applicable'revision; evaluation of replacemenV of:SSCs-as..an aging management

strategy.' !Research is also needed'forl'imp'roved-NDE methods to-evaluate'material properties,- including embrittlement, and'fracture'toughness, fatigue

strength, and evaluation of iitergranular stress corrosion cracking.

(Improved acoustic emission(a) is a strong;can'didate for improved material

assessment.) ' '

4.11- REVIEW OF INDUSTRY TECHNICAL REPORTS FOR LICENSE RENEWAL - . ! '

.Objective: Provide NRC with review'comments on the eleven Industry

Reports that identify'LR-issues for nuclear-power plant components. .'

*Contractors: ''All NPAR 'Laboratories. -

- Methodology: 'The reports are reviewed at:the laboratories, and comments

are compiled through-the NPAR'Program'Coordinator:for'transmittal to NUMARC. ; ,* -- - r -,. I ,-

... Status: See'Figure 5.2 for the-status oflinitial.reviews.- Further

iterations on'some of the NUMARC responsesjto NRC-comments may be requested..

- r'-~ ~ - i *. .-2 ,-'8,, * . 1 - ! . - -

(a) An ASME-Boiler and Pressure Vessel 'Code,' Section.-XI,'Co'de Case N-471, is

currently' in the "pending adoption" use statu's 'for acoustic emission

inservice inspection.

-4.11
Results: The NPAR participants have provided comments on all Industry

Reports (see Figure 5.2) to the NRC.

Plans: Further reviews will be conducted and additional comments provided to the NRC as requested.

4.12 AGING EFFECTS IN BALANCE-OF-PLANT SYSTEMS

Objective: To identify and characterize aging effects in balance-ofplant (BOP) systems on plant performance, to calculate the plant risk due to

systems aging, and to evaluate and assess the plant aging mitigation activities on BOP systems.

Contractor: Brookhaven National Laboratory (BNL)

Methodology: An aging assessment will be performed on the BOP systems

and components that are considered important for safe operation of a plant.

The study will analyze plant data pertaining to operating experience,

unplanned reactor scrams, and maintenance to characterize aging in BOP components. The component data will then be used to determine the reliability of

each BOP system.

Status: A preliminary evaluation of several other studies on BOP systems was completed. Unplanned reactor scrams and shutdowns reported in the

NRC graybook (NUREG-0200) were obtained in database files for further analysis. The data pertaining to Combustion Engineering plants were analyzed.

Similar efforts will be performed for the Westinghouse, Babcock and Wilcox,

and General Electric plants to characterize aging in BOP systems.

Results: The frequency of unplanned reactor scrams has been cited as an

indicator of safety performance. The main contributors to unplanned reactor

scrams caused by BOP systems are the feedwater, main turbine, main generator,

main steam (usually the steam bypass to the main condenser), and the condensate systems. The electrical distribution system and, less frequently, the

circulating water, service/instrument air, fire protection, and heating,

ventilation, and air conditioning (HVAC) systems are also frequent contributors. The electrical distribution system consists of the 120V ac power,

switchyard, large plant loads, dc power and control centers. At a component

level, the feedwater regulating valves and turbine-driven feedwater pumps are

frequent contributors, as well as the main turbine electro-hydraulic control

subsystem. Failures in the main generators are also important. Based on the

study completed for Combustion Engineering plants, the reactor scram/shutdown

rates initiated from support systems (such as the cooling water system; electrical distribution, other than the safety-bus; air systems; and HVAC, other

than the control room) increase with the age of the plant. Similar trends are

also observed for the turbine/generator systems.

Plans: In addition to analyzing the operating experience data, other

considerations involving potential effects of failures of non-safety-related

components will be analyzed. Plant-specific data relating to BOP systems will

4.12
be analyzed to calculate their aging effects on the plant risk. Plant maintenance activities will be assessed in the light of alleviating aging 'problems

that can stem from the BOP system/component degradation.

4.13 'MAINTENANCE TO MANAGE AGING

Objectives: To 1) identify how aging management principles can be'

applied to improve maintenance practices; 2) interact with the-ong6ing efforts

on risk-based prioritizationto identify plant-specific applications of.the

risk-based methodology as it applies to maintenance practices; 3),systemati-,

cally evaluate plant maintenance practices and'determine how these practices

need to be improved to effectively manage aging;'4) support'NRC initiatives,

including the Maintenance Rule, regulatory.guides, and LR activities.

Contractor: Pacific Northwest Laboratories (PNL)

Methodology: The following four-fold approach is used. 1) Results of

NPAR hardware tasks were reviewed to identify effective maintenance practices

and recommendations for'managing aging; detailed'guidance is~obt'ained'-from

NPAR'contractors'concerning how results of.'their componbent or system studies

should be translated into improved maintenance practices. .2) Surveys are.made

of maintenance practices in four.areas--Japanese nuclear, U.S. commercial

aviation, U.S. Air Force B-52 program, and the U.S. Navy Extended Operating

Cycle Program for nuclear submarines. 3) NPAR prioritization methodology is

used to identify maintenance intervals and other parameters that need to be

evaluated on a plant-specific.basis, including' working with one or more plants

to develop prototype maintenance program.improvements, based on aging con-.

siderations. 4) Information from the NPAR component/maintenance system

studies and from the plant-specific studies is used to identify.specific program modifications. (e.g., surveillance methods, intervals and trending) and to

ensure that plant maintenance practices.effectively address aging management.

Status: A draft report, Maintenance Practices to Manage Aging, A Review

of Several Technologies, has been prepared, which includes summaries.of maintenance practices in other technologies,. including recommendations on how

certain approaches may be useful to-the.nuclear industry. Preliminary discussions have .been held'between PNL.maintenance and prioritization staff to identify important areas for future interaction.

Plans': 'Proceed to 1)'compile and evaluate maintenance recommendations

emerging from NPAR systems/components'.studies;.?) conduct.,selected evaluations

of plant-specific maintenance data with.relevance to prioritization methodology; 3) systematically review current plant maintenance methods and practices

to identify areas where recommendations for changes-are needed.

4.14 AGING DEGRADATION OF CABLES

Objective: To'determine the life-extension potential of cable products

used in nuclear power plants and to determine the potential of condition,

monitoring for residual-life assessment of cables.

.4.13
Contractor: Sandia National Laboratories (SNL)

Methodoloy: Twelve cable products, representative of those currently

in use inside the primary containment of nuclear power plants, are being aged

and then subjected to simulated design basis accidents. Simultaneous thermal

and radiation aging exposure is used to age the cables to nominal equivalent

lifetimes of 20, 40, and 60 years. This is followed by a sequential accident

exposure consisting of high dose rate irradiation followed by a simulated

loss-of-coolant accident steam exposure. Condition monitoring measurements

performed during aging are assessed for their ability to predict cable degradation. Extensive monitoring during accident tests evaluates aged cable performance under accident conditions.

Status. Aging and accident tests of all cables have been completed.

Extensive electrical and mechanical properties measurements have been and are

being performed. A high temperature steam test, a submergence test, and a

test of unaged cables have also been completed.

Results: The results of the Aging Degradation of Cables Program have

shown thatra number of popular cable products might qualify for extended life

operation. However, some cable products failed during accident testing, while

others exhibited reduced insulation resistance that could be significant in

some applications. Some cable condition monitoring parameters could be useful

to assess remaining cable life.

Plans: Additional condition monitoring measurements will be performed

and the final reports for the experimental program will be completed.

4.15 FIRE VULNERABILITY OF AGED ELECTRICAL COMPONENTS

Objective: To identify and investigate issues of plant aging that might

result in an increase in plant fire risk.

Contractor: Sandia National Laboratories (SNL)

Methodoloqv: The Fire Aging Program is drawing upon Sandia's unique

fire testing and analysis capabilities developed, in part, as a result of the

USNRC-sponsored Fire Protection Research Program, coupled with Sandia's extensive understanding of fire risk analysis methodologies and results. Using

these capabilities, Sandia has identified and prioritized a list of six plant

aging issues that can significantly affect estimates of fire risk. These

issues are now being investigated in a systematic manner through testing and

analysis.

Status: To date, two of the identified issues have been resolved.

These are the impact of aging on 1) the vulnerability of electrical cables to

thermal damage, and 2) the flammability of cable insulation materials. Current investigations of the impact of aging focus on the vulnerability of other

types of plant equipment to fire-induced damage and the integrity and fire

performance of passive fire barriers.

4.14
Results: -The results'ofthese'efforts have demonstrated that two of the

initially-identified issues need not be'further considered.' Thetwo'studies

involving the fire performance of electrical cables have demonstrated that

changes in either the flammability or thermal fragility of cables that result

from aging will not result in a significant-.increase'in:fire'risk estimates.

* ;Plans: Future efforts'will continue-to~utilize Sandia's unique fire

testing and analysis capabilities to investigate the remaining identified fire

aging issues consistent with the prioritization of those issues that were

developed early in-the program.-..

4.16 OTHER-SPECIFIC TOPICS : . -

'Though not included in Table 4.1, the following have also been topics of

special topics., '.. -. :

4.16.1 !Aging/Seismic Shock Interaction

An. understanding of the vulnerability of degraded equipment to seismic

disturbances isnecessary for.extending.the design life oflan NPP. This

includes the original plant life of 40 years and any-extended operating..

period.. Current industry standards (IEEE Standards 323 and 344).and relevant

regulatory guides require pre-aging before seismic qualification'of electrical

equipment. -.However, theNRC has not determined'such aneed'for mechanical

components and is currently evaluating the'significance of aging as a factor

in qualifying mechanical equipment.- Therefore, an assessment is needed of the

potential importance'of aging in'degrading':the seismic performance of

equipment. . ;

Both the nuclear industry and the NRC'have ongoingiprograms.to assess

the aging-seismic effects. . These include the following:,-

* laboratory testing of naturally- and-'artificially-aged components

* qualifying equipment using existing test data

* using experimental data for qualifying components

..evaluating seismic-fragilities for different components

* identifying weak links in certain equipment' assemblies,

. developing surveillance and maintenance programs'to alleviate the

effects of aging on seismic performance-of-equipment. .

To avoid duplication of 'work, some studies involve coordination between:

industry and the NRC. Recently, it was determined that the qualified life of

some equipment may have to be extended when a utility submits an application

4.15
to the NRC for a license renewal. In that case, the components originally

qualified for a 40-year period would have to be reassessed for extended

operation.

4.16.2 Role of Safety-Centered Aging Management

Elements of maintenance work for NPAR aging studies on SSCs are outlined

below:

* reviewing current practices and procedures used by nuclear utilities to maintain equipment.

* reviewing nuclear-equipment-vendor recommendations on maintenance

of components or subcomponents selected for aging assessments.

* performing an evaluation, including a comparative analysis, of the

relative merits of performing corrective maintenance when a component has been discovered to be malfunctioning versus the merits of

programs concentrating on preventive maintenance. Emphasis is

placed on the relationship between failures (causes or modes)

expected to be experienced during operation and those that would

potentially occur under the stresses associated with design basis

or trigger events.

* evaluating the relative merits of predictive methods that can be

used to identify imminent failures and to enable maintenance or

replacement to be scheduled, based on actual equipment performance.

This age-related approach lends itself to reliability methods and

condition monitoring to mitigate equipment degradation.

* identifying, where possible, those component failure mechanisms

likely to be inadvertently induced through faulty preventive or

corrective maintenance actions. Specifically, identify those

failures that might be detectable through short-term postmaintenance, surveillance, inspection, or monitoring.

* developing recommendations for acceptable or preferred maintenance

practices, based on the preceding activities.

These activities emphasize the technical (e.g., mechanical, electrical)

aspects of maintenance rather than institutional, organizational, programmatic, or human factor considerations.

Evaluations are made on the role of maintenance in managing the aging of

SSCs. A review of present practices, in terms of a comparative analysis of

corrective versus preventive maintenance, and recommendations for preferred

practices are included for all SSCs in this study.

4.16
4.16.3 Status of the Maintenance Rule

The importance of effective maintenance'programs in ensuring safe and

reliable operation of NPPs has been widely recognized by both the NRC and the

industry. A high level of plant safety requires a high level of reliability

of all safety-related SSCs. This high level of reliability must not only be

designed into the SSCs and ensured by effective quality control procedures, it

must be sustained during plant operation-through sound maintenance practices

(Dey 1989).

In the early 1980s, several industry and NRC studies concluded that

there was significant room for improvement'in planning and implementing effective maintenance programs. Shortcomings in the 'established practices at some

NPPs were resulting in an unacceptably high rate of-challenges to safety systems, and unnecessarily high radiological exposure of maintenance personnel.

As a result of these findings, the NRC implemented a Maintenance'and Surveillance Plan (SECY-85-129), beginning in 1985.

The in-depth findings and conclusions from Phase I of the NRC Maintenance and Surveillance Plan (MSP) confirmed that maintenance-related safety ' problems were present to varying degrees -at different NPPs (NUREG-1212). In

response to the findings of the MSP, the NRC developed and formalized the Commission's position on maintenance at NPPs (53 FR 9430). -This policy statement

was the subject of a public workshop (NUREG/CP-0099) and, together with findings from a review of U.S. and -foreign maintenance programs (Dey 1988), formed

the basis for issuance of a proposed rule (53 FR 47822)'and a draft regulatory

guide (DG-1001) defining the essential elements of an effective maintenance

program.

Promulgation of the final Maintenance Rule was strongly opposed by the

NPP industry, who had already initiated several actions aimed at improving the

effectiveness of.maintenance practices and increasing the reliability of NPPs.

Consequently, the Commission decided to postpone rulemaking on maintenance for

a period of 18 months (until June 1991) to monitor the industry initiatives

and progress in improving the effectiveness of maintenance programs

(54 FR 50611).

Activities to resolve the status of the Maintenance Rule have resumed.

Options have been evaluated and a recommendation is scheduled to be presented

to the ACRS in April 1991.

4.17
5.0 LICENSE RENEWAL -

License renewal (LR) would permit qualified nuclear power plants (NPPs)

to operate for. a lice'nsed period beyond the current 40-year limit. Since the

mid-1980s, -the nuclear;industry has been actively developing the technical

basis for LR. In parallel,-the NRC has sponsored programs to understand the

technical issues.involved in.LR and isdeveloping the regulatory basis too'

review industry requests for LR. License renewal is a priority activity for

the NRC and has been an element of the NPAR Program (see Figure 3.1 and

Section 3.0).

Figure 5.1 shows the time frames--for--currenrt-licensed operation to

40 years and for proposed extended plant-operation beyond 40:years.. In the

figure, aging processes are shown to be ongoing in both time.frames and, hence, must be effectively understood and managed across the-entire'operating life of the plant'.

Aging

- I ; . . . . . W ; , , i . 4 .

I. I . w , % ,;

, , , . ;, .. I ! - i , !.' ; ....

a .

: .  ; . . I ,

.- .1 - ,, -11 I I I

Current Operatin

Current Operating License .. .~~~~~~~~~~~~~~~~~~

~!Extended

. . .~~~~~~~~~~~~~~

Plant Operation

-.; 5 -- . 40 i .. . . .~...----. ,. * - ,- .

- . , A, - X' ' Ens.urin

"Understanding Aging - A Keyto Ensuirig ISafety"

A"Managing Aging - A Necessity .. .. I . I.

to Ensuring Safety"

Guy A. Arlotto

FIGURE 5.1. Nuclear Power Plant Licensing Regimes and Aging Importance

5.1
Although the NPAR aging studies have applied mainly to the operating

license time frame, the principles of understanding and managing aging derived

from the studies are equally applicable to extended plant operation. Specifically, the hardware studies (Section 3.0) and the special topics (Section 4.0) provide technical bases for understanding and managing aging in

structures, systems, and components (SSCs). These studies are part of the

technical bases that provide guidance and input to the NRC's development of

the basis for LR. Some of the bases for understanding and managing aging are

discussed in Appendix B.

5.1 REGULATORY GUIDANCE FOR LICENSE RENEWAL

The NRC has proceeded to establish the regulatory basis for LR by developing a LR Rule, which has been issued for public comment (55 FR 29043). The

LR Rule will establish the regulatory guidelines for the LR process and will

require that age-related degradation in NPPs be managed effectively during the

LR period. The LR Rule will be supported by a draft regulatory guide to specify for the applicants the content and format for the LR application. This

draft regulatory guide, Standard Format and Content of Technical Information

for Nuclear Power Plant Operating Licenses, (DG-1009) also was issued for

public comment (55 FR 50065). A third important initiative of the LR regulatory process will be the Standard Review Plan for License Renewal (SRP-LR),

which is intended to guide the NRC staff in reviews of LR requests. The draft

SRP-LR (NUREG-1299) also has been issued for public comment (55 FR 50065). To

provide more detailed technical support to the regulatory guide and the

SRP-LR, a compilation of aging management information (Morgan and Christensen

1991) has been prepared, including input from NPAR research staff at several

laboratories. A draft (NUREG/CR-5562) has been issued for comment.

5.2 NPAR PARTICIPATION IN THE LICENSE RENEWAL PROCESS

Results from the NPAR aging assessments provide the regulatory staff with

a basis to evaluate the effects of age-related degradation on safety-related

SSCs both in current operations and in LR. The NPAR Program is contributing

to the development of LR by providing technical support to the LR rulemaking

and its implementation. Several NPAR initiatives support the License Renewal

Rule:

* Lead responsibility for developing the draft regulatory guide on

Standard Format and Content of Technical Information for Applications to Renew Nuclear Power Plant Operating Licenses (OG-1009)

* Lead responsibility for developing a Review of Information Useful

for Manaqing Aging, NUREG/CR-5562 (Morgan and Christensen 1991,

draft)

5.2
* Technical support to NRR . *....... . .. .

-' in preparation of the LR Rule and in resolution of public

comments on the Rule.

- input to development of the SRP-LR

- review of NUMARC industry reports (see Figure 5.2 for reports

and review schedules)

- potential input to prepare Safety Evaluation Reports (SERs

- potential support to review lead plant applications.

The 'followin

on LR, industry r

roles. .

g sections desc'ribe the industry and reguiatory!perspectives

eport reviews, lead plant implementation, and other NPAR

- * i 5 ' . ~~~~~~~~I ..!-.

5.3 INDUSTRY PERSPECTIVE.',

In the U.S. there are over 110 commercial.NPPs in operation,' constituting

nearly 20%.of.U.S.'. electrical capacity. 'Several plants have operated for more

than 20 years and a few are approaching 30'years. The nuclear industry is

actively developing the.basis to extend the life of operating reactors'beyond

their originalllicense term. From the industry perspective, a major motivation'for LR-is economic. The projected.,net benefit to the U.S: economy is

estimatedlto exceed $100 billion,,assuming a 20-year life extension for currently operating reactors.. The industry recognizes that management of agerelated degradation is a key consideration for LR.

License renewal initiatives have been developed through cooperation~among

the Department of Energy (DOE), the Electric Power Research Institute (EPRI),

and utilities, coordinated by the working groups of the Nuclear Utility Management and Resources Council (NUMARC), Nuclear Utility Plant Life Extension

(NUPLEX). . Two pilot studies weretcompleted:- a-boiling-water reactor-(BWR)

evaluation involving the Monticello-plarit (operated by Northern'States Power

Company) and a pressurized-water reactor (PWR) evaluation involving the Surry

plant (operated by Virginia Electric Power Company). Subsequently, a bidding

process was used to select two lead plants to develop formal applications for

LR. 4The two plants selected were a BWR (Monticello operated by Northern'.-

States Power Co.) and a PWR (Yankee Rowe operated by Yankee Atomic Company).

The LR applications are scheduled to be submitted to NRC inA1991 (see

Figure 5.2). ' .

From a technical perspective, ;it A is recognized that certain plant SSCs

important to LR that are not readily refurbishable or-replaceable are subject

to age-related degradation.. The'position was'stated at the'NRC Public Workshop on Technical and.Policy Considerations-for NPP License Renewal

(NUREG-1411): "The focus'of license v'enewal is mitigation-and management of

age-related degradation to ensure an adequate level of safety." ItJs .

5.3'
Ft"'tYea 198a 19 89 | 1990 1991 1992 1993 1994

.1.1.1 ~ ~ 0118 AI#ala 1.41.10,10u 0ll-u 1 Wei 111.11 .,;1., ,MbAWllIl ,t"U"Ij.II.d ..1 of 1 -.fi -l 1 -,1M.1 #r sn

Padvv Pubkc publish Publish

Ndotice WoIshop Dalt fnl V Scheduhd

Deelopmenlt of License Renewal Rule _ IF T

68 b11189 7/90 "I V Cnpteed

Development of Regulatory Guidance ll Publih

Regulatory Guide on Sid. Format and Content s N7 12190 JIM2

Pubish Pub"111

Drat Fidal

Revise RG 4.2 Preparatlon of Environmental V V

ReHorts for Nuclear Power Stations "IS 4il2

Publish Pubblbh

Draft FIn Standard Review Plan for tUcense Renewal IF

12190 4tS2

Review of Industry Technical Reports *T NRC Revi"d SEH

Recetwed Commnts IRT Issued

Screening Methodology * V V

PWR Containments 'V

BWR Pressure Vessels p 'V- -

WR.Yessel Internals Mil

PWR Pressure Vessels * _ 0p

PWR Vesse intlernals

Class I Structures

PWR Coolant System

Cables In Containment =

BWR Pressure Boundary

BWR Contalnments

kwlwl FUIal NRC FInall NRC

NRC Applicallo Format Ynkee Ro Unella Dewmnailnon Dlermnlneonon

Willing Mancdont at MetIng Applcaion Appilckslin Yank" Raow* MontIlCllo

Lead Plant Implementation It V V V V

2186 8/89 91 12/91 M3 1293

36"1014

Un

.JFa

FIGURE 5.2. Schedule of License Renewal Rule Development and Implementation
commonly viewed that reliability and operational readiness; of.,risk-signif--

icant SSCs subject to aging must be ensured ing ng.their entire operating

life. -

The industry, DOE, and.EPRI have developed industry reports.(IRs) that.

present, on a generic basis, the technical and safety issues that they con-.

sider important to LR. The reports have been subject to NRC review and

comment. The industry proposes that preparations to renew. licenses for-individual plants would be guided by the IRs,.but also-that each plant would need,.

to assess the plant-specific issues related to the components or structures

addressed by the IRs. Both the industry and the NRC are.focusing on programs

to determine residual life of major LWR components and structures. A screen-.

ing methodology to identify components that.may degrade during extended operation'has been proposed.. The role of-that screening-methodology or an alternative approach will be-an important.LR consideration.

5.4 REGULATORY RESEARCH PERSPECTIVE . -. 

Research programs conducted under theANRC Office of.Research have focused:

on understanding and managing aging in nuclear'safety-related SSC (see Figure 1.1). The Structural and Seismic.Engineering'Branch sponsors research on' - concrete structures; the Materials'Engineering Branch is focusing-on reactor

pressure boundary components; the Electrical and Mechanical Engineering Branch

conducts aging research on safety-related systems and components.under the.- .. NPAR program. These programs help to resolve uncertainties regarding degradation processes that affect key SSCs and detection and management of degradation before safety is impaired... The programs also focus on methods to

mitigate simultaneous; multiple age-related failures of components that could

occur during a transient or !accident, thereby compromising.the function of the

safety systems.

.Reviews of operating experience indicate that components can fail from

numerous causes, including corrosion,:,radiation, thermally-induced embrittlement of electrical insulation,.pitting of electrical contacts, surface

erosion, metal fatigue, oxidation, creep,-and binding-and wear. As operating ' reactors advance in age-toward the.'nominal design life.of.40 years or in -. extended operation, the major regulatory safety issues are ensuring that aged

plant systems and components do not result in failures, including.common mode

failures that could weaken the defense-in-depth strategy, lead to an accident, or render inoperable the redundant but aged safety equipmen't needed to

mitigate accidents..: ! I . ' ,. !' , '' ' '

5.4.1 Technical Issues

The technical issuesfor a plant that demonstrates.a satisfactory.operating history over a 40-year period are expected to be related to .the aging

and degradation of hardware, not to the current licensing basis of the plant

(NUREG-1362).' Past.'and current regulatory programs, as well as the NRC's

evaluation of abnormal operational events and inspection efforts, ensure that

5.5
design-related safety issues are resolved on a continuing basis, and modifications are made to the plant where necessary. These modifications, in both

the physical plant and in the plant operating procedures, become part of the

current licensing basis for the plant. The current licensing basis of the

older operating plants is assumed to be adequate, although it may differ from

the design basis of new plants.

Safety during the renewal term can be ensured by systematically assessing

and managing the aging of selected safety-significant components of the plant.

A systematic approach to assess and manage aging will ensure a uniformity in

the scope of the licensees' efforts and in the standards that they should

meet, while allowing the necessary degree of flexibility. The nuclear industry's pilot plant and lead plant studies emphasize the importance of aging

analyses to support LR. Similar analyses will likely be conducted by all

licensees applying for LR, and these analyses will take into account the

significant findings of the aging studies by the NRC and the industry.

The aging management strategies for the renewal term adopted by the

licensees will likely be plant specific, depending on plant type and vintage.

However, the development and implementation of these strategies will depend on

the quality of data and methods used in the aging analyses. Therefore, the

NRC's specification of a formal methodology for these analyses will provide a

high level of safety assurance during the LR term.

5.4.2 Plans and Schedule

The schedule for implementing the proposed License Renewal Rule (Figure 5.2) includes separate but interdependent schedules for rule development,

regulatory guidance development, IR reviews, SRP-LR development and lead plant

implementation. The progress of each activity will influence the progress of

the others. Appropriate participation by the industry has been assumed, based

on schedules formally provided by NUMARC to the NRC.

Completion of rulemaking will provide the basis for review of the lead

plants. If technical issues are resolved and regulatory guidance is developed

in sufficient time to create a stable regulatory environment for LR, then

other licensees can plan effectively for future generating capacity.

5.5 INDUSTRY REPORT REVIEWS

While NRC developed the LR Rule and regulatory guidance, the nuclear

industry (represented by NUMARC), EPRI, and DOE concurrently sponsored the

development of several IRs on important aging mechanisms that can affect

safety-significant SSCs. The NRC staff, assisted by the NPAR contractor

laboratories, have actively reviewed the IRs as shown in Figure 5.2.

5.6
5.6 LEAD PLANT IMPLEMENTATION

Yankee Atomic Electric Company's Yankee Rowe plant (PWR) and Northern

States Power Company's Monticello plant (BWR) are expected to be the first two

plants to apply for LR. The preparation-and application work at these plants

is being sponsored jointly by DOE, EPRI, and NUMARC to test the LR process

established by the NRC. A series of meetings involving the lead plants, the

lead plant program sponsors, and the NRC has been initiated. Results from the

NPAR aging assessments will contribute to the technical basis for evaluating

the LR applications.

5.7 SUPPORT FROM CODES AND STANDARDS AND INDUSTRY GROUPS

Codes and standards organizations, particularly ASME and IEEE have been

active in identifying methods to improve the codes and standards to develop

guidelines that will be valuable to manage age-related degradation.

5.7
6.0- UTILIZATION OF RESEARCH RESULTS

The results from aging studies coordinated by the Nuclear Plant Aging

Research (NPAR) Program have both technical and regulatory applications. This

section summarizes the.status.of.the results of these studies, the documentation, and the general and'specific applications of results. Additionally,

this section describes how the NPAR results contribute to understanding and

managing aging and how they support license renewal (LR) rulemaking.

