NUREG/CR-5248

PN L-6701

Prioritization of TIRGALEX

Recommended Components for

Further Aging Research

Prepared by .S. Levy, J. Wreathall, G. DeMoss, A. Wolford,

E.P. Collins, D.B. Jarrell

Pacific Northwest Laboratory Operated by Battelle Memorial Institute

Prepared for

U.S. Nuclear Regulatory Com mission
I

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NUREG/CR-5248

PNL-6701

R9

Prioritization of TIRGALEXRecommended Components for

Further Aging Research

Manuscript Completed: October 1988

Date Published: November 1988

Prepared by .S. Levy, J. Wreathall*, G. DeMoss*, A. Wolford*,

E.P. Collins*, D.B. Jarrell

Pacific Northwest Laboratory

Richland, WA 99352

*Science Applications International Corporation

2941 Kenny Road, Suite 210

Columbus, OH 43221

Prepared for

Division of Engineering Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B2865
ABSTRACT

The "Plan for Integration of Aging and Life Extension," developed by

Technical Integration Review Group for Aging and Life Extension (TIRGALEX) in

May 1987, identified the safety-related nuclear power plant structures and

components (S/C) that should be prioritized for further evaluation by the NRC's

Nuclear Plant Aging Research Program (NPAR).

This report documents the results of an expert panel workshop established

to perform the S/C prioritization activity. Prioritization was primarily based

upon criteria derived from a specially-developed risk-based methodology. This

methodology incorporates the effect upon plant risk of both component aging and

the effectiveness of current industry aging management practices in mitigating

that aging.

An additional set of criteria used to categorize the S/C is the importance

of aging research on S/Cs to the resolution of generic safety issues (GSI)

and/or to identified NRC/NRR user needs. The resultant S/C categorization was

to provide additional information to decision makers, but was not used to

calculate final S/C ranks.

iii
ACKNOWLEDGMENTS

This report documents aging-related research activities sponsored by the

Office of Nuclear Regulatory Research, Division of Engineering, of the U.S.

Nuclear Regulatory Commission. It was funded through the Nuclear Plant Aging

Research (NPAR) Program, J. P. Vora, Project Manager and Program Coordinator.

These activities could not have been successful without support from a variety

of additional sources.

EG&G-Idaho, Inc. [Idaho Nuclear Engineering Laboratory (INEL)], through

its NPAR project (managed by P. MacDonald and B. Cook) supported Science

Applications International Corporation (SAIC) in the development of the basic

Risk Significance of Component Aging (RSCA) model. It also developed the

comprehensive component failure data base from which the precalculated aging

failure data used in this workshop were derived. David Satterwhite, a

co-author of the data base report, graciously provided updated information

shortly before the workshop.

Pacific Northwest Laboratory (PNL), through its NPAR project (managed by

A. B. Johnson, Jr.) supported the expedited development and application of an

enhanced model, the Risk Significance of Component Aging and Aging Management

Practices (RSCAAMP) model, for this workshop. SAIC was the major subcontractor

to PNL on this project and was responsible for applying the RSCAAMP methodology

in the workshop environment. SAIC provided subcontract management and staff

support from its Columbus, Ohio, office as well as staff support from its

McLean, Virginia, and New York, New York, offices.

Several PNL staff members, through participation in two review activities

(a peer review of the RSCAAMP model and a test of the workshop process), provided valuable insights on areas of improvement that contributed to the success

of this workshop.

Kathy Wilson, Conference Coordinator at the Battelle Seattle Research

Center, and her support staff ensured that the workshop participants had the

requisite conference facilities, equipment, and support staff to carry out

their work efficiently.

v
EXECUTIVE SUMMARY

BACKGROUND AND PURPOSE

At the direction of the Executive Director for Operations (EDO), in April

1986 the NRC established the Technical Integration Review Group for Aging and

Life Extension (TIRGALEX) to develop a plan to integrate the NRC's aging and

life extension activities. In May 1987, TIRGALEX finalized its plan (TIRGALEX

1987); it was approved by the Office Directors of RES, NRR, NMSS, AEOD, and IE

and was reviewed by the representative from OGC. The TIRGALEX plan identified

the safety-related structures and components (S/Cs) that should be prioritized

for subsequent evaluation in the NRC Nuclear Plant Aging Research (NPAR)

Program.

The safety-related S/Cs that the TIRGALEX plan identified for evaluation

are listed in Table S.1. During the development of the technical information

for this workshop, ac/dc buses were found to have sufficient risk importance

and were added to the TIRGALEX list. In its deliberations while developing the

plan, TIRGALEX specifically chose not to identify systems for potential study,

only the components of which the systems are comprised.

Pacific Northwest Laboratory was assigned the responsibility for prioritizing the TIRGALEX components for subsequent evaluation by the NPAR Program.

TABLE S.1. TIRGALEX List of Components(a)

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

15.

Reactor pressure vessel

Containment (metal and concrete)

Other Category I concrete structures

Reactor coolant piping and safe ends

Other safety-related ing

Steam generator (PWR)I b

Reactor coolant pump casing

Pressurizer (PWR)

Control rod drive mechanism (CRDM)

Cables, connectors, and penetrations

Emergency diesel generator

Rea tQr internals

RPV c support (sliding foot) (PWR) d)

Recirculation piping safe ends (BWR)(

Snubbers

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

29.

30.

Instruments and controls

Switchgear and relays

Valves

Pumps

Motors

Turbines

Heat exchangers

Compressors

Fans/chillers

Batteries

Battery chargers/inverters

Transformers

Fuel storage racks

Accumulators/tanks

AC/DC buses e1

(a) This list was not prioritized by TIRGALEX nor intended to be inclusive.

It was anticipated that components would be added if warranted by the

prioritization studies.

(b) Pressurized-Water Reactor.

(c) Reactor Pressure Vessel.

(d) Boiling-Water Reactor.

(e) Added since TIRGALEX report as result of analyses prior to the

expert panel workshop.

vii
SCOPE

PNL conducted an expert panel workshop to prioritize the TIRGALEXrecommended set of nuclear power plant S/Cs.

Several considerations of the NRC and PNL provided the ground rules within

which the workshop was conducted. These were:

o TIRGALEX list of components, with additions as the pre-workshop

studies or expert panel analyses may warrant

* aging of current plants (e.g., during their original license period)

* incorporate an understanding of aging and its effects (e.g., define

the contribution of S/C aging to plant risk)

* assess the adequacy of current industry practices for managing component aging within acceptable levels of risk

* importance of S/C aging of individual components/component groups on

plant risk

* application of the "Risk Significance of Component Aging" (RSCA)

methodology (being developed by W. E. Vesely of SAIC under the NPAR

Program) to S/C prioritization

* use of operational failure data

* use of expert judgement through an interdisciplinary panel

* importance of aging research on S/Cs to resolution of generic safety

issues (GSIs) anl identified Office of Nuclear Reactor Regulation

(NRR) user needs"a) to aid NRC decision-makers, but not to formally

prioritize the components.

The expert panel consisted of the following membership: R. J. Budnitz,

Panel Chairman; P. J. Amico; P. L. Appignani; S. H. Bush; L. J. Chockie;

S. Kasturi; T. M. Laronge; and D. A. Wesely.

The panel membership represented expertise .in a full spectrum of relevant

technical areas: PRAs, structures, electrical and mechanical components, component reliability, materials behavior and failure analyses, in-service inspection, operations and maintenance, as well as safety, regulatory, and aging and

life extension issues.

(a) "User Need Letter--Nuclear Plant Aging Program" written by H. R. Denton to

E. S. Beckjord, dated April 9, 1987.

viii
The expert panel was supplied with the TIRGALEX list of components, prioritization criteria, prioritization methodologies, and technical support material prior to the workshop. The panel used judgement to score the S/Cs for

each criterion and to rank the S/Cs relative to one another.

Two sets of prioritization criteria were provided, "risk-based" criteria

and "other technical" criteria.

The risk-based criteria were used to assess:

* the potential increase in plant risk from component aging; and

* the adequacy of current aging management practices for maintaining

risk at acceptable levels.

The other technical criteria were used to identify:

* Generic Safety Issues (GSIs) that could directly benefit from aging

research on an S/C; and

* Identified NRR user needs that could directly benefit from research

in the NPAR program on an S/C.

This categorization of S/Cs against these other technical criteria provides additional information to NRC decision-makers, but were not used to rank

the S/Cs. An overview of the workshop methodology is presented in Figure S.1.

The outputs from the expert panel workshop were the following:

FIGURE S.1. Overview of Workshop Process

iX
structure and component prioritization for the risk-based criteria

(includes scores for each of the five criteria for each S/C and an

integrated ranking of all .S/Cs)

structure and component categorization for the other technical criteria (identification of those S/Cs for which aging research would be

expected to benefit the resolution of a GSI and/or an identified NRC

user need)

The remainder of this executive summary presents highlights in each of

these two areas.

PRIORITIZATION OF STRUCTURES AND COMPONENTS USING RISK-BASED CRITERIA

The risk-based criteria were established through the development and

application of a state-of-the-art risk-based methodology, the Risk Significance

of Component Aging and Aging Management Practices (RSCAAMP) model.

The RSCAAMP model allows the assessment of both the risk significance of

component aging and the effectiveness of current industry management practices

for maintaining an acceptable plant risk level in the presence of component

aging. The RSCAAMP model was developed by enhancing the Risk Significance of

Component Aging (RSCA) methodology, which was developed to evaluate a

component's contribution to plant risk due to aging (Vesely 1987b, 1987c). In

the basic RSCA model, the change in a component's contribution to risk due to

aging is a function of the component's importance to risk (N), the rate at

which the component's failure rate is increasing due to aging (A), and the

interval during which the component is aging (L). The basic methodology has

been expanded from a treatmen taf individual, plant-specific components to a

treatment of component groups for generic applications.

Equation (S.1) represents the basic model.

2

AR = N x A x 2i ' (S.1)

where AR = The change in plant risk due to the aging of a component.

The risk measure adopted for the S/C prioritization activity

is core damage frequency (CD/year).

N = The normal risk importance associated with the component. It is

the difference between the core damage frequency calculated in

the PRA when the component is always available and the core

(a) In the remainder of this report, we will use "component" to represent any

level of component aggregation, whether singular, as appropriate for RPV,

or multiple, as in motor operated valves.

x
damage frequency if the component is defined to be unavailable

[e.g. unavailability = 1.0; (ANSI/IEEE 1987)]. This risk

importance measure is expressed as the derivative of risk (AR)

with respect to component unavailability. Hence, the units are

the same as those used for the risk measure (e.g., CD/yr).

A = The increase in failure rate (from the rate used in the PRA)

due to aging of the component. A is expressed as failures

per unit time squared [also tejmed1 failure acceleration (Vesely

1987c)]. We use the units hr yr , or the annual increase

in hourly failure rate.

L = The interval during which the component is aging (e.g. between

overhauls); its units are months.

The product of the normal risk importance (N) and aging failure rate (A)

is termed the Core Damage Frequency Acceleration (CDFA):

CDFA = N x A . (S.2)

CDFA is also described as the "risk significance of aging" (Vesely 1987c).

The basic RSCA methodology applies directly only to PRA-based components.

It uses risk importance as data calculated from four PRAs from NRC-funded

studies and uses aging failure rates calculated from information in the INEL

data base, generated for the NPAR Program.

Precalculated values of N A, and CDFA were constructed from Equations (S.2) for the PRA-based components. The panel reviewed the PRA S/C list

and the risk importances (N) and aging failure rates (A) and made some changes

by disaggregating components, applying experience with newer PRAs and

judgements on relative failure rates of components.

Structures and components not appearing in PRAs fall into two categories:

1) those primarily intended to prevent core damage accidents and 2) those

designed to mitigate and control post-accident releases. Several components,

such as the RPV and reactor coolant piping, fall into each category. The panel

assigned the structures and components having a dual role to core damage prevention. Solely based on judgment, the panel assigned the N and A values to

the non-PRA-based accident prevention components and incorporated them into the

existing lists for the PRA-based components. For S/Cs with the major role to

mitigate accident consequences, it was necessary to adopt an equivalent risk

importance value that reflected the panel's opinion of the overall significance

of accident mitigation to safety. The panel judged it appropriate to consider

the accident-mitigation S/C (containment), after a core damage accident has

occurred, to have a risk importance equal to the highest risk importance for

any of the accident-prevention S/Cs. This recognized that the containment provides a barrier to the release of radioactive materials to the public that is

xi
of equal importance to the RPV. Based on this assessment, the panel incorporated the N and A values for the accident-mitigation S/Cs into the previously

integrated lists for the accident-prevention S/Cs. For this final integrated

list of all S/Cs, the risk significance of aging (CDFA) was then calculated.

Enhancements were made to the basic RSCA model so that the effect of current industry aging management practices on plant risk could be evaluated in an

expert panel framework. The enhanced model does this by 1) defining an acceptable or control value of the component's aging risk contribution and identifying those components for which risk contribution exceeds this value; 2) defining current industry aging management practices and evaluating their adequacy

for maintaining the risk contribution of these aged components within the control value; and 3) calculating the relative contributions to plant risk of aged

components given this defined adequacy of industry practices.

An acceptable or control value of AR was defined as being ARc. ARc is the

limit placed upon the additional contribution to plant risk due to the aging of

a component. The control value is established by considering the NRC's safety

goal core damage frequency as the measure of total plant risk. Then, a portion

of this goal value is allocated to each component such that the sum of the

individual component risk contributions will not exceed the goal value. The

allocated value for each component is the limit placed upon ARC. The panel

chose 1E-7 CD/yr as the working value for ARC.

This control value of component risk contribution was then used to screen

from further consideration those components whose risk significance of aging

(CDFA) is so low that there is no need to assess the effectiveness of current

industry aging management practices upon them. The basis for this screening is

the increase in risk that would result if aging of the component were allowed

to continue, without overhaul or replacement, for the full 40-year design life

of the plant. This screening criterion was applied to the integrated list of

PRA-based, non-PRA-based accident prevention, and accident mitigation components, for which the panel had previously developed N and A values (and for

which CDFA values were then calculated). A few components were eliminated.

The remaining components are with risk significance of aging sufficiently high

that adequate aging management practices are required to bring component risk

contributions down within ARc.

To assess the adequacy of the current industry aging management practices

in controlling the changes in risk due to aging, an acceptable or control

overhaul interval, Lc, was defined:

2 x AR

Lc CDFA (S.3)

For the panel to evaluate the adequacy of current aging management practices, a value for L that is representative of current industry practice is

required. First, La t was defined as the actual (effective) interval between

component overhauls hat is representative of current industry practice for

xii
that component. Next, the ratio of Lag to Lc (the control overhaul interval)

was defined to represent the adequacy o current practices in controlling risk:

L 2

act adequacy of current industry aging management (S.4)

(L) practices in controlling risk

The effect of industry aging management practices on the risk contribution of

an aged component is represented by:

/L \2

AR = AR x act (S.5)

If industry practices are adequate,

( act <1, and R < R (S.6)

That is, the risk increase due to aging, when aging is adequately managed,

remains within an acceptable value.

If industry practices are not adequate,

/L 2

/> 1, and AR > AR c (S.7)

That is, the risk increase due to aging, when aging is inadequately managed,

exceeds the acceptable value.

Then, Lact was defined in terms of parameters that allow the elicitation

of quantitative input relative to the adequacy of current aging management

practices, and to do so in terms of both the adequacy of aging detection practices and the adequacy of aging mitigation practices.

Lact =PD x PR/D(S.8)

xiii
where Lact = The actual (effective) interval, representative of current

industry practices, that the component ages without mitigation.

Lind = The surveillance/test interval representative of current industry

practices. The panel limited L to the surveillance/test

interval for the risk significanpdfailure mode (e.g., that used

to generate N).

PD = The probability of successfully detecting aging degradation in

the component within the surveillance/test interval. The panel

limited this aging degradation to that causing the increased rate

of risk significant failures (A), these being for the failure

mode used to define N.

PR/D = The probability of successfully mitigating this aging degradation

process, given its successful detection in the component (e.g.,

the aging clock returned to zero or "good as new").

The resultant, fully developed equation for the RSCAAMP model becomes:

AR = N x A x 1 d ) (S.9)

The factors N A, Lind9 PD, and R'D are used as the S/C prioritization

criteria. Each factor is normalized in o a scoring scheme, with a score of 5

representing the highest risk effect from that factor and 1 representing the

lowest risk effect on a logarithmic scale (see method described in Appendix B).

The scores for each S/C are accumulated (addition of logarithms) to obtain a

figure of merit (FOM), which represents the AR value for that S/C. Final ranking (prioritization) of the S/C is based upon these calculated AR values. This

ranking incorporates both the risk significance of aging and the effectiveness

of industry practices in maintaining aging within an acceptable risk level.

The process used by the expert panel for the risk-based prioritization of

the TIRGALEX S/Cs is summarized in Figure S.2. In applying the RSCAAMP model

to prioritization of the TIRGALEX components, the expert judgement was used to:

* disaggregate the TIRGALEX S/Cs

* evaluate the pre-calculated values for normal risk importance (N) and

failure rate increase due to aging (A) for the PRA-based components

* generate estimates of N and A for the non-PRA-based S/Cs

* generate estimates of the effectiveness of current industry aging

management practices (Lind, PR, and R/D) for all S/Cs

* provide an integrated ranking of the PRA- and non-PRA-based

components.

Table S.2 presents the components identified by TIRGALEX, the 4-plant

PRAs, and the final list evaluated by the expert panel.

xiv
TIRGALEX List of Compor

(Table 1.1)

Expert Panel List of Components

(Table 2.2)

Accident Prevention S/Cs

I L~~~~~~~~~.

PRA-Based S/C

I

Panel Reviews

Estimates of N

(Table 2.3)

Panel Reviews

Estimates of A

(Table 2.4)

Non-PRA-Based S/Cs

I

Panel Estimates

Values of N

(Table 2.3)

Panel Estimates

Values of A

(Table 2.4)

ients

and Structures

Accident Mitigation S/Cs

Panel Estimates

Equivalent N Value

(Table 2.3)

Panel Estimates

Values of A

(Table 2.4)

Integration of all SCs

Evaluate Values

for Credibility

& Apply Screening

Criterion

(Table 2.5)

Panel Estimates LPD

PR/ (Table 2.6)

Rankings Based on

NAL9 PD, P R/D AR

(Tabie 25 

Criteria Scores

and S/C Ranks

(Table 2.8)

FIGURE S.2. Expert Panel Process for Risk-Based Prioritization

xv
TABLE S.2. Structures/Components Evaluated by Expert Panel

TIRGALEX Groups

1. Reactor pressure vessel

(RPV)

4-Plant PRA Components

Components Chosen

by the Expert Panel

1. RPV

2. Containment 2. Containment

a. BWR

b. Other

3. Other Category I 3. Other Category I structures

structures

4. Reactor coolant piping

(RCP) and safe ends (SE)

5. Other safety-related

piping (SRP)

4. RCP & SE

a. Large LOCA

b. Small LOCA, PWR

c. Small LOCA, BWR

5. Other SRP

a. Large (10-24 In.) ibe a

b. Small (6-10.) pipe

6. Steam generator

a. S/G tube

b. S/G shell

7. RCP casing

6. Steam generator S/G)

7. Reactor coolant pump (RCP)

casing

8. Pressurizer

9. Control rod drive

mechanism (CRDM)

10. Cables, connectors, and

penetrations

8. Pressurizer

9. CRDM

a. BWR

b. PWR

10. Cables, connectors,

a. Cables

b. Connectors

(* Penetrations as part of

containment)

11. Diesel generator

12. Reactor Internals

13. RPV support (PWR)

11. Diesel generator

12. Reactor internals

13. RPV support (PWR)

14. Recirculation piping safe

ends

11. Diesel generator

14. Small LOCA, BWR

15. Snubbers

16. Instruments and Controls

(I&C)

17. Switchgear/relays

16. Instruments and Controls

(&C)

a. Thermostat

17. Switchgear/relays

a. Relay (load)

b. Circuit breaker c. Transfer switch

d. Bstable trip unit

15. Snubbers

16. Instruments and Controls

(&C)

a. Thermostat (c)

b. Transfer switcic

c. Bistable trip

17. Switchgear Relays

a. Relay

b. Circuit breaker

xvi
TABLE S.2. (contd)

TIRGALEX Groups

18. Valves

4-Plant PRA Components

18. Valves

a. Air-operated valve (AOV)

b. Check valve

c. Hydraulic valve

d. Manual valve

e. Motor-operated valve

(MOV)

f. Safety/relief valve

(S/RV)

Components Chosen

by the Expert Panel

18. Valves

a. Air-operated valve

b. Check valve

c. Hydraulic valve

d. Manual valve

e. Motor operated valve

f. Safety/relief valve

19. Pumps 19. Pumps

a Motor driven pump

b. Turbine driven pump

20. Motors

19. Purps

a. Motor driven pump

b. Turbine driven pump

20. Motors (included In valves,

purps, etc.)

21. Turbines

22. Heat exchangers

21. Turbines

22. Heat exchangers

21. Turbines

22. Heat exchangers

a. Heat exchangers

b. Air conditioners

23. Compressor

24. Fans/chillers

23. Compressor(d)

24. a. Chiller

b. Fan

24. Fan

25. Batteries 25. Batteries

26. Battery chargers/inverters

27. Transformers

28. Fuel storage racks

29. Accumulator/tanks

26. Battery chargers/inverters

a. Battery chargers

b. Inverters

27. Transformers

25. Batteries

26. Battery chargers/i nverters

a. Battery chargers

b. Inverters(e) c. Rectifier

27. Transformers

29. Tank

28. Fuel storage racks

29. Tanks

a. Medium pressure tank

b. Atmospheric pressure tank

c. High pressure tank

30. AC/DC bus 30. AC/DC buses

a. AC bus

b. DC bus

30. AC/DC buses

a. AC bus

b. DC bus

31. Bolts

(a) Represented by service water system piping for which HPI could not keep up If break occurred.

(b) Represented by piping In letdown and reactor water cleanup RWCU) systems, for which HPI could

Just keep up If small break occurred.

(c) Moved by panel from S/C-17, 4-Plant PRA.

(d) Compressors In Instrument Air System.

(e) Moved by panel from non-assigned S/C (see Table 2.1).

Xvi i
Table S.3 presents the scores for the prioritization criteria and the

final integrated ranking of the S/Cs. The S/Cs are arranged in decreasing

order of their contribution to plant risk increase. Those components whose

AR(40) values were below 1E-7 CD/yr and were not evaluated for adequacy of

aging management practice are included in this table; their rankings are placed

at the nominal value of one (1). To stay within the guidelines of considering

aging only in the context of the current 40-year license period, the calculations of Lact (Equation .8) were truncated to a value of 480 months.

The impact on component risk contribution of aging and aging management

practices, as well as the component's normal risk importance, were evaluated.

For the top-ranked components (ranks 5 and 4), all of these factors were

significant; no one factor was dominant.

The importance of including component aging and its management in assessing plant risk was dramatized by some unexpected findings. Some components not

considered in PRAs were found in the top rank (rank of 5) for risk contribution

(cables, connectors, small safety-related piping); conversely, concrete structures and diesels that are considered in PRAs to be significant contributors to

plant risk were found in the lower ranks (ranks of 2 and 4, respectively); and,

perhaps as significant, is the number of "abundant small components" (e.g.,

small other safety-related piping, cables, connectors, and S/G tubes) found in

the top ranking (four of the five groups ranked 5) when one might intuitively

have thought of the major non-redundant, defense-in-depth structures (RPV and

containment) to be top-ranked [RPV was ranked 3, containment (other) ranked 2,

with only the special case of the BWR-Mk-1 containment (due to aging)

ranked 5].

If Lact had not been truncated, several components would have Lact values

well above 80 months and, as a result, would have higher final risk increase

values. This impacts the consideration of plant relicensing. If a nuclear

plant were allowed to operate beyond 40 years, the ineffectiveness of aging

management practices represented by the components' untruncated values would

result in continuing and increasing risk. (Note: this assumes no change in

those practices.) Of particular concern would be S/C-1: RPV, S/C-2b: containment-other, and S/C-3: other concrete structures since these are permanent

structures (though the possibility of RPV replacement has been considered).

Table S.4 presents the current status of component research in the NRC

Plant Aging Research Program for the ranked components (from Table S.3).

PRIORITIZATION OF STRUCTURES AND COMPONENTS USING "OTHER TECHNICAL" CRITERIA

TIRGALEX S/Cs were categorized using the other technical criteria, e.g.,

the importance of aging research on S/Cs to the resolution of GSIs and/or to an

identified NRC/NRR user need.

xviii
TABLE S.3. Structures/Components Prioritization Criteria

Scores and Final Rankings (Truncated to 40 years)

Component

5.b.

10.a.

2.a.

10.b.

6.a.

19.b.

17.a.

11.

12.

17.b.

18.e.

4.c.

19. a.

5.a.

16.a.

24.a.

1.

25.

23.

18.a.

30.b.

9.a.

18.b.

24. b.

22.

31.

30.a.

18.f.

2. b.

3.

27.

26.b.

16.b.

15.

18.c.

21.

16.c.

18.d.

26.a.

29.b.

26.c.

29.a.

9. b.

Small other safety pipe(a)

Cables

Containment (BWR)

Connectors

S/G tube

Turbine pump

Relay

Diesel

RX Internals

Breaker

Motor operated valve

BWR pipe (small LOCA)

MNtor pump (c)

Large other safety pipe

Thermostat

Chillers

RPV

Battery

Compressor (Instr. air)

Air operated valve

DC bus

CRDM (BWR)

Check valve

Fan

Heat exchanger

Bolts

AC bus

Safety/relief valve

Containment (other)

Other concrete structures

Transformer

Inverter

Transfer switch

Snubbers

Hydraulic valve

Turbine

Blstable

Manual valve

Battery charger

Tank (atmos. pres.)

Rectifier

Tank (medium pres.)

CRDM (PWR)

N

Score

2

4

5

3

3

4

3

4

4

3

2

3

3

3

2

5

3

2

2

4

4

2

2

3

1

4

5

5

3

3

3

2

A

Score

4

2

3

3

5

5

4

5

2

2

5

3

4

2

3

4

4

4

4

2

2

4

2

4

i

4

2

5

4

5

3

3

3

2

3

3

L

4

4

3

4

4

2

3

3

4

2

3

3

3

5

3

3

3

3

3

3

3

3

4

4

3

2

3

3

3

4

3

4

2

2

2

2

3

P

Scre

5

5

4

3

3

4

3

5

3

2

4

3

5

2

2

1

4

3

3

5

2

5

3

3

1

5

5

2

3

4

3

1

2

3

1

5

3

5

3

P

SU9e

2

4

4

4

4

3

2

2

Increase

Rank

5 (b)

5 (b)

5

5

5

4

4

4

4

3

3

3

3

3

3

3(b

3(b

3

3

2

2

2

2

2

2

2

2

2(b)

2 (b)

2 (b

1

1

(d)

8.

6.b.

29.c.

4.a.

7.

13.

28.

4. b.

Pressurizer

S/G shell

Tank (high pres.)

RC P & SE large (LOCA)

RC P casing

RPV support

Fuel rack

PWR pipe (small LOCA)

I 1 __ __ -- 1

1 1 -- 1

1 1 -- -- -- I

1 1 -- -- -- 1

1 1 -- -- -- 1

I I -- -- -- I

1 1 -- -- -- 1

2 1 -- -- -- 1

(a) 6-10 In. pipe represented by letdown and RWCV systems (see Table S.2).

(b) L for these components was truncated to a maximum of 40 years (480 months).

(c) IB-24 In. pipe represented by service water system piping (see Table S.2).

