NUREG/CR-4144 - . - -- . : - - PNL-5389,

Importance Ranking :Based on

Aging Considerations "'of

Components Included in

Probabilistic Risk Assessments

Prepared by T. Davis, A. Shafaghi, R. Kurth, F. Leverenz/BCL

Battelle's Columbus Laboratories

Pacific Northwest Laboratory Operated by Battelle Memorial Institute 4 . . I I

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Prepared for

U.S. Nuclear Regulatory Commission

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NUREG/CR-4144

PNL-5389

RV

Importance Ranking Based on

Aging Considerations of

Components Included in

Probabilistic Risk Assessments

Manuscript Completed: March 1985

Date Published: April 1985

Prepared by T. Davis, A. Shafaghi, R. Kurth, F. Leverenz, Battelle's Columbus Laboratories

Battelle's Columbus Laboratories

505 King Avenue

Columbus, OH 43201

Under Contract to:

Pacific Northwest Laboratory

Richland, WA 99352

Prepared for

Division of Engineering Technology Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

NRC FIN B2865
ABSTRACT

This report presents a method for focusing additional research on aging

phenomena that affects nuclear power plant components. Specifically, the

method ranks components using a risk, aging sensitivity measure that

describes the change in risk due to changes in component failure rate.

Describing the aging phenomena and the resulting time-dependent component

-failure rate changes is beyond the scope of this study.

The applications use average component unavailability equations currently employed in PRAs to calculate the risk aging sensitivity. A more

exact calculation is possible by using unavailability equations that

include'the time-dependent characteristics of component failure rates;

however, these time-dependent characteristics are not well-known. The

-risk aging sensitivity measure presented here is, therefore, segregated

from these time-dependent effects and addresses only the timeIndependent portion of aging'phenomena'. The results identify the

component types that show the most potential for risk change due to

aging phenomena. Future research on the time-dependent portion of aging

phenomena for these component types is needed to completely describe the

risk impact due to component aging.

.~~~~~~~~~~~~~~~~~ -

i
TABLE OF CONTENTS. - . .,- . .. - v ..

.. : . . i

,- ,. Page

1.0,>INTRODUCTION . . . . . . ., .. . . ,. 1-1

2.0 -RISK

2.

:,'2.1

IMPACT OF COMPONENT AGING . . . . . . . . . . . .-. . . .

Risk Analysis . . . . . . . . . . . . . . . . .:..,. .

2-1

2-1

,I'-. 2.1.1

2.1.2

2.1.3

2.1.4 . 2.1.5

Background . . . . . . . . . .. . .

Overview of PRAs . . . . . ....

Scope of PRAs. . . . . . . ... .- .

Risk Equations . . . . . . . .

Unavailability Equations . . . .....

, .

. . .! 1_

2.1.5.3

2.1.5.2

I I t, -~~~~~., . . ., .

I Periodically Tested Components. . . . ! Continuously Monitored Components ., , 1 . ,

. -. . .2-1

. . . . 2-1

2-2

* . . 2-2

.. . . p2-3

X . . .. .2-4

..v~ 2-5

. . . . 2-6

. . . . 2-6

. . . . 2-9

2.2 Aging Analysis. . . . . . . . . . . .

2.2.1 Effect of Increases in Failure Rate.

2.2.2 Effect of Degraded Characteristics *- I

2.3 Aging Sensitivity Measure . . . . ... . . . . . . . . . . 2-9-

3.0 APPLICATION OF THE AGING SENSITIVITY MEASURE AT

SELECTED PLANTS. . . . . . . . . . . . . . . . .

3.1 Component Boundaries and Failure Modes. . . 3.2 Results for Components at RSSMAP Plants . .

. . . . . . . . . 3-1

3-1

3-1

3-4

3-4

3-4

3.2.1

3.2.2

3.2.3

Oconee . . . .

Calvert Cliffs . . . Grand Gulf . . . . . . . . . .

3.3 Combined Results. . . . . .

3.3.1 Oconee . . . . .

3.3.2 Calvert Cliffs . . 3.3.3 Grand Gulf . . . . .

3.3.4 Combined PWR's . . .

* * * * * * * * * * * * * * * . 3-4

* . . . . * . . . . . * . . . .

* . . . . . . . * . . * . . . . .

. * . * * . * * . * * . * * . @

* * . . * . * * * * * * * * * *

3-13

3-13

3-13

3-19

iii
TABLE OF CONTENTS

(Continued)

Paqe

3.4 Additional Components .............. . . . .. 3-19

3.4.1 Reactor Vessel . . . . . . . . . . . . . . . . . . . .. 3-19

3.4.2 Steam Generator Tube . . . . . . . . . . . . . . . . . 3-19

3.4.3 Snubber. . . . . . . . . . . . . . . . . . . . . . . . 3-23

3.5 Limitations and Assumptions . . . . . . . . . . . . . 3-24

4.0 CONCLUSIONS AND RECOMMENDATIONS. . . . . . . . . . . . . . . . . . 4-1

4.1 Conclusions . . . . . . . . . . . . . . . . . . . . . . 4-1

4.2 Recon uendations . . . . . . . . . . . . . . . . . . . . . . . 4-1

4.2.1 Use of Results . . . .-. . . . . . . . . . . . . . . . 4-1

4.2.2 Interfaces . . . . . . . . . . . . . . . . . . . . . . 4-2

4.3 Suggestions for Future Work . . . . . . . . . . . . . . . . . 4-3

5.0 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1

APPENDIX A . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

iv
LIST OF TABLES

Page

Table 1. Rate of Change of Component Unavailability with Respect

to Failure Rate . . . . . . . . . . . . . . . . . . . .

Table 2. Component Boundaries. . . . . . . . . . . . . . . . . . . . .

Table 3. Aging Sensitivity of Components at Oconee . . . . . . . . . .

Table 4. Aging Sensitivity of Components at Calvert Cliffs . . . . . .

Table 5. Aging Sensitivity of Components at Grand Gulf . . . . . . . .

Table 6. Aging Sensitivity of Component Groups at Oconee . . . . . . .

Table 7. Aging Sensitivity of Component Types at Oconee. . . . .

Table 8. Aging Sensitivity of Component Groups at Calvert Cliffs

Table 9. Aging Sensitivity of Component Types at Calvert Cliffs. . . .

Table 10.

Table

Table

Table

Table

Table

11.

12.

13.

14.

15.

Aging Sensitivity of Component Groups at Grand Gulf. . .

Aging Sensitivity of Component Types at Grand Gulf . . .

Aging Sensitivity of Component Groups in PWR's . . . . . . .

Aging Sensitivity of Component Types in PWR's. . . . . . . .

Aging Sensitivity Measures for Selected Components . .

Aging Sensitivity Measure Calculation for Steam

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

Aging Sensitivity of Oconee Components Grouped

by Type and System . . . . . . . . . . . . . . . . . . ...

Aging Sensitivity of Calvert Cliffs Components

Grouped by Type and System . . . . . . . . . . . . . .

Aging Sensitivity of Grand Gulf Components

Grouped by Type and System . . . . . . . . . . . . . . .

2-11

3-2

3-5

3-7

3-10

3-14

3-15

3-16

3-17

3-17

3-18

3-20

3-21

3-22

3-22

A-1

A-5

A-9

Table A-1.

Table A-2.

Table A-3.

v
EXECUTIVE SUMMARY

This study utilizes existing probabilistic risk assessments (PRAs) to gain

insights about the relationships between aging of nuclear power plant components and public risk. A method is developed and applied for determining the

potential risk significance of aging effects. This method is based on determining the sensitivity of risk to increases in component failure rates. The

partial derivative of the core melt frequency with respect. to the failure rate

of a specific component is the risk aging sensitivity measure used. Those components having the highest sensitivity have the potential for causing the

greatest change in risk if their failure rates increase due to aging or service

wear. A O~pE 

The results of the analysis indicate the most risk significant components at a

plant depend on a number of factors including plant system design, testing, and

maintenance intervals and operating procedures. Based upon the three PRAs

analyzed. (Oconee, Calvert Cliffs and Grand Gulf) many of the potentially mostly

risk significant components are in the auxiliary feedwater system, the .reactor

protection system and the service water systems. Pumps, check valves, motor

operated valves, circuit breakers, and actuating circuits are the component antItypes-that'have the most potential risk impact based on the aging se sitiity 4

measure.' Le0t5b_ J

The results of this study are intended to provide uidance for this selection

of components for further study in the aging program and as a guide toward

prioritizing resources. The results presented are subject-to several assumptionsand limitations. The risk aging sensitivity measure used does not

describe the time-dependent behavior of the failure rate. In addition no'

assumptions are made about which components are most susceptible to aging proVI

cesses. Other key limitations of this study are the limited number of plants -

analyzed and limited scope of the PRAs performed for these plants. Only the

components which appeared in the PRAs were considered in detail. Components

not analyzed in the PRAs or components assumed to have negligible failure rates

can be important to risk if their failure rates increase substantially. The

study suggests future research activities which would address many of these

limitations.

The output from this study can be combined with other studies (data, analytical

or experimental) that identify the components that are most susceptible to

aging mechanisms. The combination of identification of risk significance and

aging susceptibility will provide a good basis for effectively focusing

resources.

vii
1. INTRODUCTION

The overall goals of the Nuclear Plant Aging Research (NPAR) Programs are:

* To identify electrical and mechanical component aging and service

wear effects likely to impair plant safety.

* To identify methods of inspection and surveillance of electrical and

mechanical components that will be effective in detecting significant

aging and service wear effects prior to loss of safety function so

that proper maintenance and timely repair or replacement can be

implemented.

* To identify and recommend acceptable maintenance practices which can

be undertaken to mitigate the effects of aging and to diminish the

rate and extent of degradation caused by aging and service wear.

The NPAR program is being performed by the NRC Office of Nuclear Regulatory

Research and its contractors. Several specific research activities have been

established to achieve the overall program goals. Gpo

The objectives of this study concern only the-first goal. Our objective is to0

identify components in nuclear power plants that adversely affect risk if aging

processes decrease component reliability or degrade performance characteris-

'tics. This objective does not include identifying specific aging processes or

describing aging effects on component failure rates.

The approach taken in this study uses the results of existing probabilistic

risk analyses (PRAs) to gain insights about the relationship between risk and

component aging or wear-out. PRAs performed to date do not explicitly model

risk as a function of timen,but calculate an average risk level. This report

defines a risk importance measure that measures the sensitivity of risk to

changes in a component failure rate. This measure is the partial derivative of

the core melt frequency with respect to the failure rate of a specific component. Those components that have the highest sensitivity have the potential

.for causing the greatest change in risk if their failure rates increase due to

aging or service wear. The development of the aging sensitivity measure is

described more fully in Section 2.0. Results of application of the aging

sensitivity measure to components in selected PRAs are presented in

Section 3.0.

EThe output from this study can be combined with other studies (data, analytical

or experimental) that identify the components that are most susceptible to

aging mechanisms. The combination of identification of risk significance and

aging susceptibility will provide a good basis for effectively focusing

resources. Section 4.0 presents the conclusions and recommendations of this

study.

1-1
2. RISK IMPACT OF COMPONENT-AGING;

This section develops the aging sensitivity measure from the risk

equations of a PRA. Some background information regarding PRAs is

briefly reviewed to put.the study.in context. '-The'second part of this.

section discusses the potential impacts 'of component aging on risk.' 'The

third subsectioh presents the'aging serisitivity'measure. --

2.1 "Risk Analysis e'

2.1.1 Background

PRAs are performed in-order to assess the r-isk of nuclear power' plants

and to identify the~key'contributors to-that risk. A number of insights

developed from review of WASH-1400 (1) and other past PRAs are useful to

focus aging related research.

The Reactor Safety Study (WASH-1400) was the first comprehensive study

of the risk due to'theoperation of nuclear, power plants;;'-This study

shows that the risk to the public from-nb'rmal operation'and routine - releases-is minimal. The'lorisklis-dominated'by' low prbbability-,high

consequence~events 'where .large amountsof radi activity are;released. '

In order for'large;.amounts of-.radioactivity to be released1substantial'

fractions of the reactor core-must melt. From'a risk significance

viewpoint,'the'-a ing-processes.of concern are-th'ose that could

potentiallytaffect the likelihood of-core melt or affect the systems

that mitigatethe-consequences'of;core:imelt. .'

2.1.2 -Overview of PRAs -; :

PRAs are a method tormathematically estimate the likelihood and the

consequences of potential accidents at nuclear-power-plants. In the '

process of performing;-'aPRA, .the potential acc'ident initiators (LOCAs,

transients, loss-of-offsitecpower, etc.) earieidentified and'their '

likelihood quantified. The safety systems and their support systems

that must function to safely shut down the reactor are-thenidentifiedfor each initiator. -The safety systems and their support systems are

modeled' using event-tree and'fault tree-methodology.2' The'safety-systems

generally considered ina:PRA ar-e the reactbr pr6tection-system, main'.i'

and.'auxiliaryi'feedwater' systems,' high pressure-and -lowkpressure-injec-'-

tion systems, residual heat removal systems, containment.sprays,con-'-.' A-.

tainment coolers, and accumulators. Support systems include electric

power, service water, and engineered safety feature actuation systems.

Operator actions are also included in the models.

The event tree and fault tree model solutions determine the combinations

of component failures that lead to a core melt:for each-of the-initiators. The combination of an accident initiator and the system failures

2-1
that result in core melt is referred to as an accident sequence. The

combinations of individual component failures that cause the required

systems to fail is-referred to as a cutset.

The;probability of-eachindividual.component being;unavailable is

referred to as its--unavailability. The.probability of. the cutset is the

product of the unavailability of the individual events. The-frequency

of an accident sequence can be approximated by the sum of all the

cutsets that result in failures of the same set of safety systems. The

overall plant risk is similarly approximated by the sum of the accident

sequences, or equivalently, the sum of all the accident cutsets.

In addition, a probability of containment failure can be assigned to

each accident sequence. In-some PRAs, the consequences of accident

sequences are evaluated in terms of man rem, fatalities, or economic

impact.

2.1.3 Scope of PRAs

The scope of.PRAs'vary greatly. Some consider.internal. events only;

others includeseismic events,.floods and fires,-.etc; The depth. of the

analyses--of the systems and sequence consequencesalsof varies considerably.. The scope of: the PRA,-as well as the level of detail considered,

limits the information.that can be extracted from the analysis.

PRAs genera~lly..concentrate: on'finding.theimost r.iskosignificant.come-.>

ponents. In many cases passive components such as the containment

building, the reactor vessel, and storage tanks are considered to have,

negligible failure rates and.are omitted from the risk analyses. In

most PRAs, wires and piping segments are considered-to have failure

rates that are negligible when compared to the motors and valves with

which they are associated and.are omitted.from-further analysis.

However, the risk-significance of a particular wire or piping segment

can~be inferred from the-PRA by determining the effect of failure of the,

wire or pipe. on -the component-to which it is connected.

2.1.4- Risk Equations.

In risk analyses, risk isexpressed as a combination of frequencies of

initiating events,.probabilities that safety systems are.failed and - :

consequences of the sequence. The risk from a single accident sequence,

cutset can be expressed:

- - . - - - . Rc F.Qi.C (1)

where

Rc = risk associated with the cut set

F = initiator frequency

2-2
Q= probability the components of the cut.set i are -

all failed

C consequence of the cut set. --

In the above equations, the initiator could be a plant transient or a

loss-of-coolant accident (LOCA) and the probability the necessary safety

systems are unavailable may depend on.whichi'initiator has occurred. The

consequence term, C, is a measure of the expected consequences of the

sequence given a core melt. In this report we are limiting the analysis'

by considering core melt frequency as the measure of risk and will drop

the C from the equation.