6.1 STATUS OF NUCLEAR PLANT AGING-RESEARCH RESULTS

By the end of FY 1990, Phases I and II of the NPAR Program Strategy were

completed for eight components.and three systems. Additionally, Phase-I

assessments of two components and one system were completed in FY 1990. 'One

special topic investigation (Shippingport aging evaluation) was completed in

FY 1990. Phase-I studies on fourteen special topics were completed in - FY 1990. Results from these studies are contributing to the NPAR technical

bases for LR. Two major NPAR initiatives supporting the LR Rule were completed in-draft'form in FY 1990.

6.2 STATUS OF NUCLEAR PLANT AGING RESEARCH DOCUMENTATION

Approximately 95 NUREG/CR-reports have been issued (to April 1991).

Brief summaries the NPAR reports that'have been issued through May 1990 are

provided in Kondic and Hill (1990). 7At that-time; 72 NPAR reports had been

issued and approximately 20 more were in draft form. Numerous papers based on

NPAR results'have been presented'at the Water Reactor' Safety Information'Meetings, at technical society meetings, and-at national and international symposia. The papers generally were published in peer-reviewed journals and in the

proceedings of the respective conferences which include:

Third Workshop on Containment Inteqritv, Washinqton, D.C., May 21-23, 1986.

NUREG/CP-0076, U.S. Nuclear Regulatory Commission, Washington, D.C.

Proceedinqs of an International Symposium on Safety Aspects'of the Aqeinq and

Maintenance of Nuclear Power Plants. 29 June to 3 July 1987, Vienna, Austria.

International Atomic Energy Agency. -

1988 ASME-Pressure Vessel and;PipingtConference, PVP 16. American Society of

Mechanical Engineers, New York,-New York.

Proceedings of the Topical Meetinqg.on Nuclear Power'Plant Life Extension,

JulY 31 - August 3, 1988, Snowbird,-Utah. American Nuclear Society, La Grange

Park, Illinois.

Proceedings-of the International Nuclear Power Plant'Aging Symposium, ed.

F. A. Beranek. NUREG/CP-O100. U.S. Nuclear Regulatory Commission,-

Washington, D.C.

6.1
Sixteenth Water Reactor Safety Information Meeting, ed. A. J. Weiss.

NUREG-0097. U.S. Nuclear Regulatory Commission, Washington, D.C.

1989 ASME Pressure Vessel and Piping Conference, Honolulu. Hawaii, PVP

Vol. 171. Am. Soc. Mech. Eng., New York, New York.

10th International Conference on Structural Mechanics in Reactor Technologv

(SMiRT). August 14-18. 1989. Anaheim, California. American Association for

Structural Mechanics in Reactor Technology.

International Conference on Operability of Nuclear Systems in Normal and

Adverse Conditions, September 18-23, 1989, Lyon, France. American Nuclear

Society, La Grange Park, Illinois.

15th MPA-Seminar on Safety and Reliability of Plant Technology. Stuttgart,

FRG, October 5-6. 1989. (To be published' in Nucl. Eng. Design.)

Seventeenth Water Reactor Safety Information Meetinq, ed. A. J. Weiss. NUREG/

CP-0105, U.S. Nuclear Regulatory Commission, Washington, D.C.

Transactions of the Eighteenth Water Reactor Safety Information Meeting, ed.

A. J. Weiss. NUREG/CP-0113, U.S. Nuclear Regulatory Commission, Washington,

D.C.

The NPAR technical bases for LR and the overall research program have

been subject to peer review, including reviewers from U.S. Nuclear Regulatory

Commission (NRC) staff; the Advisory Committee on Reactor Safeguards (ACRS);

the Nuclear Safety Reactor Research Committee; the NPAR laboratories; Equipment Quantification Advisory Group (EQAG) of the Electric Power Research

Institute (EPRI); NSSS vendors; and other industry groups, codes and standards groups, and industry experts. Some NPAR results have been subject to

review in workshops convened for detailed evaluations of the results.

6.3 TECHNICAL APPLICATIONS OF NPAR RESULTS

This section addresses the general and specific technical applications of

the results from NPAR aging assessments.

6.3.1 General Technical Applications

NPAR reports and other publications on aging assessments of safetyrelated systems, structures, and components (SSCs) represent technical bases

that can assist both regulators and industry to understand further how aging

affects equipment, how to improve management of the aging processes, and how

to minimize potential impacts of aging on plant safety. Efforts are underway

to improve accessibility of these technical bases.

Inquiries from industry and from NRC staff indicate that the NPAR technical findings are being utilized. The NPAR technical bases demonstrate a

6.2
sustained NRC commitment to understanding the effects of aging.ofSSCs, both

for the current license period and for extended plant operation..

6.3.2 Specific'Technical Applications.. .

Results from the NPAR hardware-oriented engineering studies'include

several specific technical findings.and recommendations that have near-term

applications. The following are several 'examples:

* Identified deficiencies in current inservice testing (IST) program

for auxiliary-feedwater (AFW)pumps',and.published recommendations

for improvements in the IST programs'forAFW pump's'[used for the

development of American Society for:Mechanical' Engineers (ASME)-O&M

Standards,'Office.of'Nuclear. Reactor, Regulation (NRR) relief.

requests, and in the preparation'of Bullet'ins,88-04 and 89-04]. '

* Published recommendations for preferred testing and maintenance of

diesel generators', and'generated 'technical 'data to support GSI-56,

Diesel Reliability. 6,

* Identified components (fuses, capacitors) 'that are prone to aging in

battery chargers and inverters and need improved maintenance. Working to upgrade Institute ofElectrical and Electronics 'Engineers

(IEEE) Standard.650, Qualification of Class 1E Static Battery

Chargers 'and Inverters. ; -. ' -

* Published IEEE recommendations for Maintenance Good Practices for

nuclear power plant (NPP) motors, motor operated valves (MOVs),

solenoid operated valves (SOVs), and battery chargers/inverters; the

recommendations contain significant''uses'of NPAR results.

* Provided direct support to NRR in resolution of'service water system problems at a specific reactor.

* Provided major inputs from'the NPAR Program to the 'roposed IEEE

W.G.3.4 Guide on "Life Extension of'Electrical and I&C Equipment."

* Published ASME/ANSI OM-16 draft on "Inservice Testing and Main-.

tenance ofDiesel Drives' in Nuclear'Power Plants,'" which contains

key recommendations made in the-NPAR-generated NUREG/CR-5078'-

(Lofgren et al. 1988).

* Transferred data developed in the NRC Equipment Qualification/Aging

program to'DOE for use~in-developing procedures for evaluating timedependent dose-rate'effects on four.idifferent'cable materials.

* Transferred NPAR generated datalto industry for developing snubber

seal material recommendations. _- -

6.3
Identified degradation sites and mechanisms, stressors, and failure

modes, and evaluated current inspection methods for twenty-two major

light-water reactor (LWR) components: pressurized-water reactor

(PWR) and boiling-water reactor (BWR) pressure vessels, PWR and BWR

containments, PWR reactor coolant piping, safe ends, and nozzles,

PWR steam generators, PWR reactor coolant pumps, PWR pressurizer,

and surge and spray lines, PWR and BWR control rod drive mechanisms,

PWR and BWR cables and connectors, PWR and BWR emergency diesel

generators, PWR and BWR reactor vessel supports, PWR and BWR pressure vessel internals, PWR feedwater lines and nozzles, BWR recirculation piping and safe ends, BWR main steam lines and feedwater

lines and nozzles, and BWR recirculation pumps. Results are being

used to evaluate industry reports for license renewal considerations, preparing guidance related to understanding and managing

aging, and preparing Standard Review Plans for License Renewal

(SRP-LR).

NPAR results related to fatigue failures in the field provided major

input to the ASME Section XI report Metal Fatigue in Onerating

Nuclear Power Plants: A Review of Design-and Monitoring Requirements. Field Failure Experience. and Recommendations for ASME Section XI Action. The report was prepared by the ASME Section XI Task

Group on Fatigue in Operating Plants. One of the major recommendations, which was based on NPAR results, is to reevaluate the current

inservice inspection (ISI) requirements and to include in the ISI

program the inspection of highly stressed base-metal sites that are

susceptible to fatigue damage.

• Developed a procedure using a linear aging model for the modification of probabilistic risk assessment to account for the effects of

aging and used this method to carry out a risk-based prioritization

of components and structures.

* Generated a single source document that delineates the regulatory

documents governing the inspection and monitoring of reactor pressure vessels, steam generators, and major reactor piping, for use in

both aging study and license renewal.

* Developed a novel diagnostic technique for MOVs based on the analysis of motor current signatures for remotely and nonintrusively

detecting and differentiating among abnormalities, including those

resulting from aging and improper maintenance.

* Developed a novel check-valve diagnostic technique based on the

analysis of magnetic flux signatures for remote detection and trending of check-valve abnormalities, including those resulting from

aging, improper operation, or improper maintenance. The technique

has also been successfully demonstrated in industry tests conducted

by the Nuclear Industry Check-Valve Users Group and is useful in

satisfying check-valve IST requirements.

6.4
* 'Developed statistical'models.to analyze component degradation and

failure data to understand the aging process of components. The,;...'

modeling focuses on analyzing the times of component degradation to.

model how the rate of degradation.changes with the.age of the compornent. The modeling' also considers the effectiveness of maintenance -

as it applies to aging evaluations.

* Contributed to formulation and implementation 'of Generic

Letter 89-13 on SWS reliability.. .

6.4 REGULATORY APPLICATIONS OF NPARVRESULTS

This section addresses general and specific regulatory applications.of'

NPAR findings. . ;

6.4.1 General Regulatory Applications .

The NPAR Program is providing technical bases for timely regulatory decisions regarding the effects of aging.on NPPs of all ages. Informed.oversight

to ensure that aging of safety-related SSCs'is effectively understood and managed is important to fulfilling NRC's';regulatory responsibility. Through the'!

NPAR utilization' effort, NPAR-results are communicated in.a format that .can

assist the work of the NRC inspection staff.

6.4.2 Specific Regulatory Applications'

The&NPAR Program is structured'to respond to the following specific"

user-oriented needs:'

* Develop data for identifying and'resolving technical safety issues

related to plant aging and LR.

* Support NRR and Office of Nuclear Regulatory Research (RES) in.

resolving generic'safetyjissues involving ag ed plant safety systems,

support systems, and electrical'andmechanical components.

* Evaluate and recommend maintenanc'e'and surveillance'methods needed

to monitor aging of SSCs and to support LR.' ' - ', , .

* Provide information for developing inservice inspection procedures

suitable 'for aged SSCs.

* Develop data and supporting information -for revising appropriateNRC ' guidelines, techn'ical 'specifications',' and 'industry codes-and ' -'

standards. ':' - : - :

* Provide technical information needed to prepare NRC Information

Notices and Bulletins.

6.5
Revision to the body of regulations governing NPP operation is an ongoing process that applies key results from the NPAR aging research to improve

the regulatory basis. Specific results are applied to the regulations and

supporting documents in areas such as regulatory guides, technical specifications, generic safety issues (GSIs), and codes and standards. Specific cases

are cited in the following sections.

6.4.2.1 Overview of Specific Regulatory Applications

The support that NPAR results are providing to the regulatory process is

summarized in Table 6.1. The table comprises a matrix indicating each of the

ongoing and planned NPAR activities and the support (or potential support) for

various user-oriented needs. The milestones and schedules for the various

activities are presented in Section 9.0.

Specific examples of regulatory application of NPAR results are cited in

the following sections.

6.4.2.2 Generic Safety Issues and Regulatory Guides

One of the objectives of the NPAR Program is to support NRR/RES in

resolving aging-related GSIs identified in NUREG-0933, "A Prioritization of

General Safety Issues" (Emrit 1991). NUREG-0933 contains a recommended

priority list to assist in the timely and efficient resolution of safety

issues that have a high potential for reducing risk. Results from NPAR

studies are contributing to resolution of several of these GSIs, including

those listed previously in Table 3.3. For example, the NPAR Program has supported the resolution of the Generic Safety Issue B-56, "Diesel Reliability,"

by evaluating aging of emergency diesel generators and by publishing recommendations for preferred testing and maintenance of diesel generators. The

NPAR results also supported the basis for resolving Generic Safety Issue A44

on Station Blackout. NPAR laboratory staff provided input to resolve Generic

Safety Issue 51 on service water systems.

NPAR laboratory staff supported NRR to develop Generic Letter 89-13 on

service water systems; also generic letters 89-04 and 89-10. The NRR has provided "users-need requests" to RES for resolving the GSIs. The NPAR Program

is supporting NRR in resolving the Generic Safety Issue II.E.6.1, "Test Adequacy Study" and by assessing methods for monitoring motor-operated valves.

The third issue in this category is GI-70, "PORV and Block Valve Reliability."

A residual life assessment is being performed as part of NPAR to evaluate

the aging and residual life of major LWR components. The results of this task

will indirectly support NRR in resolving several GSIs related to the primary

reactor coolant system components. The results of this task will be used in

resolving safety issues related to plant LR and in developing regulatory

guidelines and review procedures for use by NRR in reviewing applications for

LR.

6.6
TABLE 6.1. Potential Use of NPAR.Results, Involving Aging Consideration,

for Systems,-Structures .and Components :.

I..icre .Generic Maintenance .Pi Plant .. .

- Lcen Seafety and Performance . . Codes and

NPAR Research Activities Renewal Issues Surveillance Indicators Insvection 'Standards

Shippingport aging evaluation X X X ,

Risk evaluation of significant aging

effects

X X X X X

Residual life assessment X

Containment structures ' .. X

Reactor vessel supports X

Snubbers X

Control rod drive system X

Piping systems . X

High pressure ECCS X

RHR/low pressure ECCS . X

Component cooling water system X

Reactor coolant pumps (seals) * X

Auxiliary feedwater system X

Auxiliary feedwater pumps X

Service water system - X

Heat Exchangers X

Electric motors X

Hotor-operated valves X

Solenoid-operated valves, X

Check valves X

.: 'X ' .

, X; ,--

t .. _ 1

x - x- - I , I,

. .X

:,, . X;...

X

. x. X X . . "_

, .. .... . . ,. 7

..I .I

Ii. . . . .

I; I . '

:; X

X

.X ,

X .. ....

X'

X - 'X

. ;'''X '

. X

'' . '~ : X" ' ':

X

x . .

I I . .X .

X

X '

X

.X.

X

' X

.-X

.X

X .7,

:~ ~ :

.x.

_x . . ..

*~ ,x ~ x - -

X X X

x .. .x. .

X X

-" 'X* x.

X Xx"

X . ~X .- X

I .

It

Power-operated relief valves

Reactor protection system' ' I * ~. - I

Diesel generators

Class 1-E distribution'system .

Transformers

Chargers/inverters

Batteries

Circuit breakers

Bistables/switches

Penetrations/connectors/cables

Compressors

Bolts, attachment devices

X

, 11

X.~

X X

* x

.X-x - . ., . x

x.. x. **I

- x ,*

- ., ( '..

.: cx,., - x 

I

f:

.w

I -

I-

., .. .. 1 C. X .....

X

I

X X X

6;7
NPAR results contribute to updates to and development of regulatory

guides. NPAR laboratory staff provided direct support to NRR in the revision

of Regulatory Guide 1.9 on Diesel Generator Selection, Design, and Qualification and published a research information letter (RIL) on aging management in

EDGs. Development of the draft regulatory guide (DG-1009), supporting the LR

Rule, and NPAR support to development of the SRP-LR were discussed in Section 5.0.

6.4.2.3 Maintenance and Surveillance

Effective maintenance and surveillance programs at NPPs are significant contributors to plant reliability. The NPAR Program supports the NRR

Maintenance and Surveillance Program by evaluating the role of maintenance in

managing aging effects, both generically under a special topic and specifically under the individual aging studies of SSCs.

Here are key recommendations from the NPAR aging assessments of SSCs that

address maintenance:

* improve the approach to scheduling major maintenance for EDGs

* improve the flow conditions and types of measurements to be taken in

conducting inservice testing of auxiliary feedwater and other

safety-related pumps

* implement a comprehensive inspection and preventive maintenance

program for inverters

* implement a preventive maintenance program which includes techniques

for testing motors that yield trendable data

* conduct systematic inspection and maintenance for LWR metal

containments.

The NPAR Program defines maintenance and surveillance needs to ensure the

operational readiness of aged NPP safety systems and components and to support

the NRR staff in reviews of requests for LR. The program also provides the

aging-related information needed for developing maintenance program criteria

and standards and maintenance indicators which can be used to monitor specific

components and systems.

The NPAR Program also provided input to the NRC Maintenance Rule. The

Maintenance Rule was discussed in a public workshop in November 1988. Subsequently, it was placed on hold for 18 months., Depending on its eventual

status, the NPAR Program is prepared to provide further input, based on

1) insights to maintenance developed in NPAR SSCs studies, and 2) recognition

that maintenance is important to effectively managing aging.

6.8
6.4.2.4 Plant Inservice Inspection"

NPAR results are applied to plant inservice inspections; for example,;the

signature analysis technique for MOYs is used to detect and differentiate

among abnormalities, including those resulting from aging and from improper.

maintenance. _~~~~~~ 'I .,, '. .:,,.*.4i*.

6.4.2.5 Plant Ins rvic6Testini'

The NPAR research'-identified deficiencies 'in the current'IST program for

AFW pumps (Adams and Makay 1986, Kitchet"al. 1988; -Casada-1989); consequently, recommendations-for improvirig:the'IST 'programs for'AFW pumpshave

been published and are being used.to~'tdev'elop ASME-O&M'Standards; NRR.Technical

bases also were provided for relief'r-e'uests 'and to prepareBulletins'85-03,

88-04, 88-57 and 89-04.. As a result of the electric motor .NPAR research, it

is recommended that certain'motor '-parameters be'monitored'while'performing'

required pump-`and valve inservice'testing'. ' -

6.4.2.6 NRC'Inspections '' ' .

An NPAR initiative is underway to'support'the foll6wing NRR programs:

the Safety System Functional Inspection Program and..the-Generic Communication

Program. In general, the Safety'System' Functional 'Inspec'tion Pro'gram'assesses

whether plant modifications~ of selec'ted"s'afety"'systems 'have degraded the

design margin to the point where the system's ability to mitigate design basis

events is impaired. This program consists of.an in-depth review of selected

safety systems and is usually'conductedlat older plants. The objectives of

the Generic'Communication- Program.are:; , '

* to inform licensees of problems, including those caused by aging,

that have developed in..individual plants

* to require action when these'problems are shown to be significant

and generic. .

These'programs apply,'to thd'pr'essure boundary hardware, driviers, actuators, electrical power,'and the instrumentation'and controls of engineering

safety features (ESFs). ' ' i '

The NPAR Program supports NRR in developing inspection procedures.that

are relevant'-to aging,'NRR inclu'des these proc'edur6s in the 'Inspection

Enforcement Manual issued''to guide:the activities'of th'e reg'ions. For example, some inspection procedures' 'stablish guidance for''ascertaining whether

inservice inspection and .testing activities are programmed, planned, conducted, recorded,.and reported'in 'acdo-rdance with''Section'XI of the ASME

Boiler and:Pressure Vessel Code. - - - '

The NPAR Program is developing initiatives to support ongoing' inspections conducted by the regional offices under the NRR inspection program.

These inspections ensure that systems and components have not degraded beyond

acceptable margins from any cause, including aging. To provide the inspection

6.9
staff with the current status of NPAR research results, the results from the

NPAR program will be summarized and provided to the regional and NRR inspection staff.

6.4.2.7 Codes and Standards

Codes and standards groups provide important input to define the technical requirements for selected NPP electrical and mechanical components to

ensure their operational integrity. Through participation in technical committees, NPAR participants contributed to developing recommendations to revise

relevant ASME and IEEE codes and standards. These codes and standards address

the safe operation of NPPs as component and system service times increase.

Examples of the application of NPAR Program results in the improvement of

codes and standards include the following:

* The proposed ASME Section XI recommendation for inspection of highly

stressed sites in base metals susceptible to thermal fatigue damage.

* IEEE recommendations for Good Maintenance Practices for NPP electrical equipment on motors, MOVs, battery chargers/inverters, and

SOVs contain significant use of NPAR results.

* The proposed IEEE W.G.3.4 Guide on "Life Extension of Electrical and

I&C Equipment" will contain major inputs from NPAR Program

participants.

* ASME/ANSI OM-16 draft on "Inservice Testing and Maintenance of

Diesel Drives in Nuclear Power Plants" contains key recommendations

made in the NPAR-generated NUREG/CR-5078, Vol. 2 (Lofgren et al.

1988).

* NPAR results are being used to upgrade IEEE Standard 650, "Qualification of Class lE Static Battery Chargers and Inverters."

* IEEE Working Groups 4.2 is revising IEEE Standard 387 to include

requirements of IEEE Standard 749 so that criteria and testing

requirements for emergency diesel generators will be in the new

document. NPAR data and information is an important resource for

this work.

* ASME/ANSI OM-4 "Examination and Performance of Nuclear Power Plant

Dynamic Restraints (Snubbers)," provides general and service-life

monitoring guidelines to the OM-4 standard.

* ASME/ANSI OM-6 "Inservice Testing of Pumps in Light-Water Reactor

Power Plants" includes recommendations regarding recordkeeping, data

collection, and acceptance criteria based on NPAR research into

auxiliary feedwater pumps (Adams and Mckay 1986; Kitch et al. 1988).

6.10
* ASME/ANSI OM-8 '"Startup and. Periodic .Performance-Testing of. Elec-'

tric Motor Operators on Valve Assemblies in Nuclear Power Plants"

includes recommendations regarding recordkeeping, data collection,

and acceptance criteria based on NPAR research into MOVs

(Greenstreet, Murphy, and Eissenberg 1985; Haynes,1989).

• ASME-ANSI OM-10 "Inservice Testing of Valves in Light Water Reactor

Power Plants" includes recommendations regarding recordkeeping and

stroke time testing criteria and methodology-based.on'NPAR research

into MOVs (Greenstreet, Murphy,-and Eissenberg 1985;lHaynes 1989).

* ASME OM Administrative'Standard 6200, "Format' and Content of Standard Parts," and Administrative Standard.9100 "Glossary of Terms"

were prepared .with major input from the NPAR program.l,.

* ASME CS 860900 "A Guide to Writing ASME Codes and Standards" was

prepared with.major input from the'NPAR program.

The Special Working Group on LifeiExtension, ASME Section XI, is coordinating activities-related to initiatives.of.,interest to the NPAR Program.., A

special IEEE'working group' was established to prepare a'guide for life assessment of Class lE electrical controls. The components and systems currently of

interest and.being considered in ASME standards are listed in Table,6.2. Some

of the relevant.IEEE standards are'listed in Table 6.3. Periodic briefings

and information exchanges with appropriate codes and standards committees are

scheduled as part of NPAR. - -n d s

6.5 APPLICATION OF NPAR RESULTS TO-UNDERSTAND AND MANAGE AGING

The body of NPAR data that is available to understand and manage aging

draws from both aging investigations.,ofiSCs, using the NPAR strategy (Section 3.0 and Figure 3.1), and from the special topics (Section 4.0 and

Table 4.1).

6.5.1 Results from Aging Investigations-on Components and Systems

Results from the Phase-II comprehensive-aging assessment on the components and systems are illustrated in this section in briefs that address how

the work was carried out and how the results' apply 'to'understanding and-managing aging. . ..

6.5.1.1 Understanding and Manaqin6 Aqing in Emergency Diesel Generators

Understanding aging in EDGs.is complex because the equipment is-large and

comprises many subsystems and components,, Also, nine EDG suppliers are-represented in U.S. NPPs. A laboratory approach to-conduct the' aginginvestigation was not practical, so an alternative, approach,-'summarized below,-was

devised.

6.11
TABLE 6.2.

Standard Number

ASME OM-1

ASME OM-2

ASME OM-3

ASME OM-4

ASME OM-5

ASME OM-6

ASME OM-7

ASME OM-8

ASME OM-10

ASME OM-13

ASME OM-14

ASME OM-15

ASME OM-16

ASME OM-19

Selected ASME Standards for Operation and Maintenance

of Mechanical Equipment

Title

Inservice performance testing of nuclear power plant

pressure relief devices.

Requirements for performance testing of nuclear power plant

closed cooling water systems.

Requirements for vibration testing of nuclear piping

systems.

Examination and performance of nuclear power plant dynamic

restraints (snubbers).

Inservice monitoring of core support barrel axial preloads

in PWRs.

Requirements for performance testing of pumps in lightwater cooled nuclear power plants.

Thermal expansion testing of nuclear piping systems.

Requirements for preoperational and periodic performance

testing of motor-operated valve assemblies.

Requirements for inservice testing of valves in light-water

cooled nuclear power plants.

Requirements for periodic testing and monitoring of poweroperated relief valve assemblies.

Requirements for vibration monitoring of rotating equipment.

Requirements for performance testing of nuclear power plant

emergency core cooling systems.

Inservice performance testing of nuclear power plant diesel

drives.

Startup and periodic testing of electro-pneumatic-operated

valve assemblies used in nuclear power plants.

Approach to Understanding EDG Aging - A panel of Pacific Northwest

Laboratory (PNL) and industry EDG experts analyzed approximately 2000 EDG

failure cases from nuclear databases according to failure cause and relation

to aging. The following actions were taken:

6.12
TABLE 6.3.

Standard Number

IEEE 308

IEEE 317

IEEE 323

IEEE 334

IEEE 336

IEEE 338

IEEE 344

IEEE 382

IEEE 383

IEEE 387

IEEE 501

IEEE 535

IEEE 572

IEEE 549

IEEE 649

IEEE 650

IEEE 749

IEEE 944

IEEE/ANSI C37.13-1981

IEEE Standards for Electrical Equipment for-Nuclear Power

Plants that Provide Information Relevant to Aging and;LR

Title - -

Criteria for Class IE power systems for nuclear power generating stations.

Electrical penetration issemblies in containment structures for nuclear power

generating stations.

Qualifying Class-lE equipment for nuclear power generating stations.

Standard for type test of continuous duty Class lE motors for nuclear power

generating stations.

Installation, inspection, and testirg :requirements for Class lE instrumentation

and equipment at nuclear power generating stations.

Standard Criteria for Periodic Testing of Nuclear Power Generating Station Safety

Systems (ANSI)

:.Recommended practicesfor seismic qualification of Class lE equipment for nuclear

*power generating stations.-

Standard for qualification of safety-related valve actuations.

Standard types-tests of Class lE electric cables. field splices, and connections

for nuclear power generating station.

Criteria for diesel generator units applied as standby power supply for nuclear

power generating stations. -

Seismic testing of relays for nuclear power generating stations.

Qualification of Class 1E lead storage batteries for nuclear power generating

stations.

Qualification of Class 'lE connection assemblies for nuclear power generating

stations.

Qualify Class IE motor control centers for nuclear power. generating stations.

Qualification of 1E motor control centers

Qualification of Class lE static battery chargers and-inverters for nuclear power

generating stations. * -

Periodic testing of diesel-generator units a applied as standby power supplies in

nuclear power generating stations...