(d) Components below this line were eliminated from further consideration by

screening criterion.

xix
TABLE .4. Status of Aging Research on Ranked Components

Components of

AR Research Interest But

Component Rank on-going Not In Scope

5.b. Small other safety pipe ) 5 x lO.a. Cables 5 xW

2.a. Containment (BWR) 5 b

1O.b. Connectors 5 x 6.a. S/G tube 5(c

19.b. Turbine pump 4 x

17.a. Relay 4 x

11. Diesel 4 x 12. RX Internals 4 0

17.b. Breaker 3 x 18.e. Mbtor operated valve 3 x

4.c. BWR pipe (small LOCA) 3 x 19.a. Motor pump d) 3 x

5.a. Large other safety pipe 3 x 16.a. Thermostat 3 0

24.a. Chillers 3 x

1. RPV 3 x

25. Battery 3 x 23. Compressor (Instr. air) 3 x

18.a. Air operated valve 2 x

30.b. DC bus 2 x 9.a. CRM (BWR) 2 x 18.b. Check valve 2 x

24.b. Fan 2 x

22. Heat exchanger 2 x

31. Bolts 2 0

30.a. AC bus 2 x 18.f. Safety/relief valve 2 x 2.b. Containment (other) 2 0

3. Other concrete structures 2 x

27. Transformer 1 x

26.b. Inverter 1 x

16.b. Transfer switch 1 0

15. Snubbers 1 x 18.c. Hydraulic valve 1 0

21. Turbine 1 0

16.c. Bistable 1 x

18.d. Manual valve 1 0

26.a. Battery charger 1 x 29.b. Tank (atmos. pres.) 1 0

26.c. Rectifier 1 x

29.a. Tank (medium pres.) 1 0

9.b. CRDM (PWR) I x

(e)

8. Pressurizer 1 0

6.b. S/G shell 1 0

29.c. Tank (high pres.) 1 x

4.a. RCP & SE large (LOCA) 1 x 7. RCP casing 1 0

13. RPV support 1 x 28. Fuel rack 1 0

4.b. PWR pipe (small LOCA) 1 x

(a) 6-10 In. pipe represented by letdown and RWCV systems (see Table 5.2).

(b) 0: no activity or plan.

(c) Completed FY87.

(d) 10-24 In. pipe represented by service water system piping (see Table S.2).

(e) Components below this line were eliminated from full evaluation

by the screening criterion.

XX
The list of potentially important safety issues comes from two sources:

1) a list of GSIs with elements of aging that may benefit from NPAR results;

and 2) additional issues discovered as a result of a pre-workshop review of all

GSIs.

The TIRGALEX S/Cs were initially screened to identify those S/Cs for which

these GSIs could directly benefit from aging research. This prescreening took

account of the technical content and available schedules associated with the

resolution of the GSIs and with the NRC Plant Aging Research Programs (Vora

1987). Where there was a direct technical connection (i.e., aging was involved

in the issue) and the time scales appeared to be compatible, aging research on

the S/Cs was identified as potentially having a direct benefit. On the other

hand, some S/Cs were clearly not associated with the list of GSIs, aging was

not directly related to the issue, or time scales were incompatible with their

resolution; aging research on these S/Cs was classified as not beneficial. All

GSI evaluations were presented to the panel for discussion and final categorization as whether aging research would benefit or not benefit the resolution of

the GSIs. A few GSIs considered relevant, but which were not identified in the

current GSI resolution schedule, were also presented to the panel for

consideration.

NRR user needs were expressed in the "User Need Letter". In the letter,

NRR expressed the need to "... know not only the effects of aging on

structures, systems and components, but also the risk significance..." of the

process. The "User Need Letter" was reviewed for components deemed to be

important by NRR.

In the panel's judgement, many of the generic issues that deal with equipment performance were expected to directly benefit from aging research (schedule permitting), and that further aging research on the components identified

from the user needs letter would be of benefit to NRR. Table S.5 shows the

components' ranking based on risk, S/Cs for which aging research would benefit

the resolution of a GSI and/or a NRC/NRR user need and those S/Cs for which

current aging research is not already on-going. Only for Rx internals (meeting

GSI and user needs criteria), pressurizer (GSIs), and bolts (GSI) and hydraulic

and manual valves (GSI and user needs) is aging research not already on-going.

CONCLUSIONS AND RECOMMENDATIONS

The expert panel workshop was conducted to prioritize the TIRGALEX set of

nuclear power plant S/Cs for further evaluation within the NRC Nuclear Plant

Aging Research Program. The prioritization was primarily based upon risk-based

criteria; other technical criteria were used to categorize those S/Cs for which

aging research would benefit the resolution of GSIs and/or identified NRC/NRR

user needs but were not used to rank the S/Cs.

From the S/C prioritization using risk-based criteria, the major conclusions from the workshop are the following: 1) the prioritization of S/Cs was

accomplished by an expert panel using the multi-factor RSCAAMP methodology.

Analysis of the results showed that all the factors of this methodology are

xxi
TABLE S.5. Status of Aging Research on S/Cs Ranked by Risk

Importance and Other Technical Criteria

Aging Research

Important to

Resolution of: Aging Research Not

AR User On-Going or

Component Rank GSI Need of Current Interest

5.b. Smell other safety pipe(a) 5(b

IO.a. Cables 5 x,x

2.a. Containment (BWR) 5 0

1O.b. Connectors 5 x,x

6. S/G tube 5 x (c)

19.b. Turbine pump 4 x

17.a. Relay 4 x,x,x

11. Diesel 4 x x

12. RX internals 4 x 0

17.b. Breaker 3 x,x

18.e. Motor operated valve 3 x,x x,x,x

4.c. BWR pipe (small LOCA) 3

19.a. Motor pump ( 3 x x

5.a. Large other safety pipe 3

16.a. Thermostat 3

24.a. Chillers 3

1. RPV 3 x,x,x x

25. Battery 3 x

23. Compressor (instr. air) 3

18.a. Air operated valve 2 x x

30.b. DC bus 2 x,x

9.a. CRDM (BWR) 2

18.b. Check valve 2 x,x x

24.b. Fan 2

22. Heat exchanger 2 x

31. Bolts 2 x 0

30.a. AC bus 2 x,x

18.f. Safety/relief valve 2 x x,x

2.b. Containment (other) 2

3. Other concrete structures 2 x

27. Transformer I

26.b. Inverter 1 x

16.b. Transfer switch 1

15. Snubbers 1 x

18.c. Hydraulic valve I x x 0

21. Turbine 1 0

16.c. Bistable 1

18.d. Manual valve 1 x,x x 0

26.a. Battery charger 1 x

29.b. Tank atmos. pres.) 1 0

26.c. Rectifier 1 x

29.a. Tank (medium pres.) 1 0

9.b. CRDM (PWR) 1 (e)

8. Pressurizer 1 x 0

6.b. S/G shell 1 0

29.c. Tank (high pres.) 1

4.a. RCP & SE large (LOCA) I x,x

7. RCP casing 1 0

13. RPV support 1

28. Fuel rack 1 0

4.b. PWR pipe (small LOCA) 1

(a) 6-10 in. pipe represented by letdown and RWCV systems (see Table S.2).

(b) More than one "x" Indicates multiple GSI and/or user need entries.

(c) Completed FY87.

(d) 10-24 In. pipe represented by service water system piping (see Table S.2).

(e) Components below this line were eliminated from full evaluation by the

risk-based screening criterion.

xxii
equally important to an assessment of the relative risk importances of aged

components. Further, the importance of including component aging and its management in assessing plant risk was dramatized by the unexpected findings in

the top-ranked components. 2) Current aging research warrants reevaluation:

many low-ranked S/Cs (ranks 1 and 2) are under study and others are considered

to be of interest; two of the components in the top two ranks (ranks 5 and 4)

are not currently being studied in the Plant Aging Research Program [containment (BWR-Mk-1) and RX internals]. 3) The panel's deliberations and findings

highlighted the need for improved PRA and aging-failure data bases, the shortfalls in current industry practices for detecting aging for risk-significant

failure modes in components and structures, and the usefulness of the methodology for focusing of research and regulatory actions and for providing to

utilities areas to address in order to reduce the risk-contribution of their

aged components. 4) This study should be viewed as a starting point for which

expert opinion was required; from this study an appropriately-focused research

program can generate the data needed for a definitive answer on the relative

risk-importance of aged components. 5) The explicit evaluation of the importance of aging-induced common-cause failures would be a useful follow-up exercise to perform; with minor development, the RSCAAMP model would be amenable to

such an evaluation.

From the S/C categorization using other technical criteria, the major conclusions from the workshop are as follows: 1) the categorization of S/Cs using

the other technical criteria was accomplished satisfactorily. 2) There are

many S/Cs for which an aging research program could have direct benefit to the

resolution of GSIs. 3) There are fewer, but still a substantial number of S/Cs

for which an aging research program could respond directly to identified user

needs. 4) Only for a few of these components is aging research not already

on-going.

xxiii
CONTENTS

ABSTRACT ............. . . .... .... . ...... *.* *....**. . i i

ACKNOWLEDGMENTS v ............... v

EXECUTIVE SUMMARY ... o.......vii

GLOSSARY ........ ...... ... xxx

1.0 INTRODUCTION 1.1

1.1 BACKGROUND AND PURPOSE ...... 1.1

1.2 SCOPE OF THE EXPERT PANEL WORKSHOP 1.1

1.2.1 Ground Rules 1.2

1.2.2 Process 1.3

1.2.3 Outputs ................. 1.5

1.3 COMPOSITION OF EXPERT PANEL WORKSHOP ..... 1.5

1.3.1 Expert Panel Membership 1.5

1.3.2 Workshop Technical Support Staff Membership 1.6

1.4 ORGANIZATION OF THIS REPORT ................. 1.6

2.0 PRIORITIZATION OF STRUCTURES AND COMPONENTS USING

RISK-BASED CRITERIA .... o..oo.oo o.. .. ...... 2.1

2.1 THE RISK SIGNIFICANCE OF COMPONENT AGING AND AGING

MANAGEMENT PRACTICES METHODOLOGY AND ITS APPLICATION

TO S/C PRIORITIZATION o~...oo .. 2.1

2.1.1 Risk Significance of Component Aging .... 2.1

2.1.2 Adequacy of Current Aging Management Practices 2.7

2.2 PANEL FINDINGS AND TECHNICAL RATIONALE 2.13

2.2.1 Final Component Selection 2.13

2.2.2 Component Scores and Ranks m....................... o.. 2.17

2.2.3 Final Integrated Ranking of the Structures

and Components ....... ......... ..... .. ....... ... 2.26

xxv
2.2.4 Detailed Rationale for the Panel's Findings .......... 2.27

2.3 TECHNICAL INSIGHTS AND DISCUSSION ............. .............. 2.30

2.3.1 The Workshop Process ............. .................... 2.30

2.3.2 Application of the RSCAAMP Methodology .. ............. 2.30

2.3.3 Unexpected Findings in the Component Rankings ........ 2.31

2.3.4 Role of the RSCAAMP Factors in the Component

Rankings ...................................................... 2.33

2.3.5 Insights on Research, Regulatory and

Industry Applications ................................ 2.34

2.4 CONCLUSIONS FROM RISK-BASED PRIORITIZATION .................. 2.35

3.0 PRIORITIZATION OF STRUCTURES AND COMPONENTS USING "OTHER

TECHNICAL" CRITERIA ................................... e. ... ............... 3.1

3.1 IMPORTANCE OF AGING RESEARCH ON S/Cs TO THE RESOLUTION

OF A GSI AND ITS APPLICATION TO S/C CATEGORIZATION .......... 3.1

3.2 IMPORTANCE OF AGING RESEARCH ON S/Cs TO AN IDENTIFIED

NRC USER NEED AND ITS APPLICATION TO S/C CHARACTERIZATION ... 3.3

3.3 RESULTS ........................................................... 3.3

3.3.1 Importance of Aging Research on S/Cs to the

Resolution of a GSI ....................................... 3.3

3.3.2 Importance of Aging Research on S/Cs to Resolution

of User Needs ............................................. 3.4

3.3.4 Conclusions ........................................................ 3.4

4.0 CONCLUSIONS AND RECOMMENDATIONS ........ ......... ............ 4.1

5.0 REFERENCES ............................................................ 5.1

APPENDIX A - DERIVATION OF BASIC RSCA MODEL AND THE

RANKING METHODOLOGY ........... . . ..................................... . A.1

APPENDIX B - DATA TABLES SUPPORTING THE PRECALCULATED RSCA

FACTORS .......................................................... B.1

APPENDIX C - PRECALCULATED N A, CDFA VALUES FOR PRA-BASED

COMPONENTS ..... .......................................... C.1

xx i
APPENDIX D - DETAILED RATIONALE FOR PANEL'S FINDINGS FOR

PRIORITIZATION USING RISK-BASED CRITERIA ................. D.1

APPENDIX E - EVALUATION OF AGING RESEARCH BENEFITS TO GSI

RESOLUTION ............. NEED LETTER .............. ** ...... ..... E.1

APPENDIX F - NRC/NRR USER NEED LETTER ................................................. F.1

xxvii
FIGURES

1.1 Overview of Workshop Process ..................................... 1.5

2.1 Basic Model for the Risk Significance of Component Aging ......... 2.2

2.2 Effect of Industry Aging Management Practices on Component

Aging Risk Contribution ................. ......................... 2.12

2.3 The Risk Significance of Component Aging Management Practices

Model ............................................................ 2.12

2.4 Expert Panel Process for Risk-Based Prioritization ...............................

2.14

xxviii
TABLES

1.1 TIRGALEX List of Components ...... ................................ 1.2

2.1 Components in the Four-Plant PRA Studies Compared to

TIRGALEX List of Components ..... ................................. 2.4

2.2 Structures/Components Evaluated by Expert Panel .................. 2.15

2.3 Structure/Component Risk Importances ............................. 2.18

2.4 Structure/Component Aging Failure Rates ........... .. *............ 2.20

2.5 Structure/Component CDFAs and 40-Year Risk Increases ............. 2.22

2.6 Adequacy of Current Industry Aging Management Practices .......... 2.23

2.7 Factor Values and Final Ranking of Structures/Components ......... 2.27

2.8 Structures/Components Prioritization Criteria Scores and

Final Rankings ... ........................ .................. 2.29

2.9 Panel's Preliminary Ranking of Components ........................ 2.32

2.10 Analysis of the Contribution of RSCAAMP Factors to the

Final Risk Contribution Rank of Top-Ranked Components ............ 2.34

2.11 Status of Aging Research on Ranked Components ..... ...... .... ... .. 2.36

3.1 NRC-Identified Generic Safety Issues, with Elements of

Aging, Potentially Benefitting from Aging Research ............... 3.2

3.2 Additional Generic Safety Issues ................................. 3.3

3.3 NRC-Identified GSIs with Elements of Aging Potentially

Benefitting from Aging Research .................................. 3.5

3.4 Additional GSIs with Elements of Aging Potentially

Benefitting from Aging Research .................................. 3.7

3.5 Status of Aging Research on S/Cs Ranked by-Risk Importance

and Other Technical Criteria ....... .............................. 3.8

xxix
GLOSSARY

AFW - auxiliary feedwater

AOV - air operated valve

ASME - American Society of Mechanical Engineers

ATWS - anticipated transient without scram

BWR - boiling-water reactor

CDFA - core damage frequency acceleration

CRDM - control rod drive mechanism

FOM - figure of merit

GSI - generic safety issues

HPI - high pressure injection

HPCI - high pressure coolant injection

HX - heat exchanger

I&C - instruments and controls

INEL - Idaho Nuclear Engineering Laboratory

IREP - Integrated Reliability Evaluation Program

LOCA - loss-of-coolant accident

LOSP - loss of system pressure

MOV - motor operated valve

NDE - nondestructive examination

NPAR - Nuclear Plant Aging Research

NPP - nuclear power plant

NPRDS - nuclear plant reliability data system

NRC - U.S. Nuclear Regulatory Commission

NRR - Nuclear Reactor Regulation

PNL - Pacific Northwest Laboratory

PORV - power operated relief valve

PRA - probabilistic risk assessment

PWR - pressurized-water reactor

RCIC - reactor core isolation cooling

RCP - reactor coolant pump

RES - Office of Nuclear Regulatory Research

RHR - residual heat removal

RPV - reactor pressure vessel

RSCA - Risk Significance of Component Aging

RSCAAMP - Risk Significance of Component Aging and Aging Management Praci

RSCAAMP factors:

AR - the change in plant risk due to aging of a component

AR - the acceptable or control value of AR

f- the normal risk importance associated with a component

A - the increase in failure rate due to aging of the component

CDFA - core damage frequency acceleration (NxA)

L - the interval during which the component is aging

Lc - the acceptable or control overhaul interval

L act - the actual (effective) interval between overhauls that is

representative of current industry practice for a component

;ices

xxx
/Lac 2

t) _ the adequacy of current aging management practices

Lind - the surveillance/test interval representative of current industry

practices

(Lind)apc - the acceptable industry surveillance or test interval

D - the probability of successfully detecting component aging

degradation within the surveillance or test interval

PR/D - the probability of successfully mitigating aging degradation,

given its successful detection in a component

RSSMAP - Reactor Safety Study Methodology Application Program

RWCU - reactor water cleanup

SAIC - Science Applications International Corporation

S/C - structure/component

SE - safe ends

S/G - steam generator

SRP - safety related piping

S/RY - safety/relief valve

TIRGALEX - Technical Integration Review Group for Aging and Life Extension

xxxi
1.0 INTRODUCTION

This report documents the scope, methods, results, technical insights,

conclusions, and recommendations of an expert panel workshop that was held on

November 17-19, 1987. The purpose of the workshop was to prioritize nuclear

power plant structures and components (S/Cs) for further evaluation within the

Nuclear Plant Aging Research Program being conducted by the Division of

Engineering, Office of Nuclear Regulatory Research (RES) of the U.S. Nuclear

Regulatory Commission (NRC). The workshop was conducted under the auspices of

the NRC's Nucle r Plant Aging Research (NPAR) Program by Pacific Northwest Laboratory (PNL) a and Science Applications International Corporation (SAIC),

under subcontract to PNL. It was held at the Battelle Seattle Conference

Center, a part of the Battelle Seattle Research Center.

For the prioritization process, a state-of-the-art risk-based methodology,

the latest compilation of component aging failure data, and expert judgement

were utilized to assess the risk impacts of component aging and of industry

practices for managing component aging.

1.1 BACKGROUND AND PURPOSE

At the direction of the Executive Director for Operations (EDO), in April

1986 the NRC established the Technical Integration Review Group for Aging and

Life Extension (TIRGALEX) to develop a plan to integrate the NRC's aging and

life extension activities. In May 1987, TIRGALEX finalized its plan (TIRGALEX

1987); it was approved by the Office Directors of RES, NRR, NMSS, AEOD, and IE

and was reviewed by the representative from OGC. The TIRGALEX plan identified

the safety-related structures and components that should be prioritized for

subsequent evaluation in the NRC Nuclear Plant Aging Research (NPAR) Program.

The safety-related S/Cs that the TIRGALEX plan identified for evaluation

are listed in Table 1.1. During the development of the technical information

for this workshop, ac/dc buses were found to have sufficient risk importance

and were added to the TIRGALEX list. In its deliberations while developing the

plan, TIRGALEX specifically chose not to identify systems for potential study,

only the components of which the systems are comprised.

Pacific Northwest Laboratory was assigned the responsibility for prioritizing the TIRGALEX components for subsequent evaluation by the NPAR Program.

An expert panel workshop approach was used.

(a) Operated for the U.S. Department of Energy (DOE) by Battelle Memorial

Institute under Contract DE-AC06-76RL0 1830.

1.1
TABLE 1.1. TIRGALEX List of Components(a)

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

15.

Reactor pressure vessel

Containment (metal and concrete)

Other Category I concrete structures

Reactor coolant piping and safe ends

Other safety-related pb ing

Steam generator (PWR)' }i

Reactor coolant pump casing

Pressurizer (PWR)

Control rod drive mechanism (CRDM)

Cables, connectors, and penetrations

Emergency diesel generator

Reactor internals

RPV c support (sliding foot) (PWR)

Recirculation piping safe ends (BWR)(d)

Snubbers

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

29.

30.

Instruments and controls

Switchgear and relays

Valves

Pumps

Motors

Turbines

Heat exchangers

Compressors

Fans/chillers

Batteries

Battery chargers/inverters

Transformers

Fuel storage racks

Accumulators/tanks

AC/DC buses e1

(a) This list was not prioritized by TIRGALEX nor intended to be inclusive.

It was anticipated that components would be added if warranted by the

prioritization studies.

(b) Pressurized-Water Reactor.

(c) Reactor Pressure Vessel.

(d) Boiling-Water Reactor.

(e) Added since TIRGALEX report as result of analyses prior to the

expert panel workshop.

1.2 SCOPE OF THE EXPERT PANEL WORKSHOP

This section describes the ground rules within which the workshop was

conducted, the process used by the expert panel, and the expected outputs of

the workshop.

1.2.1 Ground Rules

Several considerations of the NRC and PNL provided the ground rules within

which the workshop was conducted. These were:

* TIRGALEX list of components, with additions as the pre-workshop

studies or expert panel analyses may warrant

1.2
* aging(a) of current plants (e.g., during their original license

period)

* incorporate an understanding of aging and its effects (e.g., define

the contribution of S/C aging to plant risk)

* assess the adequacy of current industry practices for managing

component aging within acceptable levels of risk

* importance of S/C aging of individual components/component groups on

plant risk

* application of the "Risk Significance of Component Aging" (RSCA)

methodology (being developed by W. E. Vesely of SAIC under the NPAR

Program) to S/C prioritization

* use of operational failure data

* use of expert judgement through an interdisciplinary panel

* importance of aging research on S/Cs to resolution of generic safety

issues (GSIs) and an identified Office of Nuclear Reactor Regulation

(NRR) user needs to aid NRC decision-makers, but not to formally

prioritize the components.

1.2.2 Process

The expert panel was supplied with the TIRGALEX list of components, prioritization criteria, prioritization methodologies, and technical support

material. The panel used their expert judgement to score the S/Cs for each

criterion and to rank the S/Cs relative to one another.

Two sets of prioritization criteria were provided, "risk-based" criteria

and "other technical" criteria.

1.2.2.1 Risk-based Criteria

The risk-based criteria have two important factors as their bases:

* the potential increase in plant risk from component aging

(a) The TIRGALEX plan defined aging as the cumulative degradation occurring

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

loss of function and impairment of safety. Aging may be caused by natural

internal chemical or physical processes; external stressors and environment; service wear (cycling, vibration); testing; or improper installation, application, or maintenance.

1.3
* the adequacy of current aging management practices for maintaining

risk at acceptable levels.

Five risk-based criteria were used to implement these two bases:

N : normal risk importance of the S/C(a)

A : increase in S/C failure rate due to aging

Lind : surveillance/test interval represented by current industry

practice

PD : probability of successfully detecting aging degradation within

ind

PR/D : probability of successfully mitigating aging given successful

detection.

These risk-based criteria were developed by enhancing the Risk Significance of Component Aging (RSCA) model (Vesely 1987c) for application to this

prioritization activity. The enhancement was accomplished through discussions

between PNL and SAIC. The RSCA model had already encompassed the risk importance of components (N), the effect on risk of the increase in S/C failure rate

due to aging (A), as well as the interval (L) during which aging was continuing

unmitigated. The enhancements were to allow for consideration of the effectiveness of current industry practices for managing aging degradation and to do

so in a manner that would 1) subdivide those practices into both detection and

mitigation practices and 2) easily permit input of expert judgement (L dP

and PRID). The enhanced model is termed "Risk Significance of Component Aging

and Aging Management Practices" (RSCAAMP).

In applying the RSCAAMP model to prioritization of the TIRGALEX components, the expert judgement was used to:

* evaluate the pre-calculated values for normal risk importance (N) and

failure rate increase due to aging (A) for the PRA-based components

* generate estimates of N and A for the non-PRA-based S/Cs

* generate estimates of the effectiveness of current industry aging

management practices (Lind, PR, and R/D) for all S/Cs

* provide an integrated ranking of the PRA- and non-PRA-based

components.

(a) This is equivalent to the Birnbaum risk importance measure as discussed by

W. E. Veseley and T. C. Davis in Evaluation and Utilization of Risk

Importance.

1.4
1.2.2.2 Other Technical Criteria

In addition to the risk-based criteria, the importance of performing aging

research on the S/Cs to the resolution of GSIs and a set of the NRC/NRR user

needs identified in "User Need Letter--Nuclear Plant Aging Program." (See

Appendices F and G, respectively.)

The S/Cs were screened prior to the workshop to identify those for which

GSIs could directly benefit from aging research. These were submitted to the

panel for approval.

A similar approach was followed for the relevance of aging research on

S/Cs to the identified NRR user needs. The S/Cs were screened for those of

direct benefit, and then presented to the panel for approval.

This categorization of S/Cs against these other technical criteria provides additional information to NRC decision-makers, but is not used to rank

the S/Cs (per method in Telford et al. 1986). An overview of the workshop

methodology is presented in Figure 1.1.

1.2.3 Outputs

The outputs from the expert panel workshop were anticipated to be the

following:

* structure and component prioritization for the risk-based criteria

(includes scores for each of the four criteria for each S/C and an

integrated ranking of all S/Cs)

FIGURE 1.1. Overview of Workshop Process

1.5
* structure and component categorization for the other technical criteria (identification of those S/Cs for which aging research would be

expected to benefit the resolution of a GSI and/or an identified NRC

user need)

* technical insights of the panel.

1.3 COMPOSITION OF EXPERT PANEL WORKSHOP

1.3.1 Expert Panel Membership

The expert panel consisted of the following membership: R. J. Budnitz,

Panel Chairman; P. J. Amico; P. L. Appignani; S. H. Bush; L. J. Chockie;

S. Kasturi; T. M. Laronge; and D. A. Wesely.

The panel membership represented expertise in a full spectrum of relevant

technical areas: PRAs, structures, electrical and mechanical components, component reliability, materials behavior and failure analyses, in-service inspection, operations and maintenance, as well as safety, regulatory, and aging and

life extension issues.

1.3.2 Workshop Technical Support Staff Membership

Technical participation and workshop facilitation was provided by the

workshop technical support staff, which included the following membership:

* co-technical chairman: I. S. Levy - PNL Project Leader, "NRC Plant

Aging Research Prioritization Task"

* co-technical chairman: J. Wreathall - SAIC Project Leader, "Applications of Risk Significance of Aging Methodology to NPP Component Prioritization Subtask"

* workshop leader: D. L. Brenchley (PNL)

* methodology support: A. J. Wolford (consultant to SAIC)

G. M. DeMoss (SAIC)

* rapporteurs: E. Collins (SAIC)

D. Jarrell (PNL)

In addition, the workshop had three observers representing the NRC Nuclear

Plant Aging Research Program; these were M. Vagins, Branch Chief, Electrical

and Mechanical Components Branch, NRC/RES.; A. B. Johnson, Jr., PNL NPAR Project Manager; and B. Cook, EG&G-Idaho (INEL) NPAR Project Manager.

1.6
1.4 ORGANIZATION OF THIS REPORT

This report of the expert panel workshop is organized in the following

manner:

* Section 2.0 discusses the prioritization of the TIRGALEX components

using the "risk-based" criteria. This section discusses the RSCAAMP

methodology, the data sources used by the RSCAAMP methodology, the

pre-calculated values for the PRA-based components, the application

of the RSCAAMP methodology to the non-PRA-based structures and components, and the method for determining the adequacy of current

industry aging management practices. The section continues with the

findings and their rationale from the prioritization process, the

technical insights produced by the panel during its deliberations,

and conclusions from the process.

* Section 3.0 discusses the categorization of structures and components

using the other technical criteria, e.g., the importance of aging

research on S/Cs to resolution of GSI and/or to an identified NRC/NRR

user need. Structure/component categorization for each of these two

criteria is discussed separately and includes the methodologies used,

the results of the processes, comments produced by the panel during

its deliberations, and conclusions.

* Section 4.0 presents the overall conclusions and recommendations of

the workshop.

1.7
2.0 PRIORITIZATION OF STRUCTURES AND COMPONENTS USING RISK-BASED CRITERIA

The purpose of this section is to describe the Risk Significance of Component Aging and Aging Management Practices (RSCAAMP) model and its application

by an expert panel tasked with prioritizing the TIRGALEX S/Cs for further

evaluation by the NRC Plant Aging Research Program. The data sources used to

precalculate values for PRA-based components are identified, the data vulnerabilities are identified, and the data are displayed. The prioritization

process described in Section 1.2.2.1 is followed; the results of the interim

and final rankings are tabulated, and the rationale for the findings and the

major comments and insights of the expert panel are documented.

2.1 THE RISK SIGNIFICANCE OF COMPONENT AGING AND AGING MANAGEMENT PRACTICES

METHODOLOGY AND ITS APPLICATION TO S/C PRIORITIZATION

The RSCAAMP model allows the assessment of both the risk significance of

component aging and the effectiveness of current industry management practices

for maintaining an acceptable plant risk level in the presence of component

aging. The model and its application to prioritization of PRA-based S/C and to

non-PRA-based S/Cs is described below. The application of the basic model for

assessing the risk significance of component aging is first discussed. Then,

we discuss the enhancements made to the basic model to allow an assessment of

the adequacy of current industry management practices to control aging.