The plant risk, Rp, is the sum of all the accident sequences and is.

therefore expressed: ,

al

cut

sets

Rp !_LRc' (2)

2.1.5 Unavailability'Equations -'

The term Q in Equation '(1) 'is the'probbility ofa specific set-of '

components are failed and is expressed

K

Qi= qi (3)

j=1

where

qj = unavailability of component j

K = number of components in cut set i.

The unavailability term, qj, for each component is dependent on a number

.of factors including the type of component, the testing interval, the

failure rate, the time it takes to repair the component, the time period

in which the component undergoes scheduled maintenance, and the likeli---'

hood of human error that affects the component. The types-of components:

considered in this study-fall--into two gendral'categ'6rie's: /periodically

tested components and continuously monitored components. The unavailability equations for each type are presented below.

2-3
2.1.5.1 Periodically Tested Components

The average unavailability of periodically tested components consists of

five terms, and the formula is expressed as the following:

= . *IF q. qR qM qH)

where

s= total average unavailability of the periodically

tested component

qF= average unavailability contribution from failure

occurrence during the test intervals

qT = average unavailability contribution from test per

qR = average unavailability contribution from repair

of failure

-iod

qM = average unavailability contribution from scheduled/

unscheduled maintenance

iH = average unavailability contribution from human error.

The average unavailability contributions given that the failure rates

are constant are presented below:

qF = XS T/2

qT = qo T

qR = Xs TR

- = dM

TM

qH = C

where

XS= constant standby failure rate

T = interval between tests

(5)

(6)

(7)

(8)

(9)

2-4
qo override unavailability (the probability that the component

is inoperable during the test)

r = test duration time

TR = repair duration time

dM =,average maintenance duration time

TM = average interval between maintenance

C = human error probability.

Hence:

XsT~~~~ dM

qs =-- q 0 T +X T + T- . (10)

M

For.some components,.such as manually operated valves, the failure rate

(X )'is extremely small, and can be assumed negligible.' The formula for,

these components;becomes:

dM + (11)

It'should be noted that the negligible Xs is for a specific failure

mode. . . . - . .

2.1.5.2 Continuously Monitored Components'

The average unavailability of this.class of components is-the proportion

of time that the '-component is 'inoperable in a relatively.iong period of

time. Again, with the assumption that-the failure rate is constant, the

formula for .the average unavailability is given below: ;. .

._.oR . (12)

=O .l+XoTR .- :.- .

Approximately . .. .-. - - ..

* . . !;. ';-*-;

- 90 .oTR -- ; -' ' ~ i (13)

* . . - *..

where . -. '' .. - - '. - . -

4 'qo' =average unavailability'of'c6ntinuously ' monitored components

2-5
Xo'= constant operating failure rate

TR = repair duration time.

2.2 Aging Analysis

In order to evaluate the risk significance of aging phenomena, it is

necessary to define what is meant'by aging phenomena. For our purposes,

"aging phenomena" are phenomena that, have-one-of-the following.,two

effects:

(1) Cause the failure rate of a component to increase as a function of time, or

(2) Cause a component that was designed to meet certain standards

to degrade such that it no longer fulfills its design

'I requirements.

2.2.1 Effect of Increases in Failure Rate

The first aging effect-considered causes the failure rate.of a component

(orea set of com0onents) to increase with time asthe components,-age or'

wear out. Figure 1 shows a sample plot of the failure rate X as a

function of time for a typical component. This is the familiar

"bathtub" curve common to many components. This curve has three

distinct regions: (1) the burn-in period, (2) the period of normal

operation (where the failure rate is essentially constant), and (3) the

wear-out period..,Aging phenomena occur. in-the wear-out perfiod where. the

failure rate-is increasing. The root causei-of th'isincre-asein failtire..

rate results from any of a number of aging pheneomna, fatigue or

corrosion, for example. The increase in .the failure rate with.time can Thai

have two effects on risk: ' CIO e, -C-4

(1) The increasei n failure rate increases the unavailability

(decreases .the reliability) of a component important to safety

(2) The increase in failure rate of certain components could cause

an increase in initiator frequency. This effectively

increases the number of times safety systems must operate and

proportionally increases the risk.

An example of a component where the unavailability increases with'time

i's a pump in the low-pressure-injection system of'a PWR. Normally the

pump is in the standby mode and is tested at regular intervals. If the

failure rate is increasing with time (as in the wear-out region of

Figure 1), the unavailability history may look like that of Figure 2.

In this example, the test interval remains constant but the fraction of-;

tests detecting failures is increasing as the component ages. The

unavailability of that component, and therefore the risk associated with

2-6
A (T)

-~~ ~~ - *s

T-

------- Burn-In - --------- Maturity - A--- Wearout.----

FIGUREl1. Example of a failure rate curve.

2-7
Unavailable

Available

T O

FIGURE 2. Component unavailability history.

2-8
that component, is increasing with time' and'may b6' substantially higher'

at the end of the perioddof'interest than-at'the-beginning. ' . '

An -example 'of a component that co6uld -catise risk, to Increase- by causing

ithe initiator frequency to-increase is 'a steam 'gene1ftoi tubeJ''If'the

failure rate of tubes is increasing, the likelihood of a steam generator

tube rupture increases. Should this event occur, the necessary safety

.:,,systems have to operatercorrectly to prevent core melt. Another example

of components that increase risk:,by increasing 'the frequency of initiators is the reactor coolant system'.(RCS)'piping. Also,' components'on

the secondary-side of the'plant, such as-the main feedwater pumps, whose

failure rates increase with.time have the effect-of increasing the frequency of transient 'initiators'and'thus-the risk'. - ' ' '-

2.2.2 Effect of Degraded Characteristics

The othertype 'of;agingp phenomena-that-is-of intterest are processes that'

gradually'degrade characteristics of-the component. This could cause a

component that is-designed to meet' certainidesign'requirements to'

degrade such that it'no longer'fulfills' its design requirements.

Examples of this type of component are snubbers that'lo'se their damping

capacity as the fluid leaks through the seals or'heat exchangers that

lose heat transfer capacity as an oxidation'layerT-is'lformed on the-''

tubes. 'The reactor;vessel can-also be treated as this-type of component'

since its pressure capacity decreases as'a-function.of fluence.- Determining.the risk significance of:this degr'adation'-is more complex'than-"":

for components described.-in' the last sectioni since it"generally involves"

combining a probabilistic load-.distributionlw'ith'fragility curves' and '-"

considering the impacts of the different failure modes. Current risk

analyses generally consider all.components-to perform-as designed'unider'

conditions of load and to be operating in accordance with design specifications. It will, therefore, be difficult to-use PRAs directly to

.evaluate the risk significance~of components "of this type. However,

'"ounding calculations can be performed.- - -

2.3 Aging Sensitivity Measure

In order to ch&ricterize-the.risk impact'of;compo6nent-!aging:and service'

wear effects, it *is.necessary'to' characterize the'ti-me dependent; nature';

of-the change in-plant riskl 'That is,- i' i -. 'i

i: -j>->;:- I- ;: .; :; I ' - i .

A !t t ;i'.h : 

where

'A.= riskimpact due to aging ',

.'R =.plant risk.. .'

:2-9.
As defined in Section 2.1.4, plant risk.is a function of component

unavailability,.qj, and component unavailability-is a function of

component failure rate, A. For the study of aging, the failure rate is

a function of time, t. Taking advantage of the chain rule, changes in

plant risk are expressed as 0 :

aR a . aq" . J (15)

at ', axj at

J., J., .

The'risk impact due to aging can now be-separated into-two distinct

parts,

(1) The effects of changes in component failure'rate on risk (the

first two terms of the right hand side of Equation 15)

(2) The time-dependent effects of aging and service wear on the

component failure rate (the third term of the right hand side

of Equation 15).

This report concentrates on the-first. part,-'the change in-risk duezto

changes in component.failure.rate;.' The second part,'changes.in'the

failure rate-due to aging-and'service wear, is beyondthe.scope of this

study and should be-investigated through data evaluations, experimental

studies, or additional analytical models. Section 4.0 describes how

these two parts combine to describe risk impact due to aging.

We-define the risk aging sensitivity to failure rate as

3R aR .q (16)

J 5F = q.aj j 'I ~~~J

where the first term on. the right hand side of Equation 16 is the

partial derivative of riskwith respect to component unavailability and

the second term is the partial derivative of the component unavailability with respect to the component failure rate.

The first term, the partial derivative of risk with respect to component

unavailability, can be shown to be equivalent to the Birnbaum

measure. 2) This is a measure of the impact of a component failure on

risk and can be computed by changing the unavailability of'the component

in the risk equation to unity and determining the change in risk. ' Vesely(3) et al have calculated values of the Birnbaum measure in recent

work. The second term, the partial derivative of component unavailability with respect to component failure rate, is presented in Table 1.

The expressions in Table 1 are derived from the component unavailability

2-10
Table 1. Rate of change of component unavailability with respect to failure rate.. .

; * . nRate of Change of

Coiimponent Unavailability

: -With Respect to~Component

Component Type Average Unavailability - Failure Rate.-

Periodically

Tested

Component - = s T + qo0 T + X5TR

d

TM

+ C

11

( ! L '

. I- ...

 '. i_'jk - - ,a XS .

I , . ..

.1 .-.

T ' R .,, 1 .~ I I F-

:IP 

'I

-I

_1

X . .- . -.

Periodically Tested dM .,

Component With qs qo +M T Cq 0

Negligible Failure T TM a Xs

Rate

.*.~ ~ ~ ~ ~ ~ ~

! '

Continuously

Monitored qo= x. TR

. :' t:. -"

: . .0

! il ,

- I

. I: -'

TRI il

I-

.. r%

9 .!.I 

..' 1: -I

.II . I

I .

i. I I
equations in Section 2.1.5. This second term is related to the time a

component is unavailable when it is failed.

Table 1 also includes a risk aging sensitivity for components with

negligible failure rates. This type of component unavailability is

dominated by constant contributions, for example, human error, and

represents an essentially time-independent unavailability. In this case

the risk aging sensitivity factor is zero.

The risk aging sensitivity measure is used to rank components based on

their potential for risk change. The measure makes no-'assumptions about

the rate of component aging;'the ranking results are'valid only when all

the components age at the same rate. Differences in aging rates'between

different component types is beyond the scope of this study and must be

addressed in future research to describe the time-dependent behavior of

component failure rates.

Section 3 presents the results obtained by applying the aging sensitivity measure to the components at selected plants.

2-12
3. APPLICATION OF THE RISK AGING SENSITIVITY

MEASURE AT SELECTED PLANTS

In this section we present the results .of risk aging..sensitivity measure

calculations for plants analyzed as part of the Reactor Safety Study

Methodology-Application Program (RSSMAP)(4,5,6). Thii6`itudies

represent limited-scope PRAs'in-that they. do'not include-external events

Land do-not specifically include analysis of piping and wiring. The

.plant-includ din this'analysii -retw6`PWR'sl'(Oconee and Calvert

Cliffs) and-6n&e'BWR (Grand Gulf).' Also included in this'section are'

bounding calculations-for three'other-coriponents: ,a reactor vessel,

steam generator tubes, and snubbers..""

3.1 :Component Boundaries and Failure Modes ' -' '

The term "component" can be interpreted'differently. -In one sense,

"components" can be considered individual pieces of hardware, e.g., a

valve casing, a valve 'stem, wiring, 'etc. The"component" can als6'bbe

considered'as a'functional unit such as&a~motor operated valve that

consists of a number of component parts. 'Components§as-defined in most

'PRAs'and in this report represent-functional units.-- A motor operated

valve for instance is interpreted as consisting of'the'valve, the motor

operator, the circuit breaker, and the electrical cable and control

circuitry specifically associated with the valve. A brief description '

of'the component boundaries for each'type'of component'is included in

Table 2. -

* Frequently, components-are subject to a number of different failure

modes. For instance, motor operated valves could fail to function by

several:modes including: failure to'open,' failur e't6 close, and gross

leakage. Table 2 also includes the most important failure modes for

each component-typ'e.'"'These failure modes represent'component functional

failures and do not indicate the root cause of the failure or the

failure'mechanism.! From an aging-perspective, the'time dependent

processes that lead to a functional failure'are of the'most concern.'

3.2 Results for Components at RSSMAP Plants

The risk aging sensitivity measure is calculated for 'individual

components at each plant. The individual components are grouped by

component type and also listed in order for each plant.

., ' , . . , I! ., .I I . ' .
I

Table 2. Component boundaries.

Component Boundary' Failure Modes of Concern

Pumps

(Electric).

Pumps

(Turbine

Driven)

Motor Operated

Valves

Includes-pump, motor, and the

control circui~try.and electric

power components specificJ ly

associated with the pump. ?1

Includes pump, turbine, and.

control circuitry specifically

associated with the pump.

Includes valve, motor operator

and the-3control circuitry, and

electric.'power components

specifical lvassociated wtih

.the valve. )

* failure to start on demand

: failure to run

* gross leakage

0

0

0

failure to start on demand

failure to run

gross.leakage t

e.-failure to open on demand

* failure to remain open

Control.. Valves

(Air Operated)

~i..I

Check Valve

Rel iefj Valve

Circuit . Breaker/ .

Contactor (RPS)

Relay (RPS)

Trip Module/

Includes the valve, the-air

-actuator, and the control. circuitry specifically associated

with the valve.

Includes the check valve only

Includes the relief valve only

.The circuit breakers that

provide power to the control

rod drive mechanisms.

The relays that actuate'the

trip breakers on signal from

trip module.

Includes the sensors, cables,

bistables, and relays that

measure plant parameters such

as reactor coolant pressure

and send a trip signal to trip

breakers.

* Failure to go to the,

"fail safe" position on

signal

o failure to provide control

capability

9.failure to open

* stuck open

* failure to open

* failure to open

* failure to send trip

signal when plant parameters require

3-2
Table 2. (Continued)

Component. Boundary FailuremModes-of Concern,

Actuation - -Includes the.sensors, cables, . failure to send ESASChannel/ bistables,land,,relays that signal'when required

Subchannel - measure plant parameters and

send out an Engineered Safety

Feature actuation signal.

Battery Includes the battery and the s failure-to provide DC.

battery charger. power *to components

requiring DC power (given

loss of.AC power).

Diesel Includes the diesel and its . failure to provide AC:.

Generator support sytems (lube'oil p.-,power to components

cooling, fuel supply, etc.). requiring AC power (given

loss of off-site power)

Room Coolers . Includes the fan and cooling . failure to cool pump room

coils that.provide.room cool-- ;- :. -

. ing to pump rooms. . . . .. .

(1) The electrical components specifically associated withthe pumpor motor

operated valve would include the connector, cable,"and circuit breaker that

power the motor, but does not include the electric power distribution

system that feeds the circuit breaker.

,~~~~~~~~~~

. i I . . I I .I

" -. 'i ;- , , '.1 -.

- I  .. . .I 1 - i - .1 .  I-

- . - . ., I

.1 .- ,,, . I . .

-11 . - P. - . -

I . I  11

- . . . . I

3-3
3.2.1 Oconee

V

pj

Table .3 shoWsthe resultsfor the Oconee PRA components.* This table.

,, A' lists the individual components in order of importance as ranked by:

their risk' agi'ng'isensitivity measure. As can be seen'in Table 3 most of

f ,,thecomponents with:the-highest,.importance values-are in the reactor,

protection system, the low pressure service water system, and the low

prds'sire'-injectionlsystem. A number of the importantcomponents are'

electrical>components. including actuation channels, trip modules,

circuit breakers, and contactors.. The individual-components are also

grouped by type and system, and ranked by their aging sensitivity

> > 'measure in Appendix A.

3.2.2 CalvertCliffs 'g

-Table'4 shows the.results for the Calvert Cliffs PRA. At this plant,

the component's"with-'the highest aging sensitivity measures are

components of-the auxiliary feedwater system and the reactor protection

system.- Again; the/'components have been grouped'by type and system, and

these' results-are presented in Appendix A. -

3.2.3 Grand Gulf"' -

*..Table.5 shows-the're'sults for the"Grand Gulf'PRA.- The components with'

the highest aging sensitivity measures ,are' components of-the service

water system and-the residual heat removal-system;-, The components are

grouped by type and system in Appendix A.

3.3 Combined Results * - - ~i . , > -. . , , .- . A .

This section combines the results of the aging sen'sitivity'measure

calc'ulations for individual components-.to-provide.an-overall ranking.

Two levels of ranking are provided.

In the first ranking, components of the same type that are in the same

'system are grouped together, i.e., motor operated valves of the

auxiliary feedwater system comprise one group. The aging sensitivity

measure for the group is the sum of the aging sensitivity measures of

the components in the group. The combined results provide an indication 7 H 9 of which component groups have the greatest potential risk'impact. This

ranking of the component groups takes into account the importance of the

individual components and the number of that type of component in each

system.

The second ranking combines components of the same type but does not

differentiate between systems. The aging sensitivity measure provided

for the component type is the sum of the aging sensitivity measures of

all the components of that type. The ranking is then a measure of the

importance of a component type that takes into account the importance of

individual components and the number of components of that type.

3-4
Table 3. Plant name: Oconee - Reactor type: PWR.

- - - - - - - - - - - -- - - - -...~~~~~~~~~~~~~-- - - - - - - - -6 - - - - - - - - - - - - - ANot C'11PiNi h Lf Pa'Ia ESo IMPACT CnOI4PINENI TYPF RATE ofCAhiERSKIPC

* IVforsr PT" I III' OFV tCIMPONFP47 S2PfI')flI1CALLY OF cnmPONFNT nFI compnNENT

IIVAVAILARILITY IFSTEn UNAVAILAPTLITY ICING

(RANK) OCtONTINIInU¶ILY WITH FTAIRIIE

I ;'141 a LPIR A SrA,,hId tI.PRS ACTIIAlrn-l C"A4ANiL t.OflOE-03( 4) S 463379F-02 4.3379E-05

P L03h'a-P3A 13111I'Afy LPw P1g11'P 7.?.IUaOE-u4( -'s.) S A.11379F7112 el.?K16BE-O5

I5 .VP P iSIANt)T LPSIII VArIIII*I P11.10 .7.60fltfE-0A( A.)1 S 41*1379E-02 3*?9AME-05

,4 'CON A RIPS CIWCIIIT:IdQfAFQ 'A' ~7J -EoA 1 '4.~337r-02 2.602?F-;05

di CIA R kPS CIHCUtIsq"EAl%'R.H 6.00ooE-Oa(; 8-.) S 4.3379E-11? P.A.021I-05

A. Ls.0301P34 Ut1PATING.LPS4. P10aNP t.OuOE-Opit A) 0 .8IE0 2 2e3tE-05

I -VPl OPI-.RAIINt LOSIMYACuII.' P1114P .1.00OOE-O?( .2? ?.?831F-03 - 2.?`1131F-05

,A (a II1AFETY R~t 1FF-VALVE 3.SfIOOE-041( q) S .41.131"F-02 1.511A3F-05

*4 cm C h0P'.CTRrIlII AREAXFQS C., 3.000OE-O04lfl) S 41.137911-fl2 1.30I4F-05

toI Tqn ,HPS.CTNCIITIIQ1EAKFNls Is, 3OQEO(1I54~7ph .ItE

It AIT7 P QFO!tt.94fl L, I ,. 3.P(IOOE-o4( ?) S *4.3379FV-f2 *1.3014F-615

I? I4T4 ~ :RS RFVMTE, TRIP .*411IILP A, 3.uIIOfE-0a( 11) 5.1 '4.~33?91` 02 't.1014IF-05

.11 H Tm4j 903 QF,4flFI`MIP 401,IE'LE3, . 3.Ofl0(lE-04( 114) 'S4;3379E:02 i1;10tAE-O0S

14 RlhT4 -RIiS QmnvIcfTTMTPMnUI ILE a' 3.00013E-441( 15) ..6 3179 E 02 t.130W4-as

,t-~ 10S.F .4ps Cfl'ITACTnR.E . 3.II(IOOE-041(16) S 4.33?9E-02 '.04-o

1IA -QP3F ,NPS CnNTACTnR 'F!. 3.uonA-;oo(,I7) S 4,33?RE-02O?1141-0

I 7 LP,17 .LPIS A KtCCR A'MOTOn npEmATEIP VALVE 2.30O0E-04( 'tR1 S 41.1379E`-O? 997E

til LP PMA .L IA CCM A~ Plimp *, .3fDOfE.041t 19) i 43379FV;8? 9.9?72E-;0b

'I Ist P.105 I ECCHAR WoninOHPISATEn VALVF 2.3000E-04( '21) s ,4*33i9E-O? .97E

ci, .~~~~20 .LP PIR ,lPs8 1

AY.CC.4 IIPIIMP .,.1.50SOIfE-04( 21) 'S A4379E-02 9.9772E-06

!22 .CF,12 LPIS.AA.4FCCR A CHECK VALVE., 2lftOOE-04( 22M

254 CF 14 LPIS PAECCII 5 CHFCx VALVE ,..'2.00E-4('241) S :4.3374E-~02 9q.9772E-06

?4 AP'l1 .P015 A AtECTR.A rtIFr.K v&VAF ?.34o0E-;o6('25) 5 4033?F-~02 '9.9??2F-0;h,

*25 LP12 ~ LPIS:A P. FCTH A mnflnR OPIqAytn VAtvE '?.3A00E-I'4( 27) ~ s 4,13379-02 9.9772E-CA.

PT TFST-A LPI1,~A A FCCR4.& TPST VALVF ,2.3001Fl-414(:,3fl S 4.3379F-fl2 'i;977217-fl6

281 tLP %5, LPISA&..4,ECQ'AA~Ii0TaIH OPFwAVFU VALVE .2.3f00OE-04( :31) S 43379 7-82 9977

*-20 LO 33 -LPIS.9f.A. FCCS II.CI4F.TI( VALVE 2.CA-3~2 *',137qE.n2 -9.977217-06

3f% :LO 14 *LPI3S8AECC,1..i MnTnH (PERATEn VALVE ,?.3800E-04(.34) S 4.3379E-02 "9.9772E-O&6

'31 ..L 71 ,LP It. q;FCC4 lICI.KVLI2.300flE.14('34) S1 3,3?qV-fl? '9.9772F-06

3? 7 FST. S LPIF.,A'.FCCII.It,IFSI VAI.VF 2.3800E-fl4( 31) 5 4.33797-02 9.9772E-06

33 LO Ai LPN. A.FCCRtH mnTnT lIPFQATEn VALVE '2.3000E-04( '38) S .4*111ME-fl 9.97727. 06

34 LO ?22 .LPI`1I,HAjCCR,,m.FIITnR OPF1&TED VAI.VE 2.2000E-fl4I( 39) I 41.1391"-0? 9.541aE.06

P; .LP 311 1P15.8.x ACCN C"iKr VALVE '2.?n007-04( 40) 5 403379E-112 9.5434F-06

3,, Li' Pt LP I A A. tFCC.M A mnl00P OFIIATiL VAt VF 2.20UCE-06( 'di I 54 41~379E-02 9.56347-06

3? L P~;9 LOIS A.A,FCtCH,A..c4ErKVALVL 2.?lnIIEO04(' 4P) S 4i39.fp q543AC-06

IA% C'4.,t HD415 ACTIIDTON TWAAIN IAuE34 43 .379E-n2 6b.073117-06

34 ,; HO?4 1101.5 A N#1'fIT PF61'.JTFgJ VALVE -1.40OOE-04( 43) S 57E0 b03E0

_HP, 1401 t4Pt 10 A rNFCX.VAIVFE 1.4800DE-04(.45) '. S -7flz 6.07 1-06

iat H,~i?' N."tSi A '%i)11t flPri.ATriu VALVE 1t.4hfl~tE-04( a6) S 4.337q[-02? 6.01`31K-011

.LP-19 . e"' A1~.4116"F-ol( 49) .5 a35?97.q2 .0IE0

43 LO P2" IrcH A AII-lV VAL'It 1.4111OE-fl4(%f3l) s' .17~f2~~~ 0'* 731E7-06

44 Fl~v 3417 AI'aF .HFCR~' VALVF 1.3800CE-44( 511 I" S4 .' 13797-02 5.63q3f-06

44 Fl. I' I AF,.S CmrCK VAILVF I ? .14DEn 5.43937-06

oh Fno All? 311, nPFRAIFU AFWA VAL~~t It. il'nFE-m4( 52) I;4.~ 3 3577.l;F?2' ''569E

'49, F n.%.i' 4t sAF.s' l~~FiVLt t. II S0UlFj-ofoQI 54 So -12 Sb33. -,-Ffl~~~ .4? ?~1 ~F'~s rtjI!Cv VALVF t.8u.13745)5 ~ ~ S*S;EO

48 F'I*a 1* * AF.'S rmFC, UaLvr F.301)IOF-M4 55 I It 43374IF-0 5.63913F-06

All Fq0V 5th FF., &lo nflF,411`TF VAI.VE I .3flhiE-14 ( 56) . q4 .?E() .31EO

%nl "P It issIlt r r.411C VAt VP R.qiflnA- ~ ,)44.1379F-02 3.046WTF06
Table 3. contd.

WlANK, C(lMImPO~Jl I0f1t.PIAJE" I w~b tmPACT Cnmps'iNET TYPF SlATF (if CHAMS6E RISK IM4PACT'

0444F, u~~~~br-ITVTII~~~ls OF COMPNF''NNT SMVERIIMICALLV UF -CEWPONE#T OF CnOKPOEN L04AVAILAIIILIIY 1FE'tkD IINAVAILAI$ILIIT AGING'

(RANK) OxCU4tITINUMlISLY NIT4 FAILURE frflNIflMFD RATE

(A)

I

cn

5 i

54

5%

56 -

57

15"

59

60

bi

6?

63

hS

66.

b7

h9

70

71

7?

713

74

75

lb

17

79

80

$I

o?

as

44

Ms

87

A9

90.

HP P1i

NDA I UP

lip PIC

CI' P

HP ?

HAT A

MAT hI

EFP A

Fnl4 j31

Ffll 370

FOw 372

EF.' d.

EFln 353

Fns 3501

Fnxu 3s5?

1 C. I .

Tr. 2

HP 1AR

0Mi 91

'41% 94

", 9s

43 57

EFP-Tn

C £56b

LPSf 1 37

HP III HP.1341

HP 145

LO lb~

LP, IS.

LP. 15

LO., I I

LP P8

C 357 F~n~ AR

04% 91* thq q t

C jib

C. 575

HP, 31A

IhPJ% C 40111P 'WENAfEil VAI.Vf

CIS r CHrF~1. VALVF

Will C PlimA4)

jTANiURT 14PIl AII.SYiTtIE ACTOAFTU1r1 CHANNEL

HPIS C 041§Tillw nWFRAtF.I VALVE

FPS 11&tIERY 'A',

LPS 11AfTE9PY AR

ASt. A ELFCTRIC I'I'mP

AFOtS ACIIECK VALVE

A tS A CHFCO VALVF

AF wS A uOTtR MI'FA IFU VALVE

AFWA 1 FLFCTWTC P11I4P

AFWS. A CHECK VALVE

AFfIS RPCHFCK.VkLVF

AF-1C%9,f4aiTf1Q OPEW&IEU VALVE

TIMAIGgENERATU4 I

TIOWAUGENEQATUaR P

HPIS A nPERATING P1uMPfSI

AFES T T1RAINFAIR lPEWATF0I VALVE

AFNS T TIPtiINE OVFSPEEn VALVF

AFIA * TURhINE r.UVERink VALVE

AFII. T.TRINE AIR OPFWATFU VALVE

AFWS T TuQ.INE PUMP l

AFLS T: T4MIITIR OPFA4IFS. VALVE

AFLS T t LPSW t rtlnt. UPLRATEP VAL.VF

tiPI. C 4AN114L VALVE

HPIS C MAMIBAL VALVC.

NPI% C CMAM41AL VALVt

LPIA f I4 FCCl A M4ANIIAL VALvF

LPIS q A ECC't m"pANIIAL VAI.VF

LPIS A. A FCCR A MANIIAL VALVF

LOIS A. 4 ECCk A O ANIAL VALVF

GANIIAL VALVF FeoP LPiS A HPI%

AFA I *4A'IISAL VALVE

AFnS T TAN1AL VALVE

AFdVS I MA114L LVALVE

0AlS I M411A'L VALVF

ALnS P M4ANtAL VALVt

AFAS A* MIAhte1L VALVt

HP13 A*. MA?'IAL VALVt

4.I9flhJOE-0uI h?)

A.9fuQF.-0%( 64)1

8.9000E-05( 65)

8.9auflE-05( 573

8.90oOE-m05( 66)

6.h00CE-05(b.7),

7.00OOE-05( ha)

7.UOUOE-05(.70)

7.0000C-0OI( 1)

1. 000 CE-0OiL 7?

7.OOOOE-05( 73)

7.00Q0E-05(, 75)'

`1.01100E-05W '16)

l.O0uDE-05( 77)

3!hOOOE-05(' 78).

3 booOE.05( 79)

1:400DE-041(. 441)

2.00O0E-Qf#( SP)

P.00l00E-06(,84)

,2.uOoooE-ob(' 85)

2.0000E-uA( 89)

Z.U0QOE-Q6( '90)

lI.9OuOE-05( b3)

8.4000E-USC 60)

A.91001-tQi( 59)

Z.30DOE-04( 35)

2.30O0E-04( 33)

2.3000E-04( 2A)

2.3000E-0a(: ?h)'

3.bflu0E-03(~ 3)

2.41OOUE-0f,( A?)

?.nu uE -us.~ CAl),

1.uo04aE-US( 69)

1.400S1E-uU( 47)

4.3179E-02

A.3319E.02

4. 3379E.OZ

a. 337'3E-02'

4.3379E-02

4.3319E.O0a

4.3379F-02

4.3379E-02

*.3379E-OZ

4.*3319E-02

4.337 9E-02

4.1379E-02

4.3379E-02

4*3379E-GZ?'

4.33793E-02-;

4.3379E-02

4.33?9E-02-

,1.?831E-03

4 3379F.02

A.3379F-0?~

4. 3379E-02 4. 1379F.02

4. 337 9E-0?

4.3319E-02

4.3379E-OZ

0.00#QE*00

0. 0000E0Oo

0.0000Et00

a.0000E.OO

0.0080E#60'

0.000USF.0OO

0.oOoue*OO

0.OOOE.0oo

fl.0060EtO0

0.000DE.00

0.00015C. 00

1*0.006o, 00

o .0o00E900

0.nuoCE.00.

o.a0oor*oo

3.86C7E-06

3,6607E-08

I.AbO7E.O&

3.7306E-08

3.7 30eE.a*

3,036SE-06

I.OMhE-06'

3.0365E-06

3.03h5(-06

3.036SE.06

3;0365E-06

3.03b5E-06

LOUS6E-0Ob

1.616E-06'

I .5bt6F.O8*,

3.1963E.071

4.67 5SF-88

A .615SE-0S

4 .fs18E-08

6.67581-04

8.68758E-08

86.756108

8. 615GE-0a

0 .0000E*00

0.0000c.00,

0.OOOOE,00'

0.0000E.00'

0. 0000C.00

0.0000E.00

0*OOOOE*00

n.aooo0C.oo

0.0000E.00,

0.0000E400

0.0OOOCEO00

On.0000F.O0

0.0000E,0O,

I. 0000 .600

0.AOOOE,00
Table 4. Plant name: Colvert Cliffs -Reactor type: PWR

RA~ r~ilpiP,,.Iv~d r~nvP.auii FM WTSK TNPA(CT .. CnM~PONFNf IYPF R4TF,flF CtAE4GE RI1SKJIMPACT N44F felt lit SC410,T ~~~~~~~~(IF I.UMP(INEN1 , b:IF'TWl1(ALLY flF CnMPflNFNT flF CIII4PONENT

Iii ~~~~~~r'~~~t.'tuui'~~~~~~~~~ ,JM~I)AVAIL~dTLIrY iFNTwn tIINAVATLAflTLtTY WPMNG

(RANK() nimcnNTINimuttJLY w171 FAILUPRE , . . MO~~~~~~~~~ ~ ~~~~~~~~~~~~~~~~~~~N~tflRFf RATE,

I 11"Oll AFVI TjliltIPHF PIA 1.6e100L-c 07 ) S 4.3379E-fl2 .90F4 TP2? -4T~oF Pilmil t.60OOE:011(ael) &.-qqo&Fg-o F.1 It ~ TIOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .Q139-f2 .694fl&E.04

1 CU45II ,AFII rnh1r1hL VAl VF AofufE0?. I) *4.337qE-02 64*Q06E-04

I C V 4512 ~Alvfq N1041R11. vAI.vF ,.-Ih0Q0E-uPl( '61 i -.4.3379E-n02 *.~l

S; MnV40UTI AFIS ;41),MTII .11FIA Ml~ VALVF I .bnlufE-fl?( . ) 3 *.3379E-112 6.qa06F-oa

C, ME1VGu?1I A FWS MUTnfl'flPFRAT~iI VALVE~ t.bou0k-~07( 0) s 4.131qrE-0? .6.4406LE-04

7; P3 AFW15.rHFC' VALvr, l.6001 1E-O?( 12) A 4.337117,02 i6.9406E-04 A 55i A~w4 r~HFCW VALVF I.6fn0E-~U?( 13) A 4.3379E.'02 6.9a06F-04

q P51 A~ft.t C.NFCI ,VAI.Vf 1. IhflOflE-0( I?7) S -4.33791!.02 6.940&F-04

lo. by AFVR CHFCW VALVF 10UEf?(- A S 4,.337qF-n2 6.9106E.0i

PI 14 AFNR CHrC.CAVALVF 1 .6000E-OP( 201) 4.'33qE-fl2 .6QOE

I I Sil 'ArIs CkFCK( VALVF I hAODOE-O? C3 2214.3379EF-02 h.q4O6F.O4

51.3 A4vnS CHECKe.,VALVV. 1. t00iDfEtI 23) s a.3379Fn02 6.9006E-04

IS '.K1N I "i FLAY 1 .2100E-02( 25) S 4.337qF-fl2 5.2055E-04

Ill WFLAY, KP "PS ~~~~~~~~1.20OOE-02( 24 a.033?9f-02 S.205SE-04

17 sS PS RIEL AY.,. 1 .2100CE-02( 271 S 4.3379E-02 3.205SE-0a

In XU PS RELAY ..... ,.. I.21100E-;02(. 214) 5 ~ 4.3379E-OZ 5225E0

tq "P lrIa RA.F 9.OOOOE-03( 2q) ,4.3379E-;02 3. 9041E-04

~~J 2 'A'I2 , S 1.T1NAAW-9ooOo331) S 4.3379E-0 2 ..3.9041E-04

14 ~ *21 . 34. R pS 'CIlCUT AqFAKF4 9.00loOE-03( 3P) S 4.3379E-02 .,3.9041E-04

2? IA., . - U~~~PS ,CIUCUTT AREA4V(FR 9.onOEl-o3(, ? 4.3579E-02 *,3.4041F-04 23 "IR 7 US3CqCTI.NQE4F 24 201 ' kPs CT~rITT bREAKER 9.0OOOE-91(..31) S 4.3379E-02 .3.904tF-04

25 CTRCIITT,' 3R RPS hQEAKER 9~~~.OOUOE-03(jsS3) A 4.13?9E-02 .3.9041E-04

Ph 44 RPS CIUCUIT, OPVAkFO 9.0OflOE-03(. 36) S 4.3379E-02 -3.soaiE-.0

27 MflVbs9 H PIS 9201 4 N21 POT(IqflPFW&TFU VALVE 4.4flOfE-04( _341 S 4.1379F-fl2 1.90R7F-04

2A .7 iAnlvCbfl III 24 021 MO1TIQ nPFIATFO VALVE 4.400f)E-O1( 5q) S4.3379E-O? i.qO87E-04

.q 111eit fs AJESEL' rENEPATIJP 012 2.2000E-03( 40) 5 4.337qE-02 453E0

*30 ;Slt?. .SIAS.SItIACHANNFL tH7. I.. 1.9OOOE-U1(. at) . % 4.3379E-02 01.?42GE-05

31 -1k2? *SALT. JI?. PUMP 1 .5004'E-OI( 42) 5 4.33yqE-Ile ht,5068E-05

3?P F(121T EPS O)IESEL, r.ENPAIE)R Nit 1.5flOfE-03C 43) 14 4.137qE-0? 50175 ,. t aTP~ I MIATTLPY,.Npl ! t.ifl3 E-03( 44) -.S4.14379E-02 .A.OAEO

3 34 C Vs I 5? .ALT, N2P,,CfltITRW. VALvF J. I.500E-03( 45) 3 4.3379r-02 6.150668E-05

-3,CV%I~jI .ALT,.#2? CONIfqI)L VAI.VF 1.511OOE-u3( -46) * .3379E-02 6.1506AE-05

3e. C"PIP ..SL A'C. 1.5s10E-o3( lEN~n 47) .1 4.3379F-62 ft.SOAAF-05

37 *S~ -,?; *>5M5 X?? PI`'14. 1.5ntoe0~-3( 44) 11 , .3379E-02 4.5068E-05

*4: SAN'fIV4flPNAT4I)`JL 9 11 4. t.VIs vpI g.Hw mvtnFAF)VL ' A .600111Eg-0aI(;5 SI 44137qEO02 'P.40OF.-05

41_C4 ------- tiIPIS N.'I I It LPI'S Sp 1. F '1Mpq e iccV h.6110oE0E-oI. 52). a1-.13?QEF-02. 2.1163OFE05

4A C ? iiP1 T oI It .4P,? 121 I U*.rK A4 I h.600OFE-94( 54) A, 4*13lqF-Op ?.4630E-05

41 Cet~' tip I t 4'I it 11po ##2 cmI~r' VALVI; t1`C100fl-04( t. 6k)) S 4.133?V..02 P.11630E-fts

44 H's I t4P IS Net F MPPb WA1 PII.4P 4.I6fl0fE-04(1 61) -. -4.1337E-op? ?.;4630E-05

45 TAP ;D IA45giIqg4A%'tIFL 4P h.hfOUOE-v4( b2) . 4.337qE.-o2 7.800S-05:

-46 'WIV4 t~g, uI4'4 le'1 NA?. ,6 mPP"1 vo') ..-,III 1EIPVAIFI) VALV 4.7000El-04( 61) . 4*.3S79P..'9?2 P.036SE-05,

'17_ r0h. .PiS. vp 3It. Ii' t S 1f?) '. 110%. 4*??., CIIFi: VII.vF 4.700110'tS-I(. hi).-. .. . .4*%3?qF-#)2 .. .2.f3881-05!

194 S IN? N dA% 4.A1C.HA'R)IFL '.2 4.hOfl(f04U( hA) s1 I.Sq~l qq%4F.-I5

4', Cf.t I :*P: 1 :A3 : N UgNf: Pj. coot9.ro vAIy.9'1.iotoI.04 ( in) A 4.5 579F-op I~qPE

Sot r.¶ is1 ait ; q I4,sjo a.' IN"#rot VALvI -'. 4.%lufE-n4( hAQ I 4.1374F-fl? 1.95?1F-05
Table 4. contd. . ..

---------------------.............. ... ..... .. .. ... ..... .-......-............. . .-..

WNAIK Cn'4P1N4It* J I **l.014`14~.Il W1SXI,J0PACT., COMPmlNeET IYPF WAIF Ofl r.1A'dI. l~t3ic t4PACT. jA14F a)FSr.,IPTfln4 oF cnmpnNF.NT 3xJFwjflIflIEALLY OF FO1aPONElT OFi COMPONENT UdAVAILAHILITT TF!%Tk!' UNAVAILAHtLfTY AGING. (RANIK) ozcO'dTiNtiIUSLY PituTH FAILIINE MONITORFUJ RATE ---- ------ ------ ------ ---- --- ------- --- --- --- ---- --- -- ! --- ---- --- ---- --- ---- --- ---- --- ---- --- ---

w

5;, C2I

54 mnV4144

Sb CII

50 CUPi

60l SAI

b2 CVSIb'1

ON CV,3?Ub

64 CVSA42

b5 MtIVfS4

66 stu1

67 5211.

OR hASh,

69 CVbS7

70 mnfl)bSA

73 CV30b

7; CVSPIO

14 CVSISO

ii S"121 76 CCI

77 SAIA

741 SI f311

70, cal.

III LP??

143 HAT??

OSi Clib

AR 10j117

41 14111

943 %*

4'. Ph

qh PA

loii' C's

"PIS d'1 A HPQ4 NP? i'1p

SALT ?il gunmi cfl'h.&R

tr~CP MEig.4"FCK VAIVF,

t(CC 0?l1,&17019 IIPFNAIFU VALVE. ErCn 9.?? 141171) nPFRAEAIU VALVE

kE.CR Ne? CHECK VALVE

SALT MZP WIN13 COOuLER

CI'W 3TA14II PUmP

CC"I CHEC4 VALVE

SIAS .¶IPICHA9NNFL Al

"PIS #21 '4JI709 (1PFO.ATFfl VALVt

SALT. 921 C0'4T`IEIIL VAL.VE

SALT 021 CO'dT11O. VALVF

CC"I C04'TknL VALVF

HpJs oil mnlTOin PEWATFO VALVE

SIAS SLIRCHANNFL "1

SALT oil PlIMI

HAS 31INHiAN4NEL Al

LPII C0I4IRUL VALVE

LP13 E.01 0 4JL VALVF

LPIN CU'4fUOL VALVE

HAS 31iCt4ANNEL RI SALT #21 CuNTROI. VALVE

SALT 921 CU'4TROL VALVF

SW 1121 PUMP

SIAS SURC14A'NNEL A3

STA3.311IRCHANNEL h1t

LPIS M0. A LPISS A? CHEC9' VALVt

LPIS X2??A LPQS $?? C4ECKp VALVI,

LPIS.822 A LPIEI *?2 P14P

ILPS HATTETIS. VI?

is's ItArFiFn MP2.

LOIS NP1 A tpqS 921 LEriCA VALVE LPlA9'I 40i A I.PQ3 92I CH~rth VALVE~ LPIS 821 A LPRb A?I P11NP

CE.. MHINIAL VALVF

C110 -4ANIIAL VALVF

Criri .4AII1AL VALvF

CrIbb 4A1111A1. VAI.VF

Cr^ !%A.111AL VALVF

Ar-jr 6*A;*UAL VALVI

AFtfr MAN11I1. VALUIr

AC-.% '4A-3I1AL VAL'F.

AF.,sS 4A'A4 11 A1. VALV3t

Ar1, L 9AV1111 VAV

AF.i% *j"',a4L VA1.'JL

Ar-15 "A'41AIL VALV.

AF-%S '4A1*IAI. VAL~t

AS0 'A'II-0-1AL VALWIL

4.40001f0-4( 7 3).,

4.40VUEE!04( 74)

3.SOOCIE-04( 77).

3.3000E-04.1.1A).

3.30OuE-a4( 79)

3.1000E-041 so),

3.I000E-!u41. 83)..

2.0OoOE-04( 84).

,?. ?0E01134 ( 85).

2.S000E-041 0M)

2.SOOOE-uI 87).

2.5000E-041 be)

2.IUOOEU94('41)

2.IIIOOE-04C 9?)

1.600OE-04( 91)

1.500flE-u4( 94)

1.1000fE-04( .95).

I.IOOOE-04(. 96)

a.1000E-04( 97)

I.U0ooE-04( 94)

7.PO0OE-Oi( 99)

7.OOOOE-05(100)

7.OOOOE-0U5(I0I)

4.lflOE-04( 15)

2. 40U001-06 102)

2.40001-0fI. C103

2.2000f-06(1U7)

2.2000E-06(tonJ

2.20001-06(l I),

2.21104ICl-UAC I).

e. 00u0E-0 I( I£6)

2. 00OUE-UhI I17)

h.hPoun-Qa( Sb)

6.hou0f11,41( Sri)

b.h00(1E-04C 53I)

5.30uft-03t( 37)

l.AAOC021j? 21)

I.b6001-Q?( 19)

3.hoooh1up( IS)

l.hpunE-u?( 14)

1.b0flEJ0LII( 14)

I.60uook-UP( Q)

lv.40011r-111 I Pl

4.13379[-0 2

4.1379E-0? 4.3379F-02

4.13791-02

4.3379E-02

4.33791-02

4 *3379E-02

4.33791-02

4.3379E-02

4.3379E-02

4.33791-0?

4,3379E-02 4.3379E-02

4.3379E-02

4.3379F-02

4.3379E1-02

4.3379F-02

4.3379E-02

4.3379E-02

4.3379E-02

4.3379E-02

4.3379E-02

4.3379F-02

4.3379E-02

?2.2321-03

4.3379s-02

4.1379F-02

41.3379E-02

4. ;379F-02

4.3379E-02.

*4.3379f -0e

4.31791-0?.

4. 3379FrOL..

4. 379E-02

.3379F-0?

0. 0000f *00

o.aoaurFno

o.nonor+OO.

0.o00nuf*nu

0. 0000F.00 0. 0000F.+0

o.uunfjF*onu

n.ooaoPnuEe O * (0no! 0F10 n .0uoOF*0u

0* 1100F*01.0

o.011n0F+no

I.0*71.0S

1.90871-05

1.90*71.O5

1 .43ISF-05.

1.43151-05

I.4315E-05.

1.34471-05

I .34471-OS.

I .?146E-05

1.0845E-05

I .084SE-OS

9. I 096E-06

9010961-06

7.80*21-06

6 .506RE-06

4. 77 17E-0

4.7711E-0b

4.77171-Gb

4.3379E-06

3. 03651.06 1.0365E -06

3.03651-06

9. 36071-07

1 .041IF-07

I .1411E-07

9,5434E-06

9,.*434F-0a,

9.54341-08

9.54 34L1-08

`1.5.414fOfl

JI.&ISAE-08,

A.6754F-06

8.67%66-08.

0. 0000F*00.

0.1104*001.G0

fl.no0010E00.

4 n4nr.0F0Q

0.4no uF~ou

4f11100001.00

0A~ uouF+0ft

0.~00001 .00

a 0000*00.

AO,0nnoF.* 00-

n.0n00c*6o

0. OUOUP*00

tw.0000F,09

0.n000111100

^ sxantSs^e t9s * *_xtwornR sq^ReFst
// Tabie 5. Plant name: Grand Gulf - Reactor type: BWR

14A14K Cl-IPLIN~tlhe41 .'CtINpt~II NT t4TS11 1MPACTI Cr'4PIMENT TYPE MAtlE OFCHAAI6E R1S3 IMPACT

tA4F , .'1FSCtCOITIr OF ClIPftltintdr S:PFRIGOICALLY fF COMPONENT OF CUMPOKENT

II*AVAILAOITLITY IFSTIn IINAVAILAbILITT AGIIG

(qANK) -- :CnNTTNIJlInSLY wTTI'4 FATLIENE

MoINTTnRFD RATF

/ ~~I FOU1AA ~5w.5t A I. btiUt £ A .SPIAS A *4fl~flW tlP~TEf V~t VE 7.l100lE-U4( 1) S 4.3379E-02 3.10h7E-OS

2 F0lUIH4K 5lS It 4 PHR H A SPU/S A 4nIhN IIPERATEn VAVLF 7.3ooaE-04( 2) s 4.3379E-02 3.1I66E-OS

* COUIAA S1SS A 0114P 6.000E-04C 3) it 4.3379F-02 2.90baF-OS

4 COPIMA :-13S H PlewP 6.7000E-u4I 4) a 4.3379E-02 2.90h4E-05

S FOIISAA t rS A HiliR npFwtAfFU VALVE 6.100GE-04( 53 S 4.3379E-02 2.SQb4E-05

* F0O516 bas S A HuTUR nPFWtATF0 VALVE b.1000E-U4( 7) S 4.3379E-02 2.9064E-O5

7 SAC SSrS a ACUIATT,114 AN Ca4TIRLt eIUCIIlT b.70COE-U4( 4) 4.3379E-02 2.906aE-oS

n 5'AC' 3SS A ACT(IATIOn AlJn CONTROLC IQCICIT. 6.7louoEo04g 10 S 4.%379E-O2 2.QO*6E-OS

* FnvfA SSW.S A CHECK VALVE 6.70QOEf-4( 11) S 4.3379E-02 2.9064E-05

10 FeIUl11 S3rS P CEfCK V ALV b6.70eOE-04( 123 3 4.3379E-02 2.9064E-OS

i1 FnI4AA iR144 A ;4oflh OPERATCED VALVE S.douOE-u4( 13 S *.3379E.02 2.5160E-05 l? FO60AA t1tt A ItOT11 UPEWATEn VALVE S.800DE-04( 14) S 4.3379E-02 2.5160E-@S

13 Fnl44FR R4lt H HtAnIEn ISPE.RATEI VALVF 5.AOOfEW-4( IS) S 4.3379E-02 2.516CE-OS

14 FnhAIIK WF4M N "tMOnR UPEPATE. VALVE 5%.000E-04( lb) S *.33?qE-02 2.SI10E-05

1% HATA EPS RAATEECT A 4.SOOOE-04( 21) S 4.3379E-02 1.4521E-oS

lb LRACT LPCS A LPCIS A A RHR B A TNT TATING LllSTC CIR4CUIT 3.300E-04 22) S 4.3379E-02 1.431sE-OS

17 "HCACT LPCIS C & LPCtS A 4 RHO H, INIrTATtNG LOGIC CIRCIIIT 3.00QOE-04( 21) S 4.3379E-02 1.3014E-O5

IA FOu3AA d4II A 4n0n14 OPERATEnOVALVE 0.8o00EA-04( 24) S 4.3379E-02 1,2146E-05

1q Fo474A kmM A MKOTOR OPERATEn VALVE 2.AuoOE-04( 25) S 4.3379E-02 1.2140F-O5

20 FOuIH MNH4R A 14nTnH OPERATEn VALVE 2.80OOE-04( 2b) 3 8.1379E-02 I.?14bE-O5

21 F0478E H1t d MOTOR UIPEPATED VALVE 2.8000E-04( 21) S 4.3379E-0z 1.214bE-OS

22 C00264 LPCIS 4 & MHR H PlO4P 2.600CE-04C 28) S 4.3379f-02 1.?14bF-05

0 23 FO0484 LPCIS A 4 WHO f MTlOf OPERATEn VALVF 2.ADUOE-04( 29) S 4.1379E-02 1.2146E-OS

24 FMIN LPCIS H 4 kh14 H CHECK VALVE 2.8000E-04( 31) S 4.3379E-02 1.214bE-OS

2% COu2AA LPCIs A A R4O A PA11NP 2.bOOOE-04( 32) s 4.3379E-02 1.1279E-OS

2b FO24AA 'NO4 A- 4ninm I)PERATED VALVE 2.bOOE-04J(' 3'0) 5 4.3379E-02 1.1279E-O5

27 FO2480 14HR H 4nTnfw UPERATE, VALvF 2.bOgoE:041 34) S 4.3379E-02 I.1279E-05

20 FE4IAA LPCI3 A A WHO A MOTOR 11PERATEn VALVF 2.boonE 4l 3%) S 4.3379E-02 1.1279E-O5

24 FO44AA WIHO A JnflfhI)PERATEn VALVF 2.bOOE-0o4( 36) ' 4.3379E-02 1.127qF-OS

30 FO4Rdr 14iiw I M4nTno UpEqATtn VAi.vE 2.0000E-041 37) S 4.3379f-02 1.1279f-05

31 FOSIA LPCIS A A RHO A CHECK VALVE Z..0OoE-04( 4h) I 4.1379E-02 1.127sE-oS

32 Cnul WMC ICIPomP I.oAooEco0 44) S 8.337F342 4.3379F-06

3 F0134 A -ItrSr3 4nOn UPFATVD VALVF 1.OoE-04( 49) s 4.3379E-02 4.3379F-06

34 F04SA OCIC3-4nInW UPEMRAM VALVF I;00o0E-04( 50) S 4.337VE-0? 4.31179F-ob

3% FnhAbA wcICs m4Tn0 oPERAQEn VALVF 1.400L-f04( 51) S 4.1379C-02 4.33741-0b

3h FEInA mrICS 4n0nw OIPFPATE.,I VAlVF 1.0U0DF-U4l ?S) S 4.337qE-02 4.1379E-Ch

37 'FOb4A hwric tinloI Upt RAItn VAi VF I.OOuDE-04 153) S 4.3379F-02 4.3319E-06

34'' Fn10h wrlrs i4ol)nst IUPFPATkn VALVF l.OO000F-041 54) s 4.3379F-02 4.5379E-O,

39 TTV NrICS IRIP ThPMoTILt VALVra 1.U0u0E-04( 55) 1 4.3379E-02 4.3379E-Ob

40 TIrv RCICS TIJRINt Gf)VE1'4lNl,; VALVI l.u4OE-U4( S6 a 4*.1379E-02 4.3317E-Cb

41 COu? WrICS Tlh6thlLt ' 1.000O-04( 57) S 4.3379E-02 4.3379E-0t

4? WACI Hrlrs AIIIATINI; CioCIIl l.OnOUOE-04( 53 S 4.4M79E-02 4.3319F:-Gh

41 fO40 Rr4IrS C4LCK VALVE I.uOuOF-o04 541 5 4.3179F-02 4.3379F-ob

44 Fleh 4qrtrs C4ECK VALVE l.UOUOE-04( bOI S *4.3579f-O? 4.437qf-ob

4% #I",% ,wir. :otrCK VAI.Vt I.000OL1041 ol) S 4.3374F.02 4.3379F-o0

41, e .314 .4r.Ir ricrK VALVI C0onuOL-04( b2. S 4.37r4r-02 4.3379E-6

47 hACT hl'C. ACT11IATIG rIPeeI IT b.500UL-OSt hO)' S 4.3379E-02 P.8196f-o0

4n POOIC 41l:S MelTlP OPFRIATFIl VALVE b.5A6OE-Q5l 743 S 4.3319F-02" -2?.I46E-Ob

49 FOtt? #IPLt. rhEi:X VLVF' - - h.SOoilF-usi 7A) S 4.S37QF-02 ?.AI*F-ot.

- sn C41C firc"S Pll 1P 6.S4UUE-15( 771 q '.3379E-OP 2.8196E0ob

-'. -. - , - . '. - . '. - .. ' .": -  - . - . 'o :. .- . . ' .1- ' - -- '. -, -  .- - -1 ... - -I - -11 '. "I ' 11 I - I
-U,--

Table 5. contd.

"&Ng Cvl"I01I4tMy Lfnolrtirtw.1 VISK 1A4PA1.T rfni-PONENT tYPF RATF nF CHANGE P1INK IMPACT

" qv [ I.r'4To1l IMI (I rn&'PnFlFNT SzeFwUfOnOI6A[Ly OF CONPfl'EENT nF CbMPONENT

IINAVAILA4HTI.TfY TFSTEn UNAVAILAIILTTY ACI4G

(QANp() olxenml4Unl3tLY WTTH FaIL.IRE

mfnTI~nI4Fu RATF _____________.______________________________________________________________________________________________.__________..._._______-

t

Ip-4

51 Fo04

5? Fn14C

5t F 111i

54 P

Sb Fn0l?AA

57 Fnl^A4I

5 OttsfdeI

Sq FOt?'f

6ht FOIMAAA

hI It ESELI

to? FOIl

bi 64ACC

btt SAACC

hSi t EIELI

b7 CfOMC

69t Fill C

bq F911IC.

To -sec. .

Ilt FOI7AA.,

72 FOSAA

73 Fn2,AA

74 PF0dA

75 F4t7RR

Th. FF521H

77 .FO26144

in, .vnl~j4s

rq7 Fnh41HR ro0, FPt4?

b? C0OCR

"I F014ACA;

All Fnt41

$S, Fn44;tn

tl* F001A 015 F O'a *gI

Illt PelA nA CMONI

R9 Foi I,

s4n Fnn1SA

91 FP011h.

9P Fn4tiA

93 Fn4UAA

4Y FA?7A4

q5 F?tIn

4*. r AN I 94. F?40A

99 FPI n.

ton F; lA

liwCs r4Fc'1 VALYF

f#~ H"jfJ)Q IPESJATIF, VALVIA

tPCS CHFCK VALVF

SAFFtY QELIVF VALVtS

fPS IAIFItY 14

SP04 A "OUfl7I 3 nPFMAJfII VlLVt

bSrtS tl u"niIR nPFhA fF1 VkLVL

tr.s WESEL OtNEt47*1Y tr2

;vPR4 R *MIT(I fPFWAIFDI VALVk.

S~l A '4UtUQ WF4A fF1 VALVE

FPS I)JESEL .lE4EQATIJR ii

Qrlrs Cmtrut VALVE

SPhS A ACTUATIOJN t tAI1NTu1tL CIttt 1t1

50! R ACTtIUATOR A CO64IQI'L C1rCIClT

FPS DtTEqLI rltNERATUP 05

LPS IIAVTERY C

StSW.CPU,4P

S.tW.5C CHpFCiK VALVF -

Sb3%,oC mul'ToI nPFRA(EIn VALVE

S.q,,q C ACTUATTION1 A CONTROL ClCclIT

u1 A mnrnI4 0JPERATEn VALVF

wwf A f4ntn4l 0LQATEn, VALVF

4R. A ,MnhnR. oPFPATEn VALVE

1464 A. CHtECK- VLVE..

kHM-H44 mnIOn OPERAThI). VALVF

a"k 1 04fiTnpl UPEAYTEnC VALVF

u1HtR et 0nTnfqttlPEQAT4Fn, VALVF

Iti d. CHErit VALV -

LPCTSC.CHFCK( VALVE

LPCIS-C 140nIn EiPEPAtIEn VALVF

LPCS C- C4ECK VALVfE.

LPCIS C POO4P

LPCT3 C i4n01R oPEPATFn VALVF

LPCIS II CHECK-VALVE

LPCTS r) llnTn4 oPtPATtn VALVF

LPfTS Ht 4nrnRO nPfRAfftl VALVF

LPC3 Mill'o nPFRATFU VALVE

LPC3 P1JMP

LPCs .rt.FC, VALvF

LPrtS 4UT(QlnPU11ATFI VALVt

LPCS,,FCX VALvE.,

LPCIS A. CItCEK. VLVE.

tITTS A t1niu OllPPrATFn VALVF.,

MLS A unf1t 1IIPfqATnf VALVF

''rtri M4AIIUAL VAlVr

Lut10 A. A. h4 A hAVIIIAL .VAIVF

'44Iu l.!, A* 411AI Val Yr

'441 A * 'A'AL VAI. vy.

.446.1 .4 * AII . V At. V

: 1 A I 1 4 *A % I A I' 6 4 v r

b.snOOE-091 76 3 6.50001E-05( 75) 6.5000F-tl( 75)

6. tfOOL-O9( 4413

9.SOOOA-oS( 723

4.4n00E-09( t42)

9.4000E-09( 70)

9.4600E-09( 693

6.4tOOE-OS( 03)

t.O1OOE-O4( b?I

1.OOUIOE-o0( bh)

l.OoE-o4( 65)

4.4000E-O09( 1463

4.40fOOE-OS( 853

2.7000E-0o( 116)

?.7nonE-Os( 87)

2.7000E-05( 9?)

2.?OOOE-o9( 93)

2.70VOE-o5( 95)

Z.7000e-os( 963

I .9000E-09( 97)

1.9000E-o5C 94)

I.90ooE-05('993

1.9000E-nIMt 02)

1.9 nooE-t,9(t ION)

t t.90OOE-05(104) bflOOE.09(t 03)

l.hOOlE-OS9 t 06)

1.6000E-0S 110)

I. I.&nOOOE-05(t .fnoOIE-n~ .0 C 10)9

1.4600OE.-05MIt1)

I. 600DE-09( 123)

t .AfOOE.-O5( tAR^

1.6000E-05C I U)

I .4n0fE-05(t 1h)

I1.4010OE-O5(l II?)

t .4000E-05(t 12)

I. 40 ofEf-t20)

I S.00PE-0 (I 23).

1 .5000E-O't?24)

I .%OooE-Eot(Izi)

.onoOF-0(.f 6%3)

d.hed"E~-04C 47'

2.hf1Onf-nO(' 69

'hinonf-n04(: 44) 0.b0, t:-04( '3)

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5

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A. 3 37 tE-02

6.33?9E-02

*.3,iqF_02

4.33?9E-02

4.3379E-02

'.3379E-02

*.3379F-02

*.3379E-02

4.3379E-02

4.3379E-02

4.3379E-02

4.3379E-2z

'.3379F-02

4. 3379E-02

4.3379E-02

4.1379E-02

*.3379E-02

*.3379E-02

6.0379E-02

4.3379E-02

4.337qE-02

'.3379E-"2.

*.3379E-02.

4.3379E-02.

4.337mr-o0

*.3379E-02,.

4.3379M-02

A. 3379E-02

43* .3?9E-02

*.3379E-02

4.337?9-02.

*.3379E-02

4. 3379E-O2

4.3379E-02

A. 337E-02

4.3379E-02

4.1379f-02

4.3S79E-02

4. 3379E-O2

4.3319E-02

4.3379E.02

*.33?9sE02

4.3379E-02

fl.0000F.,OO

0.0000o*Wa

0.n00flo'F+0

0.no0OaFoo

2;4196E-06

2.8 196E-06

2.8!96Ei-ob

a.646bF-Ob 4. 07 ThF-06

1.9087E-Ob

4* 07 76E-06

4.0776E-O6

I.9087EE-04

4.3379E-06

*.3379E-06

4.3379F.06

I .9087E-06

I .9087E-Ob

t .1712E-06

1.t7t2E-O6

t.I?12OE-O

1.1712E-Ob:,

l.1712E-06

4j4?O0E-o?

n.2420E-o?

8.2420E.07

R.24?DE-07

8.2420F-07

6.2420f-o?,

R.24POE-O7

6.9b 0E.OT.

b.9q06E.07

6.906E.-07

6.940bE-07:

6.940&E-07

6.940&f-07

6.9410WE-07

6.9406E-07

6.nIF.-07:.

6.07 3E-07

.07"IE-O7

4.073tE-07

6.0731E-07

4.SO68E-nO

6.50.SE-O0

6.5ObSE-O6

o.ooo0rEoo

0.0 OflOAOo

0.00O0fE4flO 0.00noF.00,

o.nonoF.OO

*n.OflOnF.OO
Table 5. contd.

WANK Cfhplp#"-t.UI CO c1,pfLfj .ISM IMPACT' CnNPONFNT TYPF 4ATE'nF CHANGE HT3K IMPACT

NAUP uI~~~~~C.E~~m'?iO'~~~~a OF rlamPoM.~t SaPFRIIIOICALLY OF'COMPONENT (IF COMPONFNT

AINAVAILAPIhLTTy IESTLfl UNAVATLAIITLJIY AGING'

(RANK) - I3CnNTINunlj3BLY WITH FAWIRE

mONITORED, RATE

'Ut

I05

IUW..

..I10

ton0

111

Ito

III

112

113,

ua4.

I'1S.

IthE

.117

Ito

119

120 1a1

I 22

123

1PA

IRS

F I0 1A

F I 1?A

F 029,A

Ft 344H

F INqt4

Fo?qC

FlailS

F I AhA

FEI its

F I #5A

F"116

WMI It MANUIAL VALVF

4Mk n .4AP411AL -VALVF

14UW A MA1NUAL VALVE

WNW A '4ANIIAL VALVF

LPtI3 I A WHR A S4AINIIAL VAtLVF

kHN it I4WIALI VALVE

"HP it 0MAiIIAL Vat VF

NHM A MANIIAL VALVE

kHM A '44NI141. VALt F

btA N MANIUAL VALVE

s85A A WAf1IJAL VALVE

LPCIS A MANuJAL VALVF

LPCS MANIUAL VALVL

LPCIb H MANIIAL.VALVF

LPCIS C MAmUAL VALVE

LPCI3 C MANuAL VALVE

biNS C MANUAL V&LVE

S3S" C ANIIAL VALV&

S.AW C M4ANU4L VALVE

5S"3 C MANUAL VALVE

53dS C MANUAL VALVE

HPCS MANUAL VALVE

SWIS H IANUJAL VALVE

.SA~S A 4ANU1AL VALVE

NfIC5 WANIIAI. VALVF

P.hDOUIE-fl4I 41)

2.bOGOE-04C 403

2.0OuOE-U4( 39)

Z.bO00E-U0( '30)

2.8000E-04( 301)

5.60OOE-04( 20)

5.0OGOE-04(,19)

5.8000EkU4( 1A)

'j.AO00E-O4( 17)

b.?000E-04C, 6)

I .50u'E -0t(tI 2?)

1.400OFFU5C121)

I .bGo0E-OO(t 12)

I. bGuOE-05 C I O)

21000Elh-O(USt 44)

2.7000E-05( 91)

2.JOOJOE-uSC 0O)

2.YOGOE-G%( OA4)

b.%flUOE-;0%( 75)

9.4flOPE-uSC 71)

I. 0000E.G04C 603

I.UOUDE-04( b4)

S

S

S

S

S

S5

S

S

S

.5

I

S

3

S

S

S

S

S

S5S

S

.5

3

S

0.01o (E.00

0.0000E oo

0.uoonE*oG

.l.OOOOE.+0

0.00oooE+0

n008GE. Oo

0.GoonE.00

0.o0oE .00

0.00t0OFno

0. 000GEoo 0.0onoE+oo

0. 000E*00

0.0o0oE*00

0.0000E.OO

o.OOOoF 00 0.oon0GoEo

0 0n00E.00

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0. 00GE.o0

0.00noE.o0

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0.0oooE.OO

f.UoonEsOO

0.00 OE+0o

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0.00 uoE ,00

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0.oG00r-,o0

'O.0GGQE.oo

O.0000E*00

0*0OOGE#00

O.OOOOF*00

G.00o0E,00

0. 0040GE*00

0.0000E .00

O.'OQOQE*00

0;0000EoG

OG OEO*O00

a~0 ooovF*0

0. 000 GE .0

90.000E .00

1). t00 GF..OG

0.GOGOE*00

0t.0 00 E+00

0,G000EO00

0. 0000OC 40

(.A~
In the last-part of-this section, the'rbsults''of the -two PWR's are

combined to give an overall -ranking-for PWR components'

3.3.1 Oconee

Table 6 shows the combined results of components of the same type and

system at Oconee. The component groups are ranked from highest to

lowest. The table shows that the component groups with the highest

potential risk impact are service water pumps, low pressure emergency

core cooling system motor operated valves and check valveslreactor

protection system circuit breakers, and engineered safety feature

actuation system actuators. ;

Table 7 shows the ranking for component types without differentiating

between systems. The'types-of components with the most potential risk

impact are pumps, check valves, actuation channels/trip modules, motor

operated valves, and circuit breakers/contactors.

3.3.2 Calvert Cliffs

'Table 8-shows the combined results for component groups at the Calvert

Cliffs. .:The component groups with the highest potential risk significance are all in the auxiliary feedwater system (check valves, motor

loporated'valves, and'pumps) and' the'reactor protection system'(circuit

'breakers and trip" relays). -

Table 9-shows the results of aging sensitivity measure calculations for

component types. Check valves have the highest potential risk . - jsignificance followed by circuit breakers, relays/actuation'subchan'nels,

,motor operated valves','-air operated control valves, and pumps.

3.3.3 Grand Gulf

Table 10,shows the combined results for component groups at the Grand

$ulf.' Motor operated valves'of the low pressure emergency core cooling

Jsystem and service water system and actuators of the engineered s'afety

actuation system have the highest potential risk impacts as measured by

-the--aging sensitivity measure.-

-Table 11 shows the ranking of the component types. Motor operated

tvalves, check valves, actuators, and pumps have the highest values of

'~the aging sensitivity measure.

-3-13
Table 6. Aging sensitivity of component groups at Oconee.

Aging Sensitivity

Rank Type System. (per reactor year)

1 Pump, Service Water ' 1.1 x 10-4

2 Check Valve Low Pressure ECC 9.8 x 10-5

3 . Circuit Breaker Reactor Protection 7.8 x 10-5

4 Motor Operated Valve Low Pressure ECC 7.1 x 10-5

5 Actuators Safeguard Actuation 6.3 x 10-5

6 Trip Modules Reactor Protection 5.2 x 10-5

7 Check Valves Auxiliary Feedwater 3.3 x 10-5

8 Contactor Reactor Protection 2.6 x 10-5

9 Pump Low Pressure ECC . 2.0 x i0

10 Motor Operated Valve High Pressure-ECC.. 2.0 x 10-5

11 Relief Valve Reactor Pressure Control 1.5 x 10-5

12 .-Control Valve (air operated) Auxiliary Feedwater - 1.2 x 10-5

13 Batteries' Emergency Power. 8.0 x-10-6

14 Check Valves High Pressure ECC 8.0 x.10-6

15 Pump Auxiliary Feedwater 6.1 x 10-6

16 Motor Operated Valve Auxiliary Feedwater 6.0 x 10-6

17 Pump High Pressure ECC 6.0 x 10-

18 Turbogenerator . Emergency Power 4.0 x 10-6

I

3-14 WI
/I

.: .:Table 7. Aging sensitivity of component types at Oconee.

Rank Component Type Aging Sensitivity,. % Contribution

1 Pumps 1.-4 x 10-4- 23

2 Check Valves i - 1.2 x 10-4 - 19

3 - Actuation Chinnels/TripModules 1.2 x 10-419

4 Motor Operated-Valves, 1.0 x 1O-4 16

5 Circuit Breaker/Contactor i 1.0 x 10-4"-- 16

6 Relief Valve 1.5 x 10;-5 2

7 Conirol-o.Valv6 (air operated) 1.1 x 10-5 2

8 Battery 6.7. x 10-6 1

9 Turbogenerator - 3.1 x 1-6' '1 *,~ .

- .. . . .: v . S ::: ... ; : v , ' . ! & ' ; .~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

I I

-' , I ...1

C.r

-.3-15
Table 8. Aging sensitivity of component groups at Calvert Cliffs.

t,

/1 Aging Sensitivity

Rank Type System (per reactor year.)

1 Check Valve Auxiliary Feedwater 5.5 x 10-3

-2 Circuit Breaker Reactor Protection 3.1 x 10-3.t:

3 Trip Relay Reactor Protection 2.1 x 10-3-

4 Control Valves (air operated) Auxiliary Feedwater 1.4 x 10-3

5 Motor Operated Valves Auxiliary Feedwater 1.4 x,10-3-

6 Pumps Auxiliary Feedwater. 1.4 x 10-3

7 Motor Operated Valves High Pressure ECC 4.5 x a10-4

8 Motor Operated Valves Service Water - 2.9 x 10-4

9 Diesel Generators Emergency Power 1.6 x10-4

10 Actuators Safeguard Actuation 1.6 x 107

11 Pumps Service Water 1.5 x 10-4

12 Motor Operated Valves Low Pressure ECC . 9.5 x 10-5

13 Check Valves High Pressure ECC 9.4 x 105

14 Check Valves Low Pressure ECC 8.1 x lo15 Batteries Emergency Power 6.5 x 10-5

16 Pumps High Pressure ECC 4.7 x 10-5

17 Room Coolers Service Water 3.3 x 10-5

18 Check Valves Service Water 1.3 x 10-5

19 Pumps Low Pressure 1.8 x 10-7

3-16
Table 9. Aging sensitivity of component types at Calvert Cliffs.

Rank Component Type Aging Sensitivity % Contribution

1 Check Valve 5.7 x 10-3 34

2 Circuit Breaker 3.1 x 10-3- 19

3 Relay/Subchannel 2.1 x 13

4 Motor Operated Valve 1.9 x 10-3 11

5 Control Valve (air operated) 1.7 x 10-3 10

6 Pump/Turbine Pump 1.6 x 10-3 9

7 Battery 2.6 x 10-4 2

8 Diesel Generator 1.6 x 10-4 1

9 Room Cooler 3.3 x 10-5 1

Table 10. Aging sensitivity of component groups at Grand Gulf.

.Rank Type System Aging Sensitivity

1 Motor Operated Valves Low Pressure ECC 2.3 x 10-4

2 Motor Operated Valves Service Water 1.3 x 10-4

3 Actuators Safeguards Actuation 9.9 x 10-5

4 Pump Service Water 5.9 x 10-5

5 Check Valves Service Water 5.9 x 10-5

6 Motor Operated Valves High Pressure ECC 5.4 x 10-5

7 Check Valves High Pretsure ECC 5.4 x 10-5

8 Check Valves Low Pressure ECC 2.8 x 10-5

9 Batteries Emergency Power 2.4 x 10-5

10 Pump Low Pressure ECC 2.4 x 10-5

11 Pump/Tubine Pump High Pressure ECC 1.3 x 10-5

12 Diesel Generator Emergency Power 9.5 x 10-6

13 Relief Valves Reactor Coolant Pressure 2.6 x 10-6

Control

3-17
Table 11. Aging sensitivity of component types at Grand Gulf.

Rank::: Type Aging Sensitivity % Contribution

1 - Motor Operated Valves 4.1 x 10-4 52

2 Check Valves 1.4 x 10-4 18

3 Actuators 9.9-x 10-5 13

4 Pump/Turbine Pump 9.6 x 10-5 12

5 Batteries 2.4 x 10-5 3

6 Diesel Generators 9.5 x 10-6 1

7 Relief Valves - 2.6 x 10-6 1

I... 

. f3-18
i 2 A r~t-ne DLIDI'- .J.*I.t \UIIILuIU | flIU r HA (Z ' AmJ -I

This-section combines the-results of the analysis of the two PWRstot " .determine an.overallPWR.rankin'.' .The.Grand Gulf .results are-assumed

typical of a BWR since information was only available for one plant.

..Table 12-presents the-aging sensitivity.rankings.for component-groups--at

PWR's. These results are obtained by adding the results of the

component groups at the two PWR's. Check valves of the auxiliary, , feedwater system and breakers/contactors and trip relays/trip modules of

the reactor protection'system have'-the highest-potential risk impactfas

measured by the aging-sensitivity measure. .

Table 13 presents the'combined-results for'component types of the two

plants. -Check valves, circuit breakers/contactors,itrip modules/... . actuation channels, 'motor operated valves, pumps;1 and air operated

control valves have':the highest values of the aging sensitivity measure.

3.4 Additional Components

In-this.section we estimate Rthe-aging.sensitivity measure for three.

additional component types: 'the-reactor vessel,Fsteam generator tubes,

and snubbers using existing PRA's'and related studies;-t:Thb calculations

in this section are bounding calculations intended to compare the

.impor'tance of these components to other components at the plant.''Table

14-presents the results of these 'calculations.. The following-paragraphs

discuss the assumptions-and.implications of the analyses. ,. -

3.4.1 Reactor Vessel .. -..... -- .;..

The reactor vessel has the highest potential impact on risk of.any*

component in the plant.''PRA's generally make the conservative assump-

'tion that a failed reactor vessel results in an uncoolable configuration

that leads to core meltdown.. The ,aging impact as measured .bythe aging

,sensitivity measure'is high cbmpared'to the.other components'in'the

plant.- --

3.4.2 Steam Generator Tube j.'

A rupture in a steam generator, as-an initiating event, results in a

small LOCA and consequently loss of heat removal capability of one steam

generator. In this-situation,--core cooling requirements-generally are

the operation of the auxiliary feedwater system and at least one high

pressure injection pump. Table 15 gives an estimate of the tube aging

impact based on the cooling requirement for four plants. Consistent

with the aging sensitivity measure definition, these estimates are based

on simply adding the conditional failure probabilities of the auxiliary

feedwater system and the high pressure injection system. The average

value from these four plants is included in Table 14. The potential

a.3-19
Table 12.- Aging sensitivity of component groups in PWR'fs.

Rank Type System Aging Sensitivity

I Check Valves Auxiliary Feedwater 5.5 x 10-3

2 :,.Circuit Breaker/Contractor Reactor Protection 3.2 x 10-3

3 Trip Relay/Trip Module Reactor Protection 2.2 x 10-3

4 Control Valves (air .operated) Auxiliary Feedwater 1.4 x 10-3

5 Motor Operated Valves' Auxiliary Feedwater 1.4 x 10-3

6 Pumps Auxiliary Feedwater 1.4 x 10-3

7 Motor Operated Valve High Pressure ECC 4.7 x 10-4

8 Motor Operated Valve Service Water 2.9 x 10-4

9 Pumps' Service Water 2.6 x 10-4

10 . Actuation Channels .SSafeguards Actuation 2.1 x 10-4

11 Check Valve Low Pressure ECC 1.8 x 10-4

12 .-Motor Operated. Valve Low Pressure ECC 1.7 x 10-4

13 Turbo Generator/Diesel Emergency Power 1.6 x 10-4

Generator

14 Check Valve High Pressure ECC 1.0 x 10-4

15 Batteries Emergency Power 7.3 x 10-5

16 --'Pumps High Pressure ECC 5.3 x 10-5

17 Room Coolers Service Water 3.3 x 10-5

18 Pumps Low Pressure ECC 2.0 x 10-5

19 Relief Valves Reactor Coolant Pressure 1.5 x 10-5

Boundary

20 Check Valves Service Water 1.3 x 10-5

3-20
Table 13. Aging sensitivity of component types in PWR's.

Rank Type Aging Sensitivity

1 Check Valves 5.8 x 10^3

2 Circuit Breaker/Contactor- 3.2 x 10-3

3 Trip Module, Relay/Actuation Channel 2.4 x 10-3

4 Motor:Operated Valves 2.3 x 10-3

5 Pumps 1.7 x 10-3

6 Control Valves (air .operated) 1.4 x 10-3

7 Turbo Generator/Diesel Generator 1.6 x 10-4

8 Batteries 7.3 x 10-5

9 Room Coolers 3.3 x 10-5

10 Relief Valves 1.5 x 10-5

-~O~- G 4 4J- I. 0 o

1~~~th .,,,, IC' ..... ~~~ . ........ ~ ~~ 

3-21
4 - l~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Table 14. Aging sensitivity measures for

selected components.

Component Aging Sensitivity

Reactor Vessel 1

Steam Generator Tube 3 x 10-3

Snubber 1.8 x 10-5

Table 15. Aging sensitivity measure calculations for

steam generator tubes.

Plant Name Cooling Requirements Aging Sensitivity

ANO 1/2 EFWS 6.5 x 10-4 +

1/3 HPIS 4.0 x 10-4=

1.1 x 10-3

Oconee 1/2 AFWS 2.4 x 10-4 +

1/3 HPIS 1.4 x 10-3

1.6 x 10-3

Calvert Cliffs 1/2 AFWS 3.0 x 10-3 +

1/3 HPIS 1.7 x 10-3 =

4.7 x 10-3

Sequoyah 1/3 AFWS 4.3 x 10-5 +

1/3 HPIS 3.5 x 10-3 =

3.5 x 10-3

3-22
risk impact of-steam generator tubes-as-measured by the aging

sensitivity measure-is higherthan that of the standby components -'

lanalyzed in Section 3.2.' '" -

3.4.3 Snubber

In order to determine th`e aging impact of snubbers we reviewed the

results of the Seismic Safety Margins Research Program (7).' Thee case of

snubber failure is specific in that it has been done for the Zion'plant

based on the information given in Reference (7).

The risk associated with snubber failures is characterized by an

increased likelihood of a LOCA induced by an earthquake. The earthquake

also degrades the safety system that cools the core in the event of a'

LOCA. In this situation, it is assumed that snubber failure will result

in a large or medium LOCA for any earthquake with a magnitude larger

than design'basis.' The d6minant core melt-sequences for an earthquake

induced LOCA contain failure of the Safety Injection System (SIS) to

cool the core. A risk impact of the snubber failure is estimated by the

following computation:

6

BR _ i=

aq i=1 a LOCA1 SIS1 (17)

where ' * ,, , - ; *;; **; , , - 1 , I

*- ai = The earthquake frequency. .

LOCAi = The LOCA probability given an earthquake

range of as

is in the

SIS- =,The probability of SIS failure 'given an earthquake is

in the range of aj.' : - -

The summation is -over the six accident sequences identified inReference' (7). : ' ' '' 'A ' ' ' - '

Consistent with'the definition of risk 'impact, -the snubber-is'assumed

failed. >Since the purpose~of the snubber is to 'prevent:piping failure,,

this implies LOCAi = 1 in Equation (17). Now', using'the values of ai;

and SISj given in Reference (7) the risk impact of the snubber failure

is calculated from Equation (17). ') '- :'.-

I : , ,i ,  7  ;,I

-3-23
Earthquake Conditional SIS

Frequency, ai LOCAi Failure Probability

2.52 x 10-4 x 1 x 4.7 x 10-2 +

4.55 x 10-5 .; . x 1; x 1.2 x 10-1 +

6.57 x 10-7 x 1 x 2.6 x-10-1 +

1.61 x 10-7 x -' . x 5.0 x 10:4. +

5.31 x 10-8 x 1. -x 7.5 x 10-1 +

4.10 x 10-8 x 1 x 9.9 x 10-1 = 1.8 x 10-5

Hence

- = 1.8 x 10-5 per. reactor year

If snubbers are tested every year as recommended, then

year

The aging sensitivity measure for snubbers as' calculated in-this manner

is moderately high when compared to the other results in Section 3.2.

This calculation is an approximation and subject to high uncertainty.

Further, the information used is for only one plant that is not located

in a high seismic activity zone. The potential risk significance of

snubbers will be very site-dependent in general.

3.5 Limitations and-Assumptions

s 0'The:analysis > performed- for..this report- is limited bythe-available

v > information as well as time and budgeting constraints. Further, the

inherent uncertainties in PRA's are limiting factors in identifying the

most important-components., The-results presentedin-this-.section are

also subject to the uncertainties inherent in PRA's including component.

failure data uncertainties, modeling uncertainties, and uncertainties in

human actions: and response.. -The particular PRA's utilized to determine

the. component results.did not..include-treatment, of all aspects of risk

such as--seismic analyses, fires, tornados,.etc.

The most important limitations of this study are the.limited~number of

plants analyzed and limiting the.scope of components studied to those

analyzed in the PRA's. The analysis is limited'to the effects of

-complete failure'(loss of function); the effects-of degradation are not

specifically addressed." Also common-cause failures attributed to aging

are 'not specifically addressed.

3-24
This report considers only some of the components that are potentially

important to risk. We did not consider components whose primary purpose

is to mitigate the consequences of severe accidents such as containment'

spray nozzles, piping and pumps. The importance to risk of components

that mitigate accident consequences is not easy to determine in light of

the large uncertainties associated with the phenomenology and fission

product behavior of severe accidents. We did not consider structural

components such as the containment and containment lining. Piping and

wiring are not explicitly considered in these analyses and components

such as the reactor vessel, steam generator tubes and snubbers are

treated only superficially for example purposes.

3-25
4. CONCLUSIONS AND RECOMMENDATIONS -.- 7; -,

In this section we draw conclusions from the results of the aging sensitivity calculations and make.several recommendations for utilization of

the results. ,' . - '... ,. .

4.1 -Conclusions . - . . -

The information presented in arstandard PRA-does not include time.-...-

Ldependent effects. In determining-.the risk level at.aplant,-PRA'.s

iJ' generally use a time'averaged unavailability. -Aging .issuesdeal with :.t

the time dependent nature of-risk..,:This.-limits.the.nature-ofthe:t,,.

nformation that can be extracted from a PRA without extensively

tModifying-the:1PRA. -This'report suggests a-method-for determiningjthe F"'

potential-risk-significance'of aging effects that ,is based on deter-.., , mining the sensitivity.of.'risk to-increases.in-failure rate.- This-.

adaptation of:PRA-results enables-'us~to,..identify.-the.components ,that *-,

..have the most'signif.icant..impact-on risk if.:theirfailurerrates increase;

due to aging or 'service' wear.,effects witiout-itdescr *

ehvior of!,the failurei e infrmationextrac ed from>,:

s 'In can be quifeusefull-in guiding researchefforts if.,'1.

used in context. . . ' -' -

-:.-The results of the analysis indicate :themost risk significant components at a plant depend on a number of factors'including plant system

design, testing, and maintenance intervals and operating procedures.

The key components with regard to risk can be different at each plant':,

owing to differences in system design or testing, maintenance andoperating practices. -

Based on the component resu'lts in' Section' '3''any'of thd`jotfntihlly most

risk significant components are-in the.auxiliary feedwater system, the

reactor protection system and the~service~water.systems. Pumps, check

valves;'m t'orjoperated valves, circfuit breakers,', an'd, 'act-uathg''circuits'

are the'component types thiat'hav'e the'mo'st"potential'risk-impact-basied

on the aging sensitivity'mea''sur'e';"These,tesiult' m'us't be coupled with';''

time-dependent 'failuretrated cha'racteristic~s t'cobmplet& th6'rHk impact'

due to component"'aging. ,; * ,,). - - -

., .I. . a" {Ag ,i 3s t'sloe ;!-: 

Components !not "analyied 'i'n PRA's-orcompobnents 'ass'ui ed'tdthave neg-'',

ligible failure 'rates' caa nbe''lmfOritanf-to rHisk ifitheir fail'ure'rate's t

increase substantially., Research programs are already in place for some

of these components''su.chas the reactor vessel, rector~coolant piping,

and'steam 'generator 'tubes. -I '' '' 't -

4.2' Recorrmendations" " -' ii,.

4.2.1 Use of Results

The risk aging sensitivity.defined in this report is a measure of the

sensitivity of risk to changes in component' failure'r'ates'.' aa.Thos-e-'i " * - :: ! - t i ' / '' -' - ; , ; i * ) '', a' a

4-1
components with the highest aging sensitivity cause the greatest impacts

on risk if their failure rates increase substantially.

*Thes'e results:are intendedto.provideguidance.to the-selectionof

.components for further study and as a guide toward prioritizing

resources. Three levels of results are provided. We recommend using

the results of the third level (component type rankings) as a ranking of

the most important component types. To focus research further we

recommend concentrating efforts on a particular component type (such as

motor operated'valves) or the type of'operating- environment typical of

the systems that have the highest-potential impact for that component

type (the auxiliary-feedwater system for example).

> 4"These-results-make'.no-assumptions -about which-components are:most '

"s "?susceptible to aging processes. The significance-of a-aging mechanism

*can be.obtained-by combining.,the~-risk, aging'sensitivity.as'.presented'

here with estimates'of the increaseJ-inlthe:time-dependent failure rate,-

Estimates of-time dependent-failure'rates can be obtained from experimental programs, analytical models or operating history. Ideally, if an

equation for time dependent failure-rate-were obtain'able (from an

analytical model or a data correlation) the time dependent risk

associated with a component can be approximated by:

Ri(t) =G Xi(t) (18)

where

Ri(t)"= The time dependent risk and

Xi(t) = the time dependent failure rate.

The risk,increase'associated~withithe.aging processcould.be quantified

by integrating Eqatii. n,(18).oVer .the time period of 'interest. ."In'

; practice a:.goodestimate bof-time dependent failure I rate willbe..

difficult to-obtain,',- Forpr'ioritization with respect' to aging .it:is

sufficient'to focus resources on those components thatahavepotentially!

high impact on risk (as measured by the aging sensitivity measure) and

also have failure ratesthat are most affected by aging and service wear

effects (as determined by data, analytical-or experimental studies).

g v We recommend.limitedadata or analytical.studies for.,each class of

component to determine if any aging or service wear~effects are evident

a from the available data bases. A more extensive analysis can evaluate

those components that have a relatively high potential risk significance

Up and exhibit some evidence of age related degradation.

4.2.2 Interfaces

.The aging progr'am in..general.and the risk significance task in particular can benefit from the products of other NRC and industry programs

4-2
including the Accident Sequence.Evaluation Program (ASEP) and the data

gathering programs-LER'sNPRDs, 'and others). 'The' ASEP program is

designed to provide analysis of the dominant accident, sequences for most ) -i - LWR's in the United States. As-a-part of thiis program 'the cutsets' for

the dominant sequences will be identified and risk importance measures

will be calculated for a large number of components. When the-results

are made available'it will be poss~ible to apply'the methods outlined' in

this report to-a'broad,range of plants. This will provide a good basis

for assigning''priorities to'component' classes based-on 'th'e 'risk esti;-

mates at a large number of plants rather than the three' analyzed here.

The approaches used in ASEP will allow identification of the most risk

significant components and'syste-ms' based on''plant' design and other*

operating characteristics. This information will assist in'making

specific recommendations as towhat type of inspection and preventive

maintenance 'pograms 'will be mbst',effective in controlling risk atdifferent plants based on plant'design.

4.3 Suggestions for Future Work

The risk aging sensitivity measure identifies the potential risk impact

of components in nuclear power plant'PRAs. This provides direction for

evaluating aging effects; however, there are other important issues that

must be addressed to fully understand aging phenomena.

A necessary complement to the risk aging sensitivity measure is a

description of the time-dependent effects of aging on component failure

rates.. Initial estimates of these effects could possibly be estimated

from older plant operating history and component failure data. A

complete description will include:

(1) Identification of component types that are susceptible to

aging

(2) The environmental conditions and system applications that

influence component aging

(3) Time-dependent functions defining component failure rates.

This study recommends these factors be investigated first for the

components that have high potential risk impact as determined by the

risk aging sensitivity measure Sensitivity calculations employing

Weibull type aging functions(c) based on current knowledge of relative

material aging rates could further focus this research effort.

Investigation of components that do not appear in PRA dominant cutsets

is also necessary. The basic effect of aging phenomena is changes in

component failure characteristics. Components now believed non-dominant

;in PRAs can become major contributors-to risk when they are susceptible

to significant aging. Identification of sensitive component types and

important environmental conditions will provide direction for

identifying these components.

4-3.
Other areas where aging effects can influence risk include:

,(1) Common' cause failures among components that have similar aging

susceptibility:

(2) Ability"of compiponenttesting to detect,aging effects

(3) Ab'ility of repair efforts to compensate'for age-related

deterioration

'(4) Aging effects and external events such as earthquakes and

floods.

A well-defined effort to investigate these concerns'.will provide a

better understanding of the effects of aging phenomena.

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

I

4-4
5.0 REFERENCES

(1) "Reactor Safety Study, An Assessment of Accident Risks in

U.S. Commercial Nuclear Power Plants," WASH-1400. (October 1975).

(2) Birnbaum, Z. W., "On the Importance of Different Components in a Multi

component System" in Multivariatic Analysis--II, Krishnaih (ed.)

Academic Press, New York, 1969.

(3) Vesely, W. E., Davis T. C., and Saltos, N., "Measures of the Risk

Impacts of Testing and Maintenance Activities," NUREG/CR-3541, 1983.

(4) Kolb, G. J., et al., "Reactor Safety Study Methodology Applications

Program: Oconee #3 PWR Power Plant," NUREG/CR-1659/2 of 4

(January 1981).

(5) Kolb, G. J., et al., "Reactor Safety Study Methodology Applications

Program: Calvert Cliffs #2 PWR Power Plant," NUREG/CR-1659/3 of 4

(May 1982).

(6) Hatch, S. W., Cybulskis, P., and Wooten, R. O., "Reactor Safety

Study Methodology Applications Program" Grand Gulf #1 BWR Power

Plant," NUREG/CR-1659/4 of 4 (October 1981).

(7) Smith, P. D. et al. "Seismic Safety Margin Research Program," Phase 1,

Final Report Overview. NUREG/CR-2015, Vol. 1, April 1983.

(8) Burdick, G. R., Note to J. Vora, August 29, 1984.

5-1
APPENDIX A

AGING SENSITIVITY OF OCONEE COMPONENTS GROUPED BY TYPE AND SYSTEM
Table A-1. Aging sensitivity of Oconee

grouped by type and system.

components

- *, < . I1 ; *' ' i'

Component ' ,,' Component Aging Sensitivity

Type System Designator (Per Reactor'Year)

Pump LPSW LPSW-P3B

VP1

LPSW-P3A

VP2

. : 2.3

2.3

3.3

3.3

x

x

x

x

10-5

10-5

10-5

10-5

LPIS & ECCR LP-P1A

LP-P1B

1.0 x 10-5

1.0 x 10-5

HPIS

AFWS

;7 .

f % '

. I

HP-1AB

HP-1C

3.2 x 10-7

3.9 x 10-6

.1

EFP-A

EFP-B

EFP-TD

3.0 x

3.0 x

8.7 x

10-6

10-6

10-8

Valve

Motor Operated LPIS & ECCR - LP-17

- -. LP-18

" -' LP-5

LP-8 .- ~ LP-22

, LP-21

ECCR , i LP-19

-, LP-20

1.0

1.0

1.0

1.0

1.0

9.0

x

x

x

x

x

x

10-5

10-5

10-5

10-5

10-5

10-6

-I; 6.0 x 10-6

6.0 x 10-6

HPIS

AFWS

- r..-

,- e t5

.I ,,I

. _ i .~ I

HP-24

HP-26

HP-27

HP-25

FDW-372

FDS-382

C-156

6.0

6.0

4.0

4.0

x

x

x

x

10-6

10-6

10-6

10-6

10-6

10-6

10-8

-'.~ ~, 

.III I-,

r,

. . . I 3.0 x

3.0 x

8.6 x

---. - - -LPSW & AFWS LPSW-137' -- - "10-8 ---- 

.A-i
Table A-1. Continued

Component Component Aging Sensitivity

Type System Designator (Per Reactor Year)

manual LPIS & HPIS LP-28 0

LPIS & ECCR LP-11 0

LP-15 0

LP-13 0

LP-16 0

Test A 0

Test B 0

HPIS HP-101 0

HP-118 0

HP-148 0

HP-114 0

HP-111 0

AFWS C-575 0

C-576 0

MS-90 0

MS-91 0

FDW-88 0

C-157 0

Check LPIS & ECCR CF-12 1.0 x 10-5

CF-14 1.0 x 10-5

LP-31 1.0 x 10-5

LP-12 1.0 x 10-5

LP-48 1.0 x 10-5

LP-33 1.0 x 10-5

LP-14 1.0 x 10-5

LP-47 1.0 x 10-5

LP-30 9.0 x 10-6

LP-29 9.0 x 10-6

AFWS FDW-232 6.0 x 10-6

FDW-317 6.0 x 10-6

FDW-233 6.0 x 10-6

A-2
'Table A-1. 'Continued

Component Component Aging Sensit'ivity

Type System , Designator (Per Reactor Year)

Check (Continued) AFWS

(Continued)

FDW-319'

FDW-373

FDW-370

FDW-383

FDS-380

6.0

3.0

3.0

3.0

3.0

x 10-6

x 10-6

x 10-6

x 10-6

x 10-6

HPIS HP-113

HP-102

4.0 x 10-6

4.0 x 10-6

Air Operated AFWS FDW-315

FDW-316

MS-93

MS-87

MS-94

MS-95

6.0

6.0

8.6

8.6

8.6

8.6

x

x

x

x

x

x

10-6

10-6

10- 8

10-8

10-8

10-8

Relief SRS Q 1.5 x 10-5

Contactor RPS RPS E

RPS F

1.3 x 10-5

1.3 x 10-5

Circuit Breaker

Remote Trip Module

RPS CB A

CB B

CB C

CB D

2.6

2.6

1.3

1.3

RPS RTM

RTM

RTM

RTM

1

2

3

4

1.3

1.3

1.3

1.3

x

x

x

x

x

x

x

x

x

x

x

x

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-6

06.

Actuation ESFAS CH 4

CH 3

CH 1

CH 2

4.3

1.0

6.0

* 4.0

'A- 3
Table A-1. Continued

Component Component Aging Sensitivity

Type System Designator (Per Reactor Year)

Battery EPS DC BAT A 4.0 x 10-6

BAT B 4.0 x 10-6-

Turbogenerator EPS AC I TG 1 2.0 x 10-6

TG 2 2.0 x 10-6

A-4
Table A-2. Aging sensitivity of- Calvert

grouped by type and system.

Cliffs components

Component'' Component Aging Sensitivity

Type System Designator. (per reactor year)

Pump AFWS TP21

TP22

6.88 x 10- 4 ,

6.88 x 10-4 -

SWS S22

SW22

CC21

CC22

S21

SW21

6.5

6.5

9.4

1.3

8.0

3.0

x

x.

x

x

x

x

10-5

10-5

10-7

10-5

10-6

10-6

HPIS & ECCR

,S. &

LPIS & ECCR

HP21

HP23

2.8 x 10-5

1.9 x 10-5

LP22

LP21

9.0 x 10-8

9.0 x 10-8

Valve

Motor Operated AFWS MOV-4071

MOV-4070

6.88 x 10-4

6.88 x 10-4

HPIS MOV-659

MOV-660

MOV-656

MOV-654

1.9 x 10-4

1.9 x 10-4

1.16 x 10-5

9.03 x 10-6

SWS CV-5152

CV-5153

CV-5212

CV-5162

CV-5208

CV-5160

CV-5206

CV-3824

CV-5210

CV-5150

HPIS & LPIS- MOV-4143

& ECCR - MOV-4142

6.45 x 10-5

6.45 x 10-5

6.45 x 10-5

2.84 x 10-5

2.84 x 10-5

1.1 x 10-5

1.1 x 10-5

1.1 x 10-5

3.0 x 10-6

3.0 x 10-6

2.8 x 10-5

2.02 x 10-5

A- 5
Table A-2. Continued

Component Component Aging Sensitivity - Type - System . Designator (per reactor year)

Valve

Motor Operated

(Continued)

ECCR

LPIS

Manual AFWS

MOV-4144

MOV-4145

CV-657

MOV-658

CV-306

C3

C4

P1

P4

S6

P2

P6

S8

Hi

H2

Mill

M105

M106

M107

M108

Mo10

M113

M114

M116

M9A

M28A

1.9 x 10-5

1.4 x 10-5

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

0

SWS

HIPS & ECCR

LPIS & ECCR

M30

M47

M32

M51

0

0

0

0

M34

M54

M55

M28

0

0

0

0

A-6
Table A-2. Continued

Component . Component Aging Sensitivity

Type' System Designator (per reactbiY-year)

Manual

(Continued)

LPIS & ECCR

(Continued)

M42

M43

Air Operated

Check

AFWS .I CV-4511

CV-4512

6.88 x 10-4

6.88 x 10-4

AFWS

HPIS & L

& ECCF

HPIS & E

P3

S5

P5

S7.

H5.

H6

S3~

S4,

I S, C65

;C 66

:cR C37

C64

C39

C61

6.88

6.88

6.88

6.88

6.88

6.88

6.88

6.88

x

x

x

x

x

x

x

x

10-4

10-4

10-4

10-,44";.'

10-4

10-4

10-4

2.8 x 10-5

2.0 x 10-5

I

2.8

2.8

1.9

1.9

x

x

x

x

10-5

*io-5

ECCR C21

C20

1.9 x 10-5

1.4 x 10-5

SwS

LPIS & ECCR

C115

C41

C63

C35

C56

1.3 x 10-5 I- ..... ..... .

9.0

9.0

9.0

9.0

x 10-8

X--10-8'- -

x 10-8

X 10-8

Trip Relay RPS K1

K2

K3

K4

5.2

5.2

5.2

5.2

x

x

x

x

10-4

10-4'

10-4

A-7
Table A-2. Continued

Component Component Aging Sensitivity

Type System Designator (per reactor year)

Circuit Breaker RPS 1A 3.9 x 10-4

2A 3.9 x 10-4

3A 3.9 x 10-4

4A 3.9 x 10-4

1B 3.9 x 10-4

2B 3.9 x 10-4

3B 3.9 x 10-4

4B 3.9 x 10-4

Actuators ESFAS (for SWS) SIB7 8.2 x 10-5

(for HPIS) SIA2 2.8 x 10-5

(for HPIS) SIB2 2.0 x 10-5

(for HPIS) SIA1 1.2x 10-5

(for HPIS) SIBI 9.0 x 10-6

(for ECCR) RASA1 6.5 x 10-6

(for ECCR) RASBI 4.3 x 10-6

(for LPIS) SIA3 1.0 x 10-7

(for LPIS) SIB3 1.0 x 10-7

Battery EPS DC BAT21 6.5 x 10-5

BAT12 9.0 x 10-8

BAT22 9.0 x 10-8

Diesel EPS AC D12ST 9.0 x 10-5

D21ST 6.5 x 10-5

Room Cooler SWS R21 1.9 x 10-5

R22 1.4 x 10-5

A-8
Table A-3. Aging sensitivity of Grand Gulf

grouped'by'type'and'system.

components

Component Component Aging Sensitivity

-Type System Designator (per reactor'year)

Pump SSwS - :-

RHR &-LPCIS

COOlA-A

COO1B-B

C002-C

C002B-B

C002A-A,

2.9 x 10-5

2.9 x- 10-5

1.2 x 10-6

1.2 x 10-5

1.1 x 10-5

IRCICS Cool 4.3 x 10-6

HP CS

LPCIS

LPCS

COO1-C

C002C-B

COO1-A

2.8 x 10-6

6.9 x 10-7

6.0 x 10-7

Valves '

Motor' Operated SSWS FOO1A-A

FOO1B-B

F005A-A

F005B-B

F018A-A

F018B-B'

FOl-C

3.1

3.1

2.9

2.9

4.3

4.0

1.2

x

x

x

x

x

x

x

, . . .. I

. - 1. .

RHR

k - , I . I I 

F014A-A

F068A-A

F014B-B

F068B-B

F003A-A

F047A-A

F003B-B

F047B-B

'-'; F024A-A

F024B-B

F048A-A

F048B-B

F087A-A

F052A-A

F026A-A

F087B-B

2.5 x

2.5 x

2.5 x

2.5 x

1.2 x

1.2 x

1.2 x

1.2 x

1.1 x

1.1 x

1.1 x

1.1 x

8.2 x

8.2-x8.2 x

8.2 x

10-5

10-5

10-5

10-5

10-6

10-6

10-6

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-5

10-6

10-6

10-7

10-7

10-7

A- 9
Table A-3. Continued

Component Component Aging Sensitivity

Type System Designator (per reactor year)

.. .. ..R .R

Valves

Motor Operated

(Continued)

RHR

(Continued)

RHR & LPCIS

RCICS

F052B-B

F026B-B

F004B-B

F004A-A

F013-A

F045-A

F068-A

F010-A

F064-A

F063-B

TTV

TGV

6.0

6.0

6.0

6.0

6.0

6.0

6.0

6.0

8.2 x 10-7

8.2 x 10-7

1.2 x 10-5

1.2 x 10-5

x

x

x

x

x

x

x

x

10-6

10- 6

10-6

10-6

i0-6

10-6

10-6

10-6

HPCS

SPMS

LPCIS

LPCS

F004-C

F001-C

FOO2A-A

FOO2B-B

2.8 x 10-6

2.8 x 10-6

1.9 x 10-6

1.9 x 10-6

F242-B

F004C-B

F042B-B

F027B-B

F042A-A

F027A-A

6.9

6.9

6.9

6.9

6.0

6.0

x

x

x

x

x

x

10-7

10-7

10-7

10-7

10-8

10-8

FOO-A

F005-A

Manual SSWS F199A

F199B

F023A

F023B

F185A

F185B

6.0 x 10-7

6.0 x 10-7

0

0

0

0

0

0

A- 10
Table A-3. IContinued

Component - Component Aging Sensitivity

Type - System Designator (per reactor-year)

Manual

(Continued)

SSWS

(Continued)

F186A

F186B

F013

0

0

0

RHR F130A

F120A

F130B

F120B

F102A

F103A

F102B

F103B

F210A

- F210B

F083A

F083B

IS F029B

* F029A

F200

F016

0

0

0

0

0

0

0

0

0

0

0

0

RHR & LPCi

RCICS

HPC

LPCIS

0

0

0 ,. ... 0

F205 0

F239

F029C

: F039B

F039A

0

0

0

0 . I

LPCS FO . . 0

Check SSWS n- F008A

F008B

F012

2.9 x

2.9 x

1.2 x

10-5

10-5

10-6

; - ! '

RHR & LPCIS , F031BF031A I , 1.2 x 10-5

1.1 x 10-5

A-11
Table A-3. Continued

Component Component Aging Sensitivity

Type System Designation (per reactor:year)

Check (Continued) RCICS

HPCS

RHR

LPCIS

F040

F066

F065

F204

FOl1

F005

F024

F002

F054A

F054B

F241

F031C

F041B

F041A

6.0

6.0

6.0

6.0

6.0

x

x

x

x

x

x

x

x

10-6

10-6

10-6

10-6

10-6

10-6

10-6

10-6

2.8

2.8

2.8

8.2 x 10-7

8.2 x 10-7

6.9

6.9

6.9

6.0

x

x

x

x

10-7

10-7

10-7

10-8

LPCS F003

F006

P

C002

6.0 x 10-7

6.0 x 10-7

2.6 x 10-6

6.0 x 10-6

Relief SRS

Turbine RCICS

Actuators ESFAS (for'SSWS) SAC

SBC

SCC

2.9 x

2.9 x

1.2 x

10-5

10-5

10-6

(for RHR & LPCS

& LPCIS)

(for RHR

& LPCIS)

(for RCICS)

(for HPCS)

(for SPMS)

LRACT 1.4 x 10-5

BCACT 1.3 x 10-5

RACT

HACT

SAACC

SBACC

6.0

2.8

1.9

1.9

x

x

x

x

10-6

10-6

10-6

10-6

A- 12
Table A-3. Continued

Component Component Aging Sensitivity

Type System Designator (per reactor year)

Battery EPS DC BATA 1.9 x 10-5

BATB 4.0 x 10-6

BATC 1.2 x 10-6

Diesel EPS AC DIESELl 4.3 x 10-6

DIESEL2 4.0 x 10-6

DIESEL3 1.2 x 10-6

A- 13
NUREG/CR-4144

PNL-5389

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12 F. Leverenz

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NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I. REPORT NUMBER IAusigedbv TIDC. add Vol No. ItJ 7y

(2-841

3201.3202. BIBLIOGRAPHIC DATA SHEET NUREG/CR-4144

SEE INSTRUCTIONS ON THE REVERSE PNL-5389 2. TITLE AND SUBTITLE 3. LEAVE BLANK

Importance Ranking Based on Aging Considerations of

Components Included in Probabilistic Risk Assessments A DATEREPORTCOMPLETED

MONTH | YEAR

5 AUTHORISI March 1985

T. Davis, A. Shafaghi, R. Kurth, F. Leverenz (BCL) 6 DATE REPORT ISSUED

MONTH YEAR

ADril 1985

7.1PERFORMINII.ORGANIZATION NAME AND8MAILING ADDRESqVIc1d#ZPC0d. B. PROJECT/TASKIWORK UNIT NUMBER

Battelle's Columbus Laboratories

505 King Avenue Under Contract to: 9 FIN OR GRANT NUMBER

Columbus, nH 43201 Pacific Northwest Laboratory B2865

Richland, WA 99352

10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS tIrcud. ZW Code) Ila. TYPE OF REPORT

Division of Engineering Technology T

Office of Nuclear Regulatory Research Technical

U.S. Nuclear Regulatory Commission b. PERIOD COVERED dlnc'. tesl

Washington, D.C. 20555

12 SUPPLEMENTARY NOTES

13 ABSTRACT /07 w.ords oless/

This report presents a method for focusing additional research on aging phenomena that

affects nuclear power plant components. Specifically, the method ranks components

using a risk aging sensitivity measure that describes the change in risk due to changes

in component failure rate. Describing the aging phenomena and the resulting timedependent component failure rate changes is beyond the scope of this study.

The applications use average component unavailability equations currently employed in

PRAs to calculate the risk aging sensitivity. A more exact calculation is possible

by using unavailability equations that include the time-dependent-characteristics

of component failure rates; however, these time-dependent characteristics are not

well-known. The risk aging sensitivity measure presented here is, therefore, segregated

from these time-dependent effects and addresses only the time-independent portion of

aging phenomena. The results identify the component types that show the most potential

for risk change due to aging phenomena. Future research on the time-dependent portion

of aging phenomena for these component types is needed to completely describe the risk

impact due to component aging.

1, DOCUMENT ANALYSIS . KEYWORDSIDESCRIPTORS

Aging Sei

Importance Ranking Uni

Nuclear Power Plant

Probabilistic Risk Assessment

1

,vice Wear

availability

15 AVAILABILITY

STATEMENT

Unlimited

16 SECURITYCLASSIFICATION

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Unclassified

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