Application and testing of uninterruptible'power-supplies for nuclear power

generating stations. ;

Low voltage ac power circuit breakers used in enclosures.

6.13
| I Aging Failures _. _ _ _ J

I 5 1 ~ ~ ~ ~ ~ Diesel Generator

* 8% ~~,System Boundary\ - Pumps 2% j * -- -

- Coolers 2% 0 -/

- Filters 1%

*- Others 3% a I 3 a

10% 2%: a -Crankcase 2%

- Pistons 1% a - _

- Con Rods 1% _ - Crankshaft 1% * ;

§ -Others 5% J

a 9% a

Outside System

Boundary

is- Switchgear 4%

a Electrical 3% i

- Others 2% I

I | 30%

-Governor 12% +, - Relays 3% i I _ -Control Air 2% a

-Sensors 3% a 2

:- Other 10%:: hj...

._s -Exhaust 40/ 6

I I S

I _ _ _--- J--

.-JI-.

Only

14% of *4

Failures

Here

I ___-_----- -

: - Regulator 2% "z _ - - - - - - - - - - _a 

I - _ . Il

a - 2% _|_ _< -enerator Staning Cornbustion SyteSyste

..., ~~~I __ _ _ , _ St rn ; e _na Coi_ _ _ Fue l

S- Generator 2%~ a-1- 0yse II I'

________'__ r-...----- - -- -- --- ~ - ------ ...-- a 10% 10% t 13%

*Air Valve 3% / Pumps 2 a */ - Piping 5%

-Controls 3% -Intake 2% 0 : Piping 2% 1 - Pumps 4%

-OMotors 2%' * a 1 -Coolers 2%, -Injectors 2% a

- Others 2%/6 § 8 ... -Others 4% -Others 2% -- - -- - - -_ __ _ J I*--_--_--_--_--_--_-

FIGURE 6. I. Emergency Diesel Generator Subsystem and Component Failure Rates
* Development and AP~lication of a Computerized Database -a database

-was developed, sorted according-to engine type, failure cause,' and

aging.susceptibility of subsystems'and'components. .'

* Validation of Results'-- the results were validated.in a'PNL/industry,

workshop and by, extensive peer review. "

L Documentation of EDG Aqinq Assessment'- the EDG aging.assessment was

documented in NUREG/CR-4590 (Hoopingarner et'al. 1987; Hoopingarner

and'Vause 1987). .' ''-

* EDG Aging Characteristics - Figure 6.1 identifies the EDG subsystems

and components that are most-susceptible to aging....

* Co son - Key conclusions include 1) the principal aging 'mecha- ' * nismsi J in nuclear EDGs'are vibration,; vibration loosening, thermal:>'

-andphysicaVshock, 2)theEDG systems most'viulnerable toaging'are:

instrument-and control, fuel,-starting system', cooling';system, and

3) fast-start testing-contributes'to-EDG aging'(also acknowledged

earlier in NRC Generic-Letter 83-41).' PNL-participation'in ah-EDG " engine examination provided evidence of advanced'engineowear due'to'

fast starts.:; .

The-following are key outcomes from'the EDG'aging research: -

- development of recommendations to mitigate EDG aging in a PNL/

* .'EDG consultant workshop. . . - .. "

- documentation of the recommendations and extensive peer, review

- further validation of the results by.in-depth discussions'

during plant visits to three 'utilities

- - .-publication of the proposed recommendations in NUREG/CR-5057

(Hoopingarner and Zaloudek 1989) :: ..

-- :recommendations to-1) replace the fast-start statistical,;basis:

for EDG reliability assessment-with An engine "health check" - ;approach,-2) permit.slower.starts.andlonger run.-times and

monitor and trendtseveral parameters that. indicate'engine con-i-;: .-:. ditiona and 3) conduct major engine-overhauls based on need-.-'.

rather.than elapsedltime.. -; . * , . * ' A * ,- , , -

6.5.1.2 Understanding and Managing Aging-in..Electric Motors - * '

,,; I. t ' [ ;*s__!,;i * ,' I .

Despite the fact that modern electric motors are-better built.and-more.

sophisticated than older. electric'motors and.maintenance practices:.are more.-

extensive, electric motors still fail. Most motor failures described in.thel

(a) Varies with engine type, subsystem, and utility.

66.15
operating experience data bases for U.S. NPPs occurred during starting. For

smaller motors (under 200 hp), which constitute more than 90 percent of the

motor population in a typical NPP, the stator insulating system and bearing

assemblies were the subcomponents that failed most frequently. This accounts

for almost 70 percent of the reported failures. Large motors, on the other

hand, are equipped with devices to monitor the motor insulation and bearing

and oil temperatures. Corrective actions are taken immediately on indication

of excessive insulation temperature or rising bearing or oil temperature.

Therefore, insulation and bearing failures are not as pronounced in large

motors. The major factors contributing to large-motor failures include

voltage surges and mechanical stresses caused by centrifugal or magnetic

forces.

Understanding Motor Aging. A review of motor testing techniques documented in IEEE standards and related literature revealed a number of test

methods available to monitor the condition of stator winding insulation and

bearing assemblies. The study has also identified the predominant failure

modes and mechanisms caused by aging and service wear effects. The insulating

material is degraded primarily by heating cycles of the winding because of

starting and overload conditions. A humid environment will produce more rapid

degradation in both cases. The bearing failures are primarily caused by

lubrication deterioration or bearing misalignment. Using cost-effective techniques to monitor these two problems could eliminate many untimely failures

and, thus, improve the overall system reliability. The following information

is useful for understanding electric motor aging:

* Documentation of Motor Aging Assessment - NUREG/CR-4156, (Subudhi,

Burns, and Taylor 1985).

* Motor Aging Characteristics - Figures 6.2, 6.3, and 6.4 identify the

motor subcomponents that are susceptible to aging.

Managing Motor Aging. To determine the tests or analysis methods that

can aid in identifying age-related material degradation in motor components at

an incipient stage, two test programs were conducted. One involved periodic

testing of a 10-hp, 460-V induction motor with 12 years of service in a commercial nuclear plant. The motor was subjected to plug-reverse cycling to

induce age-related degradation and monitored for various test parameter trends

with age (number of reverse cycles). The second test program included diagnostic testing of a failed 400-hp motor with a service life of more than

20 years in a nuclear research reactor. The form-wound coils of the stator

windings were tested to determine their condition by comparing similar test

parameter variations. Based on these two test programs and industry operating experience trends, insulation and bearing tests are recommended as functional indicators. These indicators, when trended with the age of the motor

under surveillance, would indicate the rate of deterioration occurring as a

result of aging and service wear. Information on managing motor aging

includes:

6.16
Environment * Degradation of Mechanical

. Temperature and Electrical Strength

* Humidity with Time.

* Radiation * Shorted windings

* Chemical

* Contaninants

Vibration' * Turn-To-Turn Electrical

* Loose Laminations, Short

Wedges * WindIng-To-Ground Short

Blockings, Balance

Weights

Excess Current

* Frequent Starts

. Overload

Overheating, Burnout

Windings

Excess Losses

4 I - I~~~~~~~~~~~~~~~~~~

High Voltage Gradlent

a%

I-

* Coroina or IonIzati Ion of

Insulation

* Electrical Short and

Breakdown

Functional Indicators:

* Temperature (windings)

* Vibration signature

* Current signature

* Voltage gradient, corona

* Insulation Resistance

* Power factor/loss-faactor

* polarization Index

Improper Ventilation * Overheating or Windings

. Design * Turn-To-Turn Snidtt

,* Foreign Particles

Water/Sterm

, Shorted Electrical Circuits

Corrosion of Laminations

Poor Electrical Connections

Doml nant

Fail ure

Mechanism

4

t~~~ .1tl Performance - 4 -allure Functional b

(Motor Hazards - Effects Indicators

(subsystem) I

FIGURE 6.2. Functional Indicators for Dielectric Integrity
Excessive or Under Lubrication * Overheating

* Damage Oil Sling Ring * Cracking, Scoring, Brinelling,

* Seal or Gasket Leakage or Splitting of bearings

Water/Steam * Corrosion

BEARINGS

&

LUBRICATION

(ROTOR ASSEMBLY)

Vibration

* Mlsalignment

* Noise

* Cracking

- I

Bearing Insulation Damage * Bearing Current Leading to

Overheating

I

Functional Indicators:

* Lubrication

* Lubricity

* Viscosity

* Vibration signature

. Temperature (Bearing,

oil)

. Eddy current losses

* Alignment

* Clearances

* Speed (torque)

Improper Cooling

* Seal or Gasket Leakage

* Overheating

Mechanical Problems * Jamming/Freezing of Balls

* Oil/Foreign Particle Deposits * Stator Lamination Damage

* Rotor-Stator Interferance . Tripping of Accessories

* Excessive Axial Movement * Shifting of Rotor C.G.

* Too Rich Clearance Between

Bearing and Shaft

* Frozen Bushing

Dominant

Failure

Mechanism

Performance -,- - Failure +| - FunctIonal I

M Motor Hazards Effects Indicators

\Subsyst m) -I

FIGURE 6.3. Functional Indicators for Rotational Integrity
Vibration - * Loose or Damaged Leads

* Bolts W Weld Failure

* Welding/Brazing * Loss of Balance Weight

* Dowel pins * Bending or Distortion

of Shaft

* Shear of dowel pins

Enviromnental * Improper Electrical

Contaminatlon/Corroslon of' Contacts

Leads

Damaged Seals & Gaskets .* Oil or Water Leaks

Functional Indicators:

* Vibration signature

* Environmental conditions

* Corrosion

* Cracks

* Leaks (seals & gaskets)

. wear -

4.

Wrong Adjustments

and Settings

.a, Overload Ing, Overheating:

tn

.~

'-a

Faulty Heaters . Too Much Moisture

Mlsapplicatlon/Misoperation ' Overloading, Osierheating

* Valve Seating.

. Phase Error

* Frequent Starting

.Canmutator. Brushes (wear) * Poor Connections

Dnminant

Fal lure

Mechanism

*~~~1 4 - Performance-" '- - -- FaIlure - 'Functional -

/' Motor ) Hazards Effects -'' Indicators

ISubsyste )

I

FIGURE 6.4. Functional Indicators for Mechanical Integrity
* Documentation of Motor Aging Management - NUREG/CR-4939 (Subudhi

et al. 1987a, 1987b; Subudhi, Taylor, and Sheets 1987)

* Methods of Aging Management - Monitoring age-related degradation in

electric motors can be achieved by instituting an effective maintenance program that includes periodic testing of the subcomponents.

The two test programs discussed provide useful results on various

dielectric and bearing tests.

6.5.1.3 Understanding and Managing Aging in Battery Chargers and

Inverters

The dominant causes of failure associated with nuclear power plant battery chargers and inverters are overheating, electrical transfers, and personnel errors. Overheating can degrade the life of components and materials. In

many cases, the stresses resulting from these occurrences result in an accelerated aging of critical components. For example, electrolytic capacitors that

provide electrical filtering were found to be temperature sensitive. Their

failure in a short circuit mode results in the direct loss of the equipment's

availability.

To manage the effects of aging in battery chargers and inverters,

Gunther, Lewis, and Subudhi (1988) presented a comprehensive program based on

an evaluation of the effectiveness of current storage, maintenance, repair,

and replacement practices. This program addresses the integrity of discrete

components such as capacitors, transformers, and semi-conductors and other

entities such as cable connectors, wiring, and structural fasteners.

While emphasis has been placed on maintenance practices for managing

plant aging, the potential improvement in vital bus reliability through the

use of an automatic transfer switch should be considered. Other recommended

design improvements discussed by Gunther, Lewis, and Subudhi (1988) include

the use of equipment for detecting and suppressing electrical bus disturbances, the use of higher voltage and temperature-related components, and the

addition of forced air cooling.

6.5.1.4 Understanding and Managing Aging in Batteries

The most common aging problem for lead-acid storage batteries is thermally induced oxidation of the grids and top conductors that are made of a

lead alloy. Lead experiences a 21 percent growth as it oxidizes to lead

dioxide. This growth causes the plates (including grids) to swell, causing

poor contact between the grid and the active material in the plate, and

results in decreased capacity of the battery. In addition, oxidation causes

the lead to become brittle, which can lead to decreased seismic ruggedness.

Edson and Hardin (1987) and Edson (1990) reported the results of a comprehensive aging assessment program for lead-acid storage batteries. The program included seismic testing of naturally-aged batteries that were about

13 years old to evaluate the effectiveness of current surveillance, monitoring, and replacement practices. From seismic testing, it was concluded that

6.20
when batteries are maintained and operated in accordance with-IEEE Standard

450, Regulatory Guide 1.129, and manufacturer's recommendations, little, if

'any,.electrical capacity will be lost as a result-of seismic shaking at levels

that are typical of the most severe seismic event postulated for NPP sites in

the United States..

6.5.2 Results from Nuclear Plant Aging Research Special Topics

Several NPAR special topics from-Table 3.1 have'made significant findings

contributing to understanding and managing aging, illustrated in the following

sections.

6.5.2.1 Effects of Aging on Risk

Results from NPAR aging/risk studies are utilized in the'development of a

procedure, based on the linear aging model, for modifying probabilistic risk

assessments.(PRAs) to account for the-effects of aging on reactor safety.

This procedure has been used to conduct risk-based prioritization of SSCs and

is being used.in PRA calculations to quantify the effects of aging.

Key findings from the aging/risk studies are summarized here:

NUREG/CR-5248, (Levy et al. 1988) "Prioritization of..TIRGALEXRecommended Components for Further Aging Research," was issued in

November 1988. Thirty safety-related-plant components and structure

groups were prioritized by an expert panel using the Risk Significance of Component Aging and Aging Management Practices (RSCAAMP)

methodology, providing insights to the relative safety significance

of the equipment that was addressed.-

* The results of the RSCAAMP prioritization workshop showed that aging

* effects (e.g., aging failure rate and the weaknesses in maintenance

programs to mitigate aging) can.be major risk contributors.

The RSCAAMP methodology was enhanced by incorporating the interactions of multiple components that.had been subject to aging.

* This methodology, which 'accounts for multiple component aging - effects, was incorporated into software that enabled-analyses of the

risk contributions due to aging of components in the minimal cut

sets;of probabilistic risk assessments.

* The aging effects software were applied to PRAs for-a PWR and a BWR

to generate a parametric study that assessed the impact of aging on

total core melt frequency (C). Aging resulted in' significant

increases (factors of;10 to 1000) over the baseline plant core'melt

frequency (AC) for maintenance programs reflected by component

renewal periods of from 18 months to 72 months and for an aging

failure.rate corresponding to.one failure after 5 years... The interaction of multiple aged components was a major cause of the

increased AC at the renewal interval of 72 months.

6.21
6.5.2.2 Residual Life

Residual life assessments for major LWR components include preparing summary charts that highlight information useful for understanding and managing

aging. The summary charts identify degradation sites, aging stressors, mechanisms, and failure modes; requirements and recommendations pertaining to

inservice inspection, surveillance, and monitoring; and actions useful for the

mitigation of aging of major LWR components. Figures 6.5 through 6.15 show

summary charts-for the following components:

* Emergency Diesel Generators (Figure 6.5)

* PWR and BWR Pressure Vessels (Figures 6.6 and 6.7)

* BWR Mark I containments (Figure 6.8)

* PWR and BWR Pressure Vessel Internals (Figure 6.9 and 6.10)

* PWR Cooling System Piping and Nozzles (Figure 6.11)

• PWR Steam Generator Tubes (Figure 6.12)

* Pressurizer, Surge,- and Spray lines (Figure 6.13)

* BWR Recirculation Piping (Figure 6.14)

* LWR coolant pumps (Figure 6.15).

6.5.2.3 Shippingport Aging Assessment Results

Decommissioning of the Shippingport PWR afforded a unique opportunity to

conduct onsite assessments, acquire a wide range of naturally-aged components,

and obtain a variety of radiological and materials specimens for testing. The

scope of the Shippingport aging assessment is outlined below:

* in-situ measurements of circuit integrity on 42 electrical circuits

* in-situ measurements of ferrite contents of cast stainless steels to

select specimens for subsequent studies of thermal embrittlement

* on-site drill operations to obtain eleven 15 cm-diameter specimens

from the RPV neutron shield tank to evaluate low-dose neutron

embrittlement of carbon steel

* acquisition and shipment to participating laboratories of approximately 200 naturally-aged components, associated manuals, operating

histories, etc.

examples of technical and regulatory results that are emerging from

the Shippingport investigations are:

6.22
Understanding and managing aging

of emergencey diesel generators

P~4dipa Oweol Engne in N&tw Servic_ -

.Marufeturer - -

ALCO

Ai Ctarnr

Cooper Besern- .-- -*--'- - --

Fairbantk Morse

Etoleto-U ive O eu'on

.NowdbE l

Trasnamerca O"a

materialtpcat

'Alloy atos.wlded st el plates. caa*on kixdkhng'

gray. aliuraM atelte seas kroed atelsb

- iefo non mere gasket. hoses. '.

HP 21510 C70 per cynderr or 800 HP b 89 HP

KW rsingas 500 b 0. 30. GM0

* COOsg vmtr. kAINO wi. d ibV d . .n'ake and e9imot detariomon dy iec Ovas.

* vbadon. bA. gL weo and hash tsaq

Cytrdwt4od Crabea -

Ai bt PFam -

Al;,@l

-P-a "Sam

AsS 0 bu eas .- 

CM,~ ae4w a Ssv.CEM Pws..

am

.A

UNDERSTANDING AGING - - ., . . -.

(Stressor & EnvfromentInterations) . , .MANAGING AGING

SitesJ -AgigConcerms - Inservivelnspedion. Servellance ... ,,. Mitigation ¶

.* hrsb~att and c*old rsystes E~ wnay induced: duaL wat.t - - 1ted EDG pvgram ol sing.

. t , ,r Governor ., .. ha. ci Ctt iC5-, -C .A*GI t stsvdl e . pfel~ toeg ................................ pec. ntoftoung.- heat ok Corn& mm ~ ~ ~ *and Isntkamnace Getislies: I . * Makorince ermes: Inadequate frnkMn., ' -FIo 1.¶0A..rout. teein. -ybewliani due

,rSbI, ,ts 5tCo "to , AG irs0.Sion5j .- , * T t.,a s, rsd. an toting * Stup coEow5 Fastf t mid Oer regulatory ir&cd * Genel eer 8141_..1tt tOW - . to a 'tom San tet and

Fuel * Fwatern , , . , . _ * 10CFR-50.appendhAcattllo4.517.18 of. to Piping on engine D r i o P ^ _ b t sdir p s o

.icl hsr O pn ow Ixol *" - t*10-CFR-5.-s- B.soeon oprx Xt r t,;,-:; Sy~~~~~~~~e, .... eco ~ eqive.' I~~~B .

.S* *Opaboa . .-I~rme ,; inad"a trainingenom a atstwn io Pei bi anitec"o krnprvd 'iswejon e, *eedy.

I ' zCOE'tola .9and dull$ * 1 O-CFR-50. 55% COdes A Srd -ASME PVC . mon~f. avd y b deterrnine

Sthtrg 61iia0Y ' *VI~i~' I,-d secd on NI. K. IEEE STD 279 ron smotors moe Sterling fnot - - Fuel or Ukralon deganmfon- - TS 3Y4 8.1 stA lancebApecdon fr opefor 5..- Pffd

AUr oewnO * Gasketu. . or orgac tneateial degeneration .otb st n , , .

Sw g athneagr *ya , ;hdeqxt 4p-rer Wty. sU g. e. otdering * GSt B5S-k pow. rllablity cd En4 * Uiiao o Breakss- protsne. data and spacatfons * IEEE standard acceplaoe for us. by NRC-Rt5 1.9 a h4jpe Og. do not over hat

Relays Cowosion otdalon -- - urhios Inspedon and ben hnstwi and conlros Thwra gUt , rus

COO&V s "cona r no Maqfua - 1i t y ., : .. .i

P1m" Fadgue not rlaed to Ybradon -edtralringhrEW b 

Oxc s Metal h1 ,, Staff in on-l nsgwwe

LUbrlc&*V Worn Be avial adokr, -Systenmodlttionsto rniftot

He *dera. .. .

Lube l . .. ^ Lubo l , t ,.. ,. .. * -,. . : .

FIGURE 6.5. Understanding -and Managing Aging 'of Emergency Diesel Generators
Understanding and managing

aging of PWR pressure vessels

W*%,Ws ~~~~~~~~W&. v.404 L*wOa6.uauIMA.IA4 &1e.I "_ 30 -;

. Tr" W " 4A HS

V.4b.l * 10wgWOygh _i.

Mhl =04Mg 610

* *rdb14 - 06 i.Wd #444 WiMACwsle " 1044 0..X 64 Chm 3

-' 50001.1 HgM hiMg e_ lbe - . -CW m .I_ Isde-w

-^oe -zw * | zUp she

I Uwwsl.Id lW -

-00 -_ ln 300

111014 ~a~iI. o 4eCIIwf W" OWWq IN. a m'tsa.

-mw mid 5199L . VA*1PWI."MA !h"4 by.5d iW*dd" SlM

UNDERTANDING AGING - MANAGING AGING

(Materials. Stressors, & Environcnent Interacllons) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Sites Aging Concerns Inservice Inspections, Survellance. and Monitoring Mitigation

hlM _o .~so daiso ec_4Si.-Id *#4I. W.use"iba:,C . . . . . _ v ,ua 

O.Mgid .wqIC* pi II-SO OMIWI- ,SU _4m _ 'I-$ Mg _ .00000 so..ine

___ MgM~p14&,W*W in U se tomanOWI..so.414010. ie . D" In vp i *e sew f. Gums L.% A- "I 23 * _4 -- -5. gd .D~~~~~~~~~~~~WSI ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~.4Is~ .Me=1_

* 144 Nu4mS "B*A11y. P mdisg hiog shl SUOKI In4I14 k~mpemis Fr~ (4MCFr, M 05 DoI . _II0 14iW _de msm

I S~~~eo ~pmClr.... .q1. 0 __ 1441 w .4 WebW pou

_ 1M t__P94rrq~ A% mIi Mr8%M ft -d h _ 

but.14 Mg"h"i.di polmi WA' 4 At. .. Ogo e.P 24ao

O.5. Wm- back" (Am.l 89.. 14 pboaphorw " k.6 co9w .MeMO. 1, C., SO A9 6

.- ~~~~~~~~~~~~ . . We U 400.d404'1 64w swq*_ 06mb. A-W 14.peMso fd.

[bn -o ... 10J0P 6I-. * 0-*4- PMAft"

_RS" Sp r _-" w_ h 

.~~~~~~~~~~~~~~~s _h _ O"O UX~ 4 CF

-~~~t - a. .bb I 5 .f .~ 

luhabI sI-swiso .4.6 M~iS WIw_ $4,0e mo& 91SK 

~~~~~~~~~~~~1601. g. 0. se4:r r .- t. I

(if FR U41f . "" . .

o~~~~~~~~~~~~th. rg IbS__i_ ff 141 _ _ .ew 0q .~~~~~~~~~~~~~~~~I CAMi. 11414100^iO..iWhie saul. £ w...4.h.m.Su.. _ ~ "44M" a~1 &W OmMjig a M M dit _40 m Mg. ps14Lha .05. _1655.bs..h (O1560660of i " "W4.0 14teo"Oh

FuI s..4M.. db~oDv.6i

2G tJK haA440id*W *) WA6t Nui.Mii. o1_ Ohll_ _g4

~~~~~(f" woo of _ 4 _me "d_ wom _A" of _ _ .md

caou~~~t-k ~~~ "__ suds ___ _ _ _ _ _ _ _ _ I-d MA*"maw Bk o zz-w

________ *_____________ *1__ 4_u*e _ ..

C% glef "i w.A 0m I. g NOW .. . A" W. iiu ft id 3W _ 1,00 W -W -WM.Wow W . -. 5DM-. 

FIGURE 6.6. Understanding and Managing Aging of PWR Pressure Vessels
Understanding and managing

aging of BWR pressure vessels

Ond

En-kosdv"aa

Vasael *Low W"e Ca.ou, ams. SA65338.t. &A.202

M.Ctsfng ly3OeSS and no S

Sale Ends Typ 304 SS. rope 215 SS. Incoral Ss-ttInesl531 -- A-3*4051

Thaemal Sletws. Type 304 32S coowsu Studs * 5*40 at. 3220or02 Weldawfita SA*193O0. 8?

operational twlrss~ota Psiruon fluxj and Rfll0 90.tur. laO coolerd and preasode

UNDERSTANDING AGING

(Materials, Stressors, and Environmental Inleuuctions

Sites Aging Concerns Inservice Inspection, Surveillance, and Monitoring Mitigation

071

U'

Foadeew nw ,e NW ad t hCyC1O *ansl Matge cOUsd R R C RaQU m ni . RModdy design. sle V end sW.f by eawtar le uakag.."e m a"" M S weie penttgm

VAeb, t& 4oaoardn .isdo n 10s0n 40 VO asl., (donitorng9 pipe a s*e 'ee 'a end neb "m adt E. .o p-ntel b - . tmd - ndlu tt du 'e eoM"sd. nos. t, ted . P -Ww. etd."t"

each atapec" kparwal (WS-2w0) Oetesp ae ir a sewing hlh4 P. 1n ha

_ _ _ _ _I- _ _ _ _ _ g ome-d .

'A m a ss n a k . e a d e t o ud

RsIdeloi Net%~ K5CC iek kVed h HAZ eOY I pagate al _s b dur met h O wi OK% lspectioi D to on- .Corrosion nto ngemtry t

rus, " dululr knt base mt Am _ OIWS-2#0 E, hbns tlebsng tdlt tGSCC:: o .hydrn

ml wk* tub" kw : eh - .ma.dge

~~~~~~n l u b n w t l s is es 'n r r e lf ( IW SW 2 50 0 ) 'c ' n

Satins R egion blsdstlo n ?oInlitbnmssd Swystoane. p .0g... 5o assess shiftW R in f RZi f .se RAT RA g. GU'" tK Para. 2 to .n ruce an n 1ang

... _ .Chemca p,.Itis .1 waas .mgais s and drop in USE (10 CiR s0 AM Iac s. wospiorus f copp t ASTM 5

. (Ca.td.P) SF- R~P . 0 tittR - 2) content hi sw

.s In h rek a o r&4u tr 05* n Da _age gralu on (5 CrR 5F App a) usa Stste.ol6th-.Wt kap"cto- sang and

. , E . .. . . __ .. me i n f Ihsa @ tlef} ............ .... detect delects,(. . ,i. sihz g d - -'.

. a . _nfflft Don b p tto e eecftk. InaerIsce e h.etdm

nsut * Iselotl hogue t( IntkWave pom raqalre~~~~~~~~Se 3st w iant 10yCrosai al a lto presure 5 nteo. est GPM topcts probe ositdoing f e

: .~~~~~~~~~r s n . . . r* h w b d I c t e ,¢ t dS _ t u g s o n s 'a t n cits is nf e l md. as. . . ......CIt ptecitm pro be h ab p Kse w 

.. . .-. . .. 1 : . _ ............................... -- "a h InNob 0 w ~ pogramt 

eacith "ct a i) C 50.555. Dv p use of ,,gonltlsled and

.W 3.5K. Rag. Guida 1.150. P" 1) e*" m poem .4 and cet le

WsadL~s of reconoda d pek" on .~ .1 ~~~~~~~~~~~~~~~Ps .P .~bae . kadW-_0e

Use I _ c pract Vow s e e v AS E' Laskaga md 1 A 05t 00 P' OWWI~ 1lotte (13 CPR Soctd i It. Appandix A)

.000. .W O- . lo

d bect reck grow S (nn rmd

appends is be"n deWsaped by ASUR ._ _ _ _ _ _ _ _ _ _ _ _ .'OSection XI)

Clowew Spafdt -n *n ti gnf elude .. *Pe e s daring had -PIa1am t mit ;pred Oaet s du ng asi

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ F r e t t n k i sp a c li n i n te V l ( tW e ls4 5 0 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

E. tenal altectai sna Lo awycse therma VW mw~ethnial Vsga olu mr lor or Surtae. e aO lnM a tlon (tW e-S`2 "Ieed5 rich US abk l

up a mo __le_ _ _ _ __ _ _ _ _ _

re..S

FIGURE 6.7. Understanding and Managing Aging of BWR Pressure Vessels

I
Understanding and managing

aging of BWR-Mark I

containments

Ma16 la s Shell .Carbon steel. SASI56 GrJO

SA*212 Gr.B

Bellows Type 304 Stainless Steel

Coatings. Zinc rich. red lead and epoxy

Stressors Corrosive Internal environment

and lemperalure. humidity. oxygen

Environment content. degraded fill material moisture.

microorganisms. cyclic thermal loading.

and teak tests

I1

Embedded shell

region

<d' Waterlevel - I

P esure suppressioi char

BWR Mark I type metal conlatnment

Downcomr pipe'

BWR Mark I drywell base, concrete shield wall and sand pocket.

01

0'5

UNDERSTANDING AGING

(Materials, Stressors, & Environment Interactions)- MANAGING AGNG

Sites . . Aging Concerns, . Inservice Inspections and Monitoring Maintenance

Exterior surface near sand Pitting. crevice corosioo. Thermal. mechanical. NRC Reum RcoCmendatios

pocket (unsealed gap) Mirba Influeanced corroalon, and onvwrnonmetal NCRurmntReo edais

and uifor corrsion fatigue .Leakt tests (ID CFfl SO App. .J). Maintain surfac, coatings

Exteror swfti;e Pitting and crevice corrosion Recommendations

(degraded fill materIal present) inservice inspectlomASiE Check bellows alignment subsection IWE

Embedded shell region Pitting and crevice corro~sion Being reviewed by NRC lot maintaint caud ointsa

Drywall (deteriorated caulked joint P inclusion In federal regulations mbedmentreion

at concrete-metal interface) * Visual examination ol 25% of Evaluate use of cathodic

pressure retaining welds, and protection for exterior

Pipe penetrationa Galvanic corrosion - coated and uncoated surfaces surface of dry well Vente pies cngson= durisGavni eachInspection Interval Vent pipes Visual examinatlon ol caulked

joints at embedded regions Exteriot and Interior surfaces Uniform attack Boroscopic examination of

(deteriorated coating) exterior surface near penetnation

Surface examination ol bellows

Interior surface Pitting Thermal, mechanical. Monitor coolant leakage (deteriorated coating) and environmental from faulty bellows

Pressur fatigue, Wall thickness measurements anid

Chamber Near waterline Differential aeration ltrending . Develop methods for Inspecting

Below waterline icrobil Wluenced corrosion metal surfaces through Microbial corrosion trifluenced ~~protective coatings

. Magnetic particle testing Heat-alfecled zone Intergranuiar stess corrosion Thermal and * Eddy curnt tsing

cracking mechanical fatigue Develop methods for inspecting

Bellows drywell wal In the embedded

Cold-rotted ponion Transgranular stress corrosion shell region

cracking Electromagnetic acoustic transducer

irev. x

FIGURE 6.8. Understanding and Managing Aging of BWR Mark 1 Containments
Understanding and managing

aging of PWR reactor internals

Materials Instrument tubes (flux thimble tubes)

Bolts and pins

CRGT split pins and radial support.

key bolts

Thermal shield. core barrel, baffle

and formers. upper and lower core support structures

Flow mixer, cruciform

Instrument guides

Hold down spring

A !

Type 304 SS

Type316 SS

(W. except as riOted below):

Alloy A-286 (CE

and B&W)

Alloy X-750 (W)

Type 304 SSCast SS (Gr. CF-8)

(some early W plants)

Type 403 SS

(WN and CE)

urc* 7--A

CRA V'mI-2

V.-Co hbe

ICo, bwal

Fuel

0-0,19S Stressors & Operational transients, temperature, fluid flow.

Environment residual stresses boll preloads neutron radiation Typical Westinghouse

Reactor Internals .-

Typical CE

Reactor Internals

.UNDERSTANDING AGING

(Materials, Stressors, & Environment Interactions) . _MANAGING AGING

'Sites Aging Concerns Inservice Inspections, Surveillance, and Monitoring Mitigation

NRC Requirements Recommendations Use thicker-walted or double- Instrument tubes (flux thimble Fretting, wear, high-cycle fatigue Visual Inspection of accessible Develop ASME Code Section Xl walled thimble tubes (W plants) tubes) . subcomponents (ASME Section XI, inspection requirements

Bolts and pins G0SCC, high-cycle fatigue, stress Table IWB-2500-1 B14.10) - Reduce preloads on high. relaxation tsto etig Monitor the wear of W in-core strength bolts to mitigate IGSCC Preoperational vibration tesfng instrument housings per USNRC

CRGT split pins and radial tGSCC (USNRC R~s 1.20 1.68) Buttetin 889 Use improved heat treatments

support key bolts (\) - - Loose parts monitoring systems Use advanced inservtce Inspection for Alloy X-750 high strength - (USNRC RG 1.133) techniques for pins (e.g.; Framatome bolting materials to reduce Thermal shield High-cycle fatigue, irradiation REBUS UT method for split pins) IGSCC susceptibility

embritlement. distortion Evaluate the use of cylindrically guided

Core barrel, baffles and formers High-cycle fatigue, irradiation wave technique for bolts Remove thermal shields (or re- place with neutron shield panels embrittlement Establlsh vibration monitoring programs on exterior of core barrels) for

Upper and Irradiation lowercore embrittlement. support - ~using ex-core and/or in-core neutron designs that have experienced Upper and bower core support tnradiatbon embrilement. .usnoise detectors to detect excessive bolt tailures

structures corrosion-fatigue movements of core barrels and thermal

shields Use one-piece thermal

Flow mixer and crucilorm Thermal (ferrite phase) and ir- shields rather than multi-piece instrument guides (some early radiation (austenite phase) Use ANStI/ASME OM-5 procere bor designs (RPVs in some older

W plants) embrinttement axia pret arre W plants have inadequate

access for one-piece Hold down spring (W and CE) Stress relaxation Installationt

PAM 0,Jsmt 1"

FIGURE 6.9. Understanding and Managing Aging of PWR Reactor Internals
Understanding and managing

aging of BWR reactor Internals ,. ,,*. '.- I

Materials Attachment weld material

Core plate. core shroud

Jet pump

holddown beams

riser support braces

welds

castings

Top guide

Core spray and

leedwater spargers

Ofifced fuel supports

Peripheral fuel supports

Baffle plate access hole covers. core support cylinder

Shroud head bolts

Neutron monitor dry tubes

Alloy 182

Typo 304 SS

Alloy X-750 F

Type 304 SS

Type 308 SS

Gr. CF-8 SS

Type 304 SS

Type 304 SS

Gr. CF-8 SS

Types 304 and 304L SS

Alloy 600

Alloy 600

Alloy 600, Type 304 SS

Stwoud POW bofII*

Cwewa sparSars.

I PAV mo" _b _

BwPi.-

-Its

Ea"?, pla1

Stressors & Temperature, residual stresses. flow-induced vibration.

Environment bolt preloads, dead weight, neutron and gamma radiation.

chemical impurities and oxygen content in coolant Typical BWR/3. BWR14

Reactor Internals

Typical Jet Pump

Assembly

0r)

co

UNDERSTANDING AGING MANAGING AGING

(Materials, Stressors, & Environment Interactions) _-

Sites Aging Concerns Inservice Inspections, Surveillance, and Monitoring Mitigation

Attachment welds to reactor vessel IGSCC. corrosion-fatigue NRC Requirements Recommendations Maintain strict control cn coolant

Core plate, core shroud iGSCC. irradiation-assisted SCC Visual inspection of accessible Develop ASME Code Section XI impurities to keep condmc:'vily (IASCC) components (ASME Section XI, inspection requirements includ- below 0.2 P.S/cmf Table IWB-2500- ) ing volumetric inspection (UT) of Use hydrogen water chemistry to

Jet pumps Holddown beams IGSCC. high-cycle fatigue br~ e~ attachment welds and of compo- suppress dissolved oxygen in

Jetpumps suppot Riser braces High-cycle fac tigue Preoperation vibration testin nent welds such as the jet pump coolant from typical 200 ^;; to

Wie upobaeldsHigS-CCl ftge(USNRC R~s 1.20, 1.68) inlet riser to thermal sleeve below 20 ppb to mitigate IGSCC

Wetds IGSCC Loose parts monitoring systems welds and core shroud to core Assess effect of Hydrogen Water

Castings Thermal embrittlement, (USNRC RG 1. 1 33) support cylinder welds Chemistry on

IGSCC: (if ferrite content is low) Develop UT inservice inspection - IASCC

Top Guide IASCC, IGSCC techniques for inaccessible attachment - fatigue aack growth

welds to reactor pressure vessel -ftgecakgot Core spray and feedwater IGSCC. corrosion-latigue - radiation fields

spargers Use in-core SCC monitors - fuel performance

Oriliced fuel supports IGSCC. Thermal and irradiation For replacement components

embrittlement. IASCC

Peripheral fuel supports 1GSCC. IASCC - Reduce preload stresses

Peripheral IGSCC, lASCC luel supports - Use high temperature

Baffle plate access hole covers IGSCC (crevices) annealing and age hardening of Alloy X-750

Shroud head bolts IGSCC (crevices) material

- Remove crevices Neutron monitor dry tubes IGSCC (crevices). IASCC

FrM- e.1. tft

FIGURE 6.10. Understanding and Managing Aging BWR Reactor Internals
Understanding and managing aging of PWR RCS piping and nozzles 161efrals main Coolant pi. Cenirifgaly cost SS-Or. C75A end

CF58M i type 204SS VW 3153t

(early W ptants), SA-SIC Or. 70 ().

SA-lOb Or. C IBMW)

Fillngs .ta1tealty eat" 5S .Or. CF&A en

CrtM On IA-5ItS Or. 70, Type 309L SI .C. BSW):B Type 352L SS (0BW)

Cla~ddig TMP.3051SSIC!. Baw)

Yp 304 (S &W W ) Surge end spray I16 et Type SI. S aSt 55. Or. CFlM

- - (sri Om hIn ses" C! plants)

Chtarng. safety w tnl i.n. Tpe 316 SS

*nd residual tYat reovl

tSnes , ~ -'

WoNt$es on mtn coolent * SA t05 Of. 2 (CE). p 304N SS OM

Saet Od . 316 SS. Inswoel SB164 ICE. saW)

ThermSl sl lnosnel SB-IS$

Strisors 1 Opein t al transients temperatue tobw ntduced wvbrtions, stratilted EnIonment nl . thermal stripin. vt fekag, nd tad mat ShOcks

Steam

,g-,V*rMtor

,aover l9

r'3

%D

UNDERSTANDING AGING MANAGING AGING

(Materials, Stressors, & Environment Interactions) .

Sites . Aging Concerns Inservice Inspections, Surveillance, and Monitoring Mitigation

Nozzles end thermal sleeves Low- end htgh-cycle -NRC requirements Recommendations Maintain full flow in spray

ChsrgIng - - thermal end mechanical Igt"u Perform more frequent examinations llne and opernte It continuously r9 ne - Votumetric and surf examination of nozzle welds end horizontal to prevent strltfied flow

Safely injection of 25 of butt welds, Including the piping welds having signitlcant and thermal shock condifons following welds durIng eech Inspection fatigue damage end spray line

Surge Interval (10 CFR 50S55a, IWS-2500): Determine fatigue damage by on-line piping by replacing short

Spr - - * All terminal ends in each pipe monItorIng of pipe wall temperatures horizontal sections with sloped ___________________________ ~~~~~~connected to vessels an oln lwtmeauead sectIons to prevent stratified

Terminal end dissimilar ~~~~~~~~~~~~~~~pressures In nozzes and horIzontal flow conditlona Terminal end dissimiar Low. and high-cycle thermel * An dissimilar metal portlions of piping subject ton e meWal welds and mechanical fatigue welds operatIonal transients, Including Redesgn pIping to elminate (between carbon steel stratified flows and thermal sleeves deleterIous effects of valve components and stainless All welds ha-ing leakage steel piping) cumulative usage factor Perform nondestructive examinations equal to or greater than 0.4 and loose parts monitorIng to Minimlze velve leakage Surge and spray lines tLow. and hIgh-cyce assess status of thermal sleeves MaIntain smaller temperature

thermal and mechanical fatigue -All welds having prImary plus Develop use of acoustic emission differences between Charging, safety Injection, and secondary stress Intensity range method to detect crack growth In the pressurlzer and hot leg

residual heat removal lines equal to or greater than 2.4 Sm base metal and welds coolants during heat up

to first IsoStwon v rv I Develop techniques to monitor actual and cool down

Inspected during each Inspection degree of thermal embritllement In cast

Cast stainless steel Interval stainless steel piping: Maintain smaller temperature

piping.... Thermal embrlttlement - -Analylical degradation modelling of Inservice differences between the pressurIzer end spray line

Hot leg . Coolant temperature Flow evaluatIon - Metallurgical evaluation to coobnts

- Ferrite content and (tO CFR SO.SS, IWO-3000) characterize mlcrostructure

Cross-over leg spacIng .NOE to establish correlatIon

- Chem~al compsition Leskage and hydrostatic pessuve * between ultrasonic attenuatIon Cold leg of base metal tests (to CFR SOSS, IWASOO) and fracture toughness

Fittings Cycle counting of sisecifled Monitor valve leakage In safety Inlection Filling$ Cylel countingd sra eieed end residual heat removal lines

Surge Dine t(Tech. Spece requirement) Develop UT to delect naws In cast

stainless steel piping

FIGURE 6.11. -Understanding and Managing Aging of PWR RCS Piping and Nozzles
Understanding and managing aging

of PWR steam generator tubes

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Understanding Aging

(Materials, Stressors, and Managing Aging

Environmental Interactions)

Types Sites Aging Concerns Inservice Inspection, Surveillance, and Monitoring Mitigation

U-bods. rad t110. and PWSCC (Ptwow scq C Rld) Reeomwiodl PIbo 4t bv"n cantlO ion aftdfy 0,554. dalont arned .091.. 15. 01W. no-nlneg 5l;f 0th..uWbv s h55 5, bWiaosn

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pib. P"Llids Ubn poln ad Ownor Cmvp' 4odlinsi s bor gtlos. Un 5ftutwn 0d rnhsp .(Ontl ) te"d utd01 of g s donhoden nd sW and ftamIeft condenser lsbagr as

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aotM e egth 01 0*150 Li Inpcto res9db. 0* and ngM96. OM sgw'al Uj-bqndS lo

Cels 1. Lizl ol onw-* j stem qwsq MW nsmlr Istos r1 P ts

wo"NO Sentg9 0op099sf fw .olod " Oflsf ofd 01 Apply fo"l platig on Us Inner sudate

deposIts Li td ofd 0Fmqnc1 h 9-Itim s ns vlous sham desigb ad th kol to psnne PWSCC 0i50

nun of lub1 to beb k..en and p.0O

Inspecitd (anowes o 3% 0 PerN1 buor0ce Ispsotion

.~ lob..) .., 4.0001kM h t10 >wpiogS U.. DobS *01091 @9010ce lubr" sbotS by Tack. Sp"As' .Osld* L43) e e sik *4it 94*55 all.Ane

St.1d,dsf h sOw 0 ble P.., ellrn~st.. W065055. skidgo ls.lCkj

hdO m otop 0s99*4p1~~~51 brb4nc.Uog d Hlm~ hst sisamioe ndrsn Ow cearii.m oe "dodk

~~~~~~~~~~~~~901165_ _sf0 bolg p.psr) 10 05f0 IcAK;5CC p neo weo being _ p.*

p54 1) EXAnute copprs plkW by use of 55s15s.. Of 55sbds wtool Cond~ns

Cordsd podls botnon NW FrPon FI.. *"cpsta cdtsd e lobs. SM VW 00995f*b4Ihr g

SM snIwblon lbw datoimwd by 1.0 SpO. 1 b. S the le9 dM0 t

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6"sq .1Asht Wastasge CO0le dh l d cb.l.g" h- - Uon hot bilnds SM noosafty 01 Phiogal ftoahng. SM *s5.50 of ?4.00 ____________ ~~~~~~~~~~~~~~~degrded kho. Chew.,..cm:0Mn.000s, L ton fo

1g Gowdhe tLIl) "Su"

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FIGURE 6.12. Understanding and Managing Aging of PWR Steam Generator Tubes
Understanding and managing aging

of pressurizer, surge and spray lines

lewa1

14.0se

OwkM A-= OL 01. U. I.~ -

Cledding. Tile 04 SS£ PWI-CFe A"o

* PweS% indInewdeft Ugo

.a0elSSat51Oc CF&.d CFWA CM; 01AS &1 7. M Ty" 3M0

51 ICl. U&f Type WX SS IUM~

- iye0W SS I=Too 304 SS

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81.... too-F

D. ... :el..

gin.. lo c:ad "Ild

K- - StgoIne *7ype31SBM cadSS-C C8M

soemymns -T*?ppeStS "enines n MGMnccla ki *BA 10 Oc2 IM ?M304 M S9

Thowil 1110e -Mntn minl

Opm-a -wt %nnw~wvk fle kakd 'ftdenOM

se acama1d ft% for -UV -w~ hO-CI%4 be-dI

Em1 aeN-ww. owdkadd -wan.sd oee - hwwAg aMvsi nd owsi nla enlen

Tlydcd p.w,.aii end ~owne

o0s

l'-A

UNDERSTANDING'AGING .MNGN GN

()Reieriati. Stremicor., &Environment tnteractlon.) MNGN GN

Slem Aiging Concern. ineorvIce Inspectfone. Surveillance, and Monitoring . . Mtgto

Noead. end tVn-d aft-we Low and high-cycd. Owod NMicg~edn MAWAMI U~ ft. In .pey lin. ad

* intbe..4d~~~flen E..den . t1kana4ift dtil . aMANlOOR ci 25% OF 1`d.... PWO. freqw.4i ..andaen of Pveft %. beld ani ege0"ed so.w end eWedW se

* 3.~~~~~~g. ~~~~~bot . ".Id IneIing I* blolwi; a Aed, *a-h Molt eoucodee WnHge ledoc cod.

* Sp..y - h'e~~~~~~~~&Pftden *ua-d 10 CFFI 5055.6 PW825te Swer.. D~~~~~~~~~~~~~~~~~.* od.nen leg.d~g yc-n ..i~nc ol Papiac. bwtzw.4 section ci way.

Al dekolA, nweld o anft&d p~pNg korO.,des, wsd fl e ie hIn nored., end am Ath eloedeec tonPt.

Twino~ end dteebJ gWed Lv.4Vpel Owgd end *Al welds .hai cw.eda&* wage Ia hodotaw pctln im dpipngdu&oing vewoendsande needed Re.. MOnddc

wedi 0a0e ceibonaa *661 od~ liege. .qtud i.W g.eote Vft 0,4 abaVd fi-k. end Seed shocks

CO.o'..pe.. we tJ.. Al -*&fbS. hdn Web. fre-ney -eg ci Fhd~elg pipin in .Sekwf -de

St" pipin 14 S, Peion. iedesbWde &,,Jnalon. lead"n. pcte 6ake." mencin to asessteen dle V-Send ale.. daet

gao,. Ied, we .qu.d lo be Insoected bpawe NM~ "a~e la dat crack go-Og In ft. besa Pwedof ,pdo

dwan eachb inpco klae awed end welds wAkhww" Ne ho. ee. N.plaama.