2.1.1 Risk Significance of Component Aging

2.1.1.1 Basic RSCA Model and Definitions

The basic RSCA methodology was developed to evaluate a component's contribution to plant risk due to aging (Vesely 1987b, 1987c). In this model, the

change in a component's contribution to risk due to aging is a function of the

component's importance to risk (N), the rate at which the component's failure

rate is increasing due to aging (A), and the interval during which the component is aging (L). The basic methodology has been expanded from a treatment

of individual, plant-specific components to a treatment of component groupska)

for generic applications, as described in Appendix A. This basic model is

shown in Figure 2.1.

Equation (2.1), repeated from Figure 2.1, represents the basic model. The

derivation for the equation is included in Appendix B:

AR = N x A x 2 ' (2.1)

(a) In the remainder of this report, we will use "component" to represent any

level of component aggregation, whether singular, as appropriate for RPV,

or multiple, as in motor operated valves.

2.1
The change in

a component's

contribution to

risk due to

aging

The component's

normal risk

importance

I The increase in

Ithe component's

Xjfailure rate X

due to aging

A factor representing

the interval during

which the component

is aging (e.g.

between overhauls)

AR = N X A X -

FIGURE 2.1. Basic Model for the Risk Significance of Component Aging (RSCA)

where AR = The change in plant risk due to the aging of a component.

The risk measure adopted for the S/C prioritization activity

is core damage frequency (CD/year).

N = The normal risk importance associated with the component. It is

the difference between the core damage frequency calculated in

the PRA when the component is always available and the core

damage frequency if the component is defined tbe unavailable

[e.g. unavailability = 1.0; (ANSI/IEEE 1987)].'fd) This risk

importance measure is expressed as the derivative of risk (AR)

with respect to component unavailability. Hence, the units are

the same as those used for the risk measure (e.g., CD/yr).

A = The increase in failure rate (from the rate used in the PRA)

due to aging of the component. A is expressed as failures

per unit time squared [also te imed failure acceleration (Vesely

1987c)]. We use the unjhv hr- yr-1 , or the annual increase

in hourly failure rate. The panel limited A to the aging

mechanism that contributed to the risk significant failure mode,

e.g., the one used in the PRAs from which N is generated.

L = The interval during which the comppQent is aging (e.g. between

overhauls); its units are months.

Vesely (1987a) has shown that an estimate of the aging rate (A) may be obtained

by the method of moments.

The product of the normal risk importance (N) and aging failure rate (A)

is termed the Core Damage Frequency Acceleration (CDFA):

(a) Multiple failures of similar or dissimilar components in redundant systems, even in the same cut set, are not explicitly considered in the current prioritization activity.

(b) To convert to the units for AR (CD/yr), A is multiplied by 8760 hrs yr'

and L is multiplied by 12 yrs- .

2.2
CDFA = N x A . (2.2)

CDFA is also described as the "risk significance of aging" (Vesely 1987c).

Precalculated values of N, A, and CDFA were produced for the PRA-based

components and presented to the panel for their review. These data and their

sources are presented in the following section.

2.1.1.2 Precalculated Data for the PRA-Based Components

Data Sources. The basic RSCA methodology applies directly only to PRAbased components. It uses risk importance data calculated from four PRAs from

NRC-funded studies and uses aging failure rates calculated from information in

the INEL data base, generated for the NPAR Program. These data, which support

the calculations of N, A, and CDFA, are in Appendix B.

Four PRAs were selected to contribute to the generic risk importances (N)

needed for this study. These PRAs were selected because they were under study

in the NPAR program's RSCA effort funded through INEL; the plants represent

three of the four U.S. nuclear power plant vendors. The four PRAs are for:

* ANO-1 Unit 1: An Integrated Reliability Evaluation Program (IREP)

PRA performed on a PWR built by Babcock and Wilcox (B&W).

(NUREC/CR-2787)

* Grand Gulf: A Reactor Safety Study Methodology Application Program

(RSSMAP) PRA performed on a BWR built by General Electric (GE).

(Hatch 1981)

* Calvert Cliffs: A RSSMAP PRA performed on a PWR built by Combustion

Engineering. (Kolb 1981b)

* Oconee: A RSSMAP PRA performed on a PWR built by Babcock and

Wilcox. (Kolb 1981a)

The components considered in each of the four plant PRAs and their relationship to the TIRGALEX list of S/Cs is shown in Table 2.1. Not all of the

TIRGALEX S/Cs are considered by these PRAs. However, the panel was not

restricted to the components considered in the PRAs, and, in fact, the panel

disaggregated existing component groups and added new components as well. As

noted in Section 1.2.2.1, expert judgment is required to develop the N and A

values for the non-PRA-based components, and is used to evaluate the precalculated values for the PRA-based components.

The number of components contributing to each PRA is shown in Appendix B

(by component group in Table B.1 and by system in Table B.2). The component

risk importances from these PRAs are shown in Table B.3. These importance

values were calculated by SAIC. To provide the generic importance of component

groups, medians are taken for the importances for that component group.

2.3
TABLE 2.1. Components in the Four-Plant PRA Studies

Compared to TIRGALEX List of Components

Componer . Ca I vert Grand

TIGALEX 4-PIanL Study Cliffs Gult uconee ANO-I

1. Reactor pressure vessel

2. Conta i nnent

3. Concrete structures

4. RCP and safe ends

5. Other safety-related

piping

6. Steain q(enerator

7. KC pump casing

8. Pressurizer

9. CRD inechanisin

10. Cables, connectors and

nenetrations

11. Emeiygecy diesel yen.

12. Reactor internals

13. RPV support

14. Recirc. piping

15. Snubbers

I6. instruments and controls

17. Switchijear and relays

Di ese I gemmrators

Circuit breakers

C onl ctor s

Relays

Transfer sui 'ch

Bistable

I X v1

X

K

x

A x

X

X

X

X

18. Valves Check

Motor operated

Air operatel

Manua 1

Ilydraul ic

Safety/rel iet

Solenoid

X

X

X

X

K K

K K

X

K K

K

K

K

X

19. Pump1i s

20. Motors

21. Turbines

22. Ileat exchange

23. Compressors

24. Fans/chillers

25. Batteries

26. Battery chargers/

i nverters

27. Trd.. urmers

28. Fuel storage racks

29. Accumulators

Motor driven

Turbine driven

Air conditioner

Heat exchanger

Fans

Chillers

Batteries

Battery charger

Inverter

Others

Bus-ac

Bus-dc

Recti fier

Turbotenerator

X

K

K

K K

K

X

K

K

K K K X

K

K

K

K

K

K

K

2.4
While the median values for risk importances were presented to the panel,

the individual importances from each of the four PRAs were also made available

(as shown in Table B.3). Systems importances and the importances of the

components within systems were also calculated and provided to the panel

(Table B.4). (Only the importances of components within the systems were used

by the panel.) A S/C risk importance value was selected by the panel. In the

cases where the panel made changes from the precalculated values, it was

usually based on consideration of newer PRAs within their experience.

As shown in Equation (11) of Appendix A, the increase in failure rate due

to aging (A) is a function of the aging fraction (f), the mean time-to-aging

failure (TA), and the component failure rates (). The aging fractions were

taken from Meale and Sutterwhite (1987), the large-scale aging data analysis

performed by INEL using the nuclear plant reliability data system (NPRDs) data

base. For that study, experts from INEL evaluated the causes of large numbers

of component failures and then classified the failures to determine the fraction caused by age-related stressors.

These data were generated from a review of more than 2000 NPRDS records

for three vendors [B&W, Westinghouse Electric Corporation (WEC), GE] and fifteen safety related systems. The systems are identified in Table B.5. The

supporting data on aging failure rates are presented in Tables B.5 through B.8,

including failure rates of components within systems.

The mean time-to-aging failures (TA) (Table B.5) also comes from Meale and

Sutterwhite (1987). The aging fractions (f) are shown in Table B.6; they were

calculated for the various safety systems that the data allowed. When no significant differences between systems were noted, the fractions were actually

calculated at the aggregate level. These quantities are from straight-forward

calculations from the NPRDS data. The failure rates (Table B.7) are best

estimate values from various sources. The calculated increase in failure rates

due to aging (A) and failure rate doubling times are also presented in this

table. The panel relied heavily on the doubling times. For components not

represented in the data base, the panel developed their own estimates of aging

failure rates. Sometimes these panel estimates were based on comparison with

values for other components, or a positioning of the components based on

perceived relative aging rates.

Precalculated Values. Precalculated values of N, A, and CDFA were constructed from Equations (2.1), (12) in Appendix A, and (2.2), respectively, for

the PRA-based components and from the data sources described above; they are

presented in Table C.1. The data source for N is from the generic value in

Table B.3; the source for A is Table B.7 which in turn, is, based on data found

in Tables B.5 and B.6.

The data for N, A, and CDFA in Table C.1 are arranged by TIRGALEX component number and name. Where the TIRGALEX component represents a group of

components for which the PRAs provide subgroupings, these are identified by

a, b.

2.5
2.1.1.3 Application to Non-PRA-Based Components

Structures and components not appearing in PRAs fall into two categories:

.1) those primarily intended to prevent core damage accidents and 2) those

designed to mitigate and control post-accident releases. Several components,

such as the RPV, fall into each category.

The following TIRGALEX S/Cs (with their respective number from Table 1.1)

have an impact on the prevention of core damage but are not included in the

PRAs used in this study:

1. Reactor pressure vessel

3. Other Category I concrete structures

4. Reactor coolant piping and safe ends

5. Other safety-related piping

6. Steam generators

7. Reactor coolant pump casing

8. Pressurizer

9. Control rod drive mechanism

10. Cables, connectors, and penetrations

12. Reactor internals

13. RPV support (sliding foot) (PWR)

14. Recirculation piping, safe ends (BWR)

15. Snubbers

16. Instruments and controls

The following TIRGALEX S/Cs have an impact on the mitigation of releases

from core damage accidents.

1. Reactor pressure vessel(a)

2. Containment (metal and concrete)

3. Other Category 1 concrete structures(a)

4. Reactor coolant piping and afe ends(a)

5. Other safety-related pipingta)

7. Reactor coolant pump casing(a)

13. RPV support (sliding foot) (PWR)(a)

14. Recirculation pipiig safe ends (BWR)(a)

28. Fuel storage racks?9)

For the structures and components having a dual role, the panel assigned them

to core damage prevention. This left only the containment to be considered by

the panel as a structure used only for accident mitigation.

Solely based on judgment, the panel assigned the N and A values to the

non-PRA-based accident prevention components and incorporated them into the

(a) Associated with both core damage prevention and accident mitigation.

(b) Fuel storage racks prevent fuel damage accidents; they were arbitrarily

classified with the core damage prevention components.

2.6
existing lists for the PRA-based components. Next, the panel performed the

equivalent estimations for the structures associated with accident mitigation.

The difference between the accident prevention and mitigation S/Cs is in the

interpretation of risk importance. As usually evaluated in PRAs, the risk

importance parameter indicates the sensitivity of the core damage frequency to

increases in the component failure probability. For S/Cs whose major role is

the mitigation of accident consequences, it was necessary to adopt an equivalent risk importance value that reflected the panel's opinion of the overall

significance of accident mitigation to safety. It was recognized that there is

no uniquely correct interpretation of relative importance of accident prevention and accident mitigation, and that this is a matter of continuing NRC

policy evolution. At this point, the panel judged it appropriate to consider

the accident-mitigation S/C (containment), after a core damage accident has

occurred, to have a risk importance equal to the highest risk importance for

any of the accident-prevention S/Cs. This recognized that the containment provides a barrier to the release of radioactive materials to the public that is

of equal importance to the RPV. Based on this assessment, the panel incorporated the N and A values for the accident-mitigation S/Cs into the previously

integrated lists for the accident-prevention S/Cs. For this final integrated

list of all S/Cs, the risk significance of aging (CDFA) was then calculated.

Next, the task of assessing the effectiveness of current industry aging management practices was undertaken.

2.1.2 Adequacy of Current Aging Management Practices

Enhancements were made to the basic RSCA model so that the effect of current industry aging management practices on plant risk could be evaluated in an

expert panel framework.

Aging management practices have the potential for adequate control of

aging-enhanced plant risk. The enhancements contained within the RSCAAMP

model, described below, model this concept and allow for direct input of quantitative, expert opinion to evaluate the effectiveness of aging management

practices. The model does this by 1) defining an acceptable or control value

of the component's aging risk contribution and identifying those components

whose risk contribution exceeds this value; 2) defining current industry aging

management practices and evaluating their adequacy for maintaining the risk

contribution of these aged components within the control value; and 3) calculating the relative contributions to plant risk of aged components given this

defined adequacy of industry practices.

2.1.2.1 Acceptable (Control) Value Factors for Component Risk

Contribution

We define an acceptable or control value of AR as being ARce AR is the

limit we place upon the additional contribution to plant risk due to the aging

of a component. The control value is established by considering the NRC's

safety goal core damage frequency as the measure of total plant risk. Then, a

portion of this goal value is allocated to each component such that the sum of

the individual component risk contributions will not exceed the goal value.

2.7
Based on this, the panel chose E-7 CD/yr as the working value for ARC. The

allocated value for each component is the limit placed upon ARc.

Having this control value of component risk contribution, we can screen

from further consideration those components with risk significance of aging

(CDFA) so low that there is no need to assess the effectiveness of current

industry aging management practices. The basis for this screening is the

increase in risk that would result if aging of the component were allowed to

continue, without overhaul or replacement, for the full 40-year design life of

the plant. This would be a conservative estimate for most components; the

panel used this basis for screening. From Equations (2.1) and (2.2), the

calculated increase in plant risk for this case (e.g., that the screening

criterion is AR(40) 1E-7 CD/yr) is:

L242

AR = N x A x - CDFA 40 1 E-7 CD/yr < Ac (2.3) 2 2x~- 1 - D/r CA

Equation (2.3) was applied to the integrated list of PRA-based, non-PRA-based

accident prevention, and accident mitigation components, for which the panel

had previously developed N and A values (and for which CDFA values were then

calculated). A few components were eliminated. The remaining components are

those with risk significance of aging sufficiently high that adequate aging

management practices are required to bring component risk contributions down

within ARc.

To assess the adequacy(a) of the current industry aging management practices in controlling the changes in risk due to aging, we continue with the

following steps. With a control value defined for a component's risk contribution, ARc2 we now solve Equation (2.3) for an acceptable or control overhaul

interval, Lc* It can be shown that:

2 x AR L = Di Ac (2.4)

The control overhaul interval Lc may be viewed as a theoretical value arrived

at by applying desirable risk control and is the limit we place upon the component's aging interval with no mitigating action being taken (no refurbishment

or overhaul). Equation (2.3) may now be expressed in terms of ARc and Lc:

AR = CDFA x 2- (2.5)

(a) "Adequacy" refers to the theoretical potential for limiting the changes in

risk due to aging through its detection and mitigation.

2.8
2.1.2.2 Adequacy of Current Industry Practices Factors

For the panel to evaluate the adequacy of current aging management practices, a value for L that is representative of current industry practice is

required. First, we define Lact as the actual (effective) interval between

component overhauls that is representative of current industry practice for

that component. Next, we define the ratio of Lact to Lc (the control overhaul

interval) to represent the adequacy of current practices in controlling risk:

L 2

act' 2 _ adequacy of current industry aging management (2.6)

practices in controlling risk 2

The effect of industry aging management practices on the risk contribution of

an aged component is represented by:

AR AR c a (2.7)

From this equation, it may be observed that the adequacy of industry's aging

management practices has considerable impact on the risk contribution of aged

components since it enters the risk equation raised to the power of two.

If industry practices are adequate,

(act < 1, and AR < ARc (2.8) =~~~~~~

That is, the risk increase due to aging, when aging is adequately managed,

remains within an acceptable value.

If industry practices are not adequate,

(Lt)2 > 1, and AR > ARc (2.9)

That is, the risk increase due to aging, when aging is inadequately managed,

exceeds the acceptable value.

We now define La t in terms of parameters that will allow us to elicit

quantitative input relative to the adequacy of current aging management practices, and to do so in terms of both the adequacy of aging detection practices

and the adequacy of aging mitigation practices.

L id

Lact P~7 x RD210

2.9
where Lact = The actual (effective) interval, representative of current

industry practices, that the component ages without mitigation.

Lind The surveillance/test interval representative of current industry

practices. The panel limited L d to the surveillance/test

interval for the risk significant failure mode (e.g., that used

to generate N).

PD= The probability of successfully detecting aging degradation in

the component within the surveillance/test interval. The panel

limited this aging degradation to that causing the increased rate

of risk significant failures (A), these being for the failure

mode used to define N.

PR/D = The probability of successfully mitigating this aging degradation

process, given its successful detection in the component (e.g.,

the aging clock returned to zero or "good as new").

Substituting Equation (2.10) into (2.7), the effect of detection and

mitigation practices on a component's aging risk contribution is:

AR = Rc c x( ~L x pind L xP (2.11)

c PD X R/D/

To elicit the panel's expert opinion on the adequacy of industry detection

and mitigation practices, we posed the following three questions:

1. What is the representative industry interval for inspection/testing

for the risk significant failure mode of the component (Lind)?

2. What is the probability that the inspection/test method used will

identify the existence of the risk significant aging mechanism (PD)?

3. Given successful detection, what is the probability of successful

mitigation (repair, replacement) of the risk significant aging

degradation in the component (PR/D)?

Lact was calculated from the panel's responses to these questions, which

were input to Equation (2.10).

We can now quantify the adequacy of current industry aging management in

controlling risk practices by the detection and mitigation of component aging.

The adequacy of current industry practices for detecting aging degradation is:

A X(Lind x 1 ) (2.12)

The adequacy of current industry practices for mitigating aging, given

successful detection, is:

2.10
A = P , RID R/D (2.13)

where we are stipulating that given successful detection in the interval Lnd,

the component is immediately shut down (e.g., in accordance with relevant

regulations) and repairs are performed such that the aged component no longer

contributes to plant risk.

From Equations (2.8), (2.10), (2.12) and (2.13), we can also specify an

acceptable industry surveillance/test interval, (Lind) as:

(L ind)acc - (PD x PR/D) x Lc

or

( ind)acc - D Lc (2.14)

given the stipulation with regard to immediate component shutdown upon

successful detection of aging noted above.

This relationship shows that, as the probability of detection of aging

degradation decreases, the industry surveillance/test (S/T) interval (Lind)

must also decrease (e.g., frequency of S/T increase) in order to stay with in

the control interval (Lc) that defines acceptable risk (ARC).

An illustration of the RSCAAMP model [Equation (2.11)] is presented in

Figure 2.2. In this figure, the risk contribution (AR) of a component is

plotted against the time during which the component is aging. We've plotted

time as increments of Lc, the control overhaul interval, and have set the value

of Lc at 1 month. Two cases are shown. In the first case, we set Lact = 1.

L is not exceeded, that is, industry practices are adequate, and as a result,

Ak is not exceeded. In the second case, we set L d = 0.25 month, and both PD

and PR/D = 0.25. Thus, Lact 4 and Lc is exceede . That is, industry practices are not adequate; as a result, ARc is exceeded.

2.1.2.3 Relative Contribution to Plant Risk of Aged Components

The fully developed equation for the RSCAAMP model becomes:

AR = N x A x - 2 (PD R/D/ (2.15)

This model is illustrated in Figure 2.3.

The factors N, A, Lnd, P0, and PR/D are used as the S/C prioritization

criteria. Each factor is normalized into a scoring scheme, with a score of 5

representing the highest risk effect from that factor and 1 representing the

lowest risk effect on a logarithmic scale (see method described in Appendix B).

The scores for each S/C are accumulated (addition of logarithms) to obtain a

figure of merit (FOM), which represents the AR value for that S/C. Final

2.11
16 xARc

t

AR

AR,

0 2Lc 3K 4Lc

Time

FIGURE 2.2. Effect of Industry Aging Management Practices on Component

Aging Risk Contribution [Equation (2.11)]

.

[he change in

a component's

contribution

risk due to

aging

'The component's

normal risk

to = importance x

The increase in

the component's

failure rate

due to aging

A factor representing

the effectiveness of

X industry practices in

managing component

aging [the actual

(effective) interval

between overhauls]

.

AR N x A

( L 2

f dP/

FIGURE 2.3. The Risk Significance of

Model

Component Aging Management Practices

2.12
ranking (prioritization) of the S/C is based upon these calculated AR values.

This ranking incorporates both the risk significance of aging and the effectiveness of industry practices in maintaining aging within an acceptable risk

level.

The process used by the expert panel for the risk-based prioritization of

the TIRGALEX S/Cs is summarized in Figure 2.4. Because many of-the specific

numerical values generated for the risk based factors are the calculated

result, not the cause, of the relative positioning of the components, these

numerical values do not have a sufficient basis to justify their use to perform

analyses for which the numerical value is a necessary and dominant part of the

answer.

2.2 PANEL FINDINGS AND TECHNICAL RATIONALE

The panel's findings together with their rationale, are presented in the

following sections. While the broad rationale are discussed with the findings,

detailed rationale are provided in Appendix D.

2.2.1 Final Component Selection

The panel reviewed the TIRGALEX component list as it was amplified by the

4-Plant PRA study, which disaggregated some component groups and added S/C-30:

ac/dc buses. During deliberations on N and A, the panel developed further disaggregation of the existing groups; added a new group (S/C-31: bolts); and

deleted from further consideration the component group, S/C-20: motors. This

group was viewed by the panel as redundant since motors are contained within

the groups valves, pumps, etc., (as in motor-operated valves), and should not

be decoupled from the evaluation of the driven components.

Table 2.2 presents the components identified by TIRGALEX, the 4-plant

PRAs, and the final list evaluated by the expert panel. The panel's disaggregation of some-groups is evident. The panel's rationale for this consisted of the following:

* S/C-2: containment: to permit consideration of aging of BWR-MK-I

containments.

* S/C-4: reactor coolant piping and safe-ends: to permit consideration of variations in risk importance between large- and smallbreak loss-of-coolant accidents (LOCA) and of aging by stress corrosion cracking that occurs in BWR piping.

* S/C-5: other safety-related piping: to permit consideration of a

major failure in large piping (6 in.-24 in.), represented by the

important service water system, and a break in smaller piping (6 in.-

10 in.) such that high pressure injection (HPI) could just keep up

with the coolant loss.

2.13
TIRGALEX List of Components

(Table 1.1)

Expert Panel List of Components and Structures

(Table 2.2)

Accident Prevention S/Cs

PRA-Based S/C

I

Panel Reviews

Estimates of N

(Table 2.3)

Panel Reviews

Estimates of A

(Table 2.4)

Non-PRA-Based S/Cs

I

Panel Estimates

Values of N

(Table 2.3)

Panel Estimates

Values of A

(Table 2.4)

Accident Mitigation S/Cs

Panel Estimates

Equivalent N Value

(Table 2.3)

Panel Estimates

Values of A

(Table 2.4)

Integration of all S/Cs

Evaluate Values

for Credibility

& Apply Screening

Criterion

(Table 2.5)

Panel Estimates Lind, D,

PR/ (Table 2.6)

Rankings Based on

N(TA$bL .R/D PDe , AR

Criteria Scores

and S/C Ranks

(Table 2.8)

FIGURE 2.4. Expert Panel Process for Risk-Based Prioritization

2.14
TABLE 2.2. Structures/Components Evaluated by Expert Panel

TIRGALEX Groups

1. Reactor pressure vessel

(RPV)

2. Containment

4-Plant PRA Components

Components Chosen

by the Expert Panel

1. RP v

2. Containment

a. BWR

b. Other

3. Other Category I 3. Other Category I structures

structures

4. Reactor coolant piping

(RCP) and safe ends (SE)

4. RCP & SE

a. Large LOCA

b. Small LOCA, PWR c. Small LOCA, WR

5. Other SRP () a. Large (10-24 In.) ?&ea

b. Small (6-10.) pipe

5. Other safety-related

piping (SRP)

6. Steam generator S/G) 6. Steam generator a. S/G tube

b. S/G shell

7. Reactor coolant pump (RCP) 7. RCP casing casing

8. Pressurizer

9. Control rod drive

mechanism (CRDM)

10. Cables, connectors, and

penetrations

8. Pressurizer

9. CRDM

a. BWR

b. PWR

10. Cables, connectors,

a. Cables

b. Connectors

(* Penetrations as part of

containment)

11. Diesel generator

12. Reactor nternals

13. RPV support (PWR)

11. Diesel generator

12. Reactor Internals

13. RFPV support (PWR)

14. Recirculation piping safe

ends

11. Diesel generator

14. Small LOCA, BWR

15. Snubbers

16. Instruments and Controls

(&C)

17. SwItchgear/relays

16. Instruments and Controls

(I&C)

a. Thermostat

17. Switchgear/relays

a. Relay (load)

b. Circuit breaker

c. Transfer switch

d. Bistable trip unit

15. Snubbers

16. Instruments and Controls

(I&C)

a. Thermostat (c)

b. Transfer switc

c. Bistable trip""

17. Switchgear Relays

a. Relay

b. Circuit breaker

2.15
TABLE 2.2. (contd)

TIRGALEX Groups

18. Valves

4-Plant PRA Components

18. Valves

a. Air-operated valve (AOV)

b. Check valve c. Hydraulic valve

d. Manual valve

e. Motor-operated valve

(MOV)

f. Safety/relief valve

(S/RV)

Components Chosen

by the Expert Panel

18. Valves

a. Air-operated valve

b. Check valve c. Hydraulic valve

d. Manual valve

e. Motor operated valve

f. Safety/relief valve

19. Pumps 19. Pumps

a Motor driven pump

b. Turbine driven pump

19. Pumps

a. Motor driven pump

b. Turbine driven pump

20. Motors 20. Motors (included In valves,

pumps, etc.)

21. Turbines

22. Heat exchangers

21. Turbines 21. Turbines

22. Heat exchangers

a. Heat exchangers

b. Air conditioners

22. Heat exchangers

23. Compressor 23. Compressor(d)

24. a. Chiller

b. Fan

24. Fans/chillers 24. Fan

25. Batteries 25. Batteries

26. Battery chargers/inverters

27. Transformers

26. Battery chargers/inverters

a. Battery chargers

b. nverters

27. Transformers

28. Fuel storage racks

29. Accumulator/tanks

25. Batteries

26. Battery chargers/inverters

a. Battery chargers

b. Inverters

c. Rectifier

27. Transformers

28. Fuel storage racks

29. Tanks

a. Medium pressure tank

b. Atmospheric pressure tank

c. High pressure tank

30. AC/DC buses

a. AC bus

b. DC bus

29. Tank

30. AC/DC bus 30. AC/DC buses

a. AC bus

b. DC bus

31. Bolts

(a) Represented by service water system piping for which HPI could not keep up If break occurred.

(b) Represented by piping In letdown and reactor water cleanup (RWCU) systems, for which HPI could

Just keep up If small break occurred.

(c) Moved by panel from S/C-17, 4-Plant PRA.

(d) Compressors In Instrument AIr System.

(e) Moved by panel from non-assigned S/C (see Table 2.1).

2.16
* S/G-6: steam generator: to permit consideration of two failure modes

having significantly different risk importances and aging failure rates.

* S/G-10: cables, connectors and penetrations: to permit separate

consideration of cables and connectors due to differences in their

importance to safety (penetrations to be part of containment).

2.2.2 Component Scores and Ranks

The following subsections present the findings generated for the factors

related to both the risk significance of aging [N, A, CDFA, R(40)] and to the

adequacy of aging management practics in reducing the risk contribution of

aging [Lind, Pn, PR/D, and (LactlLc) As noted in Section 2.1.2, only N, A,

Lind P and PR/n re the prioritization criteria; scores are developed for

them. An integrated, final ranking of all S/Cs is then presented. An overview

of the rationale for these findings is presented with the findings; details are

discussed in Section 2.2.3 and in Appendix D.

2.2.2.1 Risk Significance of Aging Data

Table 2.3 presents the S/C risk importance (N) values; both the precalculated values (for the PRA-based components) and the expert panel values for all

components (the PRA-based, the non-PRA-based accident prevention components,

and the non-PRA-based accident mitigation components) are shown. In the column

for the precalculated values, the type of component is indicated for which no

values were calculated.

The panel assigned a risk importance of 1.0 to S/C-1: RPV since its

failure (gross rupture) results directly in core damage; sufficient cooling

could not be provided. The RPV value, being the highest for accident prevention S/Cs, became the source of the value assigned for the equivalent importance of S/C-2.a: containment (BWR) and S/C-2.b: containment (other). The

basis for this panel decision was the recognition of the containment's function

as a barrier to radioactive releases to the public. S/C-3: other Category I

structures were also given a risk importance value of 1.0 because of their role

in providing support of equipment necessary for core cooling (cable trays,

piping, etc.) and, in some plants, in providing a containment function.

Most of the precalculated values (for the PRA-based'components) were left

unchanged. Of those that were changed, the majority were reduced, reflecting

both the panel's insights gained from newer, improved PRAs and the improved

designs of later plants that resulted in reduced risk importance for these components. Two components, diesel generator and battery, were increased in risk

importance compared to the precalculated values. The diesel generator value

was raised because of the high risk contribution of loss of offsite power in

the recent PRAs; the panel elevated the value for batteries to reflect their

similar risk importance to the diesels.

After the S/C-1: RPV, S/C-2a: containment, S/C-3: other Category I structures, and S/C-2b: containment (other) (all of which were ranked 5), the next

2.17
TABLE 2.3. Structure/Component Risk Importances(a)

Component

1.

2.a.

3.

2. b.

30.b.

1O.a.

9.a.

12.

17.b.

17.a.

30.a.

29.b.

18.e.

11.

lO.b.

25.

27.

19.b.

19.a.

5.a.

22.

29.a.

16.a.

4.b.

5.b.

4.c.

9.b.

18.b.

24.a.

24.b.

23.

18.a.

6.a.

26.a.

8.

7.

29.c.

4.a.

31.

18.f.

16.c.

18.c.

18.d.

6.b.

16.b.

26.b.

26.c.

15.

21.

28.

13.

RPV

Containment (BWR)

Other concrete structures

Containment (other)

DC bus

Cables

CRDM (BWR)

RX Internals

Breaker

Relay

AC bus

Tank (atmos. pres.)

Motor operated valve

Diesel

Connectors

Battery

Transformer

Turbine pump

Motor pump (d)

Large other safety pipe

Heat exchanger

Tank (medium pres.)

Thermostat

PWR pipe (small LOCA) W

Small other safety pipe e BWR Pipe (small LOCA)

CRDM (PWR)

Check valve

Chillers

Fan

Compressor (Instr. air)

Air operated valve

S/G tube

Battery charger

Pressurizer

RC P Casing

Tank (high pres.)

RC P SE large (LOCA)

Bolts

Safety/relief valve

Bistable

Hydraulic valve

Manual valve

S/G shell

Transfer switch

Inverter

Rectifier

Snubbers

Turbine

Fuel rack

RPV support

Precalculated N

(C D/yr)

(b)

AM (c)

AP

AM

l.lE-l

AP

AP

AP

7.2E-2

4.8E-2

4.3E-2

AP

2.2E-2

9.2E-4

AP

9.2E-4

1.2E-2

9.3E-3

6.7E-3

AP

6.4E-3

AP

6.OE-3

AP

AP

AP

AP

1.8E-2

AP

7.6E-3

AP

3.2E-2

AP

1.lE-4

AP

AP

AP

AP

AP

3.5E-4

1.2E-5

3.3E-4

3.3E-2

AP

4.7E-6

4.2E-4

4.7E-4

AP

1.O-t

AP*

AP

Panel's N

(C D/yr)

1 .OE-O

1.OE-O

I .OE-O

1.OE-O

1.IE-1

1 I E-1

l.OE-1

1.OE-1

7.2E-2

4.8E-2

4.3E-2

2.5E-2

2.2E-2

2.OE-2

2.OE-2

2.OE-2

1.2E-2

9.3E-3

6.7E-3

6.4E-3

6.4E-3

6.OE-3

6.OE-3

1.OE-3

1.OE-3

I .OE-3

1.OE-3

8.OE-4

6.OE-4

6.OE-4

5.OE-4

3.2E-4

3.OE-4

I .IE-4

1.OE-4

I .OE-4

1 .OE-4

1 .OE-4

1.OE-4

I .OE-4

1.2E-5

1 .OE-5

1.OE-5

I .OE-5

4.7E-6

4.7E-6

4.7E-6

1 .IE-6

1.OE-6

1.OE-6

1.OE-7

Score

5

5

5

5

4

4

4

4

4

4

4

3

3

3

3

3

3

3

3

3

3

3

3

2

2

2

2

2

2

2

2

2

1

1

(a) Caution: only the relative component position (score) is Important;

do not base any subsequent studies on the numerical values presented

here.

(b) AP signifies a non-PRA-based accident prevention component.

(c) AM signifies a non-PRA-based accident mitigation component.

(d) 10-24 In. pipe represented by service water piping (see Table 2.2).

(e) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(f) AP* signifies the special case of fuel damage prevention.

2.18
highest ranked components (4) consisted primarily of electrical components; the

numerous and widely distributed electricals, such as cables, breakers, relays,

and connectors, were well represented.

Table 2.4 contains the S/C aging failure rate (A) values. As with

Table 2.3, both precalculated and expert panel values are indicated, and the

type of non-PRA-based component is also shown. The panel focused consideration

of A on the risk significant failure mode that was used to define the component's risk importance (N).

Most of the precalculated values for the increase-in-component failure

rates due to aging were left unchanged by the panel. S/C-25: battery was

increased by a factor of ten and received a rank of 4; S/C-30 b: dc bus was

increased by a factor of ten and S/C-30 a: ac bus was increased by two orders

of magnitude, but they both were ranked only at the top of 1. The aging rates

for several components were reduced compared to the precalculated values; the

greatest reduction was two orders of magnitude for S/C-18.b: check valves, for

which two failure mechanisms having differing rates (one high, one low) were

noted.

Many of the changed aging failure rate values, as well as those for other

components, resulted from the panel's consideration of the numbers of failures

likely in 40 years (or component average lifetime, if less than 40 years), the

doubling rates to achieve these, and then, the back-calculated aging failure

rates. Other component aging failure rates were the result of the panel's

relative positioning (ranking), from comparing one against another; and other

values resulted from a combination of both methods. Where the panel's disaggregation of component groups did not provide sufficient distinction of the

risk significant aging and failure mechanisms [as in the pitting corrosion of

medium- and atmospheric-pressure tanks (S/C-29:a and b) compared to the rupture

of high-pressure tanks (S/C-29c)J, the panel tried to develop effective values

for the competing mechanisms. This probably led component aging in the harsher

environment to bias the aging failure rate (or its relative ranking).

The RPV calculations started with a low known (accepted), nonaging failure

rate. This was then increased by specifying an aging failure rate that, at the

end of 40 years, would represent a credible increase in vessel failure rate.

This value was used for similar types of components at similar pressure and

temperature, S/C-29.c: high pressure tanks, S/C-6.b: steam generator shell, and

S/C-8: pressurizer. All had low aging failure rates. S/C-2.b: concrete containment and S/C-3: other Category I structures had, in the panel's view, the

lowest rates (e.g., positions in the table); and as rates of the other components were generated, the numerical values of these structures resulted.

The reader is cautioned, as noted in Section 2.1.2.3, that it is the

relative positioning of the components, not the absolute numerical values for

the N and A factors, that is important. Many of the values, particularly for

A, were the result, not the cause, of such positioning.

2.19
TABLE 2.4. Structure/Component Aging Failure Rates(a)

Component

15.

6.a.

26.b.

11.

18.a.

19.b.

24.a.

18.f.

31.

23.

18.a.

25.

5.b.

17.a.

16.b.

19.a.

24.b.

16.a.

16.c.

18.c.

2.a.

21.

26.c.

26. a.

4.c.

lO.B.

17.b.

22.

18.b.

9.a.

5.a.

10.a.

18.d.

12.

27.

30.b.

30.a.

9.b.

4.b.

28.

1.

29.b.

29.c.

6.b.

8.

29.a.

7.

4.a.

13.

3.

2.b.

Snubbers

S/G tube

Inverter

Diesel

Motor operated valve

Turbine pump

Chillers

Safety/relief valve

Bolts

Compressor (Istr. air)

Air operated valve

Battery

Small other safety pipe(c)

Relay

Transfer switch

Mbtor pump

Fan

Thermostat

Bistable

Hydraulic valve

Containment (BWR)

Turbine

Rectifier

Battery charger

BWR Pipe (small LOCA)

Connectors

Breaker

Heat exchanger

Check valve

CRDM (BWR)

Large other safety pipece)

Cables

Manual valve

RX Internals

Transformer

DC bus

AC bus

CRDM (PWR)

PWR pipe (small LOCA)

Fuel rack

RPV

Tank (atmos. pres.)

Tank (high pres.)

S/G shell

Pressurizer

Tank (medium pres.)

RC P casing

RC P SE large (LOCA)

RPV support

Other concrete structures

Containment (other)

Precalculated A

(hr-yr )

AP(b)

AP

4.9E-6

1.6E-6

2.6E-7

2.7E-6

AP

6.7E-7

AP

AP

4.OE-7

3.4E-8

AP

9.1E-8

2.3E-7

2.2E-7

2.1E-7

1.5E-7

1.4E-7

6.?g)

AM

3.7E-6

8.7E-8

3.5E-8

AP

AP

1.6E-8

1.4E-7

3.8E-7

AP

AP

AP

2.2E-8

AP

1.7E-8

1.lE-10

l.IE-11

AP

AP

AP

AP

AP

AP

AP

AP

AP

AP

AP

AP

AM

Panelts A

(hr -yr )

5.IE-6

5.OE-6

4.9E-6

3.6E-6

3,6E-6

2,7E-6

1.5E-6

6,7E-7

5.1E-7

5.OE-7

4,OE-7

3,4E-7

3.OE-7

2.5E-7

2.3E-7

2.2E-7

2.1E-7

1 .5E-7

1.4E-7

1 .3E-7

1.OE-7

1.OE-7

8.7E-8

3.5E-8

3.OE-8

2.7E-8

1.6E-8

1 .4E-8

3.8E-9

3.OE-9

3.OE-9

2 7E-9

2.2E-9

2.OE-9

1.7E-9

I l.E-9

1.1E-9

3.0E-l 1

l.OE-1 

5.OE-12

2.OE-12

1 .OE-12

1.OE-12

1.OE-12

1.OE-12

1 .OE-12

1.OE-12

1.OE-12

l.OE-12

1.OE-13

1.OE-13

Score

5

5

5

5

5

5

4

4

4

4

4

4

4

4

4

4

4

3

3

3

3

3

3

3

3

3

2

2

2

2

2

2

2

2

2

1

(a) Caution: only the relative component position (score) Is Important;

do not base any subsequent studies on the numerical values presented

here.

(b) AP signifies a non-PRA-based accident prevention component.

(c) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(d) AM signifies a non-PRA-based accident mitigation component.

(e) 10-24 n. pipe represented by service water piping (see Table 2.2).

(f) AP* signifies the special case of fuel damage prevention.

2.20
Table 2.5 presents the CFA and AR (40). values for the panel's S/Cs calculated from the panel's values for N and A in Tables 2.3 and 2.4, respectively. The S/Cs are arranged in decreasing order of R(40), the increase in

risk in 40 years; an interim rank (e.g., without consideration of current

industry aging management practices) is also calculated. A screening of S/Cs

was performed by the panel based on the value selected to represent an acceptable risk contribution due to component aging, ARc 1E-7 CD/yr [see Equation (2.3)]. The S/Cs with AR(40) values 1E-7 were eliminated from subsequent consideration of the adequacy of aging management practices. On this

basis, the following S/Cs that were eliminated:

S/C-4a: reactor coolant piping/safe ends (large LOCA)

S/C-4b: PWR pipe (small LOCA)

S/C-6b: S/G shell

S/C-7: RCP casing

S/C-8: pressurizer

S/C-13: RPV support (PWR)

S/C-28: fuel storage rack

S/C-29a: tank (medium pressure)

S/C-29c: tank (high pressure)

The remaining S/Cs are those with risk significance of aging sufficiently

high that adequate aging management practices are required to bring their risk

contribution down to ARc*

2.2.2.2 Adequacy of Current Industry Aging Management Practices

The S/Cs in Table 2.5 that remained after the screening (e.g., those

having a AR > 1E-7 CD/yr in 40 years) were evaluated by the panel for adequacy

of industry management practices in reducing the risk increase from component

aging to the control value of 1E-7 CD/yr.

Table 2.6 presents the data generated for the assessment of the adequacy

of current industry aging management practices. The panel provided inputs foS

Lind, Pn- and PR'n [see Equation (2.10)]. Values for Lc, Lact, and (Lact/Lc)

were cayculated'Oter the workshop from the panel's inputs and Equations (2.4)

and (2.10, respectively. Since AR is directly related to the adequacy factor,

(Lact/Lc) , [see Equation (2.7)], he greater the inadequacy of industry practices to mitigate aging [(Lact/Lc) >1] the larger the risk increase Therefore, the S/Cs are arranged with those having the highest (Lact/Lc)2 values at

the top. The table shows that very few S/Cs (of those S/Cs remainin after the

screening) have adequate aging management practices [e.g., (Lact/Lc) < 1.0].

Values are also shown in the table for Lc [from ARc < E-7 (CD/yr) and

Equation (2.5)], Lind, P, P'RD (generated by the expert panel), and Lact

[derived from Equation (.10 . As noted in Section 2.1.2, of the factors in

Table 2.6, only Lnd P and R/D are used as S/C prioritization criteria. A

discussion of each of te factors in Table 2.6 is presented in the following

subsections.

2.21
TABLE 2.5. Structure/Component CDFAs and 40-Year Risk Increases(a)

N A C DFA AR(40) Interim

Component (C D/yr) (hr yr- ) (C D/yr3) b) (C D/yr) Rank

2.a. Containment (BWR) 1.0E-0 I.OE-7 8.8E-4 7.0E-1 5

18.e. Motor operated valve 2.2E-2 3.6E-6 6.9E-4 5.62-1 5

11. Diesel 2.OE-2 3.6E-6 6.3E-4 5.0E-1 5

19.b. Turbine pump 9.3E-3 2.7E-6 2.2E-4 1.8E-1 5

17.a. Relay 4.8E-2 2.5E-7 1.1E-4 8.4E-2 4

25. Battery 2.0E-2 3.4E-7 6.0E-5 4.8E-2 4

6.a. SIG tube 3.0E-4 5.0E-6 1.3E-5 1.IE-2 4

19.a. Motor pump 6.7E-3 2.2E-7 I.3E-5 1.OE-2 4

17.b. Breaker 7.2E-2 1.6E-8 1.0E-5 8.1E-3 3

24.a. Chillers 6.OE-4 1.5E-6 7.9E-6 6.3E-3 3

16.a. Thermostat 6.0E-3 1.5E-7 7.9E-6 6.3E-3 3

10.b. Connectors 2.0E-2 2.7E-8 4.7E-6 3.8E-3 3

5.b. Small other safety pipe(C) I.OE-3 3.0E-7 2.6E-6 2.1E-3 3

9.a. CRDM (BWR) 1.0E-1 3.0E-9 2.6E-6 2.1E-3 3

10.a. Cables 1.1E-1 2.7E-9 2.6E-6 2.lE-3 3

23. Compressor (Instr. air) 5.OE-4 5.0E-7 2.2E-6 1.8E-3 3

12. RX internals l.OE-1 2.0E-9 1.8E-6 1.4E-3 3

18.a. Air operated valve 3.2E-4 4.0E-7 1.IE-6 9.0E-4 2

24.b. Fan 6.0E-4 2.1E-7 1.1E-6 8.8E-4 2

30.b. DC bus 1.1E-1 1.1E-9 1.1E-6 8.5E-4 2

22. Heat exchanger 6.4E-3 1.4E-8 7.8E-7 6.3E-4 2

18.f. Safety/relief valve 1.0E-4 6.7E-7 5.9E-7 4.7E-4 2

31. Bolts 1.0E-4 5.1E-7 4.5E-7 3.6E-4 2

30.a. AC bus 4.3E-2 1.1E-9 4.1E-7 3.3E-4 2

4.c. BWR pipe (small LOCA) 1.0E-3 3.0E-8 2.6E-7 2.1E-4 2

26.b. Inverter 4.7E-6 4.9E-6 2.OE-7 1.6E-4 2

27. Transformer (d) 1.2E-2 1.7E-9 1.8E-7 1.4E-4 2

5.a. Large other safety pipe 6.4E-3 3.OE-9 1.7E-7 1.3E-4 2

15. Snubbers 1.1E-6 5.1E-6 4.9E-8 3.9E-5 I

26.a. Battery charger 1.1E-4 3.5E-8 3.4E-8 2.7E-5 I

18.b. Check valve 8.OE-4 3.8E-9 2.7E-8 2.1E-5 1

1. RPV 1.OE-0 2.OE-12 1.8E-8 1.4E-5 I

16.c. Bistable 1.2E-5 1.4E-7 1.5E-8 1.2E-5 I

18.c. Hydraulic valve 1.OE-5 1.3E-7 1.1E-8 9.IE-6 I

16.b. Transfer switch 4.7E-6 2.3E-7 9.5E-9 7.6E-6 1

26.c. Rectifier 4.7E-6 8.7E-8 3.6E-9 2.9E-6 1

21. Turbine l.OE-6 1.02-7 8.8E-10 7.0E-7 1

3. Other concrete structures 1.OE-O 1.OE-13 8.8E-10 7.0E-7 1

2.b. Containment (other) 1.0E-0 I.OE-13 8.8E-10 7.0E-7 1

9.b, CROM (PWR) I.OE-3 3.0E-11 2.6E-10 2.1E-7 1

29.b. Tank (atmos. pres.) 2.5E-2 1.OE-12 2.2E-10 1.8E-7 I

18.d. Manual valve 1.OE-5 2.22-9 1.9E-10 1.5E-7 I

(e)

4.b. PWR Pipe (small LOCA) 1.OE-3 1.0E-11 8.8E-11 7.0E-8 I

29.a. Tank (medium pres.) 6.0E-3 1.0E-12 5.3E-11 4.2E-8 I

29.c. Tank (high pres.) 1.02-4 1.OE-12 8.82-13 7.OE-10 1

4.a. RC P & SE large (LOCA) 1.0E-4 1.OE-12 8.8E-13 7.0E-10 1

7. RC P casing 1.0E-4 1.0E-12 8.8E-13 7.0E-10 1

8. Pressurizer 1.0E-4 I.OE-12 8.8E-13 7.0E-10 1

6.b. S/G shell 1.OE-5 I.OE-12 8.8E-14 7.0E-11 1

28. Fuel rack 1.OE-6 5.0E-12 4.4E-14 3.5E-11 1

13. RVP support 1.OE-7 I.OE-12 8.8E-16 8.0E-13 I

(a) Caution: only the relative component position (rank) Is Important; do not base any

subsequent studies on the numerical values presented here.

(b) CDFA (CD/yr) N (C D/yr) x A (/hr yr) x 8760 (hr/yr).

(c) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(d) 10-24 In. pipe represented by service water piping (see Table 2.2).

(e) Components below this line were eliminated from further consideration by screening

criterion (ARc 4 1E-7 C D/yr).

2.22
TABLE 2.6. Adequacy of Current Industry Aging Management Practices(a)

(Truncated to 40 years)

L

Copnet(MO)

2.a.

10.a.

5. b.

10.b.

6.a.

19.b.

17.a.

11.

5.a.

12.

18.e.

17.b.

4.e.

19.a.

24.a.

16.a.

1.

25.

23.

3.

2.b.

18.a.

30.b.

9.a.

18.b.

24.b.

22.

31.

30.a.

18.f.

27.

26.b.

16.b.

15.

18.c.

21.

16.c.

18.d.

26.a.

29.b.

26.c.

29.a.

9.b.

Containment (BWR)

Cables

Small other safety pipe(c

Connectors

S/G tube

Turbine pump

Relay

Diesel (d)

Large other safety pipe

RX Internals

Motor operated valve

Breaker

BWR pipe (small LOCA)

Motor pump

Chillers

Thermostat

RPV

Battery

Compressor (Instr. air)

Other concrete structures

Containment (other)

Air operated valve

DC bus

CRDM (BWR)

Check valve

Fan

Heat exchanger

Bolts

AC bus

Safety/relief valve

Transformer

Inverter

Transfer switch

Snubbers

Hydraulic valve

Turbine

Bistable

Manual valve

Battery charger

Tank (atmos. pres.)

Rectifier

Tank (medium pres.)

CRDM (PWR)

0.1

3.3

3.3

2.5

1.5

0.4

0.5

0.2

4.1

4.1

0.2

1.7

10.5

1.5

1.9

1.9

40.5

0.7

3.6

57.3

57.3

5.1

5.2

3.3

32.9

5.1

6.1

8.0

8.3

7.0

12.7

11.9

55.1

24.2

50.3

181.3

44.2

386.6

29.2

362.6

89.7

740.2

331.0

LInd

(mo)

18.0

60.0

60.0

60.0

36.0

12.0

6.0

3.0

18.0

18.0

3.0

18.0

36.0

12.0

18.0

18.0

120.0

6.0

6.0

60.0

60.0

18.0

18.0

18.0

18.0

18.0

3.0

18.0

18.0

18.0

18.0

12.0

18.0

18.0

18.0

60.0

18.0

60.0

12.0

12.0

12.0

12.0

18.0

p p Lact

D R/D (mo)

0.90

0.10

0.10

0.20

0.50

0.50

0.20

0.30

0.10

0.10

0.70

0.50

0.20

0.50

0.70

0.70

0.90

0.90

0.20

0.10

0.10

0.50

0.50

0.90

0.10

0.70

0.10

0.50

0.50

0.90

0.70

0.50

0.20

0.90

0.50

0.90

0.70

0.50

0.90

0.10

0.50

0.10

0.50

0.80

0.90

0.90

0.90

0.50

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.10

0.90

0.90

0.50

0 50

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.50

0.90

0.60

0.90

0.70

0.90

0.70

0.90

25 .0(b)

480 0b

480.0

333.3

144.0

26.7

33.3

11.1

200.0

200.0

4.8

40.0

200.0

26.7

28.6

480.0

7.4

33 .(b)

480.0

40.0

40.0

22.2

200.0

28.6

33.3

40.0

40.0

22.2

28.6

26.7

100.0

22.2

40.0

133.3

28.6

200.0

14.8

171.4

26.7

171.4

40.0

Lt/L )

2

62500.00

21155.70

21155.70

17774.22

9216.00

4455.56

4435.56

3080.25

2379.49

2379.49

576.00

553.66

362.90

316.84

226.50

226.50

140.42

111.72

85.56

70.22

70.22

61.47

59.14

45.29

36.97

31.47

29.81

25.00

23.23

10.05

5.06

5.02

3.28

0.85

0.64

0.55

0.42

0.27

0.26

0.22

0.09

0.05

0.01

Interim

Rank

(a) Caution: only the relative component position (rank) Is Important; do not base any

subsequent studies on the numerical values presented here.

(b) L for these components was truncated to a maximum of 40 years (480 mo).

(c) 6-a¶ In. pipe represented by letdown and RWCV systems (see Table 2.2).

(d) 10-24 in. pipe represented by service water piping (see Table 2.2).

2.23
The Current Industry Surveillance/Test Factor (Lind)

In generating the values for Lind, the panel considered the surveillance/

test interval representative of current industry practice that was relevant to

the risk-significant failure mode (the one used to generate N). The panel concluded that for many mechanical and electrical components, the current surveillance/operability tests required by technical specifications were of limited

value in identifying these failure modes. Often, only those tests conducted

during the refueling outages, in conjunction with preventive maintenance

activities, were considered to be relevant. This was the rationale for the

many 18-month assignments made for Lind that can be observed in Table 2.6.

In assessing how frequently a surveillance/test relevant to a failure mode

was likely to be made, a composite value often had to be determined. Some

features of the failure mode might be observed more frequently, while others

might be observed less frequently. An example is the diesel generator, where

some panelists felt that relevant observations could be made as frequently as

once a shift, but where others felt that only during the 18-month test/

maintenance activity would information specific to failure mode be generated; a

3-month Ln was finally agreed upon. These Lnd assessments were very difficult for he panel. The Lnd value for RPV became the longest, and was based

upon the 10-year inspection requirement; those for containment and concrete

structures were arrived at by consensus on the time span within which physical

evidence of a problem (spalling, liner bulging, etc.) could be expected to be

observable.

Only one component (S/C-22: heat exchanger) has an Lnd less than its Lc

(3.0 versus 6.1 mo). However, even this interval is too ong. From Equation (2.14) and the values for P and Lc in Table 2.6, an acceptable industry

surveillance/test interval [e.g., one required to control risk to a desirable

level (Lind)acc] should not exceed 0.3 months. This is due to the low P; Lc,

being >1, is more than adequate. When, as is the case for many of the components, Lind is larger than c and P is very low, then the required reduction

in the interval is even greater. For example, S/C-10: cables would require an

(Lind)acc of 0.33 months compared to the value for Lind of 60 months.

Note that even where P and PR/ are each 1.0, if Lind is >Lc, AR will

increase. S/C-25: battery, S/C-9a: RDM (BWR) and S/C-2a: containment (BWR)

are examples of this. Therefore, an adequate frequency of surveillances/tests

that are relevant to the risk significant failure mode is necessary for acceptable management of component aging.

The Probability of Successful Aging Detection Factor (P0)

After determining the value for Lind, the panel turned to an assessment of

how successful the surveillance/test chosen as relevant to detecting the risk

importance failure mode of the component would be in detecting the aging mechanisms responsible for that failure. (We assume, here, only aging-caused

failures, using the definition of aging given in Section 1.2.)

2*24
The surveillance/test methods for many components were judged to be

ineffective in detecting aging; these were given P values below 0.5, many in

the 0.1 - 0.3 range.

The present surveillance/test methods were judged to be truly effective in

detecting aging mechanisms for less than one-fourth of the components. These

were given a P value of 0.9. Values larger than 0.9 were not used in order to

indicate that similar components might not be inspected for the aging mechanism. Values less than 0.1 were not used because of the overwhelming effect

this would have on the total risk contribution value. (Moreover, the value 0.1

represents the probability that over a 40-year period, at least once in 48

months detection would be successful. For those components that are completely

overhauled or replaced, this is reasonable. For major structures such as

piping, the P and Lind values generated by the panel suggest that this may not

be conservative.)

The impact of a low P value is that more frequent surveillances/tests are

required to ensure the detection of aging degradation within the interval Lc.

We noted this effect when we discussed the Lind values for heat exchangers and

cables in the preceding subsection.

The Probability of Successful Aging Mitigation Factor (PR/D)

The panel next determined the probability of successfully mitigating the

risk significant aging mechanism once it was detected (e.g., to turn the aging

clock back to zero or as-good-as-new). Most components were given a 0.9

value. Values larger than 0.9 were not used by the panel in order to indicate

that 1) repairs appearing to be successful may not always be successful, or

2) full operational performance might be achieved but only with an as-good-asold condition (e.g., the aging clock is not fully turned back to zero).

A few components were in the 0.5 range and RPV received a 0.1. A value of

0.5 was given for one of several reasons; the following are examples:

* The observed aging-caused defect may not be considered to be risk

significant, so it would be unlikely that a repair to as-good-as-new

conditions would be made (S/C-2b: containment-other).

* Only the observable defect would be repaired; the age of the rest of

the component would not be returned to zero (S/C-6.a: steam generator

tube).

The PR/ value of 0.1 for the RPV was based on the lack of evidence confirming the enefits of vessel annealing. Also, recent laboratory data

indicate that irradiation after annealing may result in a faster rate of

embrittlement. Mitigation of aging via an NRC-mandated plant shutdown, should

aging exceed the screening criterion (RG-1.99 Rev. 2), was not considered relevant to the intent of PR/D, namely, to return the aging clock to zero and

return the component to full functionality (aging, in this case, would cease,

but the component would no longer be functional).

2.25
The Factors Relating to Adequacy of Industry Aging Management

Practices (Lact, Lc, and Lact/Lc)

Lact, the actual (effective) interval representative of current industry

practices, was calculated directly from Lind, PD, and PR/D subsequent to the

workshop. Some very high values were obtained. To stay within the guidelines

of considering aging only in the context of the current 40-year license period,

the calculations of Lact were truncated to 480 months. As observed in

Table 2.5, many components had effective intervals in the hundreds of months,

including five components with intervals of 480 months.

A critical para eter in comparing aging management practices across components is (Lact/Lc) . This parameter is directly proportioned to AR/ARc;

hence, it is a ratio index of the aging risk under present aging management

practices to the acceptable level [see Equations (2.7) through (2.9)]. For

aging management practices to be acceptable, (La t/Lc)2 < 1. Table 2.6 shows

that many components, even those having an L t that might be considered to be

low (such as S/C-25: battery, with an L ofconly 7.4 mo.), have AR> ARC. In

the case of the battery, (L ct/Lc)2 is ff2; for S/C-2a: containment (BWR-MK-I),

it is 6.2E+4. The reasons or a high (Lact/Lc)2 can be three-fold:

1. A highly aging risk significant component (CDFA is high) will necessitate a very low Lc in order to keep AR low [see Equation (2.5)].

This is the case for S/C-2a:containment (BWR-MK-I); its Lc is only

0.1 month.

2. A component with a long Lnd and/or with a low P or P /D will have a

long Lact, This is the case for S/C-3:other Category structures;

its Lact is 480.

3. Combinations of the above 1. and 2.

Whatever the reason, a high (LaCt/Lc)2 indicates the need for improved aging

management practices.

Conversely, relatively poor aging management practices may not lead to an

increase in a component's risk contribution. This is the case when the component's Lc is high (due to a low risk significance o aging, CDFA). Such is

the case for S/C-18d: manual valve, with an (LEct/Lc) of only 0.27 even though

its Lact is 200, due to a D of only 0.5; its c is very long (386.6 mo.).

2.2.3 Final Integrated Ranking of the Structures and Components

Table 2.7 presents the values for the factors used in the final RSCAAMP

calculations [Equation (2.15)] and in the final ranking of the S/Cs. The S/Cs

are arranged in decreasing order of contribution to plant risk increase. Those

components with AR(40) values were below 1E-7 CD/yr and were not evaluated for

adequacy of aging management practice (as identified in Section 2.2.2) are

included in this table; their rankings are placed at the nominal value of one

(1).

2.26
TABLE 2.7. Factor Va

Comp onent

5.b. Small other safety pipe

1O.a. Cables

2.a. Containment (BWR)

1O.b. Connectors

6.a. S/G tube

19.b. Turbine pump

17.a. Relay

11. Diesel

12. RX Internals

17.b. Breaker

18.e. Motor operated valve

4.c. BWR pipe (small LOCA)

19.a. Motor pump p e)

5.a. Large other safety pipe

16.a. Thermostat

24.a. Chillers

1. RPV

25. Battery

23. Compressor (Instr. air)

18.a. Air operated valve

30.b. DC bus

9.a. CRDM (BWR)

18.b. Check valve

24.b. Fan

22. Heat exchanger

31. Bolts

30.a. AC bus

18.f. Safety/relief valve

2.b. Containment (other)

3. Other concrete structures

27. Transformer

26.b. Inverter

16.b. Transfer switch

15 Snubbers

18.c. Hydraulic valve

21. Turbine

16.c. Bistable

18.d. Manual valve

26.a. Battery charger

29.b. Tank (atmos. pres.)

26.c. Rectifier

29.a. Tank (medium pres.)

9.b. CRDM (PWR)

lues and Final Ranking of Structures/Components(a)

(Truncated to 40

N

(C D/yr)

1 .OE-3

l.lE-1

1 .OE-O

2.OE-2

3.OE-4

9.3E-3

4.8E-2

2.OE-2

I .OE-1

7.2E-2

2.2E-2

1.OE-3

6.7E-3

6.4E-3

6.OE-3

6.OE-4

1 .OE-O

2.OE-2

5.OE-4

3.2E-4

1 1E-1

1.OE-1

8.0E-4

6.OE-4

6.4E-3

1.OE-4

4.3E-2

1.OE-4

1 .OE-O

l.OE-O

I .2E-2

4.7E-6

4.7E-6

l.lE-6

1.OE-5

1.OE-6

1 .2E-5

1 OE-5

1 .IE-4

2.5E-2

4.7E-6

6.OE-3

I .OE-3

A

(hr -yr )

3.OE-7

2.7E-9

1.OE-7

2.7E-8

5.OE-6

2.7E-6

2.5E-7

3.6E-6

2.OE-9

1.6E-8

3.6E-6

3.OE-8

2.2E-7

3.OE-9

1 .5E-7

1.5E-6

2.OE-12

3.4E-7

5.OE-7

4.OE-7

1 .1E-9

3.OE-9

3.8E-9

2.1 E-7

1 .4E-8

5.1E-7

l.lE-9

6.7E-7

1.OE-13

1.OE-13

1.7E-9

4.9E-6

2.3E-7

5.1E-6

I .3E-7

1.OE-7

1.4E-7

2.2E-9

3.5E-8

2.OE-12

8.7E-8

l.OE-12

3.OE-1 I

years)

Lind

(mo.)

60.0

60.0

18.0

60.0

36.0

12.0

6.0

3.0

18.0

18.0

3.0

36.0

12.0

18.0

18.0

18.0

120.0

6.0

6.0

18.0

18.0

18.0

18.0

18.0

3.0

18.0

18.0

18.0

60.0

60.0

18.0

12.0

18.0

18.0

18.0

60.0

18.0

60.0

12.0

12.0

12.0

12.0

18.0

0D

0.10

0.10

0.90

0.20

0.50

0.50

0.20

0.30

0.10

0.50

0.70

0.20

0.50

0.10

0.70

0.70

0.90

0.90

0.20

0.50

0.50

0.90

0.10

0.70

0.10

0.50

0.50

0.90

0.10

0.10

0.70

0.50

0.20

0.90

0.50

0.90

0.70

0.50

0.90

0.10

0.50

0.10

0.50

R/D

0.90

0.90

0.80

0.90

0.50

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.10

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.90

0.50

0.50

0.90

0.90

0.90

0.90

0.90

0.50

0.90

0.60

0.90

0.70

0.90

0.70

0.90

Risk

Increase AR

(C D/yr) t)

2 lE-3(d)

2.1E-3

1.9E-3

1 .8E-3

9.5E-4

5.4E-4

4.OE-4

2.7E-4

2.4E-4

5.6E-5

5.6E-5

3.7E-5

3.2E-5

2.3E-5

2.2E-5

2 2 E-5(d)

1.4E-5

1.IE-5

8.4E-6

6.2E-6

5.9E-6

4.5E-6

3.7E-6

3.1E-6

3.OE-6

2.5E-6

2.3E-6

1 .OE-6 (d) 7 .OE- 7 (d)

7.OE-7

5.IE-7

5.OE-7

3.3E-7

8.4E-8

6.3E-8

5.4E-8

4.2E-8

2.7E-8

2.6E-8

2.2E-8

8.9E-9

5.4E-9

1.5E-9

Final

Rank

5

5

5

5

5

4

4

4

4

3

3

3

3

3

3

3

3

3

3

2

2

2

2

2

2

2

2

2

2

2

_(h)

1

1

1

1

85.

6.b.

29.c.

4.a.

7.

13.

28.

4.b.

Pressurizer

S/G shell

Tank (high pres.)

RC P & SE large (LOCA)

RC P casing

RPV support

Fuel rack

PWR pipe (small LOCA)

I .U-4

1.OE-5

1.OE-4

1.OE-4

1.OE-4

1.OE-7

1.OE-6

1.OE-3

1 .OE-12

1.OE-12

1 .OE-12

1.OE-12

1.OE-12

1.OE-12

5.OE-12

l.OE-12

18.0 __

__

__

__

__

__

0.90 0.90

__

__

__

__

__

__

__

1.2E-12

(a) Caution: only the relative component position (rank) Is important; do not base any subsequer

studies on the numerical values presented here. -I

(b) To obtainithe units for AR, (C D/yr), A Is multiplied by 8760 hrs yr and Lnd Is ultiplieC

by 12 yr (both P and P are probabilities).

(c) 6-10 In. pipe reprgsentedR49 letdown and RWCV systems (see Table 2.2).

Cd) L for these components was truncated to a maximum of 40 years (480 months).

(e) 1t-4 In. pipe represented by service water piping (see Table 2.2).

(f) Components below this line were eliminated from further consideration by screening criterion

(ARc 4 IE-7 C D/yr); the pressurizer was not eliminated until a post-workshop re-analyses.

int

2.27
In Table 2.7, the risk increases and the final S/C rankings are based upon

the truncation of Lact to 480 months (40 years), as explained in Section 2.2.2.

Even with truncation, the effect on component aging risk contribution of the

adequacy of industry aging management practices is marked. This can be seen by

comparing the relative positions of S/C-5b: small other safety related piping

and S/C-2a: containment (BWR-MK-I) before and after L is factored into the

component risk contribution equation. Table 2.5 shows that S/C-5b is lower

than S/C-2a by a factor of 300 in risk increase due to aging R(40), and is

ranked 3 compared to S/C-2a which is ranked 5. However, Table 2.6 shows that

Lact for S/C-5b is almost a factor of 20 larger than La for S/C-2a (e.g., its

aging management is that much less effective). Due to Et entering the

RSCAAMP equation as the square, the net result is that S-5b has a greater

risk contribution than does S/C-2a, and is now at the top of the rankings.

If L had not been truncated, several components would have Lact values

well above 80 months and, as a result, would have higher final risk increase

values. This impacts the consideration of plant relicensing. If a nuclear

plant were allowed to operate beyond 40 years, the ineffectiveness of aging

management practices represented by the components' untruncated values would

result in continuing and increasing risk. (Note: this assumes no change in

those practices.) Of particular concern would be S/C-1: RPV, S/C-2b: containment-other, and S/C-3: other concrete structures since these are permanent

structures (though the possibility of RPV replacement has been considered).

Untruncated Lac. values increase from 480 months to 1333 months for the RPV,

and to 1200 months each for the other two structures. The final AR value for

the RPV increased by one order of magnitude; for the other structures, AR

increased by one and one-half orders of magnitude.

We again caution that appropriate care must be exercised in the interpretation and use of the numerical values in this table. Only the relative

ranks of the components should be used as a basis for future studies.

Table 2.8 shows the scores for the five prioritization criteria and the

final rankings of the S/Cs. All components are shown in this table, including

those eliminated by the screening criterion from consideration of the adequacy

of aging practices.

2.2.4 Detailed Rationale for the Panel's Findings

Expert panel judgment was used to alter the precalculated values of N and

A for the PRA-based components where deemed appropriate; to supply values of N

and A for the non-PRA-based components; and to supply values for L ,Pn and

PR/D for all S/Cs having risk significance of aging over 40-years, A (40J,

greater than the screening criterion.

The rationale expressed by the panel for its decisions as recorded by the

rapporteurs. The detailed rationale is presented in Appendix D. These

rationale are a line-by-line itemization of the values given for the S/C listings in Table 2.7. Appendix D is arranged by S/C number (e.g., S/C.1 RPV,

etc.).

2.28
TABLE 2.8. Structures/Components Prioritization Criteria

Scores and Final Rankings (Truncated to 40 years)

Component

5. b.

1O.a.

2.a.

10.b.

6.a.

19.b.

17.a.

II.

12.

17.b.

18.e.

4.c.

19.a.

5.a.

16.a.

24.a.

1.

25.

23.

18.a.

30.b.

9.a.

18. b.

24.b.

22.

31.

30.a.

18.f.

2. b.

3.

27.

26 b.

16.b.

15.

18.c.

21.

16.c.

18.d.

26.a.

29.b.

26.c.

29.a.

9.b.

Smell other safety pipe(a)

Cables

Containment (BWR)

Connectors

S/G tube

Turbine pump

Relay

Diesel

RX Internals

Breaker

Motor operated valve

BWR pipe (small LOCA)

Motor pump

Large other safety pipe(C)

Thermostat

Chillers

RP V

Battery

Compressor (Instr. air)

Air operated valve

DC bus

CRDM (BWR)

Check valve

Fan

Heat exchanger

Bolts

AC bus

Safety/relief valve

Containment (other)

Other concrete structures

Transformer

Inverter

Transfer switch

Snubbers

Hydraulic valve

Turbine

Bistable

Manual valve

Battery charger

Tank (atmos. pres.)

Rectifier

Tank (medium pres.)

CRDM (PWR)

N

Score

2

4

5

3

43

4

3

4

4

3

2

3

3

3

2

5

3

2

2

4

4

2

2

3

4

5

5

3

3

3

2

2

A

Score

4

2

3

3

5

5

4

5

2

2

5

3

4

2

3

4

1

4

4

4

1

2

4

2

4

4

2

5

4

5

3

3

3

2

3

3

L

Si

P P

Pade Scgre SUPe

4 5 1

4 5 1

3 1 2

4 4 1

4 3 4

2 3 1

1 4 1

1 3 1

3 5 1

3 3 1

1 2 1

4 4 1

2 3 1

3 5 1

3 2 1

3 2 1

5 1 5

1 1 1

1 4 1

3 3 1

3 3 1

3 1 1

3 5 1

3 2 1

1 5 1

3 3 1

3 3 1

3 1 1

4 5 4

4 5 4

3 2 1

2 3 1

3 4 1

3 1 1

3 3 1

4 1 4

3 2 1

4 3 3

2 1 1

2 5 2

2 3 1

2 5 2

3 3 1

Rliaks I ncr se

Rank

5 (b)

5 (b)

5

5

5

4

4

4

4

3

3

3

3

3

3

3(b

3 (b

3

3

2

2

2

2

2

2

2

2

2

2 (b)

-

(d)

8.

6. b.

29.c.

4.a.

7.

13.

28.

4. b.

Pressurizer

S/G shell

Tank (high pres.)

RC P & SE large (LOCA)

RC P casing

RPV support

Fuel rack

PWR pipe (small LOCA)

I

I

I

I

I

I

1

1

T

I

I

I

I

1

1

1

(a) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(b) L for these components was truncated to a maximum of 40 years (480 months).

(c) 18-2 4 in. pipe represented by service water piping (see Table 2.2).

(d) Components below this lne were eliminated from further consideration by

screening criterion.

2.29
2.3 TECHNICAL INSIGHTS AND DISCUSSION

Analyses of the workshop process, the application of the RSCAAMP methodology, and the panel's findings offer some additional technical insights.

These are discussed in the following sections.

2.3.1 The Workshop Process

The workshop was deemed by the panel members to be well-organized and

executed, and successful in meeting its objectives. The varied and relevant

expertise within the panel membership led to a useful synergism; pre- and

in-workshop training processes brought the panel to an early state of readiness

to meet the prioritization objectives; and the formulation of the mathematically-derived RSCAAMP model provided the structure (through the five factors)

needed to allow the panel's deliberations to be productive. In addition, the

workshop process allowed the panel to supply relative rankings of the components within each factor, from which the corresponding estimated values were

used to calculate the final rankings. All members felt that more time was

needed; a post-workshop conference call and individual contacts to refine some

of the data confirmed this.

2.3.2 Application of the RSCAAMP Methodology

2.3.2.1 Application to TIRGALEX S/C

The application of the RSCAAMP to the TIRGALEX list of S/Cs presented

several issues that the panel had to deal with. The first issue was the need

to develop methodologies to deal with the non-PRA-based components within the

list. For the accident-prevention subset of these S/Cs, the panel had to

develop its own set of risk importances from its intellectual perceptions

rather than available hard data; out of necessity, compromises developed. For

the accident-mitigation subset of S/Cs, an arbitrary decision on their equivalent importance based upon defense-in-depth considerations was used to

facilitate the process.

The second issue resulting from the TIRGALEX S/C list was the need to disaggregate certain groups to more adequately consider component risk importances, aging failure rates, etc. While disaggregation was needed and very

useful, considerable effort was required to achieve it; more time was needed

and compromises were made. Even with disaggregation, grouping of components

still occurred (e.g., S/G 18: valves were partially disaggregated to S/C-18a:

motor-operated valves, but not to a single component like reactor pressure

vessel). In many instances, the panel developed, in essence, a single surrogate component for the group (e.g., one having the higher risk importance or

aging failure rate) to develop its N and A rankings. The summing process that

the basic RSCA model uses to determine AR is equivalent to assuming that a high

degree of correlation exists between the aging-related failures of the group.

This would tend to produce a higher AR value than would be produced for a

single component group (such as the RPV).

2.30
The most desirable situation would be that involved in performing a plantspecific RSCAAMP assessment. Specific, individual components (e.g., "check

valve 5952") would be evaluated. While the RSCAAMP model is applicable to

this, unless there was a plant-specific PRA and better aging failure data base

available, several of the problems the panel faced would still exist. Moreover, to provide guidance to a generic NRC aging research program, an

evaluation at the level of "check value 5952" is not necessary.

2.3.2.2 Application of the RSCAAMP Factors

In order to achieve its goal, the panel developed working definitions for

the factors N, A, Lnd, P P and implemented a screening of S/Cs based

upon CDFA over 40 years 4YE-7RN/yr. The working definitions focused deliberations on risk-significant failure modes (N); the aging failure rates (A),

surveillance/test intervals (Lind), and aging degradation mechanisms responsible for these failure modes; and the methods required to detect (Pe) and

mitigate them (PR/D). As a result of using this focused approach, i became

apparent that the data bases for N and A were limited, that current requirements for inspection/tests do not adequately address the risk-significant failure mode, that successful detection of aging on one component (if that were to

be achieved) doesn't guarantee that the redundant component in that system (or

similar components in other systems) will be examined, and that current repair

practices may not effectively reduce the risk contributions of aging (e.g.,

even if some S/G tubes are plugged, the aging clock for the rest of the S/G is

still ticking). The screening criterion worked well since all S/Cs below

5.1E-7 were ranked 1.

2.3.3 Unexpected Findings in the Component Rankings

The component rankings produced some unexpected findings (surprises).

The panel members, on an individual basis, performed a pre-workshop homework assignment to prioritize the TIRGALEX S/Cs after reading the pre-workshop

material sent to them. The composite results of the members' efforts are

presented in Table 2.9. A comparison of this table to Table 2.7 shows that in

doing its homework, the panel ranked too high RPV, pressurizer, snubbers, some

valves, fuel storage racks, turbines, and tanks; and it ranked too low other

safety-related piping (specifically the small piping), containment, cables,

connectors and reactor internals.

Perhaps the most significant surprise is the number of high-ranked component groups representing components that are abundant and diversely spread

throughout many plant systems (e.g., "abundant small components," as opposed to

singly redundant components such as the RPV). If we include within this class

of components S/C-5b: small other safety related piping, S/C-lOa: cables,

S/C-lOb: connectors, and S/C-6a: S/G tubes, we find that in the top rank (5),

four of the five component groups were of this abundant small components category. It is striking that while both piping and cables are ranked at the top,

they are, in fact, not explicitly modeled in the PRAs. These components

possess analogous roles; cables form the "connection" between major electrical

components, while piping forms the "connection" between critical water system

2.31
TABLE 2.9. Panel's Preliminary (omework) Ranking of Components

Rankings Between 4-5

6. Steam generator (PWR)

1. Reactor pressure vessel

Rankings Between 3-4

4. Reactor cooling piping and safe ends

8. Pressurizer (PWR)

10. Cables, connectors, and penetrations

11. Emergency diesel generator

17. Switchgear and relays

18. Valves

2. Containment (metal and concrete)

9. Control rod drive mechanism

14. Recirculation piping safe ends (BWR)

16. Instruments and controls

25. Batteries

Rankings Between 2-3

26. Battery chargers/inverters

3. Other category/concrete structures

5. Other safety-related piping

12. Reactor internals

19. Pumps

30. AC/DC buses

15. Snubbers

22. Heat exchangers

7. Reactor coolant pump casting

13. RPV Support (sliding foot) (PWR)

21. Turbines

23. Compressors

28. Fuel storage racks

29. Accumulator tanks

Rankings Between 1-2

20. Motors

24. Fans/chillers

27. Transformers

components. The piping class may be easily expanded to include S/G tubing as a

specialized collection of piping; and connectors can be considered to be the

linkages between runs of cables.

Conversely, PRAs have shown the failure of concrete structures during

seismic events to be a very important (often dominant) contributor to risk. It

is only the extremely low aging failure rate for these structures with respect

to their seismic fragility which keeps their ranking low. Again, the RSCAAMP

model considers more than one factor in assessing the risk contribution of a

component.

2.32
That diesels did not end up in the top rank may be surprising since they

are shown to be significant contributors to plant risk. This results because

the aggregate of RSCAAMP factors is even more risk significant for certain

other components (e.g., other small pipe, cables, etc.) than for diesels, as

Table 2.7 indicates.

2.3.4 Role of the RSCAAMP Factors in the Component Rankings

The RSCAAMP factors were analyzed to understand reasons for the surprises.

The interplay of the various RSCAAMP factors suggests some answers. This was

alluded to above in the discussion of the low ranking of concrete structures.

A comparison of the pre-workshop homework results to workshop results is also

enlightening. For example, the high ranking of the RPV in the homework was

probably due to its high risk importance; however, the low aging failure rate

(A) and high detection capability (PD) generated at the workshop reduce its

overall risk significance compared to the intuitive position. In many cases,

the higher ranking after the RSCAAMP activity, compared to the homework,

resulted from the considerations of the effectiveness of industry aging management practices. The forced examination of the various risk factors in the

RSCAAMP model during the workshop provided these different perspectives.

A more rigorous examination of the role that the various RSCAAMP factors

have in a component's final risk contribution rank was performed. For this

analysis, we addressed only the top two aging risk ranks (5 and 4) which contain the top two orders of magnitude of risk change.

The matrix in Table 2.10 displays the component groups ranked 5 or 4 in

the panel's f nal evaluation Beside them are three columns, one each for N,

A, and (Lack) , where (Lac ) incorporates the industry aging management factors Lnd aD and P a if the value of the individual factor (N, A, etc.) is

within the top order of magnitude for that factor, an "X" is assigned under

that column. Tabulations are then made of 1) the number of factors contributing to the component's rank (by counting horizontally across the table), and

2) the number of times a factor was the principal contributor to the top-ranked

components (by counting vertically down the table).

From N-1 and N-2, we can conclude that no single factor dominated the

ranking of these top-ranked components. From N-1, we see that only two groups

(cables and RX internals) had as many as two factors contributing to their

ranking. From N-2, we see that each factor had an approximately equal

influence on component rankings; N was a principle contributor 3 times, A 3

times and (La) 4 times. The component group relays had no principle contributing facor, but all three factors were just under the top order of magnitude. Steam generator tubing is ranked high specifically because of its high

aging failure rate (A); its risk importance (N) is extremely low and A acts to

raise the overall rank.

While this analyses shows that, on an overall basis, each factor had an

approximately equal influence on components ranked 4 and 5, there is another

perspective: the importance of including component aging and its management in

an assessment of plant risk. Table 2.10 illustrates the importance for the

2.33
TABLE 2.10. Analysis of the Contribution of RSCAAMP Factors to the

Final Risk Contribution Rank of Top-Ranked Components

N-1 (Number of

(L )2 Factors Contributing

Component Category Rank N A act to Component Rank)

Small other safety 5 X 1

pipe

Cables 5 X X 2

Containment (BWR) 5 X 1

Connectors 5 X 1

S/G tubing 5 X 1

Turbine driven pump 4 X 1

Relays 4 0

Diesel 4 X 1

RX internals 4 X X 2

N-2 (number of times 3 3 4

the factor was principle

contributor to rank 5

and 4 components)

top-ranked components. The factors A and (Lac )2 were the principal contributors to the rank 7 out of 10 times. Had these factors not been considered, some of the surprises wouldn't have occurred. Small other safety pipe

and connectors would not have had their high rank. [Cables still would have

been ranked high because of N, but would not have been ranked higher than

containment (BWR), which had an order of magnitude higher N.] In addition, the

ranking for RPV would not have been so low (mid-rank 3).

2.3.5 Insights on Research, Regulatory and Industry Applications

The RSCAAMP methodology provides a discriminating tool, one useful beyond

the ranking of S/Cs for subsequent aging research. Because it facilitates an

examination of the individual factors that contribute to the risk impact of

aged components, it has three other applications: 1) focusing NRC research by

identifying the relevant factors for each component that need to be examined

(the risk-important failure mode, aging failure rate and its mechanisms; the

adequacy of surveillance/test frequencies and methods; and the adequacy of

mitigation methods); 2) focusing NRC regulation toward the aging risksignificant components, and for these, the inspection/test frequencies and

2.34
methods necessary to detect the risk-significant failure modes, aging mechanisms and the mitigation methods required to reduce risk; 3) providing

guidance to utilities to reduce the risk contribution of their aged components

[reducing N through changes in system design, added redundancy, etc.; reducing

A, through improved materials, upgraded operating conditions and environments,

etc.; and reducing the actual (effective) aging interval (Lact) through a

combination of reducing Lind and increasing P and PR/D]

2.4 CONCLUSIONS FROM RISK-BASED PRIORITIZATION

The prioritization of S/Cs by an expert panel using the RSCAAMP model was

accomplished. In so doing, the impact on component risk contribution of aging

and aging management practices, as well as the component's normal risk importance, were evaluated. For the top-ranked components (ranks 5 and 4), all of

these factors were significant; no one factor was dominant.

The importance of including component aging and its management in assessing plant risk was dramatized by the unexpected findings: 1) some components

not considered in PRAs were found in the top rank (rank of 5) for risk contribution (cables, connectors, small safety-related piping); 2) conversely, concrete structures and diesels that are considered in PRAs to be significant contributors to plant risk were found in the lower ranks (ranks of 2 and 4,

respectively); and 3) perhaps as significant, is the number of "abundant small

components" (e.g., small other safety-related piping, cables, connectors, and

S/G tubes) found in the top ranking (four of the five groups ranked 5) when one

might intuitively have thought of the major non-redundant, defense-in-depth

structures (RPV and containment) to be top-ranked [RPV was ranked 3, containment (other) ranked 2, with only the special case of the BWR-Mk-1 containment

ranked 5 (due to aging)].

Table 2.11 presents the current status of component research in the NRC

Plant Aging Research Program for the ranked comonents (from Table 2.8). The

table provides several conclusions:

* most components have work on-going; a few additional components have

been identified as being of interest but not in the current scope

(e.g., chillers, etc.)

* approximately one-half of the components ranked 1 and 2 are under

study and several are considered to be of interest; a re-evaluation

of their continued study/interest appears warranted

* of the components in the top two ranks, two [containment (BWR-Mk-1)

and RX internals] are not being actively studied in the aging

research program, and a third (S/G) is considered to be completed;

consideration of their inclusion in the program appears warranted.

The panel's deliberations and findings highlight important areas for

further consideration:

2.35
TABLE 2.11. Status of Aging Research on Ranked Components

Component

t i1

5.b. Small other safety pipe''

1O.a. Cables

2.a. Containment (BWR)

1O.b. Connectors

6.a. S/G tube

19.b. Turbine pump

17.a. Relay

11. Diesel

12. RX Internals

17.b. Breaker

18.e. Motor operated valve

4.c. BWR pipe (small LOCA)

19.a. Motor pump

5.a. Large other safety pipe(d)

16.a. Thermostat

24.a. Chillers

1. RPV

25. Battery

23. Compressor (Instr. air)

18.a. Air operated valve

30.b. DC bus

9.a. CRDM (BWR)

18.b. Check valve

24.b. Fan

22. Heat exchanger

31. Bolts

30.a. AC bus

18.f. Safety/relief valve

2.b. Containment (other)

3. Other concrete structures

27. Transformer

26.b. Inverter

16.b. Transfer switch

15. Snubbers

18.c. Hydraulic valve

21. Turbine

16.c. Bistable

18.d. Manual valve

26.a. Battery charger

29.b. Tank (atmos. pres.)

26.c. Rectifier

29.a. Tank (medium pres.)

9.b. CRDM (PWR)

AR

Rank

5

5

5

5

5

4

4

4

4

3

3

3

3

3

3

3

3

3

3

2

2

2

2

2

2

2

2

2

2

2

1

1

1

Research

on-going

x

ox(b)

x

x

x

0

x

x

x

x

x

0

Components of

Interest But

Not in Scope

x

x

x

K

x

x

K

K

x

K

0

K

0 x

x

K

K

0

K

0

0

K

0

K

0

K

0

K

(e)

8.

6. b.

29.c.

4.a.

7.

13.

28.

4. b.

Pressur1zer

S/G shell

Tank (high pres.)

RCP & SE large (LOCA)

RCP casing

RPV support

Fuel rack

PWR pipe (small LOCA)

I

I

I

I

I

I

1

1

0

0

x

K

0

K

0

K

(a) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(b) 0: no activity or plan.

(c) Completed FY87.

(d) 10-24 In. pipe represented by service water piping (see Table 2.2).

(e) Components below this line were eliminated from full evaluation

by the screening criterion.

2.36
* Improved data bases are required, specifically:

--improved PRAs with up-to-date methodologies to provide component

risk importances, including those of passive and accident

mitigation components, and risk-significant failure modes;

--improved aging failure data bases, to provide better aging failure

data for the risk-significant failure mode.

* Improved surveillance/test methods are required to detect aging for

the risk-significant failure modes in components and structures;

however, once aging failures are detected, operational performance is

effectively restored in many cases.

* Improved focus of research activities and regulatory actions and

improved guidance to utilities regarding reduction of risk contributions of aging components can be useful outcomes of methodologies similar to the one used here.

The expert opinion process used in this study is the only way that this

activity could have been accomplished; adequate data are not currently available. This study represents a starting point, not the definitive answer.

As an appropriately focused research program is performed, the data generated

will be useful for in-depth aging risk assessments and a refined level of

prioritization of structures and components.

Finally, multiple failures of similar or dissimilar components in

redundant trains was not evaluated in this prioritization effort. Performing

such an evaluation and comparing the results to the current prioritization

results would be a useful extension for determining the relative importance of

common-cause aging. With minor development, the RSCAAMP model would be

amenable to such an evaluation.

2.37
3.0 PRIORITIZATION OF STRUCTURES AND COMPONENTS

USING "OTHER TECHNICAL" CRITERIA

The purpose of this section is to categorize TIRGALEX S/Cs using the other

technical criteria, e.g., the importance of aging research on S/Cs to the resolution of GSIs and/or to an identified NRC/NRR user needs. S/C categorization

for each of these two criteria are discussed separately; the categorization

process described in Section 1.2.2.2 is followed; the results of the categorizations are tabulated and the comments of the expert panel are documented.

3.1 IMPORTANCE OF AGING RESEARCH ON S/Cs TO THE RESOLUTION OF A GSI AND ITS

APPLICATION TO S/C CATEGORIZATION

This technical criterion identifies whether there would be a direct, positive impact of aging research on the resolution of NRC's GSIs. Current NRC

programs are in the process of, or have completed the necessary technical steps

toward solving these issues. While many issues are not affected by known

aging-related phenomena, some issues may be so affected.

The list of potentially important safety issues comes from two sources:

1) a list of GSIs with elements of aging that may benefit from NPAR results

(Table 3.1); and 2) additional issues discovered as a result of a pre-workshop

review of all GSIs (Table 3.2).

The TIRGALEX S/Cs were initially screened to identify S/Cs for which the

GSIs could directly benefit from aging research. This prescreening accounted

for the technical content and available schedules associated with the resolution of the GSIs a) and with the NRC Plant Aging Research Programs (Vora

1987). Where there was a direct technical connection (i.e., aging was involved

in the issue) and the time scales appeared to be compatible, aging research on

the S/Cs was identified as potentially having a direct benefit. On the other

hand, some S/Cs were clearly not associated with the list of GSIs, aging was

not directly related to the issue, or time scales were incompatible with their

resolution; aging research on these S/Cs was classified as not beneficial. All

GSI evaluations were presented to the panel for discussion and final categorization regarding whether aging research would benefit the resolution of the

GSIs. A few GSIs considered relevant, but which were not identified in the

current GSI resolution schedule, were also presented to the panel for

consideration.

(a) Received November 1987 from R. C. Emrit, NRC/RES/DRA; found in Appendix E.

3.1
TABLE 3.1. NRC Identified Generic Safety Issues, with Elements of Aging,

Potentially Benefitting from Aging Research (NUREG-1144, Rev. 1)

Number Title

23 Reactor coolant pump seal failures

29 Bolting degradation or failures in nuclear power plant

51 Proposed requirements for improving the reliability of open cycle

service water systems

55 Failures of Class E safety-related switchgear circuit breakers to

close on demand

70 PORV and block valve reliability

84 CE PORVs

93 Steam binding of auxiliary feedwater pumps

107 Generic implications of main transformer failures

113 Dynamic qualification testing of large bore hydraulic snubbers

115 Enhancement of the reliability of Westinghouse solid state protection system

118 Tendon Anchorage failure

120 On-line testability of protection systems

124 Auxiliary feedwater system reliability

125.I.6 Valve torque limit and bypass switch settings

125.11.2 Adequacy of existing maintenance manual valves in safety-related

systems

128 Electrical power reliability

130 Essential service water pump failures

132 RHR pumps inside containment

A.10 BWR feedwater nozzle cracking

A.11 Reactor vessel material toughness

A.17 Systems interaction

A.49 Pressurized thermal shock

3.2
TABLE 3.2. Additional Generic Safety Issues

Number Title

A.3, 4, 5 Steam Generator Tube Integrity

A.40 Seismic Design Criteria

A.44 Station Blackout

B.56 Diesel Reliability

II.C.4 Reliability Engineering

II.E.4.3 Containment Integrity Check

I.F.1 Expanded QA List

79 Unanalyzed Reactor Vessel Thermal Stress

During Natural Convection Cooldown

86 Long Range Plan for Dealing with Stress

Corrosion Cracking in BWR Piping

3.2 IMPORTANCE OF AGING RESEARCH ON S/CS TO AN IDENTIFIED NRC USER NEED AND

ITS APPLICATION TO S/C CATEGORIZATION

This technical criterion identifies components for which agin research

will respond to the NRR needs expressed in the "User Need Letter" a In the

letter, NRR expressed the need to "... know not only the effects of aging on

structures, systems and components, but also the risk significance..." of the

process.

3.3 RESULTS

The results and discussion for S/C categorization against the Other Technical Criteria are presented in the following sections.

3.3.1 Importance of Aging Research on S/Cs to the Resolution of a GSI

The panel was provided with screened information in the format shown in

Appendix E and information on the GSI resolution schedules also shown in Appendix E. After the panelists were generally made familiar with each issue, they

were asked if the technical resolution of the GSIs would benefit from S/C aging

research. The panel was not asked to establish the degree of importance or the

(a) Appendix F.

3.3
expected effect of the aging research. For the GSIs that could benefit from

aging research, the S/Cs that may affect the GSIs were listed. Technical comments and insights were recorded.

Tables 3.3 and 3.4 identify the GSIs discussed by the panel, and contain

an indication as to whether aging research may benefit resolution of the GSIs.

The specific components (from the panel's list of components in Table 2.2) for

which aging research would benefit the resolution of GSIs and the panelists

technical comments are also identified.

3.3.2 Importance of Aging Research on S/Cs to Resolution of User Needs

The "User Need Letter" was reviewed for components deemed to be important

by NRR. Aging research on the following components would be of direct benefit

to NRR.

* MOV

* Power operated relief valve (PORV)

* Block valve

* RPV

* Pumps

* Valves

* Diesels

* DC power sources

MOVs and PORVs were emphasized in several sections of the letter. In the

expert panel component list, PORVs are in the category safety/relief valves;

block valves are a specific application of MOVs; pumps were subdivided by prime

mover; valves were subdivided into 5 categories; and DC power sources (battery,

rectifier, inverter, battery charger) are all considered.

3.3.3 Conclusions

In the panel's judgement, many of the generic issues that deal with equipment performance were expected to directly benefit from aging research (schedule permitting), and that further aging research on the components identified

in Section 3.3.2 would be of benefit to NRR. Table 3.5.shows the components'

ranking based on risk (from Table 2.8), S/Cs for which aging research would

benefit the resolution of a GSI and/or a NRC/NRR user need (from Tables 3.3 and

3.4), and those S/Cs for which current aging research is not already on-going

(from Table 2.11). Only for reactor internals (meeting GSI and user needs criteria), pressurizer (GSIs), and bolts (GSI) and hydraulic and manual valves

(GSI and user needs) is aging research not already on-going.

3.4
Nunnber

23

29

51

TABLE 3.3. NRC-Identified GSIs with Elements of Aging

Benefitting from Aging Research

Will the GSI Component

Benefit from Research That Will

Title Aging Research Benefit from the GSI

Reactor coolant pump seal failures No --

Bolting degradation or failures In nuclear Yes 31. Bolts

plant

Proposed requirements for Improving the Yes 22. Heat exchanger

Potentially

Comments

Seals were not discussed within PRA or TIRGALDX

Blofoullng Is an aging Issue

reliability of open cycle service water

systems

55 Failure of class E safety-related switchgear circuit breakers to close on demand

70 PORV and block valve reliability

84 CE PORYs

93 Steam binding of auxiliary feedwater pumps

Yes

No

No

Yes

17.a. Relay

b. Circuit

breakers

Operator error and control room design Issue

Design Issue

Check valve back leakage may be

affected by aging

W

en 18.b. Check valves

107

113

115

118

120

124

125.1.6

125.2.2

Dynamic qualification testing of large bore

hydraulic snubbers

Enhancement of the reliability of

Westinghouse solid state protection system

Tendon anchorage failure

On-line testability of protection systems

Auxiliary feedwater system reliability

Valve torque limit and bypass switch settings

Adequacy of existing maintenance of manual

valves In safety-related systems

Yes

No

Yes

No

Yes

Yes

Yes

Snubbers Snubbers have a high "A"

Electric design Issue

3. Category 1

structures

19.b. Turbine

driver pump

18. All valve types

18.e. MOV

18.d. Fbnual valve

RPS doesn't have testing capability

Aging Is not considered In system

reliability studies

Important because the probability

of finding a problem Is low

Naintenance can mitigate aging
TABLE 3.3. (contd)

Number

128

Title

Wil the GSI

Benefit from

Aging Research

Electrical power reliability Yes

Component

Research That Will

Benefit from the GSI

17.a.b; 1O.a.b;

30.a.b.

19.a. Mbtor driver

pumps

Comments

Aging Is very important to this GSI

Service water pumps exhibit a very

high fraction of aging failures (f)

No comment, no Information on Issue

130 Essential service water pump failures

132 RHR pumps Inside containment

A.10 BWR feedwater nozzle cracking

Yes

7

Yes 4. Reactor coolant

piping and safe

ends

A.i1 Reactor vessel material toughness

A.17 Systems Interaction

Yes

No

1. WPV Vessel can be annealed

Wathematical analysis Is currently

performed to model this

A.49 Pressurized thermal shock Yes 1. V;

8. Pressurizer
TABLE 3.4. Additional GSIs with Elements of Aging Potentially Benefitting from Aging Research

Will the GSI

Benefit from

Number Aina Research

A.3, 4,

5

Title

Component

Research That Will

Benefit from the GSI

G.a. Steam generator

tube

Comments

Steam generator tube Integrity Yes

A.40 Seismic design criteria

2.C.4 Reliability engineering

uo 2.E.4.3 Containment Integrity check

1.F.1 Expanded QA list

79 Unanalyzed reactor vessel thermal stress

during natural convection cooldown

86 Long range plan for dealing with stress

corrosion cracking In BWR piping

A.44 Station blackout

B.56 Diesel generator reliability

No

No

No

No

Yes

Yes

Yes

Yes

Studies criteria, not change In

fragility

Checks for gross openings only

QA Is probably not aging-related

1. RPV

4. Reactor coolant

piping

10, 1, 17, 30

11
TABLE 3.5. Status of Aging Research on S/Cs Ranked by Risk

Importance and Other Technical Criteria

Components of

AR Research Interest But

Component Rank on-going Not In Scope

5.b. Small other safety pipe(a) 5 x

1O.a. Cables 5 x(b)

2.a. Containment (BWR) 5 0

1O.b. Connectors 5 x

6.a. S/G tube 5c)

19.b. Turbine pump 4 x

17.a. Relay 4 x 11. Diesel 4 x

12. RX internals 4 0

17.b. Breaker 3 x

18.e. Motor operated valve 3 x 4.c. BWR pipe (small LOCA) 3 x

19.a. Mbtor pump d 3 x 5.a. Large other safety pipe 3 x

16.a. Thermostat 3 0

24.a. Chillers 3 x

1. RPV 3 x 25. Battery 3 x 23. Compressor (Instr. air) 3 x 18.a. Air operated valve 2 x

30.b. DC bus 2 x 9.a. CRDM (BWR) 2 x

18.b. Check valve 2 x 24.b. Fan 2 x

22. Heat exchanger 2 x 31. Bolts 2 0

30.a. AC bus 2 x 18.f. Safety/relief valve 2 x

2.b. Containment (other) 2 0

3. Other concrete structures 2 x 27. Transformer I x 26.b. Inverter 1 x

16.b. Transfer switch 1 0

15. Snubbers I x

18.c. Hydraulic valve 1 0

21. Turbine 1 0

16.c. Blstable 1 x 18.d. Manual valve 1 0

26.a. Battqry charger 1 x 29.b. Tank (atmos. pres.) 1 0

26.c. Rectifier 1 x 29.a. Tank (medium pres.) 1 0

9.b. CRDM (PWR) 1 x (e)

8. Pressurizer 1 0

6.b. S/G shell 1 0

29.c. Tank (high pres.) 1 x 4.a. RCP & SE large (LOCA) 1 x

7. RCP casing 1 0

13. RPY support 1 x

28. Fuel rack 1 0

4.b. PWR pipe (small LOCA) 1 x

(a) 6-10 In. pipe represented by letdown and RWCV systems (see Table 2.2).

(b) 0: no activity or plan.

(c) Completed FY87.

(d) 10-24 In. pipe represented by service water piping (see Table 2.2).

(e) Components below this line were eliminated from full evaluation

by the screening criterion.

3.8
4.0 CONCLUSIONS AND RECOMMENDATIONS

The expert panel workshop was conducted to prioritize the TIRGALEX set of

nuclear power plant S/Cs for further evaluation within the NRC Plant Aging

Research Program. The prioritization was primarily based upon risk-based criteria; other technical criteria were used to categorize those S/Cs for which

aging research would benefit the resolution of GSIs and/or identified NRC/NRR

user needs but were not used to rank the S/Cs.

From the S/C prioritization using risk-based criteria, the major conclusions from the workshop are the following: 1) the prioritization of S/Cs was

accomplished by an expert panel using the multi-factor RSCAAMP methodology.

Analysis of the results showed that all factors of this methodology are equally

significant to an assessment of the relative risk importances of aged components. Further, the importance of including component aging and its management in plant risk assessments was dramatized by the unexpected findings in the

top-ranked components. 2) Current aging research warrants reevaluation: many

low-ranked S/Cs (ranks 1 and 2) are under study and others are considered to be

of interest; two of the components in the top two ranks (ranks 5 and 4) are not

currently being studied in the Plant Aging Research Program [containment

(BWR-Mk-1) and reactor internals]. 3) the panel's deliberations and findings

highlighted the need for improved PRA and aging-failure data bases, the shortfalls in current industry practices for detecting aging for risk-significant

failure modes in components and structures, and the usefulness of the methodology for focusing research and regulatory actions and for providing

utilities with areas to address in order to reduce the risk-contribution of

their aged components. 4) This study should be viewed as a starting point for

which expert opinion was required; from this study an appropriately-focused

research program can generate the data needed for a more definitive answer on

the relative risk-importance of aged components. 5) The explicit evaluation of

the importance of aging-induced common-cause failures would be a useful extension of the methodology, with minor development, the RSCAAMP model would be

amenable to such an evaluation.

From the S/C categorization using other technical criteria, the major conclusions from the workshop are as follows: 1) the S/Cs were categorized using

the other technical criteria. 2) There are many S/Cs for which an aging

research program could have direct benefit to the resolution of GSIs. 3) There

are fewer, but still substantial numbers of S/Cs for which an aging research

program could respond directly to identified user needs. 4) For all but a few

of these components aging research is already on-going.

Table 3.5 identified the S/Cs prioritized by the risk-based criteria, the

S/Cs upon which aging research could benefit the resolution of GSIs and User

Needs, and the S/Cs for which aging research is not already on-going or of current interest.

4.1
5.0 REFERENCES

ANS/IEEE. 1984. PRA Procedures Guide, NUREG/CR-2300, U.S. Nuclear Regulatory

Commission, Washington, D.C.

Hatch, S. W., Cybulskis, P., and Wooten, R. 0. 1982. Reactor Safety Study

Methodology Applications Program: Grand Gulf #1 BWR Power Plant.

NUREG/CR-1659/4, U.S. Nuclear Regulatory Commission, Washington, D.C.

Kolb, G. J. 1981a. Reactor Safety Study Methodology Applications Program:

Calvert Cliffs #2 PWR Plant. NUREG/CR-1659/3 of 4, U.S. Nuclear Regulatory

Commission, Washington, D.C.

Kolb, G. J. 1981b. Reactor Safety Study Methodology Applications Program:

Oconee #3 PWR Plant. NUREG/CR-1659/2 of 4, U.S. Nuclear Regulatory Commission,

Washington, D.C.

Meale, M. M. and D. G. Sutterwhite. 1987. An Aging Failure Survey of Light

Water Reactor Safety Systems and Components. NUREG/CR-4747, Volume 2, EG&G

Idaho, Sandpoint, Idaho.

Seaver, D. A. and W. G. Stillwell. 1983. Procedures for Using Expert Judgement to Estimate Human Error Probabilities in Nuclear Power Plant Operations.

NUREG-CR-2743, Decision Sciences Consortium, Inc.

Technical Integration Review Group for Aging and Life Extension. 1987. Plan

for Integration of Aging and Life-Extension Activities. U.S. Nuclear Regulatory Commission, Washington, D.C.

Telford, et al. 1986. Research Prioritization Method. U.S. Nuclear Regulatory Commission, Washington, D.C.

U.S. Nuclear Regulatory Commission. 1982. Interim Reliability Evaluation Program, Analyses of the Arkansas Nuclear One-Unit 1 Nuclear Power Plant.

NUREG/CR-2787, Vols. 1 and 2, Sandia National Laboratories.

U.S. Nuclear Regulatory Commission. 1985. A Prioritization of Generic Safety

Issues. NUREG-0933, Washington, D.C.

Vesely, W. E., and T. C. Davis. 1985. Evaluations and Utilization of Risk

Importances. NUREG/CR-4377, Battelle-Columbus Laboratories, Columbus, Ohio.

Vesely, W. E. 1987a., More Comprehensive Approaches to Determining Risk Sensitivities to Component Aging. Presented at the International Atomic Energy

Agency meeting, July 1987, Vienna.

5.1
Vesely, W. E. 1987b. Presentation to U.S. Nuclear Regulatory Commission, Risk

Evaluations of Aging Phenomena (REAP): Project Status Report. Presented at

Bethesda, Maryland, June 16, 1987.

Vesely, W. E. 1987c. Risk Evaluations of Aging Phenomena: The Linear Aging

Reliability Model and Its Extension. NUREG/CR-4769, EG&G Idaho, Sandpoint,

Idaho.

Vora, J. P. 1987. Nuclear Plant Aging Research (NPAR) Program Plan.

NUREG-1144, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, D.C.

5.2
APPENDIX A

DERIVATION OF BASIC RSCA MODEL AND THE RANKING METHODOLOGY
APPENDIX A

DERIVATION OF BASIC RSCA MODEL AND THE RANKING METHODOLOGY

The enclosed material was taken from an introductory workshop presentation

given by A. Wolford.

A.1
RSCA Methodology

• Unavailability, hence risk, is time dependent due to:

- Normal variations arising from constant failure rate with

repair (standard sawtooth behavior)

- Increasing failure rate due to aging degradation

* We are interested in quantifying the increase in risk due to

aging as separated from effects without aging

• Define:

Ron Total Risk including aging effects

R * Risk without aging contribution

q 0 - Total instantaneous unavailability including aging

q m Instantaneous unavailability without aging

contribution
RSCA Methodology

Highlights of Derivation

• The risk change due to aging may be expressed as

AR .(R - R a(R Ro) d t

t

* The derivatives on the right hand side, with use of the

chain rule, may be expressed as

R _ aRq (1) at q N

and

,R 0 - aRoqo (2)

* R. q both time dep., notation will be relaxed
RSCA Methodology

Highlights of Derivation

* We may write

a R- R) - aR3 q - qc +d R - DRq 0 (3)

a t 63q\@t_ t /\q _D/at

* The first term on the right hand side is called the Risk

Sensitivity due to a Component's Aging. It is comprised of

*> the normal risk importance of the component and the

additional rate of change in component unavailability due to

aging-driven degradation

• The second term is the contribution from the nonaging

(constant failure rate) change in component

unavailabilities

• Generally, the second term will be small compared to the

f irst
RSCA Methodology

Highlights of Derivation

A specific model of unavailability Is required for explicit

evaluation of Equation (3)

a Employing the Linear Aging Model to describe

failure rate,

increasing

* (t) 0 at o (4) e,

a The small availability approximation allows

~ qO

at

z A(t) - \o at (5)

Where an estimate of "a" is given by Equation (12)

a The final form for the basic RSCA equation becomes

a(R - ) DR . at

t c q

(6)
RSCA Methodology

Working Definitions of Terms

* Risk

Generally taken to be core melt frequency, in PRA sense

• Normal Risk mportance..is not Risk Contribution

Also called Birnbaum Measure or Reliability Importance. Is

a measure of the critical relationship to the system in

which it resides. It considers both system design (eg.

redundancy) and the components reliability.

• Instantaneous Unavailability

The probability that the component is not operating at

the instant of time, t

• Linear Aging Model

Time dependent failure rate modeled as linearly dependent on

time
RSCA Methodology

Working Def initions of Terms

• Aging Contribution

In the development of RSCA we have made the working hypothesis that figures of merit for plants, systems and

components (eg. Risk) may be separated into fractions which

are due to (systematic) aging and to those which are random

in nature. The aging contribution is the fraction of some

system variable which is modeled as aging-caused

• Aging Rate

Slope of the failure rate-time function (LAM) above

• Sensitivity

A derivative
Enhancements to the RSCA Methodology

* Agregation of component importances

How component's RSCA's may be combined within and across

plants

• Separation into Factors

How the RSCA equation is separated into physicallymeaningful factors

* Incorporation of Current Aging Management Adequacy Criteria

Method for inclusion of quantitative criteria addressing

the adequacy of current aging management practices for

maintaining aging increased risk at acceptable levels
Enhancements to the RSCA Methodology

Agregation of Importances

• Generalization of Equation (6) from a single component to

groups of components involves only the following trivial

summation

ARG E N; a t (7) - i ie G

• Where we have made the following notation changes

AlR Z(jF - Ro0 )

N 1 UaR,

aq;
Enhancements to the RSCA Methodology

Separation of Factors

* We wish to prioritize component groups, specifically the

TIRGALEX groups, by Aging Risk Increase, hence Equation (7)

should be integrated, thus:

AR lEN-ait dt

I G

(8)

which results in C-

~AR M Na. L 2 (9)
Enhancements to the RSCA Methodology

Separation into Factors

we subdivide the right hand side of Equation (9) into three

(3) factors, which are physically interpretable

AR 0 .I ;c ap A (10)

* Meanings

I-. - N -NI

C, 

- A " L

AN e 

-A- L

- Normal Risk Importance for te Group

- Average Effective Aging Rate for the

group

- Aging Exposure Period Factor

Integration causes power of two
Implementation of the RSCA Methodology

in an Expert Panel framework

• Generic estimates were taken from the four plant study by a

modified median

• Standard median was taken for odd samples, geometric mean

central values for even samples

* "This modified median was thought to provide the least biased

generic estimates from the limited four plant survey
Implementation of the RSCA Methodology

in an Expert Panel Framework

a A ranking process was developed to facilitate data handling

and presentation which also allows for less cumbersome

manipulation and interpolation

a The ranking scheme, however, is based on a logrithmic

scale, and the subtleties should be well understood. I-.

WA

a The ranking process ranks each of the factors

to the nearest integral base 10 log.

The Group with the maximum rank is scaled, by adding a

constant, so that the ranik is equal to 5, the maximum rank

for each factor.

Each group is also scaled by the same constant

a In Equation form: RANK(zi) log (zi) + Cz (11)

where Cz 5 - log Max(zifl
ESTIMATES OF AGING FAILURE RATE (A)

An estimate of the aging failure rate, A. may be obtained by the method of

moments which is:

4 Xf (2

A 3 TA (12)

where f = The number fraction of failures determined to be caused by aging

related stressors

A = The constant average failure rate. Generic values are used in this

study

TA = The mean time to aging failure

A.14
APPENDIX B

DATA TABLES SUPPORTING THE PRECALCULATED RSCA FACTORS
APPENDIX B

DATA TABLES SUPPORTING THE PRECALCULATED RSCA FACTORS

B.1
TABLE B.1. Four-Plant Component Data Summary:

Number of Components by Component Type

Number of Components

Component

Group

Pumps

lype

Motor drive pump

Turbine driven pump

ANOl

9

2

Grand

Gul f

9

Calvert

Cliffs

10

2

Oconee

11

1

Total

39

5

Valves Air operated valve

Check valve

Hydraulic valve

Manual valve

Motor operated valve

Safety/relief valve

Diesel generator

37

4

85

55

5

2

21

31

47

1

3

2

21

31

22

4 6

21 100

4

15 162

20 145

7

Diesels 2 7

Electrical Thermostat

Circuit breaker

Load/relay switch

Transfer switch

Bistable trip unit

1

65 8

4

4

2

77

4

1

3

1

3

Power Battery

Battery charger

Inverter

Transformer

DC bus

AC bus

Rectifier

2

2

4

14

7

11

12

3 3 2 10

2

4

14

7

1

12

Heat exchanger Turbine

Air conditioner

Heat exchanger

Motor driven fan

1

2

1

2

6

7

6

7

Structural

Totals

Tank 3

336 115 107

_3

78 638

B.2
TABLE B.2. Four-Plant Component of Data Summary:

Number of Components by System

Component

Group

Pumps

System

AC emergency power

Auxiliary feedwater system

High pressure core spray

High pressure injection

Low pressure coolant injection

Low pressure core spray

Low pressure injection

Reactor core isolation cooling

Service water

ANO

1

3

3

2

Grand

Gulf

1

3

1

Calvert

Cliffs

2

2

2

6

Oconee Total

1 2

2 7

1

2 7

3

1

3 7

1

4 15

1

2 3

Val ves Auxiliary feedwater system

High pressure core spray

High pressure recirculation

High pressure injection

Low pressure coolant injection

Low pressure core spray

Low pressure recirculation

Low pressure injection

Primary pressure relief

Reactor core isolation cooling

Residual heat removal

Service water

Suppression pool monitoring

Emergency cooking

Core flood

39

2

42

4

25

6

20

5

22 24 85

6

2

16 12 70

20

5

4

13 22 60

3

15

32

22 93

2

14

6

2 1

15

32

52 19

2

14

6

Diesel AC emergency power

Auxiliary feedwater system

1

1

3 2 6

1

Electri cal AC emergency power

Auxiliary feedwater system

DC emergency power

Engineered safety feature act.

Reactor protection

Service water

13

2

39

3

12 12

13

2

39

3

4 28

1

Power AC emergency power

DC emergency power

Reactor protection

12

35

4

3 3

12

2 43

4

Heat

Exchanger

Auxiliary feedwater system

Low pressure injection

Service water

Emergency cooling

Core flood

1

2

7

4

2

335

2

117 104

1

2

9

4

22 

Totals 76 631

B.3
TABLE B.3. Four-Plant Component Risk Importance

Data by Component Type

Component

TIRGALEX Group

Component of Importances

Grand Calvert Generic

PRA AN01 Gulf Cliffs Oconee Importances

11. Diesels Diesel generator 9.2E-4 2.2E-4 3.7E-3 9.2E-4

16. I&C a. Thermostat

b. Transfer

switch

c. Bistable trip

unit

6.OE-3

4.7E-6

6.OE-3

4.7E-6

1.2E-5 1.2E-5

17. Switchgear/

Relays

a. Relay (Load)

b. Circuit

breaker

4.8E-2

7.2E-2

4.8E-2

7.8E-2 1.8E-3 7.2E-2

18. Valves a. Air operate

valve

b. Check valve

c. Hydraulic

valve

d. Manual valve

e. Motor operated valve

f. Safety/relief

valve

3.2E-2 2.6E-4 3.2E-2

9.3E-2

3.8E-4

2.7E-3 1.3E-1 3.4E-3 1.8E-2

3.8E-4

1.6E-1 6.9E-3 1.2E-O

7.1E-2 9.3E-3 5.OE-2

6.9E-3 6.1E-5

5.1E-3

2.7E-3

3.3E-2

2.2E-2

3.5E-4 3.5E-4

19. Pumps a. Motor driven

pump

b. Turbine drive

pump

8.9E-3 2.1E-3 5.1E-3 2.3E-2 6.7E-3

9.3E-3 3.2E-2 2.OE-6 9.3E-3

21. Turbine 1.OE-4 1.OE-4

22. Heat

exchanger

23. Compressor

24. Fans/

chiller

25. Battery

Heat exchanger

Air conditioner

6.4E-3 6.4E-3

7.7E-4 7.7E-4

Fan 7.6E-3 7.6E-3

Battery 3.9E-3 5.7E-4 1.5E-3 1.7E-4 9.2E-4

26. Battery a. Battery 1.1E-4 1.1E-4

charger charger

inverter b. Inverter

c. Rectifier

4.2E-4

4.7E-4

4.2E-4

4.7E-4

B.4
TABLE B.3. (contd)

Component of Importances

Component

TIRGALEX Group PRA

Grand

AN01 Gulf

Cal vert Generic

Cliffs Oconee Importances

27. Transformer Transformer 1.2E-2 1.2E-2

29. Accumulator/ Tank

Tanks

2.7E-2 2.7E-2

30. AC/DC bus a. AC bus

b. DC bus

4.3E-2

1 E-1

4.3E-2

1.1E-1

NOTE: Median importances are used when an odd number importance values were

available. The geometric center (mean) of the middle two importances was used

when an even number of importance values was available.

B.5
TABLE B.4a. System Risk Importances and the Importances of Components Within the System (ANO1)

Component/System EPS AC EPS-OC AFNS

Normal Risk Impcrtances (N) ANOI

ESFAS HPCS HPR HPIS LPCIS LPCS LPR LPIS PPR RCCS RPS RHR SWS SPMS ECS CFS

Air conditioner

Thermostat

Circuit breaker

Motor - driven fan

Diesel generator

Heat exchanger

Invertor

Relay

Motor - driven pump

Turbine - driven pump

Rectifier

Transfer switch

Transformer

TanK

Bistable trip unit

W DC - bus

cn AC - bus

Battery

Air - operated valve

Check valve

Hydraulic valve

Manual (hand) operated valve

Motor operated valve

Safety/relief valve

Battery charger

Turbine

Sum of System 'N'

6E-3

5E-2 9E-3 IE-3 IE-2

1E-3 6E-3

3E-4 7E-4

2E-4 6E-3

4E-4

3E-4 9E-S

6E-3

3E-5

3E-3

2E-4 8E-3

4E-4

SE-6

1E-2 5E-4

3E-5

3E-2 ZE-4

1E-5

lE-1

4E-2 4E-5

4E-3

8E-2

4E-2

3E-3

1E-4 1E-2

3E-3

4E-4

4E-4

2E-2

4E-4 4E-3

1E-2 2E-4

6E-2

SE-2

3E-2

2E-2 2E-4

6E-3 3E-4 2E-4

8E-5

lE-! 1E-1 2E-1 1E-5 1E-4 2E-2 4E-4 3E-2 3E-4 1E-2 1E-1 4E-2 8E-4

1E-1 1E-1 2E-1 1E-5 1E-4 2E-2 4E-4 3E-2 3E-4 1E-2 1E-1 4E-2 8E-4
TABLE B.4b. [contd (Grand Gulf)]

Normal Risk Importances (N) Grand Gulf

Component/System EPS AC EPS-DC AFNS ESFAS HPCS HPR HPIS LPCIS LPCS LPR LPIS PPR RCICS RPS RHR SWS SPMS ECS CFS

Air conditioner

Thermostat

Circuit breaker

Motor - driven fin

Diesel generator 2E-4

Heat excnanger

Invertor

Relay

Motor - driven pump 7E-S 6E-4 lE-5 1E-4 lE-3

Turbine - driven pump

Rectifier

Transfer switch

Transformer

Tank

Bistable trip unit

DC - bus

AC - bus

Battery 6E-4

Air - operated valve

Check valve 2E-4 6E-4 3E-5 5E-4 4E-5 2E-3

Hyaraulic valve

Manual (hand) operated valve 7E-5 6E-4 IE-5 2E-4 4E-3 2E-3

Motor operated valve 1E-4 6E-4 3E-S 8E-4 5E-3 3E-3 9E-5

Safety/relief valve 6E-5

Battery charger

Turbine

Sum of System 2E-4 6E-4 4E-4 2E-3 8E-5 6E-5 2E-3 9E-3 8E-3 9E-S
TABLE B.4c. [contd (Calvert Ciffs)]

Component/System EPS AC EPS-OC AFNS ESFAS HPCS

Normal Risk Importances (N) Calvert Cliffs

HPR HPIS LPCIS LPCS LPR LPIS PPR RCICS RPS RHR SWS SPMS ECS CFS_

Air conditioner 8E-4

Thermostat

Circuit breaker

Motor - driven fan

Diesel generator

Heat exchanger

Invertor

Relay

Motor - driven pump

Turbine - driven pump

Rectifier

Transfer switch

Transformer

Tank

Bistable trip unit

7E-2

5E-2

1E-3 4E-6 4E-3

3E-2

DC - bus

AC - bus

Battery

Air - operated valve

Check valve

Hydraulic valve

Manual (hand) operated valve

Motor operated valve

Safety/relief valve

Battery charger

Turbine

Sum syster

2E-3

3E-2

IE-I

1E+O

3E-2

3E-3

2E-3

IE-2

8E-6

1E-S

3E-4

3E-4

1E-2

7E-3

4E-3 2E-3 IEtO 2E-2 3E-4 IE-I 2E-2
TABLE B.5. Average Age (TA) Matrix by Component and System

Air conditioner

Thermostat

Circuit breaker

Motor - driven fan

Diesel generator

Heat exchanger

Invertor

Relay

Motor - driven pump

Turbine - driven pump

Rectifier

Transfer switch

Transformer

Tank

Bistable trip unit

DC - bus

co AC - bus

to Battery

Air - operated valve

Check valve

Hydraulic valve

Manual (hand) operated valve

Motor operated valve

Safety/relief valve

Battery charger

Turbine

EPS-AC AFWS EPS-DC ESFAS HPCS HPR HPIS LPCIS LPCS LPR LPIS PPR RCICS RPS RHR SWS SPMS

10.57

4.9

ECS CFS

6.4 6.4 6.4 6.4

8.51 8.04

6.3 6.3

11.53 11.53

4.86

8.2

6.5 6.5

9

4.2 4.2

4.2

4.2 4.2 4.2 8.9

9.34

1.33

9.34

9.34

9.34

15. 73

4.9

8.02

8.02

8.02

9.4

4.9

6.5 7.4 6.5 6.5 6.'

6.5

5.9

7.6

5.9

6.5 6.5 6.5 6.5

9.1 5.9 5.9

6.5

0.3

6.5

5.9 5.9

6.5

6.5

5.9

6.5 8.9 6.5

6.5

5.9

6.5 7.6

5.9 8.5

6.5

5.9

4.3

9.01 8.9
TABLE B.6. Aging Fraction (f) Matrix by Components and System

Air conditioner

Thermostat

Circuit breaker

Motor - driven fan

Diesel generator

Heat exchanger

Invertor

Relay

Motor - driven pump(9)

Turbine - driven pump(9)

Rectifier

Transfer switch

Transformer

Tank

Bistable trip unit

DC - bus

AC - bus

Battery

Air - operated valve

Check valve(i)

Hydraulic valve

Manual (hand) operated

valve

Motor operated valve

Safety/relief valve

Battery charger

Turbine

Motor

Valve operator

Aging Survey

Aging

Fraction (f)

(No System

Dependencies)

0 .6 (a)

0.25(b)

0.26

0.6(a )

0.25

0.52

0.24

0.25

0.45

0.45

0.27(C)

0.10

0.16

0.42(d)

0.23(e)

0.09

0.09

0.32

0.49

0.85

0.49

0.49

Analysis Data

Aging

Fractions (f)

(with System

Dependencies)

CCW IE RPS

Repeated Failure

Cause Analysi 

Fractions (f)"tR ,j)

AFW IE HPIS SWS

0.75 0.48

0.47

0.47 0.77

0.19 0.19

0.63

0.87

0.63

0.75

0.77

0.83 0.87

0.77

0.49

0.49

0.32

0.16

0.36

0.25

0.51 0.46 0.39

B.10
TABLE B.6. (contd)

(a) Used blower/compressor.

(b) Used instrumentation: indicator/recorder.

(c) Used instrumentation: electronic power supply.

(d) Used vessel.

(e) Used instrumentation: controller.

(f) Used the lower bound from the repeated failure cause because it excluded

unknown failures.

(g) Few turbine and motor failures are reported relative to pump failures so

the pump aging rate is appropriate for motor- and turbine-driven pumps.

(h) The CTIS system data is unique, but the system is not included in PRA, so

it is not considered here.

(i) Check valves generally have very simple or no operator, so the failure

cause analysis data (f-0.85) was used.

(j) The f-values of the more detailed report failure cause analysis portion of

the study are noted as being higher than the f-values of the aging survey

analysis, but are only used for specific systems. The system independent

values of the aging survey analysis are generally used.

B .11
TABLE B.7. Component Failure Rates and Failure Rate Doubling Times

Doubling

Failure Rate Aging Rate Time

(Per Stby Hr) (Per hr per year) (year)

11. Diesel generator 4.OE-5 1.6E-6 25.0

16. a. Thermostat 3.OE-6 1.5E-7 20.2

b. Transfer switch 3.OE-6 2.3E-7 13.1

c. Bistable trip unit 3.OE-6 1.4E-7 21.9

17. a. Relay (load) 3.OE-6 9.1E-8 32.9

b. Circuit breaker 4.OE-6 1.6E-8 250.3

18. a. Air operate valve 4.OE-6 4.OE-7 10.0

b. Check valve 3.OE-6 3.8E-7 8.0

c. Hydraulic valve 4.OE-6 6.5E-6 0.6

d. Manual valve 3.OE-7 2.2E-8 13.8

e. Motor operated valve 4.OE-6 2.6E-7 15.2

f. Safety/relief valve 1.OE-5 6.7E-7 14.8

19. a. Motor driven pump 4.OE-6 2.2E-7 18.4

b. Turbine drive pump 4.OE-5 2.7E-6 14.6

21. Turbine 2.0E-4(a) 3.7E-6 54.8

22. Heat exchanger 3.OE-6 1.4E-7 21.9

23. Air conditioner 1.OE-5 5.6E-7 17.9

24. Fan 1.OE-6 2.1E-7 4.9

25. Battery 1.OE-6 3.4E-8 29.2

26. a. Battery charger 1.OE-6 3.5E-8 28.3

b. Invertor 1.OE-4 4.9E-6 20.4

c. Rectifier 3.OE-6 8.7E-8 34.6

27. Transformer 1.OE-6 1.7E-8 58.4

29. Tank 1.0E-9(b) 2.7E-11 37.0

30. a. AC bus 1.OE-8 1.1E-11 909.1

b. DC bus 1.OE-8 1.1E-10 89.3

NOTE: IREP Procedures Guide Generic Database was used to obtain failure rates

except as noted. Monthly tests were assumed when necessary.

(a) An IEEE-500 composite of steam driven condensing turbines was used for

turbine failure rate.

(b) The number used in the ANO-1 PRA. Probably an upper bound.

B.12
TA.BLE B.8. Component N and A Values by System

Plant: AN01 Unit

Comp

1

System N a Count

Thermostat

Circuit breaker

Circuit breaker

Circuit breaker

Circuit breaker

Motor-driven fan

Motor-driven fan

Diesel generator

Diesel generator

Heat exchanger

Heat exchanger

Invertor

Motor-driven pump

Motor-driven pump

Motor-driven pump

Motor-driven pump

Motor-driven pump

Turbine-driven pump

Turbine-driven pump

Rectifier

Rectifier

Transfer switch

Transformer

Transformer

Tank

Tank

Bistable

DC bus

AC bus

AC bus

Service water

Emer power system - AC

Auxiliary feedwater system

Emer power system - DC

Reactor protection system

Service water

Emergency cooling

Emer power system - AC

Auxiliary feedwater system

Low press injection system

Service water

Emer power system - DC

Emer power system - AC

Auxiliary feedwater system

High press injection system

Low press injection system

Service water

Auxiliary feedwater system

High press injection system

Emer power system - DC

Reactor protection system

Emer power system - DC

Emer power system - AC

Emer power system - DC

Auxiliary feedwater system

Core flood

Emer. safety feature act system

Emer power system - DC

Emer power system - AC

Emer power system - DC

Emer power system - DC

Auxiliary feedwater system

High press injection system

Low press injection system

Service water

Core flood

Low press injection system

Auxiliary feedwater system

High press recirc

High press injection system

Low press injection system

Service water

Emergency cooling

Auxiliary feedwater system

High press injection system

Low press recirculation

6.03E-03

5.47E-02

1.30E-03

8.51E-03

1.36E-02

1.14E-03

6.45E-03

2.57E-04

6.67E-04

1.90E-04

6.24E-03

4.25E-04

2.50E-04

9.32E-04

3.22E-05

1.86E-04

7.54E-03

6.42E-03

2.92E-03

4.44E-04

2.84E-05

4.70E-06

1.14E-02

4.67E-04

2.70E-02

1.50E-04

1.15E-05

1.14E-01

4.29E-02

3.50E-05

3.90E-03

7.82E-02

3.26E-03

3.75E-04

1.13E-02

1.50E-04

3.77E-04

4.04E-02

1.35E-04

1.36E-02

2.46E-02

5.91E-02

2.53E-02

1.61E-02

3.11E-03

3.81E-04

1.34E-03

1.37E-04

'.38E-04

1.38E-04

1.37E-04

9.90E-04

1.90E-03

1.39E-02

1.39E-02

1.21E-03

1.19E-03

4.33E-02

2.44E-03

2.42E-03

3.73E-03

3.76E-03

1.77E-03

1.75E-02

3.76E-02

7.66E-04

7.04E-04

2.13E-03

1 .50E-04

1.50E-04

3.70E-07

O.OOE+00

8.66E-04

9.62E-07

O.OOE+00

O.OOE+00

3.OOE-04

3.44E-03

3.44E-03

3.44E-03

2.51E-03

3.47E-03

5.73E-02

1.98E-04

2.22E-04

1.99E-04

1.98E-04

1.69E-04

1.98E-04

2.91E-03

2.91E-03

2.91E-03

2

13

2

38

12

3

4

1

1

2

4

4

1

2

2

2

2

1

1

8

4

1

2

12

1

2

3

7

10

1

2

12

10

10

3

2

4

5

2

25

7

32

14

21

7

4

Battery

Check valve

Check valve

Check valve

Check valve

Check valve

Hydraulic valve

Manual operated

Manual

Manual

Manual

Manual

operated

operated

operated

operated

valve

valve

valve

valve

valve

Manual operated val ve

Motor operated valve

Motor operated valve

Motor operated valve

B.13
TABLE B.8. (contd)

Plant: ANO1 Unit 1 (contd)

Comp

Motor operated valve

Motor operated valve

Motor operated valve

Safety/relief valve

Safety/relief valve

Safety/relief valve

Battery charger

Plant: Grand Gulf

Comp

Diesel generator

Motor-driven pump

Motor-driven pump

Motor-driven pump

Motor-driven pump

Motor-driven pump

Battery

Check valve

Check valve

Check valve

Check valve

Check valve

Check valve

System

Low press injection system

Service water

Core flood

Auxiliary feedwater system

Primary pressure relief

Core flood

Emer power system - DC

System

Emer power system - AC

High press core spray

Low press cool system

Low press core spray

Resid core isol cool system

Service water

Emer power system - DC

High press core spray

Low press cool system

Low press core spray

Resid core isol cool system

Residual heat removal

Service water

High press core spray

Low press cool system

Low press core spray

Resid core isol cool system

Residual heat removal

Service water

High press core spray

Low press cool system

Low press core spray

Resid core isol cool system

Residual heat removal

Service water

Supp pool monitoring

Primary pressure relief

N

4.39E-03

4.73E-02

1.50E-04

6.40E-03

3.OOE-04

1.50E-04

7.50E-05

N

2.21E-04

6.50E-05

5.56E-04

1.40E-05

1.OOE-04

1.37E-03

5.71E-04

1.95E-04

5.89E-04

2.80E-05

5.OOE-04

3.80E-05

1 .37E-03

6.50E-05

5.89E-04

1.40E-05

2.OOE-04

4.40E-03

1.67E-03

1.30E-04

6.07E-04

2.80E-05

8.OOE-04

4.59E-03

3.02E-03

8.80E-05

6.10E-05

2.91E-03

2.02E-03

2.93E-03

5.65E-03

1.OOE-02

1.OOE-02

2.67E-04

4

17

2

1

2

2

1

a Count

a Count

Manual

Manual

Manual

Manual

Manual

Manual

operated

operated

operated

operated

operated

operated

valve

valve

valve

valve

valve

valve

1.39E-02

3.69E-03

3.76E-03

3.57E-03

3.80E-03

1.77E-03

2.98E-04

3.03E-03

3.44E-03

3.57E-03

3.44E-03

3.42E-03

2.51E-03

1.54E-04

2.04E-04

O.OOE+00

2.OOE-04

1.98E-04

1.68E-04

1.92E-03

2.92E-03

2.86E-03

2.91E-03

2.91E-03

2.02E-03

2.95E-03

1.OOE-02

3

1

3

1

1

3

3

3

6

2

5

2

3

1

6

1

2

12

9

2

8

2

8

18

7

2

1

Motor operated valve

MMotor operated valve

Motor

Motor

Motor

Motor

operated

operated

operated

operated

valve

valve

valve

valve

Motor operated valve

Safety/relief valve

Plant: Oconee

Comp

Circuit breaker

Motor-driven pump

Motor-driven pump

System

Reactor protection system

Emer power system - AC

Auxiliary feedwater system

N

1.80E-03

7.OOE-05

7.OOE-05

a

1.39E-04

3.71E-03

2.43E-03

Count

4

1

1

B.14
TABLE B.8. (contd)

Plant: Oconee (contd)

Comp

Motor-driven pump

Motor-driven pump

Motor-driven pump

Turbine-driven pump

Battery

Air operated valve

Check valve

Check valve

Check valve

Manual operated valve

Manual operated valve

Manual operated valve

Motor operated valve

Motor operated valve

Motor operated valve

Motor operated valve

System

High press injection system

Low press injection system

Service water

Auxiliary feedwater system

Emer power system - DC

Auxiliary feedwater system

Auxiliary feedwater system

High press injection system

Low press injection system

Auxiliary feedwater system

High press injection system

Low press injection system

Auxiliary feedwater system

High press injection system

Low press injection system

Supp pool monitoring

N

2.92E-04

1.46E-03

2.15E-02

2.OOE--6

1.72E-04

2.64E-04

8.OOE-04

3.18E-04

2.28E-03

1.48E-04

4.O1E-03

9.20E-04

1.48E-04

4.58E-04

1 .82E-03

2.80E-04

a

3.76E-03

3.76E-03

1.77E-03

2.OOE-02

2.91E-04

3.52E-03

3.44E-03

3.43E-03

3.44E-03

2.03E-04

1 .97E-04

1.96E-04

2.91E-03

2.90E-03

2.91E-03

2.93E-03

Count

2

3

4

1

2

4

8

3

10

6

5

4

6

4

8

2

Plant: Calvert Cliffs

Comp

Air conditioner

Circuit breaker

Diesel generator

Load/relay unit

Motor-driven pump

Mctor-driven pump

Motor-driven pump

Turbine-driven pump

Battery

Air operated valve

Check valve

Check valve

Check valve

Check valve

Check valve

Manual operated valve

Manual operated valve

Manual operated valve

Manual operated valve

Motor operated valve

Motor operated valve

Motor operated valve

Motor operated valve

Motor operated valve

System

Service water

Reactor protection system

Emer power system - AC

Reactor protection system

High press injection system

Low press injection system

Service water

Auxiliary feedwater system

Emer power system - DC

Auxiliary feedwater system

Auxiliary feedwater system

High press injection system

Low press injection system

Service water

Supp pool monitoring

Auxiliary feedwater system

High press injection system

Low press injection system

Service water

Auxiliary feedwater system

High press injection system

Low press injection system

Service water

Supp pool monitoring

N

7.70E-04

7.20E-02

3.70E-03

4.80E-02

1 .11E-03

4.20E-06

3.97E-03

3.20E-02

1.50E-03

3.20E-02

1 .28E-01

3.35E-03

8.40E-06

3.10E-04

7.70E-04

1.21E+00

2.22E-03

1.26E-05

9.94E-03

3.20E-02

1.04E-02

3.30E-04

6.71E-03

4.04E-04

a

4.91E-03

1.37E-04

1 .39E-02

8.02E-04

3.76E-03

4.76E-03

1 .77E-03

1.75E-02

2.99E-04

3.51E-03

3.44E-03

3.44E-03

3.57E-03

2.52E-03

3.44E-03

1.98E-04

1.98E-04

O.OOE+00

1 .70E-04

2.91E-03

2.91E-03

2.91E-03

2.02E-03

2.91E-03

Count

2

8

2

4

2

2

6

2

3

2

8

6

4

1

2

10

4

6

11

2

6

3

10

1

B,15
APPENDIX C

PRECALCULATED N, A, CDFA VALUES FOR PRA-BASED COMPONENTS
APPENDIX C

PRECALCULATED N, A, CDFA VALUES FOR PRA-BASED COMPONENTS

C.1
TABLE C.1. Pre-Calculated N, A, and

Risk

Importance

11. Diesel 9.2E-4

16. a. Thermostat 6.OE-3

b. XFER switch 4.7E-6

c. Bistable 1.2E-5

17. a. Relay 4.8E-2

b. Breaker 7.2E-2

18. a. A.O. valve 3.2E-2

b. Check valve 1.8E-2

c. Hydraulic valve 3.3E-4

d. Manual valve 3.3E-2

e. M.D. valve 2.2E-2

f. S/R valve 3.5E-4

19. a. Motor pump 6.7E-3

b. Turbine pump 9.3E-3

21. Turbine 1.OE-4

22. Heat exchanger 4.4E-3

23. Air conditioner 7.7E-4

24. Fan 7.6E-3

25. Battery 9.2E-4

26. a. Battery charger 1.1E-4

b. Inverter 4.2E-4

c. Rectifier 4.7E-4

27. Transformer 1.2E-2

29. Tank 2.7E-2

30. a. AC bus 4.3E-2

b. DC bus 1.1E-1

CDFA Values for PRA-Based Components

A CDFA

(per hr per yr) (per yr cubed)

1.6E-6 1.3E-5

1.5E-7 7.9E-6

2.3E-7 9.5E-9

1.4E-7 1.5E-8

9.1E-8 3.8E-5

1.6E-8 1.OE-5

4.OE-7 1.1E-4

3.8E-7 6.E-5

6.5E-6 1.9E-5

2.2E-8 6.4E-6

2.6E-7 5.OE-5

6.7E-7 2.1E-6

2.2E-7 1.3E-5

2.7E-6 2.2E-4

3.7E-6 3.2E-6

1.4E-7 7.8E-6

5.6E-7 3.8E-6

2.1E-7 1.4E-5

3.4E-8 2.7E-7

3.5E-8 3.4E-8

4.9E-6 1.8E-5

8.7E-8 3.6E-7

1.7E-8 1.8E-6

2.7E-11 6.4E-9

1.1E-11 4.1E-9

1.1E-10 1.1E-7

C.2
APPENDIX D

DETAILED RATIONALE FOR PANEL'S FINDINGS FOR PRIORITIZATION

USING RISK-BASED CRITERIA
APPENDIX D

DETAILED RATIONALE FOR PANEL'S FINDINGS FOR PRIORITIZATION

USING RISK-BASED CRITERIA

The enclosed rationale are presented line by line for the S/Cs from the

data presented in Table 2.7. The S/Cs are arranged by their TIRGALEX S/C

numbers (e.g., 1:RPV, etc.), not by their order in the table.

D.1
1: Reactor Pressure Vessel (RPV)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-0 2.OE-12 120 0.90 0.10 1.4E-5 3

N: Failure in the RPV (defined as gross rupture) results directly in core

damage because safety inspection systems are not capable of providing

sufficient core cooling.

A: The panel requested a specific notation of these opinions that the model

does not address pressure vessel aging well because the vessel lifetime is so

long. There are no documented failures, so the age factor was considered low.

Lind: 120 months. This value was based on the ten-year in-service inspection

frequency for the pressure vessel per American Society of Mechanical Engineers

(ASME) Section XI requirements. Surveillance coupons are also pulled and

examined at this time.

P 90% -- Based on well developed technology of nondestructive examination

(NDE) inspection methodology.

PR/n: 10% -- This low value reflects the fact that a pressure vessel has not,

to date, ever been successfully repaired (annealed or otherwise stress relieved

in the beltline region). Mitigation is currently handled through flux leakage

management programs aimed at aging rate reduction, not repair.

D.2
2.a: Containment (BWR-Mk-I)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-0 1.OE-7 18 0.90 0.80 1.9E-3 5

N: The "component" in this case is the BWR Mark I containment wetwell which

was separated from other containment components for special considerations.

However, all containment structures were given the highest risk importance.

A: Aging issues for the BWR Mark I containment wetwell were felt to be more

severe than for other concrete structures due to wet well problems with

corrosion.

Lind: 12 months. Reflects visual inspection of torus and associated

components.

PD: 90% -- Corrosion and other age related manifestations are being given high

priority by BWR Mark I owners.

PR/D: 80% -- Usually radiographic documentation or other clearly discernable

correction.

D.3
2.b: Containment (other)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-0 1.OE-13 60 0.10 0.50 7.0E-7 2

N: The importance for all containment and concrete structures was set to the

highest level of risk significance as discussed in Section 3.

A: The aging factor was considered to be low for the majority of the concrete

structures since they are designed to withstand environmental and stress

conditions.

Lind: 480 months. Reflects the panel opinion that the containment is

basically uninspected following construction.

PD: 10% -- Age related information is not specifically addressed by current

surveillances.

PR/n: 50% -- The failure (defect) observed may not be significant in terms of

containment integrity.

D.4
3: Other Concrete Structures

FINAL

N A Lind PO PR/D RISK RANK

1.OE-0 1.OE-13 60 0.10 0.50 7.OE-7 2

N: Buildings and other concrete structures adjacent to equipment were of

highest risk importance since they were felt to be of equal risk significance

to the pressure vessel; e.g., failure (collapse) of a building well directly

result in core damage due to damage to equipment severing of pipes and

cables. Seismic events are the initiators.

A: The stipulation for the age factor was that it was to be the lowest on the

ranking. It was recognized that aging may even strengthen the concrete.

L d: 480 months. This indicates that these structures are basically not

effectively inspected during their nominal 40 year design life.

PD: 10% -- Reflects only exterior visual inspection.

PR/P: 50% -- Identified failure (deficiency) may not be significant to

con ainment integrity.

D.5
4.a: Reactor Coolant Piping and

Safe Ends (Large LOCA)

N

1.OE-4

A

1.OE-12

Lind

0

PD PR/D RISK

n/a

FINAL

RANK

0 0 1

N: For 20" pipe, a large break was assumed to be beyond the makeup capability

of HPCI. The conditional probability for core damage given large LOCA is

generally set in PRAs at 1.OE-4.

A: The small LOCA PWR category was set at 3E-11, and the large LOCA was

considered still less likely than that.

(Eliminated from further consideration by screening criterion

(ARc 1E-7 CD/yr.)

D.6
4b: PWR Pipe (Small LOCA)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-3 1.OE-12 0 0 0 n/a 1

N: A small break LOCA of a pipe in the size of 20" in a PWR was considered for

this grouping, a small break being one such that HPCI could keep up with it.

A: The aging factor for the PWR small LOCA was considered 3 orders of

magnitude lower than that for a BWR since BWR piping exhibits higher stress

corrosion rates than does PWR piping.

(Eliminated from further consideration by screening criterion

(ARC 1E-7 CD/yr.)

D.7
4c: BWR Pipe (Small LOCA)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-3 3.OE-8 36 0.20 0.90 3.7E-5 3

N: The risk importance of 20" pipe with a small break such that HPCI can keep

up with it was felt to have the same conditional aspects as other small LOCAs,

and was therefore set at E-4.

A: BWR stress corrosion was seen as the primary aging mechanisms so the

following formula used to obtain the aging rate: 1E-03 (which is small LOCABWR)/(10 yr. doubling time x 8000 hrs/yr) = 3E-08.

Lind: Weighted average based on detection of >50% through-wall cracks.

PR: No defined test for wall thinning currently established. Anchored on 50%

through-wall => 90% detection probability.

PR/D: Replacement is method of repair with assumption that radiography and

post-repair hydro will be required.

D.8
5a: Large Other Safety Piping

FINAL

N A Lind PD PR/D RISK RANK

6.4E-3 3.OE-9 18 0.10 .90 2.3E-5 3

N: Service water system is the surrogate system chosen; piping 10-24 in. This

system is a vital link in preventing decay heat from causing core damage in a

post-trip condition and in providing cooling to key safety equipment. Piping

only is considered in this category.

A: The failure mode considered here was a complete failure of the piping

network to deliver flow required to the component cooling and RHR heat

exchangers.

Lind: 18 months. This corresponds to flow surveillance testing of the system.

P Aging mechanisms are not, in general, addressed by the system surveillance

w ich basically requires only a minimum flow rate to be met in a given

component branch.

PR: Identified failures are corrected and tested for compliance.

D.9
5.b: Small Other Safety-Related Piping

FINAL

N A Lind PD PR/D RISK RANK

1.OE-3 3.OE-7 60 0.10 0.90 2.1E-3 5

N: Letdown and RWC piping (6-10 in.) are the surrogate. The dominant risk

contributor in this category of pipe was considered to be a small LOCA. Thus,

the risk importance is the same as for other small LOCAs. This scenario

considered a small break such that HPCI can barely keep up with it;

approximately a 1" break or larger.

A: Smaller pipe was addressed in this category, and the aging acceleration was

felt to be one order of magnitude larger than the BWR-Small LOCA case for RCP

pipe, namely 3E-7.

LW d: 60 months. This represents a system weighted average aimed at crack

de ection.

PB: 10% -- Wall thinning is not considered. Crack resolution is assumed at

% confidence level for 50% through wall or more.

PRID: 90% -- Detected problems are weld repaired and confirmed by radiography.

D. 10
6.a: Steam Generator (S/G) Tubes

FINAL

N A Lind PD PR/D RISK RANK

3.OE-4 5.OE-6 36 0.50 0.50 9.5E-4 5

N: It was assumed that the concern was tube rupture in the primary to

secondary system connection. N was felt to be moderate because there are

mitigating steps available and reasonably well-defined operator actions for

mitigation depending on the scenario and plant-specific conditions. This event

was seen as more likely to lead to core damage than a small LOCA; an importance

of 3E4 was assigned.

A: For a tube leak between the primary and secondary sides, many failure mechanisms exist and all were felt to be attributable to aging. A S/G tube leak

was estimated to occur every year; its age factor was considered high.

Li d: 36 months. This value reflects the rotational inspection of a representa ve population (typically 10%) and assumes that failures in the tested

population will trigger increased inspection.

PE: 50% -- Age detection methods are currently not reliable. This figure

obviously related only to the population that is inspected.

PI/D: 50% -- Definition is sensitive here. Repair is by the affected tube

p ugging and does not return the component to new condition. The steam generator is over designed such that, up to a point, plugging does not degrade the

S/G's ability to perform its safety function.

D.11
6b: Steam Generator Shell

FINAL

N A Lind PD PR/D RISK RANK

1.OE-5 1.OE-12 0 0 0 n/a 1

N: The risk importance of this category was considered equivalent to a main

steamline break inside containment.

A: The aging factor was set at 1 order of magnitude larger than a small

break/large pipe in a PWR.

Eliminated from further consideration by screening criterion (ARc 1E-7 CD/yr.)

D.12
7: Reactor Coolant Pump Casing

FINAL

N A Lind PD PR/D RISK RANK

1.OE-4 1.OE-12 0 0 0 n/a 1

N: Same risk importance assumed as for large break LOCA.

A: The wall thickness for the casing is twice that for pipe and the ASME code

requires inspection of pump casings. No problems have been found. Therefore,

the aging factor was felt to be extremely low.

Eliminated from further considerabion by the screening criterion

(ARc 1E-7 CD/yr.)

D.13
8: Pressurizer

FINAL

N A Lind PD PR/D RISK RANK

1.OE-4 1.OE-12 10 0.90 0.90 1.2E-12 1

N: One of the panel members had experience with a heater weld leak and felt

that the risk importance should be the same as a small break LOCA. It was also

observed that upon failure, the pressurizer cannot be isolated.

A: Since these are all clad pieces and vessels, corrosion was seen as a second

or third order effect. This would be in the same category as the pressure

vessel or reactor coolant piping and safe ends for a large break LOCA.

Lifd: Averages in ASME Section XI and continuous monitoring via leak

de ection.

PD: Detection mechanism is visual inspection, NDE methods, and leaks.

Frequency number is a composite of these.

PR/D: Any failure (leak) represents a small break LOCA so repair is assured.

Post-workshop review result in component eliminated by screening criterion

(Rc 1E-7 CD/yr.)

D.14
9a: Control Rod Drive Mechanisms (BWR)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-1 3.OE-9 18 0.90 0.90 4.5E-6 2

N: Standby liquid control is the only available backup assuming that the CRDM

has failed. 0.1 was considered a reasonable rate for the SLC system and the

required operator action.

A: The CRDMs are known to be subject to aging and there has been leaking. The

aging rate was felt to be worse than that for BWR pipe due to the smaller

diameter and the stress level involved. It should be noted that the aging

factor does not account for the solenoid valves, which were felt to be the

dominant failure mechanism.

Lind: Follows Tech Spec guidelines.

PD: Weighted for Tech Spec initiated test criteria failures and resulting

replacement of piece parts.

PR/D: Influenced by rigorous surveillance and post-maintenance test

requirements. Normal practice is to remove, rebuild or replace CRDMs each

refueling outage such that in ten years all drives will have been completely

replaced, setting the clock to zero.

D.15
9b: Control Rod Drive Mechanisms (PWR)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-3 3.OE-11 18 0.50 0.90 1.5E-9 1

N: The risk importnce for these was considered a factor of 20 to 30 lower than

the BWR CRDM because ATWS mitigation is easier (primary, secondary cooling).

A: The PWR CRDM aging factor was felt to be two orders of magnitude better

than for the BWRs because of the multiplicity and design. With the BWR

solenoid, one train is kept energized and is therefore more susceptible to

aging.

Ligd: Weighted for Tech Spec requirements to replace any component pieces that

in cate degradation through surveillance testing.

PR: Rigorous testing, but not all pieces reveal degradation in tests now

a ministered.

PR/D: Stringent post-maintenance test requirements assumed on this component.

D.16
10.a: Cables

FINAL

N A L P PR RISK RANK

1.1E-1 2.7E-9 B O.?O 0.48 2.1E-3 5

N: Risk importance was assumed to be similar to that of a DC bus. This is

based upon the assumption that there exists at least one DC cable for each DC

bus whose failure has the same effect as the failure of the bus. These are the

risk-dominant cables.

A: Considering the top 10% of the cables in terms of importance, the aging

factor was treated as nearly the same as that for a transformer, or 2.7E-9.

Lin: 60 months; no systematic inspections. A series of failures of a

par icular type would trigger an investigation of that type and manufacture

only.

PD: 10% -- Shear volume, diversity of types and the virtual impossibility of

complete surveillance was the limiting consideration.

PR/n: 90% -- Once detected, the accepted repair procedure is to replace the

cabe and with high probability on all similar applications. This historically

has a high rate of success. See comments on connectors.

D. 17
10.b: Connectors

FINAL

N A Lind PD PR/D RISK RANK

2.OE-2 2.7E-8 60 0.20 0.90 1.8E-3 5

N: The panel judged connections at a lower risk importance relative to cables

since they are used in fewer applications than cables. The circuits often have

diverse channels for functional aspects, and power cables often don't use

connectors. This is based on the assumption that there does not exist one

connector whose failure is equivalent to failure of the bus (see cables).

A: The aging acceleration was considered by the expert panel to be one order

of magnitude worse than the cables, or 2.7E-08.

Lin : 60 months. This figure is based upon the fact that there is no

systematic test or inspections but that a series of failures of a specific

connector may be viewed as a generic indication which would trigger increased

surveillance of like items with possible wholesale replacement.

Po: 20% -- This relatively low value reflects the fact that there is a

sizeable population of diverse types and the lack of age related investigation

of these various types. Orientation is aimed at failures.

PRID: 90% -- Repair probability is based on the diversity of the component and

endeavors. The lack of a systematic effort to investigate root causes and

carry out a full-scale investigation of all types regardlessof failure

exposures.

D.18
11: Diesel Generators

FINAL

N A Lind PD PR/D RISK RANK

2.OE-2 3.6E-6 3 0.30 0.90 2.7E-4 3

N: Recent PRAs have found loss of offsite power as among the highest

contributors to risk whereas the 4 PRAs considered in this study do not

accurately capture the LOSP risk.

A: The aging rate is high and the primary reason is believed to be testing,

especially fast-start testing. Research is needed in this area and should

continue to be focused on the large body of diesel information (nuclear and

non-nuclear) available to uncover other aging causes within and outside testing

and to document them so action can be taken.

Lind: 3 months. This is based on major tests which include load sequencing of

the vital bus. This is a weighted figure which integrates the complex multisystem nature of the emergency diesel.

PD: 30% -- This is an integrated value which recognized various sub-system

data frequencies.

PR/D: 90% -- This value acknowledges post-maintenance surveillance testing.

D.19
12: Reactor Internals

FINAL

N A Lind PD PR/D RISK RANK

1.OE-1 2.OE-9 18 0.10 0.90 2.4E-4 4

N: For a significant but partial coolant blockage, the importance was

considered less than that for the reactor vessel but greater than that for ATWS

(but same as ATWS for BWRs).

A: Stainless steel swelling and stress corrosion cracking do occur and partial

support failure requires less severe conditions than the RPV so the age factor

was established at an order of magnitude greater than the reactor pressure

vessel.

L 18 months. This is based on inspection during refueling outage coupled

wih on-line instrumentation of vessel condition (principally AP).

PD: Difficult to examine at an age detection level -- few existing

surveillance parameters which reflect aging mechanisms.

PR D: Once detected, a failure will require the reactor to be shut down until

repairs are completed and testing is satisfactory.

D.20
13: PWR Pressure Vessel Support

FINAL

N A Lind PD PR/D RISK RANK

1.OE-7 1.OE-12 0 n/a 1

N: The definition of this category was PWR sliding foot. If the sliding foot

is removed, it was believed that the vessel would be supported by the pipes,

since steam generators have been known to be held up this way.

A: Not seen as an issue.

Eliminated from further evaluation by this screening criteria

(ARc <1E-7 CD/yr.)

D.21
15: Snubbers

FINAL

N A Lind PD PR/D RISK RANK

1.1E-6 5.1E-6 18 0.90 0.90 8.4E-8 1

N: Snubbers were not considered to play a risk-significant role except for

those on the steam generators under seismic conditions. The risk importance

was felt to be around the same level as for the turbine.

A: Aging considerations were high since the snubbers won't do their job if

they are not maintained, and the aging rate was seen to be similar to that for

bolts.

Lind: Accounts for visual walkdown outside containment, and assumes that

surveillance failures lead to a higher inspection frequency.

PD: Estimate may be somewhat high due to lack of standardization in testing,

but assumption that failures trigger an increase in surveillance scope still

makes it reasonable.

PR/D: Extensive post-maintenance testing ensures successful repair to the

extent possible.

D.22
16a: Thermostat

FINAL

N A Lind PD PR/D RISK RANK

6.OE-3 1.5E-7 18 0.7 0.9 2.2E-5 3

N: This is a subgroup of Instruments and Controls (I&C), which were considered

to be highly correlated but diverse. The generic 4-plant PRA value was deemed

to be satisfactory.

A: Considerations of equipment qualifications since TMI; field monitoring

veruss control environment, which should be considered separately, and doubling

times suggested that the precalculated values were satisfactory (unless field

mounted and control environment I&C would later be considered separately).

Lind: 18 mo

PD: 70%

PR/D: 90%

D.23
16b: Transfer Switch

N A Lind PD PR/D RISK

4.7E-6 2.3E-7 18 0.2 0.9 3.3E-7

N: A subgroup of I&C which are highly correlated but diverse.

value was deemed satisfactory.

A: Same rationale as for 16a. Thermostat.

Lind: 18 mo

PD: 20%

PR/D: 90%

FINAL

RANK

1

The 4-plant PRA

D.24
16c: Bistable

FINAL

N A Lind PD PR/D RISK RANK

1.2E-5 1.4E-7 18 0.7 0.9 4.2E-8 1

N: A subgroup of I&C, which are highly correlated but diverse. The

precalculated value was deemed satisfactory.

A: Same rationale as for Thermostat (16b)

Lind: 18 mo

PD: 70%

PR/D: 90%

D.25
17.a: Relay

FINAL

N A Lind PD PR/D RISK RANK

4.8E-2 2.5E-7 6 0.20 0.90 4.OE-4 4

N: Relays were considered vulnerable to common mode failure, especially in a

seismic event, and due, to their involvement in critical paths for nearly all

functions, they were seen as highly risk significant.

A: Relays experience degradation due to heat and humidity and are more

susceptible to seismic common mode failures when in an aged condition. Contact

degradation (pitting and arcing) also occurs as an age related failure

mechanism.

Lind: 6 months. This is a weighted estimate for all safety significant relays

based on the periodic testing, combined with system response-time testing done

once every refueling on key protection systems.

PD: 20% -- Reflects the basic lack of aging determination in this component

until substantial numbers of failures have been experienced, and may also be

limited to a particular type and manufacture. Repair is usually accomplished

by replacement.

PRD: 90% -- Shows the above weighting of differing relay types and

ma ufacture, and repair by replacement techniques.

D.26
17b: Breaker

FINAL

N A Lind PD PR/D RISK RANK

7.2E-2 1.6E-8 18 0.50 0.90 5.6E-5 3

N: Risk importance of breakers was considered high due to their fire

sensitivity, lack of backup, and far-reaching impact on many critical

components should they fail. Int his case, it was assumed that there was more

than one breaker associated with each major AC bus whose failure would have the

same effect as failure of the bus. Thus, the risk importance of breakers as a

group would be somewhat higher than AC buses as a group.

A: Aging mechanisms are associated with contact burnout and the close/open

coils, but the modes are detectable. Breakers were considered accessible,

maintainable, and a static technology.

Lind: A weighted average based on quantity and diversity of type and

manufactures of safety related breakers.

PR: High incidence of breaker failure connotes possible undetected aging

p enomena.

PR: Reflects long historical experience (nuclear and non-nuclear) and state

of echnology.

D.27
18a: Air Operated Valves (AOV)

FINAL

N A Lind PD PR/D RISK RANK

3.2E-4 4.OE-7 18 0.50 0.90 6.2E-6 2

N: Redundancy exists to back these up, in the form of MOVs, other AOVs, etc.

AOVs are in one of three or four systems available to mitigate core damage, so

the importance rank reflects this.

A: AOVs are in critical functions, and their air lines, due to their size,

have failure rates of E-01 to 1E-02 per year, so the age factor was set to

reflect these considerations.

Li d: 18 months. Normal satisfactory operation or partial stroke tests are

no considered here.

PD: 50%--Based on surveillance testing during shutdown (disassembly and

inspection as well as functional checks).

PR D: 90%--Due to the safety importance of these valves thorough postma ntenance testing.

D.28
18b: Check Valves

FINAL

N A Lind PD PR/D RISK RANK

8.OE-4 3.8E-9 18 0.10 0.90 3.7E-6 2

N: Two failure mechanisms were considered: 1) extremely high back-pressure

from, for example, the reactor coolant system to a low pressure system such

that a safety valve will not relieve the pressure, and 2) failure to open or

close upon demand (or disk falling off, which would produce same result). The

risk importance of the first mechanism is high and the second low, so the

selected importance is based only on the first mechanism.

A: The aging influences involved in the failure mechanisms cited above were

considered lower for the first item and higher for the second; the aging rate

balanced out to a median level.

Liny: This is a composite figure which covers all safety-related check valves

inc uding testable checks.

PD: Back leakage gives highest indication of aging problems.

PR/f: Best you can attain because of common mode failure. Assumes complete

posE -maintenance testing.

D. 29
18c: Hydraulic Valves

FINAL

N A Lind PD PR/D RISK RANK

1.OE-5 1.3E-7 18 0.50 0.90 6.3E-8 1

N: These valves are not used in a BWR for feedwater or main steam isolation

and the PWR closing rate requirements are not risk-significant.

A: The NPRDS data base used for the initial ranking represents a small

population and a relatively high number (2) of failures so the original ranking

was seen as being artificially high. The hydraulic actuators have a proven

history, but the accessories that go with them may affect the aging factor.

These valves are used to protect the turbine and generator from runaway and

should be considered to be reliable since there is a significant experience

base. If they were not reliable, there would be two valves in place rather

than one, because when one valve fails the turbine is lost.

Lipd: No credit is given for partial stroke tests, so only major surveillance

in ervals are counted.

P : Current surveillances give little aging information per se. Failure of

s roke time testing generally results in valve dissassembly which reveals the

valve's age status.

PR/D: Post-maintenance testing gives a clear picture of repair success.

D.30
18d: Manual Valve

FINAL

N A Lind PD PR/D RISK RANK

1.OE-5 2.2E-9 60 0.50 0.60 2.7E-8 1

N: The present ranking reflects administrative failure (e.g. operator error in

locking valves shut), not actual valve internal failure. Also, 162 valves were

modeled in the original data set so the assumptions made are multiplied.

Still, the the failure mechanisms are not believed to significantly affect

risk.

A: Although these valves are subject to aging mechanisms such as

erosion/corrosion and stem breakage, there was not sufficient impetus to

warrant changing the ranking.

Lin : Incorporates Tech Spec requirements, operational checks, and utility

ini iatives.

P: Includes all safety-related valves, and represents current inspectionm

e hods.

PR/D: Based on evidence of recurring problems.

D.31
18e: Motor Operated Valves (MOV)

FINAL

N A Lind PD PR/D RISK RANK

2.2E-2 3.6E-6 3.9 0.70 0.90 5.6E-5 3

N: The MOV risk importances vary over many orders of magnitude; it was dec

to assign a moderate importance to reflect this.

A: There are as many failure mechanisms for MOVS as there are for other va

types plus additional failure mechanisms associated with the valve operator

intermittent duty cycle that are omnipresent and key to MOVs.

Lind: 3 months. This is a composite figure of all safety related MOVs bas

on stroke testing.

PD: 70%--Use of systems like MOVATS during stroke testing is significantly

increasing aging detection.

PR/D- 90%--Valves must meet stringent post-maintenance stroke testing

requirements.

ided

lve

and

ed

D.32
18.f: Safety/Relief Valve (S/RV)

FINAL

N A Lind PD PR/D RISK RANK

1.OE-4 6.7E-7 18 0.90 0.90 1.OE-6 2

N: In BWRs, two S/RVs sticking open have caused a LOCA in the past. RCIC can

compensate for one stuck open S/RV, but not two. Therefore, the risk

importance rank is greater than three, but not as high as five (on the relative

scale).

A: The expert panel felt that S/RVs do age so the ranking in this category

should be high. The "saving grace" was considered the maintainability

accounted for in the P values of the model.

Lind: 90% of valves every 18 months. Tech Spec regulated.

PD: Very high due to sophisticated equipment which is available to test the

actuation, blowdown, and closure specifications as a result of previous safety

concern resolution.

PR/D: High due to extensive operating and test experience.

0.33
19a: Motor Driven Pump

FINAL

N A Lind PD PR/D RISK RANK

6.7E-3 2.2E-7 18 0.70 0.90 2.2E-5 3

N: Despite the fact that MD pumps are in both mitigation and support systems

with no backup, when compared with the turbine-driven pumps and when determined

that station blackout AFW and RCIC needs drove the turbine pump importance up,

it was decided to leave the MD pump rank at a medium level.

A: Motor driven pumps were considered relatively simple in nature and a long

experience base exists in comparison to the turbine pumps. The aging rate was

not felt to be as high. Aging effects exist none the less due to the

surrounding environment, thus keeping the MD pumps relatively high in ranking.

L-nd: A weighted average of safety related units which considers various types

o surveillances.

PD: A wide range of opinion which seemed to be based on variations in

experiences. The midpoint seemed appropriate.

PR/D: Fairly well defined repair practices and tests following repair.

D.34
19.b: Turbine Driven (TD) Pump

FINAL

N A Lind PD PR/D RISK RANK

9.3E-3 2.7E-6 12 0.50 0.90 5.4E-4 4

N: Station blackout-related needs for AFW (PWR) and RCIC (BWR) were considered

the driving factors for the relatively high T Pump risk importance.

A: The turbine driven pump was considered more prone to aging-related degradation due to the governor control system and its function as a "water wheel" in

the AFW and RCIC systems. It is therefore more susceptible to mechanical and

auxiliary system damage (than the motor driven pumps, for example) and merits

the high ranking.

Lind: 12 months. This reflects a weighted factoring of full surveillance

texts (full flow) with more frequent, but less in-depth surveillance

information.

PD: 50% -- Factors include predictive testing (oil analysis, vibration

measurement, etc.) with other surveillance information.

PR D: 90% -- Complete repair is assumed for these components with postma ntenance surveillance testing.

D.35
20: Motors

FINAL

N A Lind PD PR/D RISK RANK

N: Motors were considered under their specific motor operated equipment groups

such as motor operated valves and motor driven pumps.

D.36
21: Turbine

FINAL

N A Lind PD PR/D RISK RANK

1.OE-6 1.OE-7 60 0.90 0.50 5.4E-8 1

N: Turbine failure was considered separately from turbine missiles. The

original turbine ranking was based on turbine driven pumps. Additionally,

though turbines fail regularly, the risk importance of the failures was still

felt to be low. Even in the case of turbine missiles, they would have to

strike the core directly or go into the control room to cause core damage.

A: The original ranking reflected a data set which implied a failure (not

trip) of every turbine each year, and this seemed unrealistically high.

Turbine problems were seen to be centered in a couple of poorly-monitored

systems and poorly-constructed turbines. The consideration of this category as

being the turbine generator was not agreed to by the group due to the

introduction of susceptability of stray currents rather than mechanical aging

mechanisms.

Lind: Basically, this represents a tear down frequency to determine

degradation.

PD: Complete examination for degradation is assumed.

PR/n: Even with post-maintenance surveillance testing, experience with

turbines has been similar to diesels, not real great.

D.37
22: Heat Exchangers

FINAL

N A Lind PD PR/D RISK RANK

6.4E-3 1.4E-8 3 0.10 0.90 3.OE-6 2

N: The high risk importance is due to the assumed loss of heat exchanger

function for a key heat exchanger (like CCW) leading to loss of one train of

nearly all safety systems, and loss of room cooling for one train of RHR, HPCI,

and RCIC; it was assume that one train of RHR was available.

A: Cracking, flaws, and leaks are failures that compromise the HX function.

Due to the 22 year doubling time of the failure rate, the aging factor was

considered to be on the low side.

Lind- Composite number that averages operational use with surveillance

tes ing.

P: Most tests do not directly measure the heat transfer capability or extent

o corrosion of the unit.

PR/D: Complete restoration would be proven through post-maintenance testing.

D.38
23: Compressor (Instr. Air)

FINAL

N A Lind PD PR/D RISK RANK

5.OE-4 5.OE-7 6 0.20 0.90 8.4E-6 3

N: These include service and instrument air compressors, but not diesel air

compressors. These were considered to have a medium risk importance because:

(a) older plants usually have a non-safety-related design and (b) the loss of

instrument air can cause many transients not provided for in system design.

A: Oil is the predominant failure mechanism for compressors and it is agingrelated due to seal failure so the ranking was seen as being relatively high.

NOTE: Compressors include dryers as part and parcel of the system and consider

the aging-related failure mechanism of rust.

Lind: 6 months. Complete functional checks.

PD: 20%--Aging principally detected through oil analysis and vibration

measurements. Aging usually found when correcting a failure.

PR/D: 90%--Assumes confirmation through post-maintenance testing.

D.39
24a: Chillers

FINAL

N A Lind PD PR/D RISK RANK

6.OE-4 1.5E-6 18 0.70 0.90 2.2E-5 3

N: Risk importance for fans and chillers were considered equivalent. These

essential equipment are considered as room coolers; they were felt to impact

containment integrity only when all essential equipment except for room coolers

has failed. Also, it was agreed that the risk importances should be lower than

those of heat exchangers and batteries and about equivalent to air compressors.

A: The aging rate was felt to be three-quarters of an order of magnitude

higher (worse) than that of the motor-driven pumps.

Lind: 18 months. Nominal major preventive maintenance frequency for this

component.

PD: 70%--Considerable aging evaluation of the standby unit is performed.

PR/D: 90%--Based on extensive post-maintenance surveillance testing.

D.40
24b: Fans

FINAL

N A Lind PO PR/D RISK RANK

6.OE-4 2.lE-7 18 0.70 0.90 3.1E-6 2

N: Risk importances for fans and chillers were considered equivalent. As room

coolers, they were felt to impact containment integrity only when all essential

equipment except the room coolers had failed.

A: The fans themselves age due to corrosion and the fan motors also experience

aging. This combined with a five-year failure rate doubling time contributed

to the aging rate determination.

Liqd: Based on surveillance requirements and failure indicators (no flow

in cation on effected piece of equipment).

PD: Aging leads to degraded performance which will be detected by surveillance

testing.

PR/D: Performance testing following repair.

D.41
25: Battery

FINAL

N A Lind PI) PR/D RISK RANK

2.OE-2 3.4E-7 6 0.90 0.90 1.1E-5 3

N: The consideration involved just the batteries themselves and it was felt

that the risk importance should be in the same range as the diesel generator (3

to 4) (because of similar roles in station blackout); so a high 3 rank was

agreed upon.

A: The panel agreed that batteries are susceptible to aging failure mechanisms

so the ranking had to be rather high. One panel member pointed out that the

doubling time would work out to a number of months if the rank was set to 5,

and that was considered unrealistic (as was the 29-year doubling time given as

the original data set value). It was thus agreed that a rank of 4 for aging

acceleration, corresponding to a doubling time of 3 years, was most

appropriate.

Lind: This is a composite based on frequent non-aging specific daily checks

coupled with longer intervals involving more in-depth testing.

PD: Recognizes the recent increases in testing for aging deterioration on this

component.

PR/D: Repair is usually by replacement, followed by testing.

D.42
26.a: Battery Charger

FINAL

N A Lind PD PR/D RISK RANK

1.1E-4 3.5E-8 12 0.90 0.90 2.6E-8 1

N: The batteries, chargers, inverters, and rectifiers were compared in risk

importance to aid in their ranking. Chargers, then, were considered less risk

significant than batteries and were left at the original ranking.

A: No impetus to change original ranking.

L nd: This figure is a weighted average of all monitoring incuding daily

checks through full surveillance load testing during refuel outages.

PD: Due to attention this device has received, aging detection is considered

to be very complete.

PR/D: High post-maintenance testing requirements ensure complete repair.

D.43
26b: Inverter

FINAL

N A Lind PD PR/D RISK RANK

4.7E-6 4.9E-6 12 0.50 0.90 5.OE-7 1

N: The importance to risk of the inverter was debated heavily between a

preference for a lower ranking based on the greater importance of the DC bus in

the ability to change equipment states, and the contention that the AC bus

provides plant status monitoring through instrumentation and is therefore

essential to evacuation planning. A decision was ultimately made to retain a

median ranking.

A: Age-related degradation is a factor due to the inside-the-cabinet

environment, and this justifies the high rank.

Lind: Same as battery chargers.

PD: Aging phenomena are difficult to detect, resulting in a medium detection

probability.

PR/D: Once detected, part replacement and post-maintenance testing provide an

excellent probability of repair.

D.44
26c: Rectifier

FINAL

N A Lind PD PR/D RISK RANK

4.7E-6 8.7E-8 12 0.50 0.90 8.9E-9 1

N: The rectifier risk importance was generally agreed as being the same as

that for inverters and was thus set at a median rank.

A: Considered to be about the same as that for the battery chargers.

Lind: Same as inverters.

PD: Same as inverters.

PR/D: Same as inverters.

D.45
27: Transformer

FINAL

N A Lind PD PR/D RISK RANK

1.2E-2 1.7E-9 18 0.70 0.90 5.1E-7 1

N: The 4160 volt and offsite power transformers were considered in the

ranking. The risk importance was seen to be high since a transformer loss

leads to multiple simultaneous equipment losses, similar to, but not as serious

as, loss of an AC bus (since backup power can usually be provided by diesels or

crosstie breakers).

A: Industrial experience shows that transformers have a constant failure rate

over many years of service, so age was not considered to be a significant

factor in transformer failure.

Lind: Gives some weight to ability of frequent (daily) checks to detect

trends, coupled with long term oil analysis.

PD: Environmental variations usually mask temperature variations as an aging

indicator. Oil analysis is most reliable.

PRD: Very high due to equipment importance and post-repair surveillance

requirements.

D.46
28: Fuel Storage Rack

FