A-3270-12-86 NUCLEAR SAFETY LIBRMRY

Technical Report

AGING AND LIFE EXTENSION ASSESSMENT PROGRAM (ALEAP)

SYSTEMS LEVEL PLAN

Contributors: Ralph Fullwood

James Higgins

Mano Subudhi

John Taylor

Brookhaven National Laboratory

Upton, New York 11973

:',- -. I ' . c.: ... .'; . .1 4:0 r. I- ;; I :-,. a . . I

December 1986

Prepared for

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC Project Manager: S. K. Aggarwal

Under Contract No. DE-AC02-76CH00016

NRC FIN A-3270

d, ! - !eziv-b ' fv§,t'Al- i>?f
-iiiCONTENTS

Page

TABLES AND FIGURES ..... of.he........... ........................ iv

FORWARD ............................... 3t f o............. ........... v

ACKNO2.1DGoalsT ................................. 9

2UMA Y.2. Obj ctive.. .. ...................... 1

1. INTRODUCTION&OFRESEARCHRESULT................................ 23

1.1 Statement of the Pults..*.... ...... 2 3

1.2 Definitions of Au... 3

1.3 Scope of Aging and the Need for Life Extension .... 25 4

2. GOALS AND OBJECTIVES ............. .2....... 9

2.1 Goas..........l.. .... s 9

2.2 ObjectivesB....... i......a.......i o .. ......... ............ 9

3. STRATEGY ................ ..... .... ..................... ........ 12

4. MAJOR TASS........... ............. 16

4.1 Phase I Tasyste Sre... . ..... O-* ....... 16

4.2 Phase II Tass..........sk............. 21

5. UTILIZATION OF RESEARCH RESLS S............TS..... 23

5.1 Phase I and II Reslse.................... 23

5.2 Interim Resls .. .................. 23

6. RELATIONSHIP TO ONGOING WOR .................. 25

7. REFERENCES ................................. #................... 29

APPENDIX A - Aging Questions and Ansesw 0......... e rs... 33

APPENDIX B - Preliminary Prioritization of Indian Point 3 ...... 35

CCW System

APPENDIX C - Sample System Survey ................... 40
-ivFIGURES

Figure No.

1-1

3-1

3-2

3-3

Title Page

Nuclear Power Date Plant License Expirations ....... ... 7

Overall Strategy... ........ .*.....0 11

Typical Schedule for a NPP System Evaluation .......o. 13

Detailed Task Structure of System Level Plan........ 14

TABLES

Table No. Title Pag e

1-1

6-1

Stresses on Componets in a Nuclear Power Plant..... 5

Interfaces with Other NRC Projects .................. 26
FOREWORD

This plan for aging assessment at the systems level has been prepared in

a generic fashion such that it can be used by any of the NRC aging program

participants, if so desired. It will be used by Brookhaven National Laboratory for their FY 1987 aging assessment of the following two systems: 1) Component Cooling Water, and 2) Residual Heat Removal (RHR) low pressure emergency core cooling.

The follow-on work, referred to as Phase II, is not authorized by the NRC

at this time. However, it is a logical extension of the Phase I tasks, and is

in accordance with the goals and objectives of the NPAR program plan - NUREG1144.

To avoid duplication of effort, and to maintain the high quality output

of technical work, it is essential that close coordination be maintained among

all of the NPAR national laboratory participants. This is particularly important in the areas of failure data analysis and PRA modeling.

Finally, a discussion of life extension issues has been included. This

was done because an understanding of the aging process is essential in any

plant life extension analysis, and conversely, an awareness of the ultimate

utilization of NPAR program outputs (e.g., for life extension) will be helpful

in shaping the tasks that are presently at hand.
-viiACKNOWLEDGEMENT

The authors wish to thank our NRC technical monitor Satish Aggarwal, and

the previous monitor Jit Vora, for their review and input.

The guidance of Robert Hall, and our colleagues in the Department of

Nuclear Energy is appreciated, as is the excellent manuscript preparation by

Ann Fort.
EXECUTIVE SUMMARY

This Systems Level Program Plan for ALEAP presents and explains the BNL

structured approach to assessing the effects of the aging of nuclear power

plant (NPP) components and systems on the safe operation of NPPs and the extension of plant operation beyond the originally planned plant life. Note

that this plan is prepared in a generic fashion and could be used by anyone

for a systems assessment.

The plan discusses the criteria for the prioritization of plant, system,

and component selection for analysis to determine the effects of aging. The

use of Failure Modes and Effects Analysis in conjunction with the results of

natural and accelerated aging tests are discussed as means to understanding

and predicting the phenomena. The effects of aging on the failure rates of

components will be determined principally from plant data with physical and

phenomenological models used for interpolation of areas not included in the

data base. These results will be integrated with a plant risk model to be

used in addressing the question regarding "how old is old enough." I

The NRC Nuclear Plant Aging Research (NPAR) program has completed several

component level aging assessments which include the identification of dominant

component failure modes based on plant operating experience. The studies provide recommendations for monitoring, as well as mitigating, age-related component degradations.

Utilizing results from the component level studies, and work performed by'

other NRC contractors for systems data assessment and systems level risk analysis, this program will evaluate the impact of component failures on plant

system performance. The study will perform in-depth systems level failure

data reviews, reviews of current industry practices for system maintenance,

testing and operation, and probabilistic risk assessment (PRA) techniques to

understand and to predict the impact of aging on system availability. Recommendations for improving the system performance by means of degradation monitoring and timely preventive and corrective maintenance will be addressed.

This project will integrate its products with the BNL programs for Operational

Safety Reliability Research and Performance Indicators.

The first phase of this research effort will concentrate on understanding

various system designs from plant safety analysis reports, evaluating failure

data from plant operating experience data bases, applying PRA analyses, assessing industry wide surveillance and maintenance practices, and identifying

system functional indicators which are used to monitor the rate of system degradation resulting from aging and service wear. The program will separate

failures on demand from time-dependent failures. It will categorize separately, age-related failures from random and design type failures. It will produce results useful for resolution of pertinent unresolved safety issues and

for review and inspection of operating NPPs. The second phase, if authorized

and performed, will provide recommendations for improving the system performance through enhanced maintenance practices and reliability monitoring which
-2-

will be focussed on the most risk sensitive areas of a system. Recommendations will be made for improvements in pertinent Regulatory Guides, Industry

Standards, etc. This program plan delineates the goals and major tasks to be

completed in each phase. The current version of the program plan is considered

to be a draft and will be revised and updated as the first few systems are

* completed using this methodology. This will produce a final proven methodology, which can be used for all remaining systems.
-3-

1. INTRODUCTION

1.1 Statement of the Problem

As Nuclear Power Plants (NPP) age, the likelihood of common cause failure

due to age-related degradation increases. As a result of the aging of components in a nuclear power plant steps must be taken to: a) assure that the

level of safety on which a plant was originally licensed has not degraded below an acceptable level, b) identify modifications, procedures, and maintenance that will arrest or suppress aging effects and restore the reliability

to that on which the original license was based, c) recommend incipient failure detection methods to provide early warning of impending failure, and d)

determine those factors that must be implemented or evaluated for consideration of NPP life-extension requests.

The Technical Integration Review Group for Aging and Life Extension

(TIRGALEX) defines an effective NRC program presenting a structured approach

to integrating aging research and regulation. This plan indicates the closecoupling between aging and life extension. The ALEAP plan will be updated as

necessary to agree with the TIRGALEX goals.

1.2 Definitions of Aging

To clarify th3e 6ALEAP scope and approach, a definition of aging is needed.

NPAR (NUREG-1144) defines aging as the "cumulative degradation occurring

within a component, structure or system which, if unchecked, may result in

loss of function and impairment of safety." Factors causing aging/degradation

may include:

natural internal chemical or physical processes,

* external stresses and environment,

service wear (cycling, vibrations),

* testing, and

improper installation, application, and maintenance.

Note the emphasis is on actual physical changes in the properties and

performance of the plant and equipment.

Backfits and design changes may introduce new aging mechanisms or special

aging concerns, or unforeseen common cause and system interaction problems.

An example might be the installation of added electronic equipment in a

controlled air conditioned environment resulting in an additional heat load

that compromises the temperature control. Similarly, replacement of a pump

with one of larger capacity in an auxiliary system may lead So degradation of

interrelated piping components due to water hammer or erosion .