0Mg uianen dad pping cid ovwOd end machawd. call 0*0 swer PL-4 Cal fisgu Flow dcet'n amd aneleede 10 CF1 Omlop laddoqe. in anine act degee of Sewwd the inewe de pecinG

* ~~~~~~Thanca entegd75506 VVI-WO eg.d.Ic npwEd aSmoodednd 4lC~d waatais

Stess pacai&jg an NW.5515. FAS-WW.

d. da*5 A M" -ajr.t. d ekg

D4Agoc We ASWA Saeto U. i$.50 and AWEA 5aeOn baS.. d..daeen XI.0 Dae.lUT tlo doted Reee en cad ebita steelda pipin

Cycle coAv et Ni a11ild dadiga Senseants

Heo meal. Wks 8-eaLOCAdobseet. fe.n - Qc Space, ftq**ean

Cycle cwas bt dpcfedsicgn

Vessl -A H i-gh end lo eveie *-TweedO" e

.5w A .,le

FIGUREE6.13. Understanding and Managing Aging of Pressurizer, Surge and Spray Lines
Understanding and managing aging of BWR recirculation piping

Materials Piping

Fittings. pumps

and valves

Sale ends

Pipe-to-pipe and

pipe-to-elbow welds

- Types 304, 316L. and

316NG SS - Statically cast SS - Gr. CF8

and CF8M - Types 304.316, 316NG SS

and Alloy 600

- Type 308L SS

Nozzle butter

Weld metal Low alloy

Crevice cracking \ \ ~sleet nozzle.

Sale end'w

Type 304 SS thermal sleeve

Crack at recirculating waler inlet nozzle

thermal sleeve junction.

Safe end-tonozzle weld

Weld material - 3086.308 SS - Alloy 1-1 821-82

Nozzle butler - 308L SS o Alloy 1- 1 82

Sale end butter - None - None or

Alloy 1-182

h < Pump

\ z R 4 | < ~~suction line

'- $ Recirculation W i _ 

arge Pump suction

valve

Pump

discharge valve

Stressors & Operational BWR recirculation piping system. transients, residual tensile stresses. applied stresses.

Environment oxygenated coolant. impurities in coolant. and contaminants on

the outside surface. Typical weld overlay reinforcement.

"3

UNDERSTANDING AGING MNGN GN

(Materials, Stressqrs, & Environment Interactions) |_MANAGING AGING

Sites Aging Concerns Inservice Inspections, Surveillance and Monitoring Mitigation

Weld sensitized heal-affected IGSCC NRC Requirements Recommendations Employ weld overlay, mechanical

zonen Furnace sensitized sale - Inside surface Ultrasonic inspection of weldments as Develop improved methods to inspect stress improvement, induction

end. bimetallic and trimetallic -tentside surfae.sif high residual Reduer d GesnericLetter 8d8-01* weld overlay repairs. heating stress improvement, and welds tensile stresses (may be Reduced inspection period andclmigdvefo pain

(Types 304. 304L. 316, 316L SS) introduced, for example. by inCre siztpn For leakdetecion: welded piping

stres impovemntnd made of non-IGSCC resistant

contaminants are present materials (e.g.. Type 304 SS) - Instafl moisture sensitiveltape on Use Type 316NG SS.

susceptible region of piping. corrosion-resistant cladding. Sites with crevices and small IGSCC - For weldments without crack, solution heat treatment, and heat

amount of cold work depends on when stress - Use acoustic emission monitoring sink welding for replacement improvement implemented method piping

Sites with highly stressed regions Transgranular stress corrosion - For weldments with cracks. Conduct on-tine monitoring o1 1 7NG SS

and large amount of cold work cracking (TGSCC) depends on whether cracks have coolant chemistry pvaluate the use o 34 as (Type 31 6NG SS) been reinforced by weld overlay piping material or mitigated by stress For effective Hydrogen Water Assess the effects o1 Hydrogen

Sites subject to cyclic stresses Fatigue i C c Water Chemistry on overall plant - Inspection procedure for weld - Maintain electrochemical operation

overlay repair should be capable corrosion potential on standard

of detecting cracks that were 75% hydrogen electrode scale to - radiation fields

Duplex (austenitic-ferritic) Thermal embrittlement of high of the original wall thickness <-230 mV. - fuel performance

stainless steels dela lemte (>1 0. ferrite) - Inspection period can be as short - Maintain coolant conductivity - acceptable length and

- Fittingspumpandvalve regions. IGSCC of low delta as one refueling cycle, and frequency of short outages of

castings ferrite (<10O% ferrite) regions sample size can be as large as all -. pcmhydrogen waler chemnistry

webs l

Revt. 0. Jan 1991'

FIGURE 6.14. Understanding and Managing Aging of BWR Recirculation Piping
Understanding and managing aging of LWR coolant pumps

Materials Casing * Staftlly cost SS Or. CFO CM5 , CFS5I or BA-S0S with ustenhte SB clad Closure Studs * SA193 Dr. 07 fr SA-40 Or. 823

Cover Gasket Type 304 SS flexttaffic (staInless steal. grsphite-asbeatos malettra)

Shaft .ype30 or 3l5 SS

Pump 1ps Westinghoure Type F

Bsbcock & Wilcox .TYPF.'pr E Combustion Type P* Type E Engineerng..

General Electric Type C

StBssors Temperatunr. operating transients, residual and stresse gasket leakage, alternating

Evirmnment bending stresses, turbulent mixIng of

hot end cael coolan. an svibrtion

- - Vettet cross sectin

PWR Type F cootant pump PWR Type E coolant pump (locations of maximum streta IntensIty are Circled)

CA)

Z.)

a

Understanding Aging

(Materials, StressoTs, and Managing Aging Environmental Interactions). - _..

. Sites Aging Concerns Inservice Inspection, Surveillance, and Monitoring MitMgation Cast SS base metal Thermal embittlement "RC qumnnenta Reommendatons

Casing * Coolant temperature

Ch Feritec ctn9 andexcnation shon Vohtmttrt of .n Chgetevze ferrite distribution Use Improved coe gqskets

- Fattgue at hIgh stress - elds and visusl examinatiOn and existing ftws In pump casing with beteh spring-back IntensIty octionsm of Internal surfaces. in at least and welds characterfsties, proper gasket one pump casing during each Installation. and cleanlIness L:'ow-alloy base Fatigu at high stress Inspection Interval Develop standards for allowable control to prnv nt borie cdd metal ntenty locatons5 (10 CFR 5B355s tWB-2501) flaw sIzes for pump casing base corrosion of closure studs

Welds Fatigue at welds havIng high Visual eamination of tntntrndl Lea leak-off fines between residual stresses or hIgh surfaces during ystem Develop techniques for monitoring Inner and outer gaskets siress IntensItIes leakage tests and hydrostatic actual degree of thermal unplugged - Thermal embrb~mint toeas; these tests aer enbriftiement In pump casing Install Instruments on leak. Thermal embrittlemint performed during each off Ones to detect leakage

Welds with low IGSCC f sensitIzed and lower refuenng outage and each Porm exminations of high reactor coolant

ferrite content atIgu strength If Inspection Interval, stress regions mlcroflssures ere present respectIvely Remove chrome plating on pump shafts

Include visual examlnatlons In Closure Studs ' Bode acid torrnsln caused volumetric examination of sit Insertce Inspdtion requirements

by gasket leakag closure studs (on pump being to detect corrosion waslage examined) during each

Inspection Interval Use cylIndrically guIded wvef (tO CFR 50 55J, IWB2500) technique to detect both corrosion

wastage and ereks

Shd't Sites wrlh high High-cyce mechanicel fatigue Evaluate use of modifIed

stressconContrston caused by satrneing bending cynindrically guided wvel or netdudl stress" stresses technique for shaft Inspection

Near thermal igh-cycle thermal fatigue barrier caused by turbulent mixing

Chrome-plated Crocks In chrome plating may

shafts propageae In shaft by high. cycde mechanical fatIgu

rev. 4

FIGURE 6.15. Understanding and Managing Aging of LWR Coolant Pumps
Cast Stainless Steel - An ANR investigation of the microstructural characteristics of cast stainless steel from

the Shippingport Station primary system check valves has

contributed.to clarification of the thermal embrittlement

processes that can occur at LWR operating temperatures.

The phase changes that had occurred in this naturallyaged material were similar to those observed in artificially-aged laboratory specimens providing validation of

aging projections based on the extrapolation of laboratory

data.

Neutron Shield TankSamples - Mechanical property measurements were conducted on samples of the inner wall of

the Shippingport Station neutron shield tank. Preliminary

results of an ANL evaluation of these samples (including

base metal and weld.materials exposed to different neutron

flux levels) suggest that the transition temperature

changes resulting from low-temperature low-flux irradiation are less-severe than a previous study had indicated.

Differences with the previous study appear to be caused by

neutron spectral differences and not to flux rate effects.

Invertor/Battery Charger - Naturally-aged inverters and

battery chargers from the Shippingport Station were tested

by BNL under the NPAR Program. Although some aginginduced changes were noted, it was concluded that aging

had not substantially affected equipment operation.

Nuclear Protection System Panel - An INEL evaluation of a

naturally-aged 1960s vintage nuclear protection system

panel and rack showed that the system was operating close

to the original response time of the equipment when it was

new.

Check Valves - An ORNL evaluation of a piston-lift check

valve from the Shippingport Station found significantly

more wear than would be expected based on the valve's

normal service environment.

Motor-Operated Valves - An 8-in. diameter gate valve and

operator from the Shippingport Station were refurbished

and requalified at-INEL. No operational problems had been

observed during periodic testing of the valve during

Shippingport Station operations. The refurbished valve

and operator were subsequently tested as part of an internationally sponsored seismic research program. The structural integrity of the valve and operator were not

affected by the seismic excitations. However, the studies

did reveal a previously unrecognized cable sizing problem

6.34
that resulted in the issuance of NRC Information Notice

No. 89-11: "Failure of DC Motor-Operated Valves to

Develop Rated Torque Because of Improper Cable Sizing."

Radiological Assessment - A PNL assessment of the corrosion film on primary coolant piping samples from the

Shippingport Station disclosed comparatively low concentrations of long-lived activation products and very low

concentrations of fission products and transuranic radionuclides, reflecting the high integrity of the fuel cladding during reactor operation.

A list of publications from the Shippingport studies is provided in

Appendix E.

6.5.2.4 Regulatorv Instrument Review

Under the NPAR Program, a single-source document has been generated

(Werry 1990) that delineates the regulatory documents that govern the inspection and monitoring of reactor pressure vessels, steam generators, pressurizers, and major reactor piping, for use in both aging studies and the LR

process.

6.6 NPAR SUPPORT TO LICENSE RENEWAL

The NPAR technical bases were developed principally to address aging

issues for plant operation up to 40 years. However, these technical bases

have been valuable to NRC staff in the analysis of issues related to LR.

Utilization of NPAR results in development of the basis for LR is addressed in

Section 5.0. Research that contributes to resolution of technical safety

issues could be accommodated in the NPAR Program if issues warrant further

investigation.

6.35
7.0 NPAR REVIEW AND INTEGRATION ACTIVITIES

The Nuclear Plant Aging'Research (NPAR) Program'has been the subject of a

series of integration and review activities; these activities are summarized

in this section.

7.1 TECHNICAL INTEGRATION REVIEW GROUP FOR AGING AND LIFE EXTENSION

Prompted by a November 1985 memorandum from the U.S. Nuclear Regulatory

Commission (NRC) 'Executive Director for Operations, the Technical Integration

Review Group for Aging and Life Extension-(TIRGALEX) panel was organized in

1986, with L. C. Shao as chairman, involving representatives from each NRC

office. An early activity involved"a survey of NRC programs with relevance to

nuclear power plant (NPP) aging.' The Review Group identified and documented

technical safety and regulatory policy issue's associated with NPP aging and

license renewal (LR), reviewed 'relevant NRC and external'programs, and provided recommendations for future NRC actions.' A report was published summarizing TIRGALEX work, Plan for Integration of Aging and Life-Extension

Activities (NRC 1987).

7.2 AGING AND LIFE EXTENSION COORDINATING COMMITTEE

Following completion of the TIRGALEX initiative, further coordination and

critique of aging-related studies in :NRC was carried out under the Aging and

Life Extension Coordinating'Committee:(ALEXCC) 'in 1987 and 1988. The general

consensus of these critiques was that the aging program was a vital element in

developing both the LR Rule and the Maintenance Rule.

7.3 NPAR RESEARCH REVIEW GROUP

Before the formal organization of the'review group, the NPAR Program

Managers' Technical Review Meeting was held February 24-26, 1987 at Oak Ridge

National Laboratory. :A'major'NRC-wide NPAR review session took place in 1988,

with organization of the Review'Group, drawing from'the NRC Offices and the

Regions. After-an extended 'effort; .the organization was completed in'August

1988. The first major activity was a'3-4day Review Group meeting in Rockville,

Maryland, March 21-23, 1989, where the ongoing NPAR research at the laboratories was reviewed. The 'Review Grbup_'fully endorsed the NPAR Program. 'The

second review by the 'Review'Group was held in Bethesda, Maryland, March 20,

21, 1990. The third review was held in Rockville, Maryland, March 26-28,

1991. .

7.1
7.4 OTHER NPAR REVIEW ACTIVITIES

In addition to these extensive review activities, the NPAR Program or

elements of the Program have been reviewed as follows:

* In October 1987, the NPAR Program was reviewed by the Senior

Research Program Steering Group on Component Integrity. The NPAR

Program was approved as presented.

* In November 1987, the NPAR Program was reviewed by the Nuclear

Utility Plant Life Extension (NUPLEX) Working Group of the Nuclear

Utility Management and Resources Council (NUMARC). The presentation

was received as being well conceived and productive.

* In December 1987, the NPAR Program-was presented to the Equipment

Quantification Advisory Group (EQAG) of the Electric Power Research

Institute (EPRI) meeting in Charleston, South Carolina. General

comments were made to the effect that the NPAR Program was generating data on aging that would be very useful to the industry.

* In January 1988, an NPAR in-depth briefing for NRC's Office of

Analysis and Evaluation of Operational Data personnel was held.

Overall evaluation was that the program was proceeding to develop

needed aging data.

* In February 1988, Electrical and Mechanical Engineering Branch staff

and contractors participated in the EPRI/Industry Workshop on Power

Plant Cable Condition Monitoring and discussed the NPAR cable

studies. Industry strongly supports and gave high priority for

these studies.

* In March 1988, the Advisory Committee onReactor Safeguards (ACRS)

Subcommittee on Reliability reviewed the NPAR work on motor operated

valves and check valves. The ACRS Subcommittee endorsed the need

for continuing this research effort.

* In March 1988, Commissioner Rogers, his staff, and the staff of the

other Commissioners, were given an in-depth briefing on the NPAR

Program. He and his staff voiced strong approval of the program and

the need to maintain it, and if possible, to accelerate it.

* In August 1988, the NRC conducted the highly successful International Nuclear Power Plant Aging Symposium. Many NPAR papers were

presented; also papers from the industry and from foreign speakers.

For the first time, NUMARC and EPRI attendees raised the question as

to the "correct" definition of aging.

* In September 1988, at a coordination meeting with the newly formed

NUPLEX group, the NPAR Program was again presented. Though there

7.2i
was a general consensus that work being done was very worthwhile,

the issue of the "correct" definition of aging was again raised. No

consensus was reached.

* In'November 1988, the NPAR Program was reviewed by the Nuclear

Safety Research Review (NSRR) Committee. The NSRR generally

endorsed the NPAR Program.

* In April 1989, a meeting was held with NUPLEX to exchange information on NRC's aging program and steps being taken in the development

of nuclear plant licence renewal criteria.

* In May 1990, the ACRS reviewed progress on the NPAR Program.

In addition to the review activities listed above, various elements of

the NPAR Program have been reviewed and critiqued on at least five occasions

by the members of ALEXCC before it was replaced by the NPAR Research Review

Group.

It should also be noted that as part of the cooperative effort developed

between NPAR and EQAG, NPAR draft reports are reviewed by members of EQAG, and

the industry insights thus gained, where compatible with the NRC's safety

mission, are incorporated into the final published report.

Another coordinated activity is organization of a committee to develop

common aging terminology. The committee members were selected from EPRI

NRC/NPAR, NPAR laboratory staff, NUMARC, and .tilities. The effort was funded

by EPRI and was coordinated by MPR Associates. A draft report containing the

recommended definitions has been circulated for committee comment; the comments are currently being resolved. The draft will be issued for wider

industry and regulatory review before it is published.

7.3
8.0 COORDINATION WITH OTHER PROGRAMS. INSTITUTIONS. AND ORGANIZATIONS

Various institutions and industry organizations have performed studies

and instituted programs related to aging research. .Several ongoing programs

*are also producing significant 'results that cannot and should not be

duplicated. .Therefore,,a major emphasis in-the Nuclear'Plant Aging Research

(NPAR) Program Plan is that proper coordination and integration'of research on

plant aging are achieved at various'levels. This approach will help achieve

overall program goals and objectives and ensure the efficient use of available

resources.

Ongoing'.Nucle'ar Regulatory Commission (NRC) programs related to aging and

license renewal (LR).are being.conducted by the Office'for Analysis and Evaluation of Operational Data (AEOD)i Office of-Nuclear Reactor-Regulation (NRR),

and Office of Nuclear Regulatory Research. '-The'NPAR'Program coordination and

technical integration with.other agency programslare in place or will be

implemented, facilitated by inputs.from the NpAR Review Group. In addition to

NRC-sponsored research, aging and LR programs are being sponsored.in the U.S.

by-nuclear industry groups including Electric Power Research Institute (EPRI),

nuclear steam supply system (NSSS) vendors,.'utilities,-architect engineers,

and the Department of Energy (DOE). Also nuclear-plant aging and extended

operation programs'are being conducted in a number of foreign countries.

Where the scopes of these programs'are available, they are summarized in

Appendix C.'

The International Atomic-Energy Agency (IAEA) in Vienna, Austria, has

organized a program on nuclear power plant (NPP) aging. A major symposium,

held in Vienna in June 1987,-'attracted approximately 140 representatives from

30 countries. Other IAEA activities include.an.advisory group to assist in

planning activities, a consultant'group on recordkeeping, a consultant meeting

on."Safety Aspects of Nuclear'Power Plant Aging (November-1989), and a technical committee meeting on "The Evaluation and Management of the Safety Impact

on Nuclear Power Plant Aging" (November 1989). Additional meetings were held

in June and November of 1990. An *IAEA coordinated research program on NPP

aging and life extension has been organized and is conducting pilot studies on

four key safety-related components (see Appendix C).

Figure 1.1 illustrates the coordination 'with ongoing NRC programs, and

with external programs, involving both domestic and.foreign organizations. It

is very likely that new aging-related prdgrams will be sponsored in the-future

by outside organizations. An important and continuing activity in the NPAR

Program is identifying new projects and establishing appropriate interfaces.

A listing of the more relevant programs, external to the NPAR Program, is

given in Table 8.1. Appendix C contains descriptions of ongoing aging-related

programs.

8.1
TABLE 8.1. Selected Programs Relevant to NPAR Aging.Studies and Programs

Office of Nuclear Reaulatorv Research

* 10 CFR Part 54-LRR - To develop and publish a

License Renewal Rule (LRR). 10 CFR Part 51 - To

develop and publish an environmental rule for

License Renewal (LR).

* Numerous projects addressing generic safety

issues and specific SSCs have aging and LR 1mplications. In most cases. however, aging and

extended operation are not the dominant considerations in the work.

Office of Nuclear Reactor Requlation

* Maintenance and Surveillance Program--Maintenance

and surveillance are key mechanisms for control-

'ling aging. Phase I of the program his been completed, and Phase II is under way. The work

addresses many other safety and regulatory issues

in addition to aging and LR.

* Safety System Functional Inspection Program--The

objective of this program is to ensure that plant

modifications have not impaired the effectiveness of selected safety systems. This includes

the detection, control. and prevention of

selected safety system degradation, whatever the

cause; age-related degradation is covered along

with other casual factors not related to aging.

* Safety System Outage Modifications Inspection

Program--The objective is to verify that modifications and repairs are conducted properly so

that safety is not compromised.' This covers

modifications and repairs that may be aimed at

managing aging or permitting LR.

* Electrical Distribution Inspection.Program--the

objective is to verify that the electrical distribution system still meets the original or

modified design and to ensure component performance meets requirements.

Office of Analysis and Evaluation of Operational

Data-

* NPROS Analyses--AEOD analyzes data for the

Nuclear Plant Reliability Data System (NPRDS),

producing statistical, and engineering evaluations of component failure modes, time to fail-'

ure, operating conditions that affect failure,

and chemical and physical conditions affecting

component-wearout rates.

* Licensee Event Report Database and Analyses--

AEOD has developed the Sequence Coding and

Search System (SCSS). which can be used in

studies of component life and component aging.

SCSS is a comprehensive, computerized LER-based

system with features that are highly useful in

aging studies, e.g., cause codes specifically

identifying 'age-related degradation.

NRC REGIONS

* Field Inspection andNMaintenance--Dissemination

of NPAR report summaries to NRC regional. inspection personnel is currently being explored. The

format of these report summaries will be tailored

to 4rovide concise information of direct value

to the inspectors in the conduct of their

activities.

EPRI/DOE/INDUSTRY

* EPRI/OOE/Industry Cooperative Program--This

cooperative program is addressing the full spectrum of policy; safety,'regulatory, economic, and

technical' issues associated with extended operation. The BWR and PWR pilot LR studies are com-*

plete., Development of lead plant LR requests are

underway, targeted for 1991. Specific LRrelated programs sponsored separately by DOE/SNL

and EPRI are summarized in Appendix C.

CODES AND STANDARDS COMMITTEES

* ASME committees for operation and maintenance(see

Table 6.2) are developing standards that address

aging management of key safety-related equipment,

with input from NPAR in selected topics.

*ASME Section XI--A Special Working Group on, Plant

Life Extension meets quarterly. They have active

participation from EPRI, NRC. DOE, NSSS suppliers, and utilities. ASME XI is just beginning

to develop code changes based on LR

considerations. I

IEEE Working Group 3.4. under the IEEE Nuclear

Power Engineering Committee, is developing recommendations for changes to IEEE standards based

on aging and extended operation considerations.

*IEEE Working Group 3.3, under the IEEE Nuclear

Power Engineering Committee, is developing recommended methods for the' Industry to mitigate

equipment aging and to preserve equipment

operability.

FOREIGN COUNTRIES

* Several countries, including Canada, France

Germany, Japan, Spain and Taiwan have programs

that address aging and extended operation.

8.2
9.0 SCHEDULES AND RESOURCE REQUIREMENTS

The current schedule for the Nuclear Plant Aging research (NPAR) program

has been adjusted to support the regulatory activities anticipated for nuclear

plant license renewal'(LR). The estimated schedules and milestones for completing specific'research activities for aging assessments and LR'are provided in this section. These general schedules, and particularly. schedules

for evaluating specific systems, structures, 'and components (SSCs), depend on

funding, assignment of priority, and degree of coordination and participation

by other institutions and organizations. The NRC/RES staff and its contractors are actively pursuing participation' from domestic-and foreign institutions and organizations. Requirements for LR will have increased'priority

in NPAR resource allocations.

The schedule and resource requirements reflect the need for early identification and prioritization o'f the technical safety issues, based on'their

risk significance.' Other major elements affecting these requirements are the

availability of topical reports from industry programs, the work scope

required to resolve essential safety issues, and the development of proposed

regulatory review procedures.

The schedules for major elements of the'NPAR Program are given in Figures 9.1 to 9.4. The scheduling of the Phase'-I and Phase-Il'assessments for

components and systems is shown for each major activity. Also shown are the

schedules for utilizing the research results, completion periods, and delays

caused by budget cuts or reprioritization.

Table 9.1 summarizes the planned NPAR' accomplishments for fiscal years'

1991 and 1992. The schedules and major Activities are based on the current

NPAR research priorities. Note that the activities'and schedules can change

as information is developed in the program and as additional inputs are

provided and program needs'are identified.

Note also that the number of systems and components and the degree and

depth of assessments and analyses that can be conducted effectively will

depend on the availability of funds and the period of time over which the

results are required.

Participation will be needed from NRC staff at an average level of five

full-time professionals per year for the duration of the program. When participation by outside organizations is' achieved, the resources provided from

outside NRC will be identified in future revisions of this plan.

9.1
NPAR Milestones and Schedules--Systems

|FY-89 |FY-90 |FY-91 1FY-92

High Pressure - ECCS_

(INEL)

RHR/Low Pressure - ECCS

(BNL) l ______8

Service Water System

(PNL) mu

Component Cooling Water REMO

(BNL) 1

to

r\)

* Complete M Phase 1 3 Phase 2 011 Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization 39103033.1

FIGURE 9.1. NPAR Milestones and Schedules--Systems
NPAR Milestones and Schedules--Systems

I 'FY-89' |I FY-90 ' I FY-91 I FY-92 I FY-93

, ,, Reactor Protection System I

(INEL) . E-itlhJl I -~~~~

. r _ A

1111111111111= t '* _

I, I - 4.

6

Class 1E Distribution System

(INEL)

Auxiliary Feedwater System

(ORNL)'

Contrbl Rod Drive Systems

(BNL:-'PWR -'W)

(BNL- PWR - B&W, CE)

(ORNL- BWR)

. .......... ......... 3 ................ .............................. .. . . .. ... . .

5 4 .-------- -

ZZZ ,-.I - - - I

, YY xx YY YY YY YY - Y Y Y x

-------------------- ~~ 0-0 - - Q3OX0QO0X - l -- - - - = t. K=1Q -. .... ....I 

11111,1111111.11p I .........

I.-

a I

U Complete . 9Phase 1 EaPhase 2 111 Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization - - - ~~~~~~~~~~~~~~~~~~~~~~391 03033.2

FIGURE 9.1. (contd)
NPAR Milestones and Schedules--Systems

Motor Control Centers

(BNL)

Instrument Air System

(BNL)

Containment Cooling System

(BNL)

89 FY-90 FY-91 FY-92 FY-93

i

W99990

H IM

ON

MUH

to

Engineered Safety Features

(PNL)

* Complete a Phase 1 VIPhase 2 [11 Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization 39103033.3

FIGURE 9.1. (contd)
NPAR Milestones and Schedules--Systems

FY-89 FY-90 FY-91 FY-92 FY-93

Instrument & Control

(ORNL)

Automatic Depressurization System (PrNL),- , --- |

.Standby Liquid Control System , , (PNL)'''' ' '' '''''' ''' ,. , , t %

to~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o

Reactor'Co're Internals

(ORNL)

Turbine, Main Generator, &

'Controls (ORNL)

* Complete 3 Phase 1 3 Phase 2 .- Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization -te. .- - -- . . - - ~~~~~~~~~~~~~~~39103033.4

FIGURE 9.1. (contd)
NPAR Milestones and Schedules--Components

FY-89 FY-90 FY-91 FY-92_I . FY-93

Motor Operated Valves

(ORNL)

Check Valves

(ORNL)

Solenoid Operated Valves

(ORNL)

to Air Operated Valves

61 (ORNL)

Auxiliary Feedwater Pumps

(ORNL)

Large Electric Motors

(BNL)

Chargers/Inverters

(BNL)

I zz -, *IWXTHTIY I _ _.

I_ ............... ULUA1

1s1ls . ~~l

I

I 0 -

I - - l- - -I - - 001 __1 _ uuLuLwniul-

-Y...- .-...

V-zzzzzzAIIIIIIIIIIIII III

r...c.t.~.c.c.c ( . . .. I * * I * I I. _

I

r/7277z/--- i . *"I*~

I 11111111u11111111 111- _ .I ......

- . a a.

- . _______________ - -4 JVS)S.)V(.XX.X

.~~ , ~ ~~~~~~~~~~~~~~ ,9 A A A A A, i^A 

4llllI11111M

III1111iII IIjU

I I h i 1.

* Complete 3 Phase 1 E Phase 2 1M Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization 39103033.5

FIGURE 9.2. NPAR Milestones and Schedules--Components
NPAR Milestones and Schedules--Components

FY-89 _ FY-90 FY-91 I FY-92 I FY-93

Batteries

(INEL)

FFW -F1-T

. .... ..

Power Operated Relief Valves K

(Operating Experience Review & Analysis

(ORNL)

Snubbers

(PNL)

Circuit Breakers/Relays

(BNL, Wyle)

Penetrations -

(SNL)

Connectors

(SNL)

Chillers

(PNL)

I

4 -

- - - - f fifififififififififiF i

III'1111 111111111iii 11111111111iii IL.LLE. -

IIIIIIIIIIIIHIIIIIIIIH 1111M

I

tiger.

89e lHIIIIIIIIIIIIIIIIIIIIII H

DoI I I.

* Complete P Phase 1 v Phase 2 El Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization 39103033.6

FIGURE 9.2. (contd)
NPAR Milestones and Schedules--Components

I FY-89 I FY-90 |I FY-91 | FY-92 - FY-93

Cables

(SNL)

Diesel Generators

(PNL)

Transformers

(INEL)

Heat Exchangers

(ORNL)

Compressors

(ORNL)

Bistables/Switches

(ORNL)

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

7Z/r

______________...I ~~~....................u ..u.. Ii1

Ii

.1.1.. .4. 0. ......... ...... , # ------

1' ~ ~ ~ I- IN

I v .4

A .Ist~l~nfql 

II 111111111IIIII11111111 111111111111_ FL0

00

...................

FAA AAAQQQSA ~. f ^t .1mtit) rn

I

-~ ~ -Y yyy-y

i ~ ~ ~ ~ I~, I----------- ~ m m m

l l 11111 11111

A Complete M Phase 1 rVAPhase 2 1m Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization

(*)Carried on from EQ Program 39103033.7

FIGURE 9.2. (contd)
NPAR Milestones and Schedules--Components

Main Steam Isolation Valves

(ORNL)

FY-89 FY-90 FY-91 FY-92 .. FY-93

E Complete ~~lPhase 1 0 Phase 2 mI Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization

FIGURE 9.2. (contd)

39103033.8
NPAR Milestones and Schedules--Special Topics

I FY-89 I FY-90 I FY-91 I FY-92 I FY-93

Risk Evaluation of

Significant Aging Effects

(INEL.PNL-SAIC-BNL)

-Evaluation of Systems Level Aging

Effects from Operating Experience

-Risk Evaluation of Aging

Methodology

Application

Guidelines

sx^M^x^^%bMi__J_________,,^, I~~~~~e m//////// 1--------- ........ t 

1111111 11111111111111111111111111 . .......v ............. IIIIIIIIHII

-. 3 - ---

Wi..li||XE1||§§§§1§ll~ll|~ .r1isll~ 

111111111111111111 111111111iii1111111111111111111111111111111111 UALLLLLLLLLLLLLA ...........

r ,,,, ,,,,,. -...,,,,I ...,,,,,,,

-Aging Risk Based Prioritization } oi.

l toCD0

I.--nnnnn.nnnnnnnnnnnnn. 31nnn......

I I~~~.... ia . I I'lI IL a 

Shippingport Aging

Evaluation

(PNL)

Passive Component Aging

Risk Evaluation

(INEL)

Residual Life Assessment

(INEL)

I,, ,, ; . . . . .

__-f ----- - .- 111111111111111I 1111111111111111111111111I I11-

ILLLILW..j a_

---------- --- I

| ,- fiflffifififikifimnfniiiinifii

L1i.LLI.LLLLIJ.L'''''''''''''''''''' IIIIIIIIIIIIIIIII IIIIIIIIIIIIIIIIIIIIIIIIII

- - - -I - -

I ---------- l I .... ...............

i Complete M Phase 1 2 Phase 2 im Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritization 39103033.9

FIGURE 9.3. NPAR Milestones and Schedules--Special Topics
NPAR Milestones and Schedules--

Aging Program and Issues Coordination

Technical Bases for

Developing Regulatory

Guides and

Review Procedures

Recommendations for

Technical Specifications

Modifications from Aging

Viewpoint

Recommendations for

Data and Recordkeeping

Modifications from Aging

Viewpoint

Aging of Mothballed NPPs

FY-89 FY-90 FY-91 FY-92 FY-93

* Complete 3 Phase 1 '77Phase 2 1E Utilization of Research Results

Signifies Delays Due to Budget Cuts or Reprioritizatlon 39103033.10

FIGURE 9.4. NPAR Milestones and Schedules--Aging Program and Issues Coordination
TABLE 9.1. Planned NPAR Accomplishments for FY 1991 to 93

* Issue a draft Master Regulatory Guide (Format and Content) to address technical issues relating to aging degradation for license

renewal.

* Issue a manual on management of aging.

* Develop guidelines for recordkeeping and good maintenance practices for license renewal, if required.

* Develop a step-by-step procedure for treatment of aging in probabilistic risk assessment.

* Develop a procedure to account for testing and maintenance, influencing aging degradation, in PRA.

* Complete the regulatory instrument review of major LWR components from aging perspective and for license renewal

consideration.

- Containment Basement

- Cables

- Selected Pumps and Valves

* Complete the review of the technical specifications from aging perspective and generate recommendations to Office of Nuclcar

Reactor Regulation (NRR).

* Complete the reviews of industry (NUMARC) technical reports; assist NRR in developing staff technical positions; assist NRR in

developing/modifying Standard Review Plans for license renewal considerations.

* Complete the development of model(s) for residual-life assessment of selected major LWR components and structures.

* Complete assessment of selected emerging inspection, monitoring, and material testing techniques.

* Complete Phase-I1 NPAR studies on the following components and systems:;

- Residual Heat Removal/Low Pressure System

- Reactor Protection System

- Class 1 Distribution System.

- Instrument Air System

- Snubbers

- Cables

- Fire Susceptibility/Electrical Components (complete FY 1993)

• Complete Phase-I NPAR studies on the following components and systems:

- Control Rod Drive Systems. m

- BOP System Aging Effects

- Automatic Depressurization System

- Reactor Core Internals

- Air Operated Valves

- Large Electric Motors

- Penetrations

- Connectors

- Chillers

- Transformers

- Heat Exchangers

- Bistables/Switches

- Main Steam Isolation Valves

- Engineered Safety Features

* Complete Phase-I and initial Phase-II NPAR studies on the following systems and components:

- High Pressure-ECCS

- Auxiliary Feedwater System

- Containment Cooling System

- Instrument and Control

- Standby Liquid Control System

- Turbine, Main Generator and Controls

* Apply NPAR lessons learned to advanced LWRs.

. , b~~~~~~~~~~~~~~~~~~~1

9.12
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Identify Aging'Trends. NUREG/CR-3543, ORNL-NSIC-216, prepared by Oak Ridge

National LABoratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Murphy-, G. A. (ORNL) to J. Vora (NRC). -1985. Accident Precursor Events.

Involving Age-Related Component Degradation. Letter Report: available in the

NRC Public Document Room, 1717 HG Street N.W., Washington, D.C., prepared by'

Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Muscara, J.' 1985.. Research Program Plan--Non-Destructive Examination. '

NUREG-1155 Vol 4, U.S. Nuclear Regulatory Commission, Washington, D.C.

Muscara, J., and C.-Z.'Serpan."

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Washington, D.C.

1985. Research Program Plan--Steam

U.S. Nuclear Regulatory Commission,

10.4
Naus, D. J., M. F.-Marchbanks, C. B. Oland, and E. G.KArndt. 1989.

Structural Agina '(SAG) Proqram Five-Year Plant- FY 1988-1992.'-. ORNL/NRC/

LTR!89/1, prepared by Oak Ridge National Laboratory for the U.S. Nuclear

Regulatory Commission, Washington,' D.C.''

NUREG-0200. Improving Regulatory Effectiveness in Federal/State Siting

Actions: Federal/State Regulatorv Permitting Actions in Selected Nuclear

Power Station License Cases. U.S. Nuclear Regulatory Commission, Washington,

D.C.

NUREG-0800. 1987. 'Standard Review Pla'n for the Review of'Safety Analysis

Reports for Nuclear Power Plants. U.S.-Nuclear Regulatory Commission,

Washington, D.C. i

NUREG-0885. 1986. U.S. Nuclear Regulatorv Commission 1986 Policy and,

Planning Guidance. U.S. Nuclear Regulatory Commission, Washington, D.C.

NUREG-1317. 1988. ,Regulatorj ODtions for Nuclear Plant-License Renewal.

U'.S.'Nuclkar Regulatory-Commission; Washington,'D.C.;'

NUREG-1212, Vol 1. 1986. Status of Maintenance in the U.S. Nuclear Power

Industrv-1985. -Volume 1:-Findings and Conclusions.. U.S. Nuclear Regulatory,

Commission,.Washington; D.C. :'

NUREG-1212, Vol 2. 1986. Status of Maintenance in the U.S. Nuclear Power

Industrv 1985. Volume 2: Description of Programs and Practices. U.S.

Nuclear Regulatory Commission, Washington, D.C..

NUREG-1299. Draft for Comment: December 1990. Standard Review Plan for

License Renewal. U.S. Nuclear Regulatory Commission, Washington, D.C.

NUREG-1362. Draaft.for, Comment: July 1990., Regulatorv Analysis'for'Proposed

Rule on Nuclear Power PlantLicensetRenewal.' U.S. Nuclear Regulatory Commission, Washington, D.C.

NUREG-1411. 1990. Response to Public Cohments Resulting from the Public

Wov'kshbp on Nuclear Power Plant License Renewal. U.S. Nuclear Regulatory

Commission; Washington, D.C.''

NUREG/CP-0099. 1988. -Proceedings of the Public Workshopnfor NRC Rulemaking

on Maintenance of Nuclear Power Plants. U.S. Nuclear Regulatory Commission,

Washington, D.C.! -

Regulatory Guide 1.9, Rev. 2. December 1979. Selection. Design, and Qualification of Diesel-Generator Units Used-as Standby (Onsite) Electric Power.

Systems at Nuclear Power Plants. 'U.S.' Nuclear Regulatory Commission,-

Washington, D.C.

Regulatory Guide 1.29, Rev. 3. August 1978. Seismic Design Qualification.

U.S. Nuclear Regulatory Commission, Washington, D.C.

10.5
Regulatory Guide 4.2, Rev. 2. July 1976. Preparation of Environmental

Reports for Nuclear Power Stations. U.S. Nuclear Regulatory Commission,

Washington, D.C.

Regulatory Guide 4.7, Rev. 1. November 1976. General Site Suitability

Criteria for Nuclear Power Stations. U.S. Nuclear Regulatory Commission,

Washington, D.C.

SECY-85-129. April 12, 1985. Maintenance and Surveillance Program Plan.

U.S. Nuclear Regulatory Commission, Washington, D.C.

Shah, V. N., and P. E. MacDonald. 1987. Residual Life Assessment of Malor

Light Water Reactor Components-Overview. NUREG/CR-4731 Vol 1, EGG-2469,

prepared by Idaho National Engineering Laboratory for the U.S. Nuclear

Regulatory Commission, Washington, D.C.

Shah, V. N., and P. E. MacDonald. 1989. Residual Life Assessment of Malor

Light Water Reactor Components-Overview. NUREG/CR-4731 Vol 2, EGG-2469,

prepared by Idaho National Engineering Laboratory for the U.S. Nuclear

Regulatory Commission, Washington, D.C.

Shier, W., and M. Subudhi. 1988. Operating Experience and Aging Assessment

of Motor Control Centers. NUREG/CR-5053, BNL-NUREG-52118, prepared by

Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Subudhi, M., E. L. Burns, and J. H. Taylor. 1985. Operating Experience and

Aging-Seismic Assessment of Electric Motors. NUREG/CR-4156, BNL-NUREG-51861,

prepared by Brookhaven National Laboratory for the U.S. Nuclear Regulatory

Commission, Washington, D.C.

Subudhi, M., W. E. Gunther, J. H. Taylor, A. C. Sugarman, and M. W. Sheets.

1987a. Improving Motor Reliability in Nuclear Power Plants. Volume 1.

Performance Evaluation and Maintenance Practices. NUREG/CR-4939 Vol 1,

BNL-NUREG-52031, prepared by Brookhaven National Laboratory for the U.S.

Nuclear Regulatory Commission, Washington, D.C.

Subudhi, M., J. H. Taylor, R. Loffaro, A. C. Sugarman, M. W. Sheets, and

K. M. Skreiner. 1987b. Improving Motor Reliability in Nuclear Power Plants;

Volume 2: Functional Indicator Tests on a Small Electric Motor Subjected to

Accelerated Aging. NUREG/CR-4939, Vol. 2, BNL-NUREG-52031, prepared by

Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Subudhi, M., J. H. Taylor, J. H. Clinton, C. J. Czajkowski, and J. R. Weeks.

1987c. Indian Point 2 Reactor Coolant Pump Seal Evaluations. NUREG/CR-4985,

BNL-NUREG-52095, prepared by Brookhaven National Laboratory for the U.S.

Nuclear Regulatory Commission, Washington, D.C.

10.6
Subudhi, M., J. H. Taylor, and M. W. Sheets. 1987. Improving Motor Relia-.

bility in Nuclear Power Plants. Volume 3. Failure Analysis and Diagnostic

Tests on a Naturally Aged Large Electric Motor. NUREG/CR-4939 Vol 3, BNLNUREG-52031, prepared by Brookhaven National Laboratory for the U.S. Nuclear

Regulatory Commission, Washington, D.C.

Toman, G. J. 1986. Inspection. Surveillance, and Monitoring of Electrical

Equipment in Nuclear Power Plants. Vol. 2: Pressure Transmitters. NUREG/

CR-4257, Vol 2, ORNL/SUB/83-28915/3/V2, prepared by Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, D.C.

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Aging and Life-Extension Activities. Prepared by the Technical Integration

Review Group for Aging and Life Extension (TIRGALEX) for the U.S. Nuclear

Regulatory Commission, Washington, D.C.

Vagins, M., and J. Strosnider. 1985. Research Program Plan--Piping.

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Vagins, M., and A. Taboada. 1985. Research Program Plan--Reactor Vessels.

NUREG-1155, Vol 1, U.S. Nuclear Regulatory Commission, Washington, D.C.

Vesely, W. E. 1987. Risk Evaluations of Aging Phenomena: The Linear Aging

Reliability Model and Its Extensions. NUREG/CR-4769, EGG-2476, prepared by

Idaho National Engineering Laboratory for, the U.S. Nuclear Regulatory Commission, Washington, D.C.

Vesely, W. E., R. E. Kurth, and S. .M. Scalzo. 1990. Evaluations of Core Melt

Freauencv Effects Due to Component Aging and Maintenance. NUREG/CR-5510,

SAIC-89/1744, prepared by Science Applications International Corporation for

the U.S. Nuclear Regulatory Commission, Washington, D.C.

Vesely, W. E. et al. 1983. Measures of Risk Importance and Their Applications. NUREG/CR-3385, BMO-2103, prepared by Battelle Columbus Laboratories

for the U.S. Nuclear Regulatory Commission, Washington, D.C.

Vora, J. P. 1987. Nuclear Plant Aging Research (NPAR) Program Plan.

NUREG-1144, Rev. 1. U.S. Nuclear Regulatory Commission, Washington, D.C.

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Vessels: Aging and License Renewal Considerations.- PNL-6848, Pacific Northwest Laboratory, Richland, Washington.

Werry, E. V. 1990. Regulatorv Instrument Review: Management of Aging of LWR

Major Safety-Related Components Vol 1. NUREG/CR-5490, PNL-7190, prepared by

Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Wu, P. C. 1989. Erosion/Corrosion-Induced Pipe Wall Thinning in U.S. Nuclear

Power Plants. NUREG-1344, U.S. Nuclear Regulatory Commission, Washington,

D.C.

10.7
APPENDIX A

NPAR PROGRAM PARTICIPANTS
APPENDIX A

NPAR PROGRAM PARTICIPANTS

This appendix contains a brief summary of the NRC contractors functioning

under the Nuclear Plant Aging Research (NPAR) Program. The research is currently being conducted by the following U.S. Department of Energy (DOE)

national laboratories:

* Brookhaven National Laboratory (BNL), Long Island, New York

* Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho

* Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee

* Pacific Northwest Laboratory (PNL), Richland, Washington

* Sandia National Laboratory (SNL), Albuquerque, New Mexico.

Also supporting and participating in this research are the National Institute

of Science and Technology (NIST), Scientific Applications International Corporation (SAIC), and Wyle Laboratories. In addition, work is subcontracted to

other private engineering firms and private and academic consultants when specialized expertise is required. The laboratories frequently interact with

utilities operating nuclear power plants in conducting and validating the

research on aging of components and systems. Tables 3.1, 3.2 and 4.1 summarize the specific roles of the laboratories and the schedules for completion

of the individual work elements.

The major format for documentation of the results of the NPAR aging

assessments is NUREG/CR reports. Some studies are documented in technical

reports from the individual laboratories. Other results are reported in proceedings of meetings and symposia, in proceedings of the annual NRC Water

Reactor Safety Information meeting, and in peer-reviewed journal articles.

Kondic and Hill (1990) summarized the 72 NPAR reports published through

May 1990. Since then, approximately 20 more NPAR reports have been issued in

final or draft form.

A.1
APPENDIX B

UNDERSTANDING AND MANAGING AGING
APPENDIX B

UNDERSTANDING AND MANAGING AGING

In this appendix, the nature and processes of aging in nuclear power

plants (NPPs) are reviewed. Aging processes'are described, including degradation mechanisms. Also identified are 'methods to manage aging'in NPPs.

Finally, the potential impact of aging'on .NPP safety is explored.

B.1 NATURE OF AGING PROCESSES

From the regulatory perspective, the term "aging" can denote either

favorable(a or unfavorable changes in properties. As an NPAR working definition, age-related degradation is defined as the cumulative degradation

occurring within a reactor system, structure, or component, which, if unmitigated, may result in loss of function and impairment of safety.' Aging is a

complex process that begins as'soon as'a'component or structure is produced

and continues throughout its service'life.' Aging'plays'a significant role in

the operation of a nuclear plant and must be factored into the determination''

of safe operating lifetime limits. It also is important in evaluations for

license renewal. Each nuclear plant, including those still under construction

or being mothballed, should be systematically evaluated for the effects of

aging. Factors that can cause-aging include the following:

* Natural internal chemical or'physical processes - Examples include

the embrittlement of metals and plastics because of structural or

chemical changes, high-temperature sensitization causing susceptibility to intergranular stress corrosion cracking in austenitic

stainless steels, and oxidation or cross linking of polymers with a

resultant loss in toughness and dielectric strength.

* External stressors and environments '- -Irradiation, primary and

secondary coolant chemistry,'a'nd vibratory loads'are the typical

examples of stressors introduced by the operating environment.

Freezing and thawing, brackish water, and humidity are typical

examples of stressors introduced by external environment. Synergistic influences of'electrical'and mechanical stressors in c6mbination with other internal and external environments also contribute'

to degradation processes. ' ' -

(a) Certain aging processes, such as curing of concrete, can lead to improved

properties.

B.1
* Service wear - Service wear of rotating equipment and wear of the

drive assembly in a control rod drive mechanism are typical

examples.

* Excessive testing - Frequent fast-start testing of emergency diesel

generators is a prominent example.

* Improper installation, application, or maintenance - During Phase-I

of the NRC Maintenance and Surveillance Program, (NUREG-1212) the

investigators concluded that 48% of the licensee event-reports

(LERs) were maintenance related, and that over 60% of the forced

outages were due to component failures.

These factors and others can act either singly or synergistically to

degrade a system, structure, or component.

B.2 UNDERSTANDING AGING

The effective management of aging begins with an understanding of the

aging processes. The requisite understanding may be either empirical or mechanistic, depending on the nature and potential consequences of a particular

aging process.

B.2.1 Degradation Mechanisms

Stressors and environments act in concert on materials to cause agerelated degradation. Many mechanisms can potentially contribute to deterioration processes. However, frequently the deterioration occurs slowly such

that it is not expected to interfere with plant operations 'o safety over

plant life. Extensive analytical and experimental efforts by both government

and industry have characterized effects of numerousfdegradation mechanisms

that operate in NPPs. These mechanisms vary widely in terms of their potential effects. Some mechanisms are operative in numerous systems, structures,

and components (SSCs) over wide ranges of environment and stressor levels;

effects of other mechanisms are limited to specific components or materials

over narrow ranges of conditions. Degradation mecha isms'of'concern in NPPs

include the following:

Corrosion is a normal phenomenon in NPPs'; bulkVcorrosio'n in air, water',

or steam causes buildup of corrosion products, losso of 'materials, and activity

transport in'irradiated systems; however, it seldom results in major thinning

or wall penetration; several localized corrosion processes occur in NPPs, such

as crevice corrosion, pitting corrosion, galvanic corrosion, various types of

stress- or irradiation-enhanced corrosion, etc. These processes can lead to

significant local wall penetration and to various, high-consequence failures.

Erosion caused by high'velocity steam, water, or two-phase mixtures has

contributed to major failures'in NPP SSCs. Special processes of importance

include erosion/corrosion (local erosion causing a wearing away of a protective oxide film, allowing corrosion to proceed) and cavitation.

B.2
-Embrittlement of metals and-plastics because-of structural or chemical

changes.induced by radiation', high temperature,'or atmospheric-contaminants -'

can-lead to fragility and failure-under dynamic'loading. -Radiation embrittlement-and loss of toughness in metal.s.is'a consequence of intense neutron radi:

ation fields; thus, only-componentsliniclose proximity to the'reactor cor . such-as~pressure vesselsiand'vessel'supports are -affected.- Organic-and elec<;

tronic materials 'are particularly susceptible to ionizing radiation'damage' and

to degradation- causedby attack of aggressive' chemical'speciesc:produced in-''

ionizing radiation-fields.' For example,'--ionizing radiation reduces the-tough:

ness and dielectric-strength of plastic insulators used in cables' motor windings, and transformers. Thermal embrittlement4is'associated with-chemical'or;

metallurgical changes and results from processes such as high temperature

thermal aging leading to.reduced toughness of ferrous alloys', high temperature

sensitization.to intergranular stress corrosion cracking in austenitic stain-:

less steels, and oxidation or cross-linking of. polymers with a resultant loss

in toughness and.dielect'lic strength.:.:Hydrogen-absorption by ferrous alloys

can also lead to loss of toughness and.-brittle fracture.' ' :;; -

.Wear is a general concern for-rotating:or other sliding surfaces'where'.-'

tolerances can affect-performance. -Lubricant'loss'or degradation,'.e.g., ' because of contaminants'or chemical breikdown-,:can greatly accelerate wear.:E' ;,~~ ~ ~~~~~ ;*- ---a t t, ' -

Fat'igue is a common degradation process that-occurs in rotating or:'

vibrating equipment orounder other service conditions-that place periodic-or *

cyclic loads on'a system or.component.- Associated failures may occur at

either high or low cycles in response to various kinds of loads, e.g., vibrational loads, thermal cycles, or pressure transients.;'

Chemical/biological degradation-involves oxygen-induced -degradation of

organicmaterials, -micro-biologically influenced corrosion (MIC), and'plugging and corrosi6n'effects from macro-biological species.-; - .*;:

Shrinkage or creep canoccur'in.most materials-and are'common phenomena ''

in metals at high temperatures.--Polymers-;and:composites used-asielectrical i!

insulators, supports, and protective coatings'may. exhibit dimensional- insta-- -

bility as a result of exposure to-high temperatures',;moisture, mechanical .v<

stresses, or radiation.'-These~effects:can deg'ade-insulating and structural-t

properties. Shrinkage of 'co'ncrete in NPPs is primarily-caused by-long-term' u

dehydration. Dimensional changes in concrete as it ages do not degrade the

properties of concrete;:-however,-whenfthese dimensional-changes-cause interference, e.g., with other components in prestressed reinforced concrete structures, degradation-can occur. Shrinkage is' the main contributor to the loss ' of prestressing forces in prestressed concrete-containment.-. -

B.2.2 Degradation Sites' . - - ; - - - -

Certain sites on or within a given component will normally exhibit more

degradation than other sites; and, for many components, degradation is limited

to specific locations. 'Factors-that affect vulnerability to degradation -

include localized chemical or metallurgical variations, geometry with respect

to fluid flow or chemical potential gradients, proximity to mechanically or,

B.3
chemically incompatible materials, and high local stresses. Examples of sitespecific degradation include 1) localized erosion/corrosion in ferritic steel

piping because of local high fluid velocities, 2) intergranular stress corrosion cracking in heat-affected zones near welds in austenitic stainless steel,

3) excessive hinge pin wear in check valves subject to flutter, 4) rapid degradation of pump impeller blades when cavitation occurs, 5) wear. or galling of

sliding contacts, 6) crevice corrosion, and 7) fatigue cracking in regions

experiencing tensile stresses. Knowing which sites degrade, by what mechanisms, and at what rates is necessary for selecting monitoring methods and

for determining where, how, and with what frequency monitoring must be applied

to reliably trend and mitigate aging.

B.3 MANAGING AGING

When the mechanisms that cause aging are understood, they can be managed

to ensure that aged SSCs will adequately perform their designed safety functions. Inspection, surveillance, condition monitoring, recordkeeping, trending, maintenance, and refurbishment/replacement programs are effective tools

for managing aging in operating NPPs. Depending on their intended function,

these programs take various forms, such as corrective maintenance, preventive

maintenance, predictive maintenance, reliability centered maintenance, condition monitoring, inspections, tests, and surveillance programs. Together,

these.programs comprise a comprehensive maintenance program that can effectively mitigate potential consequences of age-related degradation.

B.3.1 Inspection. Surveillance. and Monitoring

Inspection, surveillance, and monitoring methods (ISMM) should reflect

mechanistic-and empirical assessments performed by qualified staff in their

efforts to understand and mitigate aging. These.'methods which also include

testing and condition monitoring should employ state-of-the-art nondestructive examination,,;such as ultrasonic testing, signature analysis,

vibration analysis,:dielectric performance measurements, and other measuring

techniques performed by qualified or. certified staff. Measurement results

should be documented, trended, and-analyzed with respect to frequency and

nature of preventive-and corrective maintenance, and for indications of their

effect on residual SSCs lifetimes.

The methods of inspection and surveillance, non-destructive examination

and condition monitoring are described in more detail as follows:

Inspection and Surveillance

Detailed and comprehensive requirements for monitoring degradation

in SSCs are conveyed by the inservice inspection requirements in

B.4
Section XI of the.ASME.Boiler and Pressure Vessel-Code and the surveillance testing requirements.statedl in'the plant technical specifications; which are part'of the'final safety'analysis report. 'hese

oversight programs can provide'useful indications of.'aging-related'.

deterioration. These programs are.suppl1mented'by non-mandatory" ' inspection, surveillance, and'testing. .. '

Non-Destructive Examination

A-variety of non-destructive'eexamination (NDE) techniques are

employed as'part of the inservice inspection and testing programs to

detect and characterize flaws.or.other-' evidence of.degradation that

may precede'failure. Commonly used methods include' visual inspection,.dye-penbtrant and magnetic 'particle treatments, radiography,

eddy current-testing, ultrasonic testing,.electrical signature

analysis, and acoustic emission monitoring. Each ofthese methods

has advantages and limitations. The limitations.derive mainly from

the fact that NDE techniques were developed primarily'as quality

control tools for detecting manufacturing flaws. New or improved

NDE methods are continuously being-developed.:.-,It-is expected that

techniques that will -be-:available'in-,the future will provide the

.quantitative characterizations'of flaws'required for fracture

mechanics analysis and will allow on-line monitoring'of -deterioration in mechanical'properties during long-term inservice exposure.

* Condition Monitoring .

Condition.monitoring and trending of degradation form the basis for

predictive maintenance...-The overall goal of. the predictive maintenance program is to provide information concerning degradation rates

and residual lifetimes that.can be used to predict'and prevent failures. Tools used in doing'this.include'NDE, re'siduial-life assessment, and information:analysis and-trending. Trends~iand defined

action levels-provide-guidance needed by the preventive maintenance

program.to schedule services'with a-frequency-that will avoid failure of.critical;:SSCs..:-Observed-degradation provides opportunities

* for identifyingiand'eliminating sources of unnecessary deterioration

* through root-cause analysis and.corrective action. . -

B.3.2 Recordkeepin6, .

Recordkeeping is an-essential element of both'monitoring and.maintenance.

programs.' The producttof aging-monitoring programs'is-information that must

be translated into effective maintenance.practices.~' This'requires:1)Athat the

information obtained by monitoring activities be recorded in adequate; unam-.

biguous detail in a form that is readily retrievable and 2) that the information be easily applicable to specific maintenance practices. that effectively

mitigate age-related degradation. Records that meet these requirements are

needed to prioritize maintenance resources.and~to correlate-actual operating environments and stressor's with design assumptions and- -computed lifetimes so

that actual SSCs lifetimes and maintenance'intervals can be-anticipated.-.

B.5'
Maintenance records serve primarily to establishper'formance histories

for the SSCs that comprise the'plant. This information and'its continuous

feedback are useful in specifying 'what, how, and when equipment should be

maintained; what information 'should be collected; and.how it should be

recorded. Maintenance histories and equipment performance trends should be

documented consistently with regulatory requirements, and the licensees' goals

and objectives should be kept current.

A well-developed plant records system, coupled with effective access to

experience from similar plants, provides the basis'for.accurate condition or

failure profiling.''Effective profiling is important-in'making 'proper decisions regarding maintenince actions, that may mitigate the relatively' large

number (30%) of abnr6mal''occ'rrences'that have been attributed'to maintenance

shortcomings. Reqiirements'for records-'retention- and retrieval should be

established by the-maintenance program and be consistent'with quality assurance program requirements.

B.3.3 Trending'

Trending of information obtained: by monitoring'activities can lead

directly to maintenance recommendations. Often1- however, the selection of the

information to be'trended'must be based on a thorough'understanding of the

important mechanisms, if the trending results are'to be'meaningful.

Records of component failure data can be trended and monitored to assess

maintenance program effectiveness. Process indicators, such as postmaintenance test results, surveillance test results, ratio of preventive to

corrective maintenance,' maintenance backlog, and rework'frequency, can also be

trended-to indicate-problem areas and overall maintenance effectiveness.

B.3.4 Maintenance, Refurbishment, and Replacement

Maintenance methods range from simple, straightforward tasks to complex

activities that require extensive coordination, training, and technical expertise. The level of oversight and resources devoted to'these methods should

reflect their complexity'and importance to plant safety and reliability. A

maintenance programjhas-several 'important elements. Those relevant to-aging

management include corrective maintenance, predictive maintenance, preventive

maintenance, and refurbishment and replacement. Some of these elements are

interdependent and are related to aging monitoring activities-discussed in the

preceding subsections. In addition, the scope and nature of the various maintenance elements should reflect the as-built plant-design specifications;

manufacturers recommendations; operating experience--both internal and

external; relevant recommendations and information from the NRC, the nuclear

power industry, and its vendors; and general good engineering practices.

* Corrective Maintenance.

Corrective maintenance is performed'to restore failed or malfunctioning equipment to service. For some types of equipment or components, where failure does not present severe consequences, a

B.6
corrective,.as opposed to preventive,.,approach is.preferred (e.g.,

replacement, oflight'bulbs'and'certain categories"of non-critical

components).' As with other'maintenance activities, corrective maintenance priorities should be based bonthe relative importance of the

equipment and on' plant safety 'and reliability objectives. An added

function of corrective maintenance is todetermine root causes of

malfunctions and'carry out 'appropriate 'corrective action to prevent

unnecessary recurrences.

Predictive' Maintenance..

Predictive maintenance involves the actions necessary to monitor,

find trends,'and' analyze parameter/property and performance char-

'acteristics-or'signatures as'sociated.with.a piece of equipment which.

would indicate that the equipmten't'may'be approaching'a state'suich

that it may no longer be capable of pedforming'-its intended function. The predictive maintenance program should be effective in

reducing the failure of SSCs'by.using techniques that indicate the

need for preventive maintenance prior to equipment failure. The

data gathered should be analyzed,' .trends'should be identified, and

action levels 'sh'ould be defined. Action should be taken to provide

feedback to the maint'enance''program'in t'im'e to'preclude equipment

failure. The'pre'dictive maintenance program should provide data to

the preventive maintenance program and provide and retrieve equipment history .data. Root causes should be dete'rmined,,if possible,

and action taken and' results fed back'into'the program.

* Preventive' Maintenance'

Preventive maintenance includes scheduled actions performed to prevent equipment failure. Malfunctions that-represent significant.

challenges to plant-safety'or reliability-should be prevented., A'

major responsibility of the maintenance organization is' to.be cog- . nizant.of-the'significance 'of potential malfunctions and to ensure

that severe consequence,'events are-normally averted by adequate preventive mainten'ance... Preventive.maintenance relies heavily on

information generated by monitoring programs to define necessary

activities and the frequency' at which.they.should.be performed. In

.addition.to input from monitor.ing .p'gr'ams, .prevefitive'maintenance.

.action sh'ould'.be based'on equipment histories, 6ther plant'performance experience,''vendor recommendations, .and.good engineering practice, including as low 'as reasonably..achievable'considerations..

Planned actions.'and'sch'edules'.should be documented, and departures

from .these.-plans should be~justified on'technical.grounds and subject to management reviewvand',approval. 'Clear,.comprehensive procedures'are vital forpreventive maintenance-and other oversight and

maintenance activities.

* Refurbishment and Replacement

Certain classes of nuclear equipment.are subject to refurbishment or

.8B.7
replacement on prescribed schedules. It is important that timely

detection of refurbishment and replacement needs that arise in key

safety-related equipment due to premature aging are accounted for,

both in categories on prescribed schedules and those not on periodic

programs.

Environment Modification

Modifications to water chemistries in BWR reactor coolant systems by

hydrogen addition and changes from phosphate to volatile treatments

in PWR secondary systems are examples of major measures to decrease

corrosion of materials in safety-related areas of NPPs.

B.4 DETERMINING POTENTIAL IMPACT OF AGING ON SAFETY

Plant safety could be compromised if degradation of key SSCs is not

detected before a loss of functional capability, and timely corrective action

is not taken. If unmitigated, age-related degradation of SSCs can result in

an undetected reduction in the defense-in-depth concept. The defense-in-depth

concept requires that the public be protected from the accidental release of

fission products by a series of multiple barriers and engineered safety

systems.

A survey (Murphy et al. 1984) of licensee event reports (LERs) conducted

by Oak Ridge National Laboratory (ORNL), as part of the planning for this

aging research plan, offers examples of numerous instances of aging-induced

failures of equipment that have been reported. The reported events indicate

that essentially all types of safety-related systems have been affected by a

variety of degradation processes. Also, ORNL described the background of

selected age-related LERs in detail to provide a better perspective regarding

the safety significance of aging (Murphy and Vora 1985). Results of numerous

NPAR studies further augment the perspectives of aging impacts. Based on

these studies, aging effects can contribute to both the probability of initiation of transients and accidents, and the probability of failure of the

mitigating equipment during operation.

The regulatory concern is to recognize and correct or avoid age-related

degradation before it decreases the reliability of systems and components to

function properly. For example, if maintenance programs are deficient, valves

may fail to open or close on demand, sensors may send faulty signals, electrical relays may not operate properly, or diesel generators may fail to start.

Individually, such events may not have substantial effects on plant safety;

however, the cumulative result may be to decrease the margin of safety. Moreover, failure of a system or component to function properly may subject other

components to increased stress and increase the likelihood of their failure.

A possible consequence is that a malfunction of one component could trigger a sequence of component failures, or result in the simultaneous, commoncause failure of degraded components in other systems.

B.8
Of particular concern is the potential for common-cause failure of redundant components in safety systems, where simultaneous failure .during a transient or accident'could lead to loss of functional performance of parallel'

safety systems. -Aging-related degradation'that could.lead to common-cause',

failures needs'to be understood as a'basis to preclude accident initiation or

loss of safety function'and to maintain capability for accident mitigation.

Another type of common-cause'failure'is the,'simultaneous or sequential

failure of multiple'.SSCs due'to an external':trigger event.' A classic hypo-

'thetical example might be'a-seismic event.that trigg'ers'the failure of

supports or snubbers restraining corroded piping.'' Although'such occurrences

are extremely unlikely, the consequences could be quite severe. The first

age-related safety analyses tended to concentrate on this'type of major event,

which has low probability but potentially severe consequences. In these early

investigations,'a deterministic approach was used to attempt to.calculate the

risk associated with plant aging. More'recent investigations hav'elused an

approach that'Aincorporates the probabilities that.'va'rious sequences might."

occur. Both types'of analyses include consideration of the'likelihood of the

sequence or sequences occurring, and both define'"risk" to be the product of

the probability.of '6ccurrence of the sequence and its consequences. 'The

primary difference'lies in the methodology'used to.define'which sequence, or

sequences, should be considered in the analysis. .'

B.4.1 Deterministic Safety AnalVsis'

In a deterministic analysis, the choice of sequence and the sequence

probability is developed through the technical judgments of an interdisciplinary panel of experts. An example would be the Appendix K analyses of.effectiveness of emergency core cooling systems (10 CFR 50)-.- As other examples of

these kinds-of judgments, we have 1) the choice of..a "guillotine" pipe break

for-large loss-of-coolant analyses', 2) the more recent acceptance.of the "leak

before break",criterion,. 3) the complicated rules detailing which mitigating

equipment is available after the-initiating event, and 4) the single failurecriteria. The types of decisions and choices selected are typically. based on

an intuitive understanding 'of which accident sequences are likely to justify

analysis, that is, to be significant contributors to the overall risk associated with the plant. .

B.4.2 Probabilistic Risk Assessment '. . ' ' '

In contrast to the deterministic approach, where the.choices of.sequence

are based on expert -judgment, the.probabilistic risk assessment (PRA) approach

attempts to use the likelihood of occurrence and the consequences associated

with the occurrence.to determine which sequences.to analyze, and their probable contribution-to overall.plant risk.: The methodology used in a PRA

focuses on the risk associated with each'potential,'sequence to determine which

sequences justify analysis. In addition, PRAs normally include fault tree

analyses, which provide enumeration of the specific sets of component (and

other basic) failures that can cause a particular sequence of events leading

to an accident.

B.9
In practice, because of limitations in the availability of precise failure rate data and the impracticality of performing extensive Monte Carlo type

calculations involving all plant SSCs, expert judgment is used to constrain

the types of sequences considered in a PRA. Likewise, the expert judgment

that is used to select sequences to be considered in a deterministic safety

analysis is based on a knowledge of historical failure rates and an implied

perception of risk. Thus, it can be argued that both "deterministic" and

"probabilistic" safety analyses are hybrids, each encompassing elements of the

other's methodology, differing only in the time and manner in which time elements enter into the overall process.

B.4.3 "Hybrid" Analysis

The term "hybrid" has a special meaning in regard to safety analyses for

license renewal. Specifically, the hybrid safety analysis approach uses the

deterministic safety analysis methodology to quantify risk associated with

age-related degradation of vital components, where the most significant accident sequences are well defined. The PRA methodology is primarily used to

screen nonvital systems and components to determine if failure sequences

within them could adversely affect the performance of vital systems and components, thus increasing overall plant risk. This combination, or hybrid,

approach takes advantage of the lower costs associated with the deterministic

safety analysis, while assuring against unforeseen accident sequences that

might not have been recognized without application of the probabilistic safety

analysis methodology.

B.5 HUMAN-MACHINE INTERFACE

Human factors have contributed to accidents and are recognized as a major

factor in nuclear plant safety. The effects of personnel on the aging of

nuclear equipment have been identified in Phase-I assessments conducted under

the NPAR Program. For example, the diesel generator aging study recognized

classes of failures in the historical database that were attributed to

inappropriate personnel practices.

Human factors relate to plant aging in two ways. First, they can contribute to the aging of plant systems and components by affecting materials,

stressors, environments and their interactions. For example, inferior maintenance of diesels will contribute to their aging, or faulty application of

procedures during inservice inspection will fail to detect material flaws.

Second, human factors are important in the response to aging phenomena. For

example, by understanding how aging affects plant performance, plant personnel

are better prepared to make decisions relating to the operation and maintenance of a plant. Also a plant staff that is alert and perceptive to unusual

phenomena become a defense against the consequences of aging.

B. 10
APPENDIX C

ONGOING PROGRAMS RELATED TO NPAR
APPENDIX C

ONGOING PROGRAMS RELATED TO NPAR

C.1 INTRODUCTION

The U.S. Nuclear Regulatory Commission (NRC) and regulatory agencies in

foreign countries are sponsoring research and development programs to evaluate

the impact of aging on the safe operation of nuclear power plants (NPPs). The

nuclear industry, including the Electric. Power Research Institute (EPRI),

nuclear steam supply system (NSSS) vendors, utilities, and architect engi-'

neers, is interested in aging because of the significant economic advantage

that can be gained from extending the'operating license of aged power plants.

Therefore, the nuclear industry, EPRI, and the U.S. Department of Energy (DOE)

are sponsoring pilot studies to evaluate the issues involved in renewing

operating licenses for BWR and PWR plants.

Several NRC and industry-sponsored programs related to aging of NPPs are

currently underway. Many'of these programs are complementary to each other

and, therefore, it is essential that they are coordinated. Coordinating these

programs eliminates duplication of'effort and provides more complete coverage

of aging-related issues in a timely and cost-efficient manner. Such coordination and integration has'a special'significance for NRC if the criteria for

license renewal (LR) are to be formally established in the near future.

The Executive Director for Operations established a Technical Integration Review Group for Aging and Life Extension (TIRGALEX) to ensure effective

utilization of NRC resources. TIRGALEX recognized the importance of the coordination effort and conducted a'preliminary review of relevant programs and

activities already under way. Further coordination across NRC offices was

provided by the Aging and Life Extension Coordinating Committee (ALEXCC).

Currently, the NPAR Review Group, comprised of representatives from the NRC

offices and regions, provides review and coordination for age-related research

programs.

Interfaces between the NPAR Program and other ongoing NRC programs have

been established and will be maintained. Similarly, external programs involving both domestic and foreign organizations have been contacted.' This section

outlines the aging-related programs sponsored by NRC. It also outlines the

similar programs sponsored by industry'-in the United States and foreign countries. The outline of these programs is structured into the following

subsections:

C.2 NRC Programs and Activities Related to NPAR

C.3 Programs Jointly Sponsored by industry, EPRI, and DOE

C.1
C.4 Other EPRI Programs Related to NPAR

C.5 Ongoing Aging and Life Extension Programs at DOE

C.6 Ongoing Life Extension Activities in Codes and Standards Committees

C.7 Aging and Life Extension Progra ms in Foreign Countries.

The list of age-related programs external to NRC will be updated periodically. The updating will be made as information exchanges take place and

information for new programs is made available to the NPAR Program.

C.2 NRC PROGRAMS AND ACTIVITIES RELATED TO NPAR

In this section, a summary of ongoing NRC aging and LR related programs

and activities is presented.

C.2.1 Office for Analysis and Evaluation of Operational Data (AEOD)

The AEOD collects, screens, evaluates, and reports on operational experience and data from'the entire nuclear industry. Through this process the

AEOD identifies components and systems with repetitive, unusual, or possibly

generic failure modes, some of which are aging:related.

The AEOD also conducts evaluations of components, systems, procedures,

and management structures in response to specific operational events. In

performing these evaluations, AEOD is concerned with acquiring and analyzing

data that lead to identifying the root cause(s) of the events. The AEOD does

not specifically focus on aging issues per se, unless the incident involves

failures or degradations due to aging mechanisms. When an event indicates

age-related degradation, the information is forwarded to the appropriate NRC

office for action. The AEOD then publishes case studies that contain detailed

descriptions of the failure parameters, failure causes, and engineering

evaluations.

The AEOD has developed, through the Oak Ridge National Laboratory (ORNL),

a comprehensive computerized data base to aid in collecting and evaluating

licensee event reports.. This data base, the Sequence Coding and Search System

(SCSS), contains established procedures and codes for collecting operational

data. The SCSS also provides a means of storing and rapidly retrieving rootfailure-cause data that can be used in studies of component life and component

aging. It also has resulted in codes specifically identifying age-related

degradation.

The AEOD has an ongoing project to analyze the Nuclear Plant Reliability

Data System (NPRDS) data base. Jointly with the Institute of Nuclear Power

Operations, which maintains the data base, AEOD has developed a list of

critical components on which to focus attention. The trends and patterns

analysis of NPRDS data focuses on these key components. The AEOD's analysis

of NPRDS data results in statistical and engineering evaluations of component

C.2
failure modes, times to failure, operating conditions that affect failure, and

chemical and physical conditions affecting component-wearout rates.

C.2.2 Office of Nuclear Reactor Regulation (NRR)

A substantial number of aging-related programs are in progress in NRR.

These programs are discussed below in terms of Program'categories, Regional

activities, and Initiatives.

Program Categories

The NRR programs fall under three broad program categories: Operating

Reactors, Casework, and Safety Technology.

The Operating Reactors categories includes licensing actions and safety

assessments of both currently operating-reactors and new reactors coming on

line. Although aging concerns are generally addressed in the licensing process, to some extent, the licensing process does not consider the aging processes for all-SSCs. The capability of equipment to perform satisfactorily for

its specified lifetime is NRR's principal area of attention.

Casework is one-of the larger program categories in-NRR and represents

projects that are conducted to-support individual licensing actions. These

projects may be initiated by licensee requests for license amendments or by

events that occur at the plants that may require a regulatory response from

the NRC. Most aging-related projects are included in this program category.

NPAR expertise has been utilized in certain case work areas, e.g. service

water systems.

The Safety Technology program category is divided into five subgroups:

unresolved and generic safety issues; risk assessment; regulatory requirements; code analysis and maintenance; and human factors program issues. Some

of these programs have aging-related aspects. For example, unresolved and

generic safety issues are sometimes related to aging issues, e.g., Pressurized Thermal Shock. In the past probabilistic risk assessments (PRAs) have

not considered all potential effects of age-related degradation; however, NPAR

is developing methods to incorporate effects of aging into PRAs.

Within the human factors program area, the crucial role of maintenance in

predicting and correcting age-related degradation has been reflected in the

Maintenance and Surveillance Program.. Phase I of the Maintenance and Surveillance Program (NUREG-1212, Vol 1 & 2), was designed to survey current

maintenance practices in the U.S. nuclear industry and to evaluate their

effectiveness. Phase II is working toward resolutions of the issues identified and assessed in Phase I.

Regional Activities

Several NRR programs guide ongoing regional inspection activities

relevant to aging, aging detection, and mitigation of aging consequences.

C.3
These programs include the Safety System Functional Inspection Program, the

Safety System Outage Modifications Inspection Program, and the Generic Communication Program.

The Safety System Functional Inspection Program, in general, assesses

whether plant modifications of selected safety systems have degraded the

design margin to the point where the system's ability to mitigate design basis

events is impaired. This program consists of an in depth review of a small

number of safety systems and is usually conducted at older plants. The major

objectives of the program are to ensure that:

* testing is adequate to demonstrate that the safety systems are capable of performing the safety functions required by their design

bases

* system maintenance (with emphasis on pumps and valves) is adequate

to ensure system operability under postulated accident conditions

* operator and maintenance technician trainingis adequate to ensure

proper operation and maintenance of the systems

* human factors considerations relating to systems and supporting procedures are adequate to ensure proper system operation under normal

and accident conditions.

The objectives of the Safety System Outage Modifications Inspection

Program are to verify, through sampling inspections, that:

* licensees have instituted effective controls for conducting modification and repair activities during outages

* activities are accomplished in accordance with established procedures and commitments

* completed repairs and modifications have been properly designed,

installed, inspected, and tested

* affected systems are ready for safe startup and operation of the

plant following an outage.

The objectives of the Generic Communication Program are to:

* inform licensees of problems, including those due to aging and wear,

that have developed in individual plants

* appraise licensees of required action when these problems are shown

to be significant and generic.

These three programs apply to the pressure boundary hardware, drivers and

actuators, electrical power, and the instrumentation and controls of engineered safety features.

C.4
The NRR also guides the activities of the regions by issuing the Inspection and Enforcement Manual.- Portions of this manual establish inspection

procedures that are relevant to aging-and life extension. For example, some

inspection procedures establish guidance for ascertaining that inservice

inspection and testing'activities are'eprogrammed, planned, conducted,

recorded, and reported in accordance with Section XI of the ASME Boiler and

Pressure Vessel Code (ASME 1986). Where applicable, these procedures prescribe inspection of the licensee's recordkeeping for modification, maintenance, and repair activities. ' '

The components selected for specific inspection efforts are containment,

instruments and components (specifically, breakers and relays), and reactor

coolant system piping. The programs in-the'comporent'area are directed at

breakers and relays that may cause anticipated'transients without scrams

and/or-loss of safety-related'equipment.~'' The containment programs address

corrosion, fatigue, and crack-propagation 'in steel, reinforced-concrete, and

prestressed-concrete containments. The reactor coolant system piping program

addresses stress-corrosion cracking of welds, and'erosion and general corrosion of pipe bends and elbows.

NRR Initiatives

NRR has undertaken two major initiatives in preparation-for the planned

receipt of LR applications for the'lead nuclear power plants (Yankee Rowe in

June 1991 and Monticello in December 1991). 'The-first-major initiative

involves participation'with RES'in development of the LR Rule'and the 'supporting rulemaking package on LR. The Rule (10 CFR Part 54), "Requirements

for Renewal of Operating Licenses for-Nuclear Power Plants'," is scheduled to

be published inJune 1991. The second major initiative involves the development of a Standard Review Plan for License-Renewal (SRP-LR) which the staff

will use in the review of LR applications. Also in support of the Rule, the

Office of Regulatory:Research has issued'a draft regulatory guide (DG-1009),

Standard Format and Content of-Technical Information for ApDlications to

Review Nuclear Power Plant Onerating Licenses. Both the draft SRP-LR (NUREG1299) and the draft regulatory;guide were issued for comment in December 1990;

publication of the final-documents is scheduled for''April 1992.

C.2.3 Office of Nuclear Regulatorv Research (RES)

RES sponsors a number of large and important aging-related programs. The

general objectives of these-programs are to'identify aging mechanisms, evaluate their safety and regulatory impacts, -assess detection methods for particular aging degradation mechanisms, and develop mitigating and corrective

actions.

Major program areas with aging-related projects include Mechanical and

Electrical Equipment Qualification; Instrument Integrity; Primary System

Integrity, which includes the Heavy Section Steel Technology program, Degraded

Piping, and Steam Generator Integrity projects; Equipment Operation and

Integrity, which includes the Nuclear Plant Aging Research Program; and Nondestructive Examination.

C.5
RES has an initiative underway addressing 10 CFR Part 54 to develop and

publish (jointly with NRR) a License Renewal Rule. The Rule is scheduled for

publication in June 1991. Another RES activity addresses 10 CFR Part 51, to

develop and publish an environmental Rule for LR. The proposed Rule is scheduled for publication in August 1991, and the final Rule is scheduled for July

1992.

C.3 PROGRAMS JOINTLY SPONSORED BY INDUSTRY. EPRI. AND DOE

A joint industry, EPRI, and DOE program was initiated in 1984 to identify issues associated with light-water reactor (LWR) license renewal. In

1984, DOE and EPRI agreed to co-fund studies and developed a joint R&D plan in

1985. Two pilot studies involving Surry-1 (PWR) and Monticello (BWR) were

initiated. After completion of the pilot studies, a bidding process was used

to select two lead plants to develop requests for LR for submittal to NRC.

Northern States Power conducted the BWR pilot plant study at its

Monticello Nuclear Generating Station, co-funded by EPRI and DOE. The utility

was the successful bidder for developing a lead plant LR request, also for

Monticello. The request is expected to be submitted to NRC in December 1991.

The PWR pilot plant study was conducted by Virginia Power at its Surry

Unit-1 plant. Yankee Atomic Electric was the successful bidder for developing

the PWR lead-plant LR application, for the Yankee Rowe Nuclear Power Station.

The LR application is scheduled to be filed with the NRC in September 1991.

In June 1988, the Nuclear Management and Resources Council, composed of

representatives from all U.S. nuclear utilities, formed its NUPLEX (Nuclear

Utility Plant Life Extension) Working Group, and charged NUPLEX with the lead

responsibility for all nuclear industry activities related to LR. The collaborative efforts of NUPLEX, EPRI, and DOE have resulted in the issuance of

eleven industry reports (IRs) in support of the LR process. These IRs, which

have been submitted to the NRC for review (see Figure 5.2), are intended to

provide the generic technical bases for evaluating the effects of age-related

degradation in specific NPP structures and systems important to LR.

C.4 OTHER EPRI PROGRAMS RELATED TO NPAR

In addition to the joint programs noted in section C.3, EPRI has a number

of programs related to plant life cycle management that address age-related

degradation. Examples include:

1. Nuclear Plant Life Extension

2. Corrosion Control

3. Component Reliability

4. Steam Generator Reliability.

C.6
The Nuclear Plant Life Extension Program approaches~aging and life-extension from the technological perspective of-understanding the aging process for

systems and components. It does not address institutional or legal issues.

The Corrosion Control Pr'ogram addresses environmentally caused cracking and

pitting. It specifically emphasizes-BWR water chemistry and understanding and

mitigating pipe cracking due to corrosion. The Component Reliability Program

is related to the Corrosion Control Program. Structural reliability and

safety improvement are the main thrusts of the Component Reliability-Program.

Specific examples of activities include material characterization, flaw

detection and assessment, inspection hardware'development, and reliability

methodology. While the central program focus is materials, the program also

addresses piping, reactor pressure vessel, and.steam generator tubes.

Three other EPRI initiatives, are of'special, interest to the NPAR Program; these are the NCIG (formerly, the Nuclear Construction Issues Group),

the Nondestructive Evaluation (NDE)-Center,7and'the Nuclear Maintenance Applications Center (NMAC). The primary activity of NCIG of interest to NPAR is

the preparation of guidelines for recordkeeping related-to procurement, qualification, and replacement of nuclear components. The NDE center has compiled

an impressive record in transforming advanced NDE methods from promising

techniques into proven, field-qualified inspection technologies. The NMAC

develops guidelines for-maintenance and maintenance-related-activities (e.g.,

pre- and post-maintenance equipment monitoring and diagnostic techniques).

EPRI LR-related activities in CY 1991 include:

* participation in lead plant initiatives (-50% of funds)

* participation in industry technical reports (-30% of funds)

including resolution of review comments

* data and record guidelines

* cable condition monitoring

* in-plant versus artificial-aging.

C.5 ONGOING AGING AND LIFE EXTENSION- PROGRAMS AT DOE

Plant license renewal is being pursued by DOE in several ways. -One of

three major efforts within the LWR safety area is the cooperative industry/

EPRI/DOE program'to extend the productivity of existin'g and future LWRs. DOE

has conducted a'study on delineatingtthe key'LWR structures and-components - '

capable of continued service beyond their design life without refurbishment or

replacement. Further, DOE has evaluated items requiring replacement in terms

of safety significance, licensing,-and cost implications. The DOE programs

are intended to support relicensing'by influencing the federal regulatory

process and evaluating the safety and economic impacts of improved plant

performance. - ;

C.7
Sandia National Laboratory (SNL) is the lead DOE laboratory for research

related to LR. Principal activities sponsored by DOE/SNL in FY 1991 are:

* participation in lead plant initiatives (Yankee Rowe and

Monticello), including identification of effective programs to

manage aging

* participation in preparing 10 industry technical reports (first in

August 1989) (issued October 1989 through October 1990), including

resolution of review comments

* emergency diesel laboratory at SNL

* development of screening methods

* evaluation of flaw distributions in reactor pressure vessels

* evaluation of sub-sized Charpy specimen performance

vessel support embrittlement studies

* development (with GE) of in-core electrochemical potential probes

* development (with GE) of an in-reactor crack monitor

* distribution of cable aging report

* annealing studies on reactor pressure vessels.

C.6 ONGOING LICENSE RENEWAL ACTIVITIES IN CODES AND STANDARDS

The following is a summary of the status of code activities related to

license renewal:

Institute of Electrical and Electronics Engineers (IEEE). Working

Group 3.4, "Nuclear Plant Life Assessment," is chartered to investigate the

codes and standards aspects of plant license renewal as they pertain to Class

IE electrical equipment. The working group has developed an action plan and

has provided several drafts of a guide document to function in place of

standards.

Instrument Society of America (ISA). There are no current activities.

They take their lead from IEEE. When relevant items arise, IEEE will make the

necessary contacts.

American Society of Mechanical Engineers (ASME). Under the auspices of

the ASME Section XI subcommittee, the Special Working Group on Plant Life

Extension is active. They have participation from EPRI, NRC, DOE, utilities,

several laboratories and NSSS suppliers. To date, the special working group

has reviewed a broad spectrum of aging and license renewal programs.

C.8
Several important changes and additions'were incorporated in Section XI

of the ASME Boiler and Pressure Vessel Code during 1988 and 1989.

C.7 AGING AND LIFE EXTENSION PROGRAMS IN FOREIGN COUNTRIES

Extensive programs in materials degradation, NDE, fracture mechanics,

structures, and other'aging-related areas exist worldwide. A few of these

have been detailed in NUREG/CR-3040. The NRC operates large cooperative

international research programs','several of which are directly related to

aging and life extension.

Other programs identified in the-analyses of several selected components

and structures include:

* Westinghouse and Framatome--performing work separately and jointly

on NDE for cast stainless steel.;

* JAERI--evaluating the effects of separate and simultaneous application of environmental stressors on cables (NUREG/CP-0071).

* France--developing predictive maintenance programs.

Additional details on foreign activities on NPP aging and life extension

are summarized below.

CANADA

The AECL (Atomic Energy of'Canada Limited) approach to'managing aging

degradation of nuclear power plant components and maintaining adequate plant

safety embodied in their Local In-Plant Fatigue Evaluation (LIFE) System. The

LIFE system involves the recording,'assessment, and-evaluation of the history

of the in-plant conditions. Particular emphasis is placed on fatigue evaluation, rehabilitation and replacem6nt of key CANDU reactor and heat transport

system components.

As applied to currently operating CANDU {Canadian Deuterium Uranium)

reactors, major features of the LIFE approach include:

* Operating policies and procedures prescribe practices that'are

'designed to minimize aging effects. Components, particularly

pressure-retaining components,'important to safety are'subject to

inservice inspection.

* Deficiencies are systematically detected and recorded; causes

(including aging degradation) are determined, and appropriate corrective actions are taken.

C.9
* Licensees are required to monitor and periodically evaluate equipment and system performance against the reliability targets set by

the Atomic Energy Control Board (AECB), allowing detection of aging

trends.

• Research and development, by both the industry and the AECB, provide the basis for predicting the behavior of critical plant components during plant operation, developing better components and

developing monitoring methods capable of detecting component degradation before loss of safety function occurs.

* Aggressive monitoring is utilized, employing.monitoring instruments, periodic testing, inspection, maintenance, and field patrols.

It is recognized that age-related degradation can be detected only

if appropriate time-variant parameters are monitored.

* Significant Event Reports are prepared for events that have a significant negative effect on reactor safety, worker or public safety,

and cost. This...system has the following important features:

- Recording of equipment and operation deficiencies in a

specified, systematic manner to allow event review and

analysis.

- Multilevel diverse screening process.

- Trend lines of equipment and system deficiencies.

- Communication of lessons learned to other CANDU stations,

owners, designers, and equipment manufacturers.

* Reactor Safety Reliability Assessment is an important part of the

annual comprehensive and systematic review of nuclear power plant

operation and maintenance by both the licensee and the AECB. It

gives the actual past-year performance and the predicted future performance in terms of system unavailabilities and serious process

failure occurrences. These can be compared to the AECB reliability

targets mentioned earlier.

Information obtained from the LIFE program is. also incorporated in new

(CANDU-3) plants during the initial design stages. Special attention is paid

to maximize the longevity of the components and where required make provisions

for economic and expedient replacement procedures.

Ontario Hydro (OH) also has an internal Nuclear Plant Life Assurance

(NPLA) program aimed at:

maintaining the long-term reliability, availability, and safety of

OH's nuclear plants during the normal service life of 40 years (life

assurance).

C.10
* preserving the option of extending the life of OH's nuclear plants

beyond the currently assumed service life of 40 years (life

extension).

* ensuring that future activities in these areas'in OH will conform to

an overall plan.

The NPLA program which is quite similar to that of NPAR, includes initiatives a) retrieve samples and information from the Nuclear Power'Demonstration (NPD) reactor, and b) inspect and assess the condition of critical

components of the Pickering Station reactors during retubing operations.

GERMANY

The Geselschaft for Reakorsicherheit, m.b.H. (GRS) has been engaged, on

behalf of the Federal Minister of the Environment, Nature'Conservation, and

Regulatory Safety, in a data collection program at power plants for 15 years

(since initial operations). This program has demonstrated how long-term data

collection and trending of component performance parameters and functional

indicators (a strategy/approach similar to that-recommended in the NRC'S

program plan for Nuclear Plant Aging Research NUREG-1144) provide protection

against failures resulting from age-related degradation.

Age-related degradation has been observed in emergency diesels; power,

measuring, and control cables; pump motors; valves; and a variety of electric

and electronic equipment. The safety and availability problems that may

result from aging are counteracted by a system of inspection and planned preventive maintenance measures. The GRS has implemented a long-term systematic

collection of data so that the accumulation of defects on certain components

can be recognized.

FRANCE

From the outset of the nuclear program, the French Safety Authorities and

the tlectricit6 de-France (EdF) (the main licensee) took into consideration

the effects of aging on the installed equipment'.

In French designs, qualification-is-one means used to check equipment

design. In most cases, it includes testing designed to evaluate the behavior

of'theequipment'with time.

Examples of aging experiences in French reactors include:

For mechanical equipment--diesel generators are exposed to untimely

starting; steam generators are subject to unforeseen corrosion or

erosion due to foreign matter.

C.11
For electrical equipment--isolating switches are operated under

unscheduled loading conditions and batteries whose autonomy sometimes changes unexpectedly.

In addition, materials such as coatings (paints) or lubricants (oils or

greases) have a great effect on the behavior of the equipment with which they

are associated.

In conjunction with the safety authorities, EdF has initiated a program

of investigations to:

* develop a method of measurement, in theory and practice, of the

aging of equipment

* determine the influence of NPP operating procedures on aging.

In addition, a file of events is kept up to date for each plant and is

analyzed to evaluate the behavior of the installation.

Since the beginning of the French nuclear program, means have been set up

that ensure the periodic monitoring of operation or of the intrinsic properties of the equipment or systems. Some typical examples are:

* periodic testing

* inservice inspection

* preventive maintenance.

Finally, procedures have been set up to evaluate primary circuit performance for each plant (measurement of primary circuit pressure and temperature

values above a certain threshold) and allow comparison to design features.

JAPAN

Technology development for NPP life extension is a priority effort in

Japan. A 7-year technology development plan was implemented in FY 1985.

The major tasks in plant life extension technology development are

related to the diagnosis of NPP aging deterioration, prediction of remaining

plant life, and replacement and improvement of plant equipment. The following

tasks have been implemented for comprehensive evaluations:

* development of life diagnosis and prediction methodology (developing

life evaluation methods, creating aidata base on aging and degradation phenomena, developing monitoring techniques)

• development of technology for replacing and improving large equipment (developing methodology, verification).

C.12
The selection of equipment and structures to be studied in detail was

made in 1985. Structures and equipment that could be easily replaced were not

subject to-further evaluation. The remaining structures and components were

ranked according to evaluation criteria that included ease of replacement and

impact on safety. -

Research on age-related degradation of key components (e.g., RPV, concrete structures, and cables) is being conducted by the Nuclear Plant Aging

Research Team (NPART) at the Japan Atomic Energy-Research Institute. Studies

on irradiation embrittlement of RPV steels encompass the effects of water

chemistry and temperature on crack growth, flux intensity effects, and

destructive examination of samples cut from the pressure vessel of the Japan

Power Demonstration Reactor (JPDR). Research on age-related degradation of

concrete structures includes the development of monitoring methods, testing of

samples bored from the biological shield of the JPDR, and separation of

neutron- and thermal-aging effects onconcrete strength. The cable research

program includes development of monitoring methods, accelerated aging of

cables, and correlations between aging of test specimens and actual NPP

components.

TAIWAN

A Component Life Assessment for NPP (CLANPP) program in-Taiwan involves

staff from the Materials Research Laboratories (MRL) and from Taiwan Power

Company (Taipower). The program began in July 1987. The project goals are to

establish a data base on material properties, failure histories, and operational parameters for selected critical components; also to provide component

life prediction criteria.

Current research activities conducted under the CLANPP Program include:

• stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) studies

SCC: Establishment of SCC crack growth rate prediction model and

development of on-line monitoring techniques

IASCC: Participation in the tests to study IASCC

• Life assessment of reactor vessel and its internal components

- Evaluation of fatigue usage factor

- Radiation induced degradation of support skirt

- IGSCC/IASCC of Top Guide/Shroud/Core Plate

C.13
Aging study of safety-related valves

- Motor current monitoring of aged motor operated valves

- Aging study of solenoid operated valves

- Improvement of stem packing performance

In addition to participation in the CLANPP Program, NRL has or is

planning research activities in the following areas:

* Life assessment on turbine rotors

* Pipe integrity study of LWRs

• Component degradation study of PWR feedwater system

* Aging assessment of cables.

SPAIN

A plant life extension program is underway in Spain, emphasizing surveillance. The program deals with programming and reporting inspections and

tests in the following areas:

single- and double-phase erosion corrosion

low-pressure turbine disc cracking

intergranular stress corrosion cracking of piping, and

control rod drive housing welds.

Fracture mechanisms and procedures for evaluating remaining life are

addressed.

IAEA Activities

Since 1986 the IAEA has been developing initiatives in nuclear plant

aging and life extension. Highlights of the activities are outlined below.

1. In 1986 an IAEA technical committee meeting of "Safety Aspects of

Nuclear Power Plant Aging," was held in Vienna, Austria, September

1-5. Delegates from 12 countries attended.

2. In 1987 the IAEA sponsored an international symposium on "Safety

Aspects of the Aging and Maintenance of Nuclear Power Plants," held

in Vienna, Austria, June 29-July 3. There were 140 attendees from

about 30 countries.

C.14
3. In 1988, J. Pachner, representing IAEA,..outlined-IAEA plans for 'a

comprehensive program on nuclear plant-aging and life extension at

the International Symposium on Nuclear Power Plant Aging, Bethesda,

August 30, 31, September 11.

4. In 1989 and 1990, IAEA sponsored the following-activities in NPP

Aging:

- June 1989, consultants meeting on NPP recordkeeping, held in

Vienna, attended by representatives from four countries.

- November.:13-17, 1989, consultants meeting in "Safety Aspects of

NPP Aging," held in Vienna.

- November 20-24, 1989, technical committee meeting on "The Evaluation and Management of the Safety Impact of NPP Aging," held

in Vienna, attended by representatives from 16 countries.

- June 23-28, 1990, consultants meeting to finalize the technical

report on, "Selection of NPP components for Ageing a Management Pilot Studies."

- November 5-9, 1990, technical committee meeting to define

existing knowledge and gaps relating to understanding and managing of aging and to propose pilot studies, on selected components. Representatives from 14 countries and the committee

of European communities attended.

IAEA working groups have been organized to conduct pilot studies on four

safety significant NPP components, selected on the basis of their safety significance and their susceptibility to different types of ageing degradation.

These components are: 1) the primary nozzle of a reactor pressure vessel,

2) a motor operated isolation valve, 3) the concrete containment building, and

4) instrumentation and control cabling within the containment.

The objectives of the pilot studies are, for each component, to identify

dominant ageing mechanisms, to identify or develop an effective strategy for

managing ageing effects caused by the identified mechanisms, and to validate

or improve the methodology for ageing management studies developed by the

IAEA.

A document summarizing the pilot studies is expected to be published in

1991. Implementation of recommendations from the studies will proceed in

1992.

(a) IAEA spelling.

C.15
The basis for cooperation between IAEA and NEA/OECD has been established.

Joint participation in a meeting in 1991 has been planned. The duration of

the IAEA program in Safety Impact of NPP Ageing is expected to be approximately five years.

The results of the pilot studies will have application in monitoring the

degradation and in preventive maintenance of the selected components, including the development of criteria for decisions on the type and timing of preventive maintenance actions; and in predictions of component performance and

remaining service life under all expected service conditions, including postulated accident and post-accident conditions.

C.16
APPENDIX D

ALTERNATIVE APPROACHES FOR CONDUCTING NPAR

PHASE-II COMPREHENSIVE ASSESSMENTS
APPENDIX D

ALTERNATIVE APPROACHES FOR CONDUCTING NPAR

PHASE-II COMPREHENSIVE ASSESSMENTS

Alternative approaches have been used to conduct comprehensive aging

assessments of naturally-aged components and systems under the NPAR Program.

These approaches are determined by the size and complexity of the components

and systems, relating to the equipment availability and potential expense of

the aging assessment. The first approach,' in-depth laboratory investiga-'

tion, focuses on laboratory assessments of age-related degradation of components, such as small'electric motors. The second and third approaches,

systematic applications of experts aand in situ evaluations, focus on age-.

related degradation of systems and components that are too large, complex, or

costly for laboratory study. Examples of these components include emergency

diesel generators, snubbers, and service water system components. These

alternative approaches to comprehensive aging assessments are described below

in more detail.

D.1 APPROACH I - EXAMPLES OF IN-DEPTH LABORATORY -INVESTIGATIONS

Case 1 - Small Electric Motors

The small motors are generally sufficiently compact, portable, and available to accommodate laboratory investigations within the available budget.

Motors representing a range of sizes and modelswere investigated at BNL,

including seismic tests. Phase-II results are reported by Subudhi, Burns, and

Taylor (1985).

Case 2 - Motor-Onerated Valves

Suitable valves were also available and portable, permitting extensive

laboratory studies at ORNL for both-aging characteristics and monitoring methods. Phase-II results are reported by'Crowley 'and Eissenberg (1986).

D.2 APPROACH II - EXAMPLES OF SYSTEMATIC APPLICATION OF EXPERTS-AND IN-PLANT

ASSESSMENTS ;

Case I - Aging Assessment of Emergency Diesel Generators (EDGs)

The EDGs represent a class of equipment that is large and complex, having

approximately 25 models and nine suppliers. Nucle'atr'diesels in the size range

at current plants have either not been'retired or-have been diverted to other

uses. Even given one or more naturally-aged EDGs, an in-depth laboratory

study (at considerable expense) would address aging characteristics of one

D.1
model that may have limited application to other models. The alternative

approach selected by PNL investigators (Figure D.1) was to establish a panel

of three EDG experts, collectively representing -80 years of experience with

nuclear and non-nuclear EOGs, including experience with design, operation and

failure analysis. 'The panel focused on a statistically selected set of -2000

EDG failures described in four databases. The panel was directed to systematically interpret the failures (generally minimal) in-terms of their root

cause and whether their cause was age-related. The panel's interpretation, of

the aging characteristics of the nine EDG types were entered into a dBASE III

program that accommodated numerous correlations. The results facilitated

identification of the' EDG' components most vulnerable to aging and the prominent aging mechanisms. 'The results were then subjected to review in a workshop that included EDG operators from several utilities. Results are reported

in NUREG/CR-4590, Vols.'.1 and 2 (Hoopingarner et al. 1987; Hoopingarner and

Vause 1987). A second workshop', involving the EDG-experts established the

basis for recommendations to mitigate aging of EDGs. Information collected

during site visits and observations during a major EDG overhaul validated the

Understanding

h - Aging *1+ Managing_ _

Aging

EDGs;

foX

Expert

*3 EDG

Consultants

* 2 PNL Staff

FIGURE D.1. ,

* 15 Utilities

* Manufacturers

.DOE Labs

* EDG

Consultants

.Utility Staff .2 EDG * 4 Utilities

* PNL Staff Consultants * PNL Staff

. EDG * PNL Staff

Mechanics

* Reg G.1.9

* GSI B-56

* USI A-44

* IEEE

Standards

- RIL

39003008.2K

Application of NPAR Strategy to Understand and

Manage Aging of EDGs

D. 2
research results. Several regulatory applications resulted from the EDG aging

research. The recommended approach to-aging mitigation is reported in NUREG/

CR-5057 (Hoopingarner and Zaloudek 1989).

Case 2 - Aging Assessment of Snubbers

Snubbers are components that involve two prominent designs (hydraulic and

mechanical) and a wide range of sizes,'and applications. A large number

(>1,000) are used in many plants, and they are exposed to a wide range of

environments. Laboratory studies were'feasible but it would be costly to

address statistically-significant tests of various snubber sizes and designs

under simulated plant conditions. Snubbers occasionally'become contaminated

with radioactive material during service; thereby increasing'examination costs

and requiring difficult laboratory procedures and disposal. 'Therefore, the

Phase-II comprehensive snubber aging assessment was pursued using an in-plant

snubber aging study (Figure D.2)._ Thejin-plant approach involved plant visits

to cooperating utilities by snubber-experts.- The utilities provided access to

data from a relatively large number and variety of snubbers. Experts familiar with..snubber designs, operating conditions, and failure characteristics

participated in acquisition and'interpretation of'the data.

Case 3 - Aging Assessment of Major LWR Components and Structures

The major LWR components' and structures are'designed'and fabricated by

several vendors'and-architect engineer's. They include twenty-two components

and structures: PWR and BWR primary pressure boundary-components, primary

containments, feedwater lines,-emergency'diesel generators, and cables and

connectors. The in-depth laboratory study of aging of these components is not

feasible because of the large amount of funding and time required; however, a

significant number of laboratory and in-field studies have been performed for

these components and structures. The major element of the strategy adopted by

INEL is to integrate, evaluate, and update the aging-related information from

the completed and ongoing NRC, EPRI,-and industry research programs and from

several aging related databases.' The'other elements of the strategy include

direct contribution from twenty-five experts having state-of-the-art agingrelated knowledge of a particularcomponent or structure and include peerreview of technical reports by the experts at the USNRC, national'

laboratories, and industry. The experts and INEL staff have worked together

with experience in assessing age-related degradation-of nuclear SSCs have

worked with the INEL.st'affito identify'- degradation'sites and mechanisms, -

stressors, relevant'field 'failures,'.and'potent'ial failure modes and to evaluate the current inspection, surveillance, and monitoring methods for each

major component and structure. Thes& results are presented in NUREG/CR-4731,-

Volumes 1 and 2. Figure's 6.5 --6.'15 are concise summaries of the systematic

and in-depth aging'assessments that have been conducted. A similar strategy

is followed to develop life.assessment'procedures'.formajor components and

structures and to evaluate''em'ergingr inspection,,monitoring, and material

testing techniques to characterize aging damage. -

D.3
h Understanding

I Aging

|4 | Managing Aging -

I ~ ~In Progress 0

|Laboratory I

Snubbers H

|Expert|

x

F Ac

Assej

* PNL Staff * Contractors * Contractor * PNL Staff . 6 to 8

* Consultants - Wyle Labs - Wyle Labs * Wyle Labs Utilities

* NRC - Lake - Lake Eng. * Lake Eng. * Evaluate

Oversight Engineering * PNL Staff * Utilities Database

* PNL Staff * Utilities * NRC Direction Improvement

* Interface with * NRC Oversight and . NRC Direction

- SNUG Participation

- OM.4

Committee

* NRC Oversight

* ASMEIOM.4

Standard

* Standard

Tech. Specs.

. Industry

Interface

(e.g., SNUG) . NRC Oversight

39003008.1K

FIGURE D.2. Application of NPAR Strategy to

and Manage Aging of Snubbers

Understand

Case 4 - In-Plant Evaluation of Inspection, Surveillance and Monitoring

Methods

In-plant evaluation of inspection, surveillance, and monitoring methods

(IS&MM) practiced at a representative pressurized water-reactor power plant

was performed. The evaluation covered standard and advanced techniques in use

and assessed the applicability of each technique for detecting and monitoring

age-related degradation in reactor protection and 1E power system components.

A comparison was made between current and advanced IS&MM to determine if the

use of advanced methods is practical in the field and, if so, whether they

provide information for trending aging that is not provided by current

methods. The advanced IS&MM techniques included a state-of-the-art infrared

thermography data acquisition and analysis system, the Redundant Instrument

Monitoring System (RIMS'), and the Electronic Characterization and Diagnostic

D.4
(ECADtm) System. The evaluations determined that the advanced IS&MM investigated do provide information and methodologies for trending that is not currently provided by routine surveillance procedures.

D.3 APPROACH III - IN SITU EVALUATIONS

Pacific Northwest Laboratory has performed several in situ evaluations of

operating power reactor service water systems. These evaluations were done

using a modeling approach to determine root causes. This approach was developed during the RES program to assess the effects of operating and maintenance

methods and procedures and to assess the measurement technique's used to determine the thermal hydraulic capabilities of an active service water system.

The need for these evaluations developed from 1)' functional failure events in

critical service water components, and 2) utility efforts to comply with

generic letter 89-13, Service Water System Problems Affecting Safety-Related

Eguipment, July 18, 1989, which requires thermal performance verification of

safety-related service water system components. Reports were prepared and

delivered to Nuclear Regulatory Research, which presented an evaluation of the

safety significance of operational aging on service water system thermal

hydraulics. These in situ case studies affirmed both the methodology and the

practical application of the NPAR approach to research on component

degradation.

In situ assessments also were conducted on electrical circuits and

selected components (e.g., cast stainless steel) at the Shippingport reactor

(see Appendix E, Allen & Johnson 1990).

D.5
i

APPENDIX E

SHIPPINGPORT STATION AGING EVALUATION

NRC CONTRACTOR PUBLICATIONS
APPENDIX E

SHIPPINGPORT STATION AGING EVALUATION

NRC CONTRACTOR PUBLICATIONS

Allen, R. P., and A. B. Johnson,

Shippingport Aging Evaluation."

Jr. 1990. "Lessons Learned to Date from the

Nucl. Eng. & Design 118, pp. 319-327.

Allen, R. P., and A. B. Johnson, Jr. 1990. -Shippincpod-Station Aging Evaluation Final Report. NUREG/CR-5491,- PNL-7191, prepared by Pacific Northwest

Laboratory for the U.S. Nuclear Regulatory Commission, *Washington, D.C.

Allen, R. P., A. B. Johnson, Jr., and J. P. Vora. 1988. "Shippingport Atomic

Power Station - A Source of Power Plant Aging Information." In Proceedings of

an International Symposium on Safety Aspects of the Aging and Maintenance of

Nuclear Power Plants, IAEA-SM-295/39, pp. 129-137, International Atomic Energy

Agency, Vienna, Austria.

Allen, R. P., and A. B. Johnson, Jr. 1988. "Lessons Learned to Date from the

Shippingport Aging Evaluation." In Proceedings of the Sixteenth Water Reactor

Safety Meeting, NUREG/CP-0097, vol. 3, ed. A. J. Weiss, pp. 41-58, prepared by

Brookhaven National Laboratory for. the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Bowers, D. L., and L. R. Greenwood. 1989. "Analysis of Activation Products

in the Neutron Shield Tank of the Shippingport Reactor." 31st Conference on

Analytical Chemistry in Enery Technology, October 10-12, 1989, Gatlinburg,

Tennessee. (Abstract)

Chopra, 0. K., and H. M. Chung. 1990. "Studies of Aged Cast Stainless Steel

from the Shippingport Reactor," NUREG/CP-0113, presented at the Eighteenth

Water Reactor Safety Information Meeting, October 22-24, 1990.

Chopra, 0. K., S. T. Rosinski, and-W. J. Shack. 1990. "Embrittlement of the

Shippingport Reactor Neutron Shield Tank." The American Society of Mechanical

Engineers, PVP-Vol. 195

Chopra, 0. K., and W. J. Shack. 1989. "Embrittlement of the Shippingport

Reactor Shield Tank." In Proceedings of the Seventeenth Water Reactor Safety

Meeting, NUREG/CP-0105, ed. A. J. Weiss, prepared by Brookhaven National

Laboratory for the U.S. Nuclear Regulatory Commission, Washington, D.C.

Chopra, 0. K., and W. J. Shack. 1989. "Irradiation Embrittlement of the

Shippingport Neutron Shield Tank." -In Proceedings of the 15th MPA-Seminar on

Safety and Reliability of-Plant Technology, vol. 1, 14:1-8.

E.1
Chung, H. M. 1989. "Thermal Aging of Decommissioned Reactor Cast Stainless

Steel Components and Methodology for Life Prediction." In Proceedings of the

1989 ASME Pressure Vessel and Piping Conference, vol. 171, eds.

T. V. Narayana, et al. pp. 111-115.

Chung, H. M., and O. K. Chopra. 1988. "TEM, APFIM, and SANS Examination of

Aged Duplex Stainless Steel Components from Some Decommissioned Reactors."

Trans. Am. Nucl. Soc. 56:249-252.

Chung, H. M., and T. R. Leax. "Embrittlement of Laboratory-and-Reactor Aged

CF-3, CF-8, and CF-8M Duplex Stainless Steels." In Proceedings of an International Workshop on Intermediate Temperature Embrittlement Processes in

Duplex Stainless Steels, August 1-2, 1989, Oxford, England. (in press)

Dinsel, M. R., M. R. Donaldson, and F. T. Soberano. 1987. In-Situ Testing of

the Shippincport Atomic Power Station Electrical Circuits. NUREG/CR-3956,

EGG-2443, prepared by EG&G Idaho for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Gunther, W. E. 1988. Testing of a Naturally Aged Nuclear Power Plant Inverter and Battery Charger. NUREG/CR-5192, BNL-NUREG-52158, prepared by

Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Nuplex News. January 1989. "Shippingport Aging Evaluation." Published for

the Electric Power Research Institute/U.S. Department of Energy by Multiple

Dynamics Corporation.

Robertson, D. E., and C. W. Thomas. 1988. "Radionuclide Source Term Measurement for Power Reactor Decommissioning Assessment." In Proceedings of the

Fifteenth Water Reactor Safety Meeting, NUREG/CP-0091, vol. 6, ed.

A. J. Weiss, pp. 67-88, prepared by Brookhaven National Laboratory for the

U.S. Nuclear Regulatory Commission, Washington, D.C.

Robertson, D. E., C. W. Thomas, N. L. Wynhoff, and D. C. Hetzer. 1989.

"Radionuclide Characterization of Reactor Decommissioning Waste and Spent Fuel

Assembly Hardware." In Proceedings of the Sixteenth Water Reactor Safety

Meeting, NUREG/CP-0097, vol. 1, ed. A. J. Weiss, pp. 73-118, prepared by

Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission,

Washington, D.C.

Robertson, D. E., C. W. Thomas, N. L. Wynhoff, and D. C. Hetzer. 1988.

Radionuclide Characterization of Reactor Decommissioning Waste and Spent Fuel

Assembly Hardware. PNL-6806, Pacific Northwest Laboratory, Richland,

Washington.

Rosinski, S. T., 0. K. Chopra, and W. J. Shack. 1989. "Shippingport Neutron

Shield Tank Sampling and Analysis Program." In Proceedings of the 1989 ASME

Pressure Vessel and Piping Conference, vol. 170, eds. D. L. Marriott,

T. P. Mager, and W. H. Bamford, pp. 109-113.

E.2
Rosinski, S. T., W. J. Shack, and 0. K. Chopra. 1989. "Irradiation Embrittlement Investigation of the Shippingport Neutron Shield Tank." In Proceedings of the 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 6-10, 1989, Jekyll

Island, Georgia. (in press).

Shack, W. J., 0. K. Chopra, and H. M. Chung. 1989. "Shippingport Aging

Studies - Results and Plans." In Proceedings of the Sixteenth Water Reactor

Safety Information Meeting, NUREG/CP-0097, vol. 3, ed. A. J. Weiss, pp. 547-

562, prepared by Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, D.C.

Steele, R. Jr., J. G. Arendts, and G. H. Weidenhamer. 1988. "Performance of

a Piping System and Equipment in the HDR Simulated Seismic Experiments." In

Proceedings of the Fifteenth Water Reactor Safety Information Meeting,

NUREG/CP-0091, vol. 3, ed. A. J. Weiss, pp. 389-413, prepared by Brookhaven

National Laboratory for the U.S. Nuclear Regulatory Commission, Washington,

D.C.

Steele, R. Jr., P. E. MacDonald, and J. G. Arendts. 1987. "Dynamic Load

Effects on Gate Valve Operability." In Proceedings of the Fourteenth Water

Reactor Safety Information Meeting, NUREG/CP-0082, vol. 3, ed. A. J. Weiss,

pp. 31-40, prepared by Brookhaven National Laboratory for the U.S. Nuclear

Regulatory Commission, Washington, D.C.

E.3
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12.89) !Assloned by NRC. Add Vol.. Supp., Rev..

NRCM 1102, end Addendum Numbers. If any.)

3201.3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG-1144

2. TITLE AND SUBTITLE Rev. 2

. Rev~~~~~~~~~~~~~~~~~~~~~. DAERPR2ULSE

Nuclear Plant Aging Research (NPAR) Program Plan 3. DATE REPORT PUBLIShED

Status and Accomplishments 4, FIN OR GRANT NUMBER

5. AUTHORMSI 6. TYPE OF REPORT

Research Program Plan

7. PERIOD COVERED IncluslveDaresl

name and mailIng address.)

Division of Engineering

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

9. SPONSO R I NG ORGAN IZATI ON - NAM E AND ADDR ESS fit NRC, type "Same as above: lf contractor, provide NRC Division, Office or Region U & Nuclerv Regulatory Commission,

and mailing nddressl

Same as 8, above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or lekss

A comprehensive Nuclear Plant Aging Research (NPAR) Program was Implemented by the U.S. NRC Office of Nuclear Regulatory

Research In 1985 to identify and resolve technical safety Issues related to the aging of systems, structures, and components In

operating nuclear power plants. This Is Revision 2 to the NPAR Program Plan. This plan defines the goals of the program, the current

status of research, and summarizes ufflization of the research results In the regulatory process. The plan also describes major

milestones and schedules for coordinating research within the agency and with organizations and Institutions outside the agency, both

domestic and foreign.

Currently, the NPAR Program comprises seven major areas: 1) hardware-oriented engineering research Involving components and

structures; 2) system-oriented aging Interaction studies; 3) development of technical bases for license renewal rulemaking;

4) determining risk significance of aging phenomena; 5) development of technical bases for resolving generic safety Issues;

6) recommendations for field Inspection and maintenance addressing aging concems; 7) and residual lifetime evaluations of major

LWR components and structures. The NPAR technical database comprises approximately 100 NUREG/CR reports by June 1991, plus

numerous published papers and proceedings that offer regulators and Industry important Insights to aging characteristics and aging

management of safety-related equipment. Regulatory applicatons Include revisions to and development of regulatory guides and

technical specifications; support to resolve generic safety Issues; development of codes and standards; evaluation of diagnostic

techniques (e.g., for cables and valves); and technical support for development of the license renewal rule.

12. KEY WORDS/DESCR!PTORS (List words or phrases that will assIst researchers In locating the report.) 13. AVAILABILITY STATEMENT

Aging, age-related degradation License renewal Unlimited

Component degradation Nuclear system, component 14.SECURITYCLASSIFICATION

Failure mode, cause, mechanism Residual lifetime Irhispagel

Defect characterization Reliability Unclassified

Inspection and condition IThis Reporr)

monitoring, qualification Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 12-891
THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER
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