Life extension is defined as a set of actions and activities aimed at

increasing the useful lifetime of a plant or of specific equipment beyond the

time originally envisioned. This would be partially based on developed Aging

Technology.
-4-

The pragmatic concept of aging used in ALEAP is that aging is a process

that causes the failure rate of equipment to increase with time. This distinguishes between the "new" performance of equipment when the failures are random (Poisson process uncorrelated with time) and the non-Poisson aging process

where failures increase with time. Thus this investigation involves separating age-related phenomena from "like new" failure phenomena, which is taken as

the base case. This requires determining the aging fraction of the failure

rate as well as the rate of increase. Such information is sought on all

levels: plant, safety function, system and component but most of the information is currently available at only the component level as a result of NPAR

and data base activities. However, given the age-dependence of components,

the dependence of the higher structures formed of the components may be found

through PRA system models.

Further analysis of failures at the component level does not necessarily

get one to the root cause of failure because components may fail from various

causes as they are operated upon by operating and environmental stresses. An

understanding of these root cause aging failure mechanisms provides the key to

modeling the rate of increase of the aging phenomena as well as the necessary

information for mitigating the aging effect. NPAR studies haves identified

stresses for each component reviewed. Table 1-1 taken from Drago , also correlates component types with the stresses that singly or collectively may lead

to failure.

These stresses suggest a connection between the environmental qualification of equipment (EQ) as defined in IEEE-Std-323 for equipment in general and

other guidance provided in IEEE-Std-382 for valves, IEEE-Std-334 for motors,

and IEEE-Std-317 for electrical penetrations as well as Regulatory Guides

1.40, 1.63, 1.73, and 1.131. The connection is that environmental stresses

provide many of the degrading mechanisms that cause aging. Process upsets

such as addressed in EQ may result in rapid aging for a short time period

while normal aging results from operation in a nominal environment for a much

longer time but cumulative effects of the normal environment may be much

greater than the upset effects.

1.3 Scope of Aging and the Need for Life Extension

tORNL ideent~ifies-about -17%"of the abnormal operating events reported to

the -NRC-as wh'iving'age-related causes. - AtoutT8Zof these-events-resulted from

insemiinnt -"drift" -of the setpoint or calibration outside of the technical

specifications. _The other-nine percent were-attributed to aging-causes such,

as wear, corrosion, oxidation, crud deposition and fatigue. Ofthe-components

that-failed,(9% of above. events),due toaging,-20% were valves,;-14%-pumps,-5% w

diesel generators, -3%-steam generator tubes, -3% heat -exchangers .and less -than

1% each forabout 120 other. components. - It should be noted that these data

are not normalized to the number of components at risk, but represent the

population distribution that would be observed in examining plant data.
-5-

Table.1-1- Stresses-on-Comiponents in a Nuclear Power Plant

- ~ 1 Ic.~-I . -

Component Types Stressess

I

i

1;

I

kI

I

1,

t

1.I

i

III

I

iI

i

L

II

Accumulators, tanks

Air dryers

Annunicator modules

Batteries

Blowers, fans, compressors

Battery chargers

Circuit breakers, motor starters, fuses

Control rods

Control rod drive mechanisms

Demineralizers

Electric connectors (cable, bus, wires)

Internal combustion engines

Filters, strainers, screens

Fuel elements

Generators, inverters

Electric heaters

Lifting devices (cranes,hoists,jacks)

Heat exchangers (coolers, heaters, steam

generators, evaporators)

Instruments, controls, sensors

Mechanical function units (gear boxes)

Motors (electric, hydraulic, pneumatic)

Penetrations, air locks, hatches

Pipes, fittings

Pumps

Recombiners

Relays

Shocks suppressors and supports

Switchgear, load control centers, motor

control centers, panel boards

Transformers

Valves

Pressure vessels (reactor vessels,

pressurizers)

T,M,C,H

TM

E,H

E,C,H

M,E,V

E,M

E,H

M,R,V

M,R,W,V

C

E,R,C

M, T

W

R,V,T,M,C

E ,M,V

T,E

M,T,R

T,M,C,R,H

E,H,M

H

E,M

H

C,M,V

M,V

C

E,H

M

II

i.i

t

i

II

I

i

I

I.II

t

t

II

I

i

I

4 i,

i

i

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E

1

1:

m,c

R,T,M,C .

*Stress codes - thermal (T), mechanical (M), radiation (R), humidity (H), wear

(W), electrical (E), vibration (V), chemical reactions (C). Additional

stresses beyond these are also possible.
-6-

The data base used in the ORNL study was limited but more extensive information is available from individual plant and industry-wide data collections. Continuing support of industry-wide data collections providing the information for determining the effects of aging must be encouraged. However,

in data analysis there is the tradeoff between specificity and statistical

accuracy. To aid in overcoming these data deficiencies, it may be possible to

use physical and phenomenological models for extrapolating data.

To gain a perspective regarding potential requests for life extension,

the paper by Marnn indicates that the first license expiration will occur in

1997 and between 2005 and 2010 about 50 GWe or about 75% of the 1983 generating capacity will expire. Figure 1-1 presents a plot from this paper showing

the cumulative expirations. Some safety systems and support systems may not

be issues in life extension as they may be reparied, replaced, etc.

A7overall document on aging mechanisms and aging rates is Carfagno and

Gibson . This document provides a valuable review and compilation of physical

aging theories and scaling parameters. It also presents phenomenological

scaling rules such as those in MIL-HDBK-217.8 More specialized work addressing

the agin 0of metals fr presented by Simonen on embrittlement, by Vignes and

by Sanoh . Hinton provides work on the life extension of piping systems

and Moelling and Gallucci on probabilistic analysis of stress corrosion

cracking in BWRs. 3 Similar aging effects in pipe cracking are provided by

Gordon and Gordon on the basic properties of types 304 and15316 stainless

steel by Horak , and on main feedwater spray heads by Spond . Additiona

work on the irradiation aging of pressure vessels is pr f9 ented in Odette

aging of piping by Banford , and fatigue aging by Server . More comprehensive work on the aging and Tfrvice wear effects on hydraulic and mechanical

snubbers is reported by Bush 20 Wear measurements of nuclear power plant components is reported by Duframe

Vause21 reports on the operating experience relating to the aging of diesel generators. Similar work is reported by Dingee and Johnson2 , by Higgins

and Subudhi , and by Vesely and DeIoss . The aging effects on electric

motors has been studied by S2zbudhi , and recyfmended maintenance practices

for life extension by Subudhi . Taylor, et al presents the results of Si -

ulated seismic testing of naturally aged small electric motors. Subudhi

also presents compiled operating experience and an aging-seismic assessment of

electric motors. Thp9 results of a correcjtion study on Class 1E equipment is

reported by Sugarman . Bonzon and Hente present their work on seismic fragility tests of Class 1E battery c511s. The aging and service wear of check

valves is described by Greenstreet . Subudhi and Taylor report their work

on reactor coolant pump seals.

On the electrical side, Toman33 and his associates, present their work on

the interactive effects of relay and circuit breaker aging. Gunther indicates the results of work on operaSng experience and aging-seismic of battery

chargers and inverters. Stuetzer is a status report on electrical cable

failures due to aging and service wear.
140

C

0

-a

w

-J

z

120 -

1'z00_

80

60

40 -

20 _

0

58 ,

I _

2000 2005 2010 2015 2020

Figure 1-1 Nuclear Power Date Plant License Expirations
-8-

Much of 36the previously described work was performed under the NPAR

(USNRC, 1985 ) program. This work will be used in ALEAP for identifying

failure modes and mechanisms of component failures. This will be related to

the systems they are contained in, in order to develop system level insights.

Fullwood37 provides a review of incipient failure detection methods.

This report is aimed at reliability improvement but these methods are equally

useful for the detection of aging. Sliter and Cgfy38 describes EPRI wirk to

this end as do the papers by Plumstead and Cady ; Engh and Figlhubur4 ; and 41 by Weber

A general review of aged power plant facilities is presented in Rose 2

In Leverenz importance measures ag presented for aged components. A report

closely related to ALEAP is Vesely 4 in which the rate of aging of selected

components is determined from plant failure experience.
-9-

2. GOALS AND OBJECTIVES

2.1 Goals

Commensurate with the NRC Nuclear Plant Aging Research (NPAR) program

plan, the ALEAP system level plan has the following top level goals:

I. To assess aging impact on system, plant safety, and risk,

II. To develop recommendations to mitigate aging effects to assure

public health and safety consistent with plant optimum performance,

and

III. To provide initial technical basis for evaluating plant life extension and support regulatory actions.

2.2 Objectives

To achieve the above goals, the scope of the system study is divided into

two distinct phases. The objectives of each phase are given as:

Phase I Objectives

1. To identify and characterize the aging impacts on system performance and

hence plant safety and risk.

2. To produce interim aging and system related outputs in a form useful for

NRR, I&E, and the NRC regions. (See paragraph 3.2.)

3. To address in an interim fashion generic issues related to the systems

under study.

4. To assess current inspection, surveillance, and monitoring programs for

systems.

5. To assess current maintenance, storage, and mothballing programs for systems and components.

Phase II Objectives

1. To support regulatory actions, as necessary for NRR.

2. To aid in plant life extension decisions.

3. To produce aging and system related outputs in a form useful for NRR,

I&E, and the NRC regions.

4. To develop appropriate recommendations to improve inspection surveillance, and monitoring programs.
-10-

5. To develop recommendations to improve maintenance, storage, and mothballing programs.

6. To finalize applications to generic issues related to systems under

study.
I

SYSTEM

MODELS

SPECIFIC

PLANT

PRA

MODEL I_

SYSTEM

FAILURE

DATA .4.~~~~~~~~~

SPECIFIC

PLANT

PHYSICAL

MODELS

PRA MODEL(S)

WITIH

AGING

EFFECTS

PRA

IMPORTANCE

MEASURE

STUDY .

-

RECOMMENDATIONS

AND

APPLICATIONS

I

I.--

I

3

AGING

IMPACT

ON

SYSTEM

CURRENT

PRACTICES

ASSESSMENTS

Figure 3-1 Overall Strategy
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3. STRATEGY

The ALEAP Systems Level Plan is structured in a two phase approach, with

the first phase characterizing the aging effects on system performance and the

second phase developing mitigating actions for arresting these adverse effects, prior to system degradation. For the selected system, the phase I

study consists of three major tasks: 1) a review of the various systems

designs in US nuclear power plants to determine the general applicability of

aging analyses performed on a specific plant to other plants, 2) modification

and adaptation of existing PRAs such as the NUREG-1150, IREP, RSSMAP, ASEP,

and industry-performed to include aging effects, and 3) perform a detailed

review of the several data bases and the NPAR results to obtain the primary

failure modes, causes, and mechanisms. To perform the multiple computations

that will be necessary in the study of aging effects and to calculate the component importances, it is advantageous to have the complete PRA implemented on

a computer. An example is the BNL NSPKTR code which models Indian Point and

is one of the reasons for choosing this plant for the demonstration. Another

example is the PRISIM code that performs calculations on the IREP model of

ANO1 which has been used for aging studies of the Auxiliary Feedwater System

by INEL. As indicated in Figure 3-1, the failure data analysis results will

be fed into both probabilistic and actual system models to characterize the

aging effects in the system performance. Once the system behavior is determined, the Phase II study will review and analyze current industry practices

for monitoring system performance and mitigating aging effects. Recommendations will be developed for appropriate system monitoring and mitigation techniques in order to improve the system reliability and alleviate aging. A

generic schedule for a typical system based on Figure 3-1 is included as

Figure 3-2.

In order to achieve the defined goals and objectives the above strategy

will be implemented for two systems (CCW and RHR) in FY 87.

Figure 3-3 illustrates all tasks to be carried out in each phase of the

program plan. The tasks to be completed and which are included in Figure 3-3,

are listed below for further clarification:
-13-

PHASE I

* Specific Plant Physical

Models

* Specific Plant PRA

Model

* System Failure Data

* PRA Model(s) with Aging

Effect

* Aging Impact on System

"IB"

* Current Practices

PHASE II

* Practices Recommendations

PRA Importance Measure

Study

* PHASE II Report

FY 87 FY 88

O N |D|J|FM A _J JM AIS O1NID JF MIAIM JIJ A S

O lu

x ALx

xx. xx: x :x :x x

x :x: x: :xxxxIi

Figure 3-2 Typical Schedule for a NPP System Evaluation
-14-

APPLY IDENTIFY EVALUATE

PRA SYSTEM STRESS ACTUAL SYST

MODEL ( S ) PARAMETERS MODELS

DEVELOP AGED FAILURE PRA --- DATA,

MODEL ANALYSIS [

PLANT LEVEL PERFORM SYSTEM

RISK/AGING LEVEL AGING

ASSESSMENT ASSESSMENT

EVALUATE EVALUATE EVALUATE

CURRENT MAINTENANCE TESTING & 1

PHASE I REGULATIONS PROGRAMS MONITORING

+ L_ l _ ~~~~~~~~PROGRAMSL

PHASE II

STUDY IMPACT ASSESS CURRENT PRACTICES

OF COMPONENT TO MITIGATE SYSTEM

FAILURE ON AGING FAILURES

SYSTEM AVAIL

| RECOMMEND TEHIQE I MAINTENANCE IIAPPLICABLE TO I

I PRACTICES I PLANT LIFE EXT.I

Figure 3-3 Detailed Task Structure of System Level Plan
-15-

Phase I Tasks

Phase IA

1) System Definition

2) PRA System Model Application

3) Operating and Environmental Parameters Identification

4) System Analysis

5) Aged PRA Model Development

6) Operating Experience (Failure) Data Analysis

7) System Analysis at Selected Plant

8) System Level Aging Assessment

9) Plant Level Risk Aging Assessment

Phase IB

1) Current Regulations and Guidance Assessment

2) Current Maintenance Practices Assessment

3) Assessment of Current Techniques for Testing and Monitoring

4) System Operating Procedure Evaluation

5) Integration of Tasks 1) Through 4)

Phase II Tasks

1) Plant Risk Assessment

2) Recommended Practices

3) Utilization for Plant Life Extension
-16-

4. MAJOR TASKS

This section provides a detailed description of each task. The discussion contains the objectives of the task, input information needed in order to

perform the task, and the product of the task. The interrelation among various tasks is illustrated in Figures 3-1 and 3-2. These tasks will be performed for each UIPP system. The first two systems to be analyzed by BNL will

be the PWR Component Cooling Water System and the BWR Residual Heat Removal

System. Upon completion of two systems studies, the draft program plan will

be revised and updated to incorporate lessons learned.

4.1 Phase I Tasks

4.1.1 Phase IA Tasks

Task (1): System Definition

The boundaries and the interfaces of the selected system with other plant

systems must be clearly defined in order to perform a complete aging assessment. All components and sub-systems that will be studied within each system

will be identified. Schematic diagrams showing the interfaces with other

mechanical, electrical, instrumentation, and control systems will be developed. Structures supporting components within the system will also be discussed. Assessments in other tasks relating to system performance will be limited to those components within the system boundary. Of particular interest are

the way components interact within the system. Design implications at the

system interfaces will be discussed, both at the load side and the input or

support system side. The interaction between systems at these interfaces will

be studied in the subsequent tasks.

A representative plant will be selected for each in-depth systems analysis which will include: review of all design information, review of operating

and maintenance procedures and practices, use of plant specific PRA model, and

review of all actual plant failure data. The plant selected should be at

least 10 years old, have a full scope PRA, be reasonably close to BNL, and be

operated by a utility willing to cooperate with the study and share information.

Task (2): PMA Systen Model Development

New PRAs will not be developed in ALEAP but existing ones such as the

recently completed models reported in NUREG-1150, utility performed or sponsored PRAs, WASH-1400, RSSMAP, IREP and/or the ASEP models. Generally speaking, all of these require adaptation to the aging analysis to permit the

investigation of the aging-caused change in the failure rates and to determine

the importance of the various components. Some PRAs have included importance

measure calculations but some of these importance measures are not suitable

for aging investigations because the importance measure is not affected by the

age of the component being investigated. Extensive work has been done in this

area by Vesely for NPAR and will be utilized as a starting point for further

work in this program. Calculating the effects of aging and importance calculations, requires repeated. calculations of the PRA which suggests a complete
-17-

computer implementation of the PRA. With some exceptions, PRAs have consisted

of multiple separate calculations that are pieced together to obtain a final

result for the base case. The exceptions are the BNL PC-code NSPKTR that implements the Indian Point PRA and the PRISIM code, operable on a special PC

that implements the IREP model of ANO. The availability of the NSPKTR code

in conjunction with past experience with Indian Point systems, are the reasons

for suggesting the use of the CCW at Indian Point for a demonstration. The

selection of a plant for the RHR demonstration analysis has not been done but

a plant having a PRA that facilitates repeated calculations will be a consideration.

Because of plant and system complexity, it will not be possible to apply

aging analysis to all components. A group of components will be selected on

the basis of their importance to safety and on the importance of aging effects

on their reliability. There are many measures of importance but it seems that

the measure "Inspection Importance" previously used in prioritizing NRC inspections is most suitable because it is weighted both by the effect on plant

risk if a given component fails and the probability of the component failing.

This carries the tacit assumption that components with a high failure rate

also have a high aging rate. This is used as a first order selection criterion on which to iterate as the results of aging analysis is incorporated. If

this assumption is not completely correct, it will be modified in the reanalysis. In summary, the PRA is used to initially calculate the Inspection

Importance of the systems and components. This leads to the systems and their

components on which to focus the aging analysis. In the case of Indian Point,

the NSPKTR code has preliminarily determined the importance of the systems and

shown that the CCW is one of the most importance systems. Subsequent work

shown in Appendix B calculates the non-aged importance of the components

making up this system. This is followed by an Aging Failure Modes and Effects

Analysis (AFMEA - See Table B-5) which is used in conjunction with NPAR work

to identify the aging mechanisms which when used with field data or aging

phenomenological models provides the age dependence of the component failure

rates. These results are fed back into the computerized PRA model for a

second iteration so that now the importance measures as well as the reliability and risk assessment contain the effects of aging.

Task (3): Operating and Environmental Parameters Identification

For the aging assessment it is imperative that system physical parameters

such as temperature, pressure, humidity, radiation, mechanical and electrical

stresses, that affect system performance are identified for both the component

and the system level. Typical mechanisms which cause component and system degradations include fatigue cycles (thermal, mechanical, or electrical) wear,

corrosion, embrittlement, diffusion, chemical reactions, cracking or fracture,

and other overstress mechanisms. This task will analyze all operating modes

of the system under normal, abnormal, accident, and post-accident conditions

including plant mechanical and electrical transients which contribute significantly to the aging process.

In addition to the operating parameters, environmental conditions are

equally important for component degradation. Since the system level analysis

will include components both inside and outside the containment, and during

-y
-18-

normal and accident situations, parts of the system will experience different

environmental conditions. Sometimes atmospheric conditions due to plant location require additional analysis for assessing the system failures.

The source of this information will be taken from the plant FSAR, PRA

studies, and other plant specific design drawings. Each piece of equipment

within the system boundary will .be analyzed for the internal and external conditions to assess the aging deterioration of its subcomponents. The output of

this task will be the aging characteristic of each component when subjected to

its particular operating conditions. NPAR studies on systems and components

will be used for the final assessment.

Task (4): System Analysis

This task will review, in detail, the design and specifications of the

system under study. Included will be system function, components, and instrumentation. For a support system (such as Component Cooling Water) where there

are significant variations between plants, the various system designs must be

catalogued and understood. See Appendix C for a sample system survey. The

effect of system failure on supplied loads and on overall plant safety/risk

must also be taken into consideration. Thereby when the failure data is analyzed both the system design variations and consequences of component failure

and degraded system operation will be appropriately treated. Relationship of

the system and system problems to pertinent unresolved safety issues must be

defined at this stage for later resolution.

The required performance of the system in the various postulated accidents and transients must be clearly understood. Potential failure modes in

these scenarios must be considered carefully since they may not be adequately

represented in the failure data bases.

Task (5): Aged PRA Model Development

Upon completion of the AFMEA which includes the NPAR work, environmental,

service and other causes of aging, it is necessary to quantify the aging effect. This will be done by: a) determining the aging effects that may be observed in field data, or b) physical and/or phenomenological models of the

aging process. This work will be closely coordinated with work at other

laboratories to avoid duplication. Especially valuable should be the root

cause analyses and the investigations of aging dependence that may be obtained

from generic data at INEL. These data as well as data developed in this work

will be used to determine the contributions of each of the aging processes.

These aging processes will be incorporated with the non-aging processes

to provide aging failure rates. These will then be used in the PRA models to

determine if the prioritization changes when the aging effects are included in

the model. If so, new importances will be calculated that will result in different prioritization. Then the aging FMFA, will be repeated and new agedependent failure rates calculated. When the age-dependent failure rates are

determined to be valid, the plant risk and system availabilities will be recalculated to exhibit the aging effects. If the plant risk is increased above

the safety goals or other criteria, the PRA will be re-examined to determine

steps for arresting the aging effect.
-19-

Task (6): Operating Experience (Failure) Data Analysis

System and component level failure data bases from LER, NPRDS, IPRDS,

NPE, completed studies by INEL or ORNL, and other sources will be obtained and

evaluated for identifying all failure modes, causes, and mechanisms of components contributing to system failure. This labor-intensive effort will be

used to identify critical components, dominant failure sequences, failure

mitigation processes, and other relevant information available in the data

bases.

Results from this analysis will be fed into the PRA models, as well as to

the specific plant models to evaluate the overall system performance and its

effect on plant risk. Interim results useful for NRR and I&E systems analysis

and inspection will be produced at this stage.

Task (7): System Analysis at Selected Plant

This task will consist of a very detailed analysis of the system under

study at a selected, representative nuclear power plant (NPP), plus shorter

reviews at one or more than NPPs. BNL will develop close working relationships with one or two local NPP utilities in order to exchange information,

and to further the research in the areas under study. Specifically, this

analysis will consist of a review of system design, failure data, maintenance

records, system testing and operation, and procedures. This will allow BNL to

understand how the systems are actually operated and maintained and will correct for deficiencies in the various data bases used in the failure analysis.

This task will also allow BNL to learn developing problems in systems under

study, and what actions utilities may be taking to ensure proper system operation. Input from actual current plant experience is vital to any study of

this nature and this task will provide the needed information. Some information gathered, particularly on maintenance, surveillance, and condition monitoring will be directly used in Phase II of the project.

Task (8): System Level Aging Assessment

-Both PRA and plant design models in conjunction with the failure data

evaluation will be utilized to assess overall system level aging deterioration. This should provide a prioritized list of components within the system,

which may require indepth engineering analysis and better monitoring

programs. Critical components based on system unavailability, will be identified. The impact of interface system or components on the subject system's

performance will be discussed. This assessment will include all associated

categories under mechanical, structural, electrical, instrumentation and control components which contribute to the system failure. Functional indicators

for monitoring system degradation will be established.

Task (9): Plant Level Risk Assessment

Using the PC-based plant PRA model, the impact of system unavailability

is easily generalized to overall plant risk for assessment using both core

melt frequency and offsite consequences as measures of risk. Coordination

will be maintained with Vesely-and INEL. The component prioritization based
-20-

on plant risk or core melt frequency, will be obtained considering age of the

components as a factor. As the plant ages, the component prioritization may

change because of the age-related degradation of the equipment. Plant level

performance indicators will be developed to monitor the system health as the

plant ages. Consideration must be given to generalizing the conclusions

developed here (with the PRA model) to other NPPs, that do not have PRAs.

4.1.2 Phase IB Tasks

Task (1): Current Regulations and Guidance Assessment

This task will evaluate the current status of existing regulatory requirements and industry guidance (including IEEE and ASME Standards) related

to the system under study. Included are standards, guides, and NRC regulatory

and inspection procedures relating to the subject system. Other NRC related

activities such as the maintenance and surveillance programs including plant

tech spec requirements, inspection and enforcement activities, plant audit

reports, and I&E bulletins and information notices will be evaluated for

improvements to mitigate system failures identified in the Phase I study.

In addition to NRC activities, different industry and engineering society

activities and standards such as ASME, ASH, IEEE, INPO, EPRI, etc. will be

searched to evaluate the system monitoring techniques as well as testing programs.

Task (2): Current Maintenance Practices Assessment

An industry survey will be conducted to assimilate various plant maintenance practices and procedures in relation to the subject system. The work

performed in Task 7 of Phase I and the USNRC Maintenance Program will be valuable here. Both corrective and preventive/predictive maintenance programs

will be reviewed. Other relevant programs to be reviewed include the plant

maintenance management program, human reliability, training, QA/QC, and spare

parts. Based on the review of the above, the advantages and deficiencies in

the current industry practices will be assessed.

Task (3): Assessment of System Testing and Condition Monitoring

In this task BNL will review and analyze current NPP practices for each

selected system in the area of system testing and condition monitoring. This

will concentrate on integrated system functional testing but must also include

testing of key components. Included in the review will be preoperational

testing, inservice testing, surveillance testing, condition monitoring, inspection practices, and training.

Task (4): System Operation Evaluation

This task will collect and assess NPP operating techniques and procedures

associated with the selected system. Included in the review will be normal

and emergency operating procedure, actual operational methods, and associated

training and qualification of the plant staff. For a normally operating system, the method of system operation can have a significant effect on how the

system ages and how it will perform under the stress of an accident situation.
-21-

NPAR component aging studies revealed that frequent starts and stops,

sudden valve closures, and many other operational related activities accelerate the aging degradations of both mechanical and electrical components in a

system. For example, too much switching causes electrical surges and sparks

on the contact surfaces, quick starts on motors overheat insulation causing

accelerated aging of the insulating polymers, and thermal transients could increase the potential for crack growth in piping leading to eventual pressure

boundary failures.Therefore, this task will review various plant component

start up and operating procedures for the subject system.

The product of this task will be a list of the current, system specific,

operating practices, their impact on equipment aging and recommendations for

future practices.

Task (5): Integration of Tasks 1-4

Tasks 1-4 products will be integrated to establish the present state-ofthe art in regulation and guidance, plant inspection, surveillance, monitoring, and maintenance programs, and system operating procedures. This section

will discuss each of the above activities and develop a matrix to illustrate

various procedures. The task will identify all high and low points of each

program and their suitability for the system operational readiness.

4.2 Phase II Tasks

Task (1): Plant Risk Assessment

Using the PRA model developed for the specific plant, an importance

measure study will be conducted to identify the critical components based on

plant risk. The study will predict the probability of system failure as a

function of system age in systems composed of components having diverse timedependent hazard rates. The output of this study completed mainly in Phase I,

will be the point or instantaneous system unavailability and its uncertainty

as a function of time. The BNL-developed FRANTIC II code, and/or the MIT

modification in conjunction with other codes will be used to propagate the uncertainties in the component hazard rates through the system models to determine the overall uncertainty in the system unavailability.

Reliability techniques as developed in the PETS program, will be used to

optimize the AOTs and STIs for components under tech spec requirements. This

will aid in reducing unnecessarily frequent tests on equipment as required by

the present plant surveillance programs.

Task (2): Recommended Practices

This task is the culmination of all the work performed in Phase I and

Phase II studies. With the knowledge of this current industry wide age-related

system problems and the mitigation programs in effect, recommendations will be

provided to the nuclear industry, as well as the regulating agency for improving the system reliability, through enhanced maintenance practices and monitoring techniques which will be focussed on the most risk sensitive areas of a

system.
-22-

Task (3): Application to Plant Life Extension

The age-dependent plant risk analysis provides the basis for evaluating

requests for life extension. It is imperative that the plant risk be maintained at or near the risk level that was the basis for the original licensing

action. When the licensee submits a plan for life extension, ALEAP analysis

will serve as a basis for considering and evaluating how the aging effects are

managed for the specific systems under study. This may be done by equipment

replacement and/or installing incipient failure detection devices. The ALEAP

model would be re-evaluated, including these anti-aging provisions, but with

consideration that the age arresting devices and procedures may fail. After

suitable analysis if it is found that the risk is within NRC requirements,

recommendations could be made regarding the specific systems under study being

suitable for extended life.
-23-

5. UTILIZATION OF RESEARCH RESULTS

5.1 Phase I and II Results

The system level aging assessment will provide a better understanding of

the system aging characteristics under normal, as well as accident and transient conditions. The study will prioritize critical components for maintenance and monitoring activities (both with and without aging considerations).

Techniques to predict the expected life of certain equipment subcomponents

will be established to assess the operational readiness of the system. Following is the list of some uses for research results:

* To support the NRC in review, development and inspection of maintenance and surveillance programs.

* To support the NRC to monitor and inspect systems.

* To identify failure modes, causes, and mechanisms associated with a

particular system under study, and to identify the dominant modes

affecting the system availability.

* To identify system and component level functional indicators for

monitoring system performance.

* To provide recommendations for updating rules and regulations, regulatory guides, industry standards, etc.

* To provide a technical basis for assessing life extension issues for

NPPs.

* To aid in evaluating storage and "mothballing" issues.

To determine the risk associated with aging of components, systems,

and plants.

* To aid in the resolution of pertinent unresolved safety issues and

Generic Issues.

5.2 Interim Results

The following items will be produced as interim results as the program

proceeds:

Products to help resolve generic issues associated with each of the

systems under study. As an example Generic Issue 65 relates to one

of the first systems selected for study, the component cooling water

system. Also Unresolved Safety Issue A-45 "Shutdown Decay Heat Removal Requirements" relates to one of the other initial systems to

be studied at BNL, namely BWR Residual Heat Removal Systems. Generic Issue C-9, "RHR Heat Exchanger Tube Failures also related to the

RHR system.
-24-

System Inspection Guidance for I&E/Regional Offices to include system failure modes, failure causes, effects on the plant of failures,

aging and service wear effects, system functional indicators, mitigation and detection techniques, recommendations for testing, inspection, or surveillance methods, and methods to prevent or mitigate system failure.

Information on a system basis to aid NRR in licensing decisions, related to system design variations, system failure modes, and effective failure mitigation techniques.

Technical Specification insights related to LCO's, AOTs, STIs, and

required surveillance testing.

Input on system reliability insights to the Operational Safety Reliability Research Program.

Input to the Accident Sequence Evaluation Program (ASEP).
-25-

6. RELATIONSHIP TO ONGOING WORK

The nature of this investigation is such that it may effectively draw

from the work performed and being performed in many other NRC and industry

programs. Conversely its successful execution will greatly assist other programs. The primary interactions are in the nature of root-cause determination, time dependence of failure rates, PRA plant modeling and completeness

thereof, value-impact decisions relating safety and operability, plant management, actions to mitigate the effects of aging, and life extension analysis.

Within BNL a seminar was held with participants from several related programs.

Pertinent questions and answers generated are included in Appendix A. During

the performance of this program, the project team will maintain contact with

the NRC program offices of Research, NRR, and I&E in order to receive input as

to how this program can serve their needs with interim products related to

work completed.

As an example of one interface, this research project must take cognizance of the work to improve technical specifications which are now generally

considered to be complex and difficult to implement and may adversely impact

safety. The Surveillance Test Intervals (STIs) and Allowed Outage Times

(AOTs) as specified by the Technical Specifications are not directly based on

risk nor do they consider the possibly increasing risk of plant operation as

the components age. Major single point passive failures (such as the pressure

vessel and to some extent, the piping) are not properly considered in PRAs but

may control the criteria of plant life extension because of the costs and difficulty of their replacement. It may be that the STIs and the AOTs will need

re-interpretation to include age considerations for possible inclusionin the

Program for Evaluating Technical Specifications (PETS). The PETS program is

designed to utilize risk-based techniques to establish a firmer basis for AOTs

and STIs while also supporting other potential technical specification needs

as indicated by TSIP. Table 6-1 summarizes interfaces with other NRC

projects.
-26-

Table 6-1 Interfaces With Other NRC Projects

Input to/from Programs

1. Probabilistic Evaluation of

Technical Specifications (PETS)

2. Nuclear Plant Aging Research (NPAR)

(of which this document is a part)

Input/Output-Applicable evaluation

models and data, age consideration in surveillance intervals and

allowed outage times.

Input-Identification of aging related parameters to be measured

and trended and applicable

monitoring techniques.

Output-Equipment reliability data,

measures of effective monitoring,

strategy for using condition monitoring for managing aging.

3. Root Causes of Component Failures

(RCCF)

- Materials

- Stresses

- Environment

Input-Identification of important

aging parameters to be monitored,

root cause failure data and alert

levels.

Output-Prioritizing root agecauses for degraded performance

and analysing corrective actions.

4 . Technical Specifications Improvement

Program (TSIP)

Input-Aging data and aging significance in tech specs.

Output-Basis for performance-based

tech specs by identifying impacts

of surveillance requirements and

allowed outage times on component

performance and measuring age

changes in reliability.

5. Maintenance & Surveillance Program

Plan (MSPP)

Input-Effective of age corrective

action techniques and maintenance

management approaches.

Output-Evaluation of age mitigating activities. Measure of maintenance & surveillance effects on

aging.
-27-

Table 6-1 (Cont'd)

Project Input to/from Programs

6. Vendor Inspection Programs

0

Input/Output-Determination of

which components and vendors are

currently experiencing problems

and a determination of which areas

of systems and components age

significantly and need inspection

attention.

Input-Identification of systems or

components that have propensity

for aging and hence, need agingdependence assessment.

7. Quality Assurance Programs

Output-Guidelines for age considerations in reliability maintenance as part of Quality Assurance.

8. Human Factors/Reliability Programs Input-Operational, test, and maintenance effectiveness, identification of human performance aids to

arrest time dependent degradation

of performance.

Output-Identification of management, test, and maintenance needs

to maintain high plant, system,

and component performance.

9. Performance Indicator Program (PI) Input-Possible indicators to use

at the system level and methods

for evaluating selected indicators

to tell when alert levels have

been reached.

Output-Possible additional items

at the system or plant level that

could add to the PI program.

10. Research to Support NRC Inspection

Prioritization

Input-Methods to prioritize safety

importance of equipment with consideration for time dependent

degradation.

Output-Suggested methods to inspect performance-based regulations for maintaining licensed

level of safety.
-28-

Table 6-1 (Cont'd)

Project Input to/from Programs

11. Operational Safety Reliability

Research (OSRR)

Input-Reliability techniques and

methods useful for both aging assessments and ongoing condition

monitoring.

Output-Items susceptible to aging

that should receive attention in

OSRR. Methods to alleviate premature aging to be implemented in

an OSRR type program.
-29-

7. REFERENCES

1. Golay, M.J. and Moinzadeh, J.H. "Extending the Life of Nuclear Power

Plants: Technical and Initiation Issues," MIT-E686-003, April 1986.

2. IEEE Std. 323-1974, "IEEE Standard for Qualifying Class 1E Equipment for

Nuclear Power Generating Stations," Institute of Electrical & Electronic

Engineers, New York, 1974.

4 3. Marriott, P.W. "Nuclear Plant Life Extension: A Comprehensive Approach to

Plant Improvements," Trans. Am. Nucl. Soc., 46, 576-577, June 1984.

4. Drago, J.P. "The In-Plant Reliability Data Base for Nuclear Power Plant

Components: Data Collection and Methodology Report," ORNL/TM-8271, July

1982.

5. ORNL "Aging Trends in Nuclear Power Plants," Oak Ridge National Laboratory Review, 16, No. 4, p. 76, 1983.

6. Marin, P.S., et al., 1983 "Reactor Plant License/Lifetime Extension," ANS

TANSAO 50, p. 328, 1985.

7. Carfagno, S.P. and Gibson, R.J., "A Review of Equipment Aging Theory and

Technology," EPRI-NP-1558, September 1980.

8. Simonen, E.P., "Predicting Irradiation Effects on Alloy Aging Kinetics,"

Proceedings of International Conference on Nuclear Power Plant Aging,

ASM, July 8-12, 1985.

9. Vegnes, A, "Understanding of the Phenomena of Materials Degradation by

Aging and Embrittlement -- Margin of Safety and Solutions to Limit or

Avoid Such Degradation", ASM op.cit pg. 165, 1985.

10. Sanoh, J. et al, 1985, "Irradiation Embrittlement Behavior of Some Japanese Reactor Vessel Steels," ASM op cit pg 169, 1985.

11. Hinton, B.M., "Plant Life Extension Program for Nuclear Piping System,"

ASM op cit pg. 253, 1985.

12. Moelling, D.S. and Gallucci, R.H.V., "Probabilistic Analysis of Integranular Stress Corrosion Crack Growth in BWR Recirculation Piping," ASM

op cit pg 261, 1985.

13. Gordon, B.M. and Gordon, G.M., "Mitigation of Pipe Cracking for BWR Plant

Life Extension," ASM op cit pg. 287, 1985.

14. Horak, J.A., et al, "Review of Effects of Long-Term Aging on the

Mechanical Properties and Microstructures of Types 304 and 316 Stainless

Steel," ASM op cit pg. 301, 1985.

15. Spond, D.F. et al., "Degradation and Replacement of Main Feedwater Nozzle

Spray Heads in Two Once-Through Steam Generators," ASM op cit pg. 283,

1985.

J
-30-

16. Odette, G. et al, 'Irradiation Effects on Aging of Pressure Vessel

Steels" ASM op cit pg. 375, 1985.

17. Banford, W.H. et al, "Thermal Aging Issues in Nuclear Power Plant Piping

Integrity," ASM, op cit pg. 370 1985.

18. Server, W.L;, 'Life Extension of Components with High Cumulative Fatigue

Usage," ASM op cit pg. 375 1985.

19. Bush, S.H. et al, "Aging and Service Wear of Hydraulic Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power

Plants," NUREG/CR-4279, February 1986.

20. Duframe, K. et al, "Wear Measurements of Nuclear Power Plant Components,"

EPRI NP-3444, May 1984.

21. Vause, J.W. et al, "Aging of Nuclear Station Diesel Generators: Evaluation of Operation and Expert Experience," NUREG/CR-4590, April 1986.

22. Dingee, D.D. & Johnson, Jr., A.B., "Aging and Service Wear of Diesel

Engines Used for Emergency Power at Nuclear Power Stations," ASM op cit

pg. 367, 1985.

23. Higgins, J.C. & Subudhi, M., "A Review of Emergency Diesel Generator Performance at Nuclear Power Plants," NUREG/CR-4440, 1985.

24. Vesely, W.E., & DeMoss, C.M., "Evaluation of Diesel Unavailability and

Risk Effective Surveillance Test Intervals," SAIC report for BNL PETS

Program, April 1986.

25. Subudhi, M. et al, "Operating Experience and Aging-Seismic Assessment of

Electric Motors," NUREG/CR-4156, June 1986.

26. Subudhi, M. et al, "Recommended Maintenance Practice for Improving Motor

Reliability in Reactor Power Plants," BNL FIN A-3270, September 1985.

27. Taylor, J.H. et al, "Seismic Endurance Tests of Naturally Aged Small

Electric Motors," BNL Technical Report A-3270-11-85, November 1985.

28. Subudhi, et al, 1985 (Aging/Seismic).

29. Sugarman, A.C., "Aging-Seismic Correlation Study on Class 1E Equipment,"

NUREG/CR-3808, September 1984.

30. Bonzon, L.L. & Hente, D.B., "Test Series 1: Seismic-Fragility Tests of

Naturally-Aged Class 1E Could No. CA-2250 Battery Cables," NUREG/CR-3923,

September 1984.

31. Greenstreet, W.L.et al, "Aging and Service Wear of Check Valves Used in

Engineered Safety-Feature Systems of Nuclear Power Plants," NUREG/CR4302, Vol. 1, December 1985.
-31-

32. Subudhi, M. & Taylor, J.H., 'Indian Point 2 Reactor Coolant Pump Seal

Evaluation," BNL A-3270/87719-7-86, 1985.

33. Toman, G.J. et al, "The Interaction Effects of Relay and Circuit Breaker

Aging in a Safety-Related System," NUREG/CR-4715, July 1986.

34. Gunther, W.E. et al, "Operating Experience and Aging Seismic Assessment

of Battery Chargers and Inverters," NUREG/CR-4564, June 1986.

35. Stuetzer, O., "Status Report: Correlation of Electrical Cable Failure

with Mechanical Degradation," NUREG/CR-3263, April 1984.

36. USNRC, "Nuclear Plant Aging Research (NPAR) Program Plan," NUREG-1144,

July 1985.

37. Fullwood, R, "Incipient Failure Detection Methods for Nuclear Power

Plants: Reality and Possibilities," BNL Report in printing, 1986.

38. Sliter, G.E. & Carey, J.J., "EPRI Research on Component Aging and Nuclear

Plant Life Extension," ASM op cit pg. 105, 1985.

39. Plumstead, W.C. & Cady, D.R., "NDE Application in Power Plant Life

Assessment Studies," ASM op cit pg. 53, 1985.

40. Engl, G. & Figlhubur, D., "Use of Nondestructive Evaluation (NDE) Methods

to Improve Power Plant Availability Factors," ASM op cit. pg. 57, 1985.

41. Weber, R.M., "Use of Nondestructive Evaluation Methods to Improve Power

Plant Availability," ASH op cit. pg. 61, 1985.

42. Rose, J.A. et al, "Survey of Aged Power Plant Facilities," NUREG/CR-3819,

June 1985.

43. Leverenz, et al, "Using PRA to Rank Importance of Aging on NPP Components," ASM op cit. pg. 639, 1985.

44. Silver, M. and Vasudevan, R., "Pilot Assessment: Impact of Aging on the

Seismic Performance of Selected Equipment Types," EQE Report for BNL Contract 223951-S, December, 1985.

45. Davis, T., et al, Importance Ranking Based on Aging Consideration of

Components Included in Probabilistic Risk Assessments," NUREG/CR-4144,

April 1985.
-32-

APPENDIX A

AGING QUESTIONS AND ANSWERS

The following questions and answers arose during a seminar at Brookhaven

National Laboratory when ALEAP was introduced to a constructively critical

audience. They are presented for further understanding of the program.

Q. How do you decide how old is old enough?

A. A very difficult question because with replacement of aged components,

operation and safety can be sustained. It seems that this is eventually

an economic problem concerning the replacing of equipment to maintain the

level of safety at or near the level on which the plant was originally

licensed. Also related is the question of when to replace components.

A hybrid approach is probably necessary combining operating experience,

expert opinion, in depth engineering studies, inspection, condition monitoring, vendor input, trending, etc. This study fills part of the need.

Q. How much can the age-caused risk rise before action must be taken to

mitigate the aging effects?

A. This is basically the question of how safe is safe enough. We do not

feel that it is our mission to establish these safety criteria but to

provide the NRC with the necessary tools on which to base the regulations.

Q. How many models will be needed to represent the nuclear industry in the

US?

A. This is unknown and depends on the system. There are something like 60

models of the Service Water System to represent the industry used in

ASEP, but only a few for more standard systems. One of the tasks will be

to establish an aging correlation matrix to determine permissible plant

grouping without serious loss of specificity.

Q. How do you propose to reduce the plant complexity to a do-able problem?

A. We propose to prioritize the plant systems and components according to

the safety significance of the aging using PRA and fault tree techniques.

Q. How do you measure this safety significance?

A. We propose to use a PRA model of the plant and to rank the systems and in

turn their components according to their importance to the plant safety

at any given time. We will place the primary effort on those components

that are the most important and address the ones of lesser importance on

a resources availability basis.
-33-

Q. What measure of importance will be used?

A. This is not settled but for the time being it appears that the Age Importance measure will be defined as:

Frequence of Age-Caused Failure x Change in Risk if the component under

consideration fails.

Q. Is Age Importance independent of time?

A. No. It must be evaluated for various time periods.

Q. Will this change the prioritization of components for aging investigation?

A. Possibly. If the assumption that the components that have the highest

aging failure rate are also those that are the most important to aging is

true, then the initial prioritization should be maintained. One must

realize that the Age Importance measure is complex. The frequency of

age-caused failure may be time dependent but so is the change in risk if

it fails because the components involved in this latter quantity are also

changing.

Q. How can you be sure that the components most significant to aging were

included in the PRA?

A. A reexamination of the PRA must be done to assure the consideration of

things such as piping and the pressure vessel until it can be shown that

they are not significant.

Q. How will common cause dependencies be addressed?

A. They will be addressed in the PRA model. It may be necessary to perform

additional analyses to characterize the operating environment of the components at least in so far as this contributes to aging. There are several codes available for this, such as INEL's COMCAN, SETS, and the WANX

series.

Q To what extent can industry-wide data be used?

A. One of the tasks is to examine its applicability. It may be so heterogeneous that an age dependence cannot be extracted. Knowing that a component is failing at some rate as is usually the best that can be extracted from say, NPRDS but aging analysis requires knowledge regarding

how long it was in service, the test and maintenance practices and what

was the root cause of failure.

Q. ~Why do you need to know the root cause?

A. To determine if the cause of failure was age-dependent. This may be

fairly broad and include age dependence in human error.
-34-

Q. How do you factor trigger events (accident initiators) into the study?

A. This must be by analyses, such as PRA, and special review and extension

of the failure data bases.

Q. What about new initiators which are not currently in PRAs, but may be

important at a later lagged/time?

A. One can never be 100% complete, but every effort will be made to identify

new initiators creeping into the failure data bases. Also, analyses can

identify potential new initiators for aged systems.

Q. What about cutsets (for PRA system fault trees) that are not now dominant

but may become dominant as the systems age? Or new failure modes due to

aging that are not currently included in fault trees and hence for which

there would be no cutsets?

A. Must be careful not to truncate low probability cutsets. Also one may

need to modify fault trees with new failure modes (due to aging) and

hence, generate new cutsets.

Q. What information do we want from selected plants to be visited?

A. Information such as:

Maintenance History Technical Specifications

Failure Data Drawings

P&IDs ISVI/ST Information

Procedures Reliability Data

Test Information Spatial Layouts

PRA Information Materials Used

Stresses Operating Environment
-35-

APPENDIX B

PRELIMINARY PRIORITIZATION OF INDIAN POINT 3 CCW SYSTEM

B.1 Safety Ranking of Components

As a demonstration of the methodology for prioritizing the safety significance of component aging, the process begins with the CCW system cutsets.

Table B-1 presents the component cooling water first and second order cutsets

from the Indian Point Probabilistic Safety Study (IPPSS). The rule that has

been followed is that the total importance of a component is the sum of its

importances in the accident sequences in which it appears. Table B-2 presents

a factored grouping of these added cutsets with the leading term being the

component of interest. This component identifier multiplies the sum of additional terms that contribute to its total importance. Thus a single cutset

appears alone and by itself as does for example, UPPLEAKS. Some terms are

paired with single terms such as UTK0031L*UTKO032L so there is one listing for

the first term and another for the second term. This table provides the algebraic representation of the Inspection Importance of each of the components.

To obtain a numerical value requires the substitution of the appropriate nonaged failure rate data. This is provided in Table B-3 in which the component

is identified as before, the non-aged failure rate from the IPPSS is presented

and a brief description of the component and the type of failure. Using the

failure rates in this table and the algebraic representations of the Inspection Importance presented in Table B-2, the individual component importances

may be calculated and ordered by-descending importance. TAble B-4 provides

such a presentation. It will be noted that the Service Water is the most

important "component" of the CCW system because it must remove the heat taken

up by the CCW. This study will address the interface only, and not examine

the details of how/why service water fails. Next in importance is switchgear

bus 2A because its failure is a common cause failure of the CCW. Least in

importance are the surge tanks because of the low likelihood of failure and

their redundancy.
-36-

Table B-1 CCW System Cutsets

First Order

1 UPPLEAKS

2 TXV31--C

3 TSW1NOFL

Second Order

1 UTKO031L UTKO032L

2 UXV759AC UXV759BC

3 UHE0031L UXV759BC

4 UXV765AC UXV759BC

5 TXV034AC UXV759BC

6 TXV035AC UXV759BC

7 UXV759AC UHE0032L

8 UHE0031L UHE0032L

9 UXV765AC UHE0032L

10 TXV034AC UHE0032L

11 TXV035AC UHE0032L

12 UXV759AC UXV765BC

13 UHE0031L UXV765BC

14 UXV765AC UXV765BC

15 TXV034AC UXV765BC

16 TXV035AC UXV765BC

17 UXV759AC TXV034BC

18 UHE0031L TXV034BC

19 UXV765AC TXV034BC

20 TXV034AC TXV034BC

21 TXV035AC TXV034BC

22 UXV759AC TXV035BC

23 UHE0031L TXV035BC

24 UXV765AC TXV035BC

25 TXV034AC TXV035BC

26 TXV035AC TXV035BC

27 TXV33--C UXV759BC

28 TXV33-1C UXV759BC

29 TXV33--C UHE0032L

30 TXV33-1C UHE0032L

31 TXV33--C UXV765BC

32 TXV33-1C UXV765BC

33 TXV33--C TXV034BC

34 TXV33-1C TXV034BC

35 TXV33--C TXV035BC

36 TXV33-1C TXV035BC

37 UES-35AC 4BS-333C

38 4BS-331D 4BS-333C

39 UXV760AC 4BS-333D

40 UM00031S 4BS-333C

41 UXV762AC 4BS-333D

42 UPM0031S 4BS-333D

43 UCV761AQ 4BS-333D

44 UCCO031F 4BS-333C
I

-37-

Table B-2

UPPLEAKS TXV31--C TSWINOFL UTK0031L*UTK0032L UTKZO032L*UTKO031L

UXV759AC* (UXV759BC+UHE0032L+UXV765BC+TXV034BC+TXV035BC)

UXV759BC*(UXV759AC+UHE0031L+UXV765AC+TXV034AC+T.XV035AC+TXV33--C+TXV33-IC)

UHE0031L* (UXV759BC+TXV035BC)

UXV765AC*(UXV759BC+UHE0032L+11XV765BC+TXV034BC+TXV035BC)

TXV034AC* (rXV759BC+UHE0032L+UXV765BC+TXV034BC+TXV035BC)

TXV035AC*(UXV759BC+UHE0032L+UXV765BC+TXV034BC+TXV035BC)

UHE0032L*(UXV759AC+UHE0031L+UXV765AC+TXV034AC+TXV035AC+TXV33--C÷TXV33-lC)

UXV765BC*(UXV759AC+TXV034AC+UXV765AC+TXV034AC+TXV035AC+TXV33--C+TXV33-IC)

TXV034BC*(UXV759AC+UHE0031L+UXV765AC+TXV034AC+TXV035AC+TXV33--C+TXV33-lC)

TXV035BC*(UXV759AC+UHE0031L+tJXV765AC+TXV034AC÷TXV035AC+TXV33--C+TXV33-lC)

TXV33--C*(UXV759BC+UXV759BC+UHE0032L+UXV765BC+TXV034BC+TXV035BC)

TXV33-lC*(UXV759BC+UH1E0032L+UXV765BC+TXV034BC+TXVO35BC+UXV759BC)

UES -3 5AC*4BS -333C 4BS-333C* (UES-35AC+4BS -331D+LTJ40003 1S+UCCO031iF)

4BS-333D*(UXV760AC+UXV762AC+urM0031S+UCV761AQ) UXV760AC*4BS-333D

UM00031S*4BS-333C UXV762AC*4BS-333D UPM0031S*4BS-333D UCV761AQ*4BS-333D

UCC0031F*4BS-333C 4BS-331D*4BS-333C
-38-

Table B-3

Component Identifier, Non-Aged Failure Rate and Description of the Component

and Failure Mode

Identifier Failure Rate Component Description and Failure Mode

UPPLEAKS

TXV31--C

TSW1NOFLa

UTK0O31L

UTKO032L

UXV759AC

UXV759BC

UHE0031L

UXV765AC

TXV034AC

TXV035AC

UHE0032L

UXV765BC

TXV034BC

TXV035BC

TXV33--C

TXV33-1C

UES-35AC

4BS-333C

4BS-332Db

UXV760AC

UM00031S

UXV76'2AC

UPM10031S

UCV761AQ

UCCO031F

4BS-331D

1.46E-8/H

9.15E-8/H

4.2E-5/H

8.6E-10/H

8.6E-10/H

9.15E-8/H

9.15E-8/H

9.73E-7/H

9.15E-8/H

9.15E-8/H

9.15E-8/H

9.73E-7/H

9.15E-8/H

9.15E-8/H

9.15E-8/H

9.15E-8/H

9.15E-8/H

Component does

5.24E-5/H

5.24E-5/H

9.15E-8/H

1.36E-3/D

3.26E-6/H

9.15E-8/H

1.36E-3/D

3.26E-6/H

6.91E-5/H

This component

17 major sections of CCS piping

SW supply vlv 31, transfers closed

No flow from SW supply header (conventional)

CC surge tank 31 leak or rupture

CC surge tank 32 leak or rupture

Heat exch 31 inlet valve, transfers closed

Heat exch 32 inlet valve, transfers closed

CC heat exch 31, loss of cooling cap. (leak

or rupture)

Heat exch 31 outlet valve, transfers closed

SW inlet to heat exchanger 31, trans. closed

SW outlet to heat exchanger 31, trans clsd

CC heat exch 31, loss of cooling cap. (leak

or rupture)

Heat exch 32 outlet valve, transfers closed

SW inlet to heat exchanger 32, trans. closed

SW outlet to heat exchanger 31, trans clsd

SW supply vlv 33, transfers closed

SW supply vlv 34, transfers closed

not appear in fault trees - no data

No control pwr at switchgear bus 2A

No control pwr at switchgear bus 2A

Pump 31 suction valve transfers closed

CC pump/mtr 31 does not start/run

Pump 31 discharge valve transfers closed

CC pump/mtr 31 does not start/run

Pump 31 discharge check vlv transfers closed

was not found in fault trees - no data

No control pwr at switchgear bus 5A

Notes: a)

b)

IPPSS page if added gives 1.E-3/24 hrs.

Apparently a misprint. The fault trees show 4BS-332D which is

assumed to be correct.
-39-

Table B-4

Identifier Insp. Impt. Component Description and Failure Mode

TSW1NOFL

4BS-333D

UXV760AC

UCV761AQ

4BS-333C

UPM0031S

UM00031S

TXV31--C

4BS-331D

UPPLEAKS

TXV035AC

UXV762AC

UHE0032L

UXV765BC

TXV034BC

T.XV035BC

UXV759BC

UXV765AC

TXV034AC

UXV759AC

TXV33--C

TXV33-1C

UHE0031L

or rupture)

UTKO031L

UTKO032L

3.02E-2

1.96E-3

1.96E-3

1.84E-3

1.5E-4

1.4E-4

1.37E-4

6.58E-5

1.06E-5

1.05E-5

2.6E-6

2.5E-6

3.3E-8

3.3E-8

3.3E-8

3.3E-8

3.OE-8

2.64E-8

2.6E-8

2.2E-8

2.2E-8

2.2E-8

9.24E-9

2.26E-13

2.26E-13

No flow from SW supply header (conventional)

No control pwr at switchgear bus 2A

Pump 31 suction valve transfers closed

Pump 31 discharge checkvalve trans closed

No control pwr at switchgear bus 2A

CC pump/mtr 31 does not start/run

CC pump/mtr 31 does not start/run

SW supply vlv 31, transfers closed

No control pwr at switchgear bus 5A

17 major sections of CCS piping

SW outlet to heat exchanger 31, trans clsd

Pump 31 discharge valve transfers closed

CC heat exch 31, loss of cooling cap. (leak

or rupture)

Heat exch 32 outlet valve, transfers closed

SW inlet to heat exchanger 32, trans. closed

SW outlet to heat exchanger 31, trans clsd

Heat exch 32 inlet valve, transfers closed

Heat exch 31 outlet valve, transfers closed

SW inlet to heat exchanger 31, trans. closed

Heat exch 31 inlet valve, transfers closed

SW supply vlv 33, transfers closed

SW supply vlv 34, transfers closed

CC heat exch 31, loss of cooling cap. (leak

CC surge tank 31 leak or rupture

CC surge tank 32 leak or rupture

i

.1
-40-

APPENDIX C

SAMPLE SYSTEM SURVEY

As discussed in Phase I, Task 3 support systems such as Component Cooling

Water (CCW) vary considerably between NPPs and hence the design variations

must be surveyed and analyzed as a part of the overall system analysis. Below

is an example of a system survey for one plants CCW system.

Component Cooling Water (CCW) System Summary

Plant: Recent 2 Unit Westinghouse PWR System Designer: Sargent & Lundy

Info Source: FSAR, System Description, P & IDs

Pumps:

Number: 5 Head: TDG - 250'

Flow Rate: 4800 gpm Motor Horsepower: 460

Elec. Source:5 pumps on 4 - 4160 volt, ESF, busers

One motor each on Bus # 141, 142, 241, 242

The fifth motor (pump-0) on any of the 4 buses.

Heat Exchangers (ax):

Surge Tanks:

Cooling by:

Loads:

Number: 3

Number: 2

Essential Service Water (ESW)

RHR pumps, RHR HXs, Rx Coolant Pump Motor and Thermal

Barrier, Positive Displacement Charging Pump, Spent

Fuel Pool Rx, Letdown Hx, Seal Water Hx, Excess Letdown

Hx, Miscellaneous Loads.

Notes: CCW is a shared, normally cross-connected, system between Units 1 & 2.

Indication Alarms Interlocks

Instrumentation: CCW pump suction temp.

CCW HX outlet temp.

CCW pump disch. press.

CCW pump suction flow

RCP Flow & Misc. Flows

Same as ind. Auto pump start

on low discharge

pressure.
J. e,

I

I

SIMPLIFIED CCW SYSTEM

Jts

KEY

S.T. = Surge Tank

CCW= Component Cooling Water

Hx = Heat Exchanges

ESW = Essential Service Water

Ul = Unit One

U2= Unit Two

Trn = Train

S.R.= Safety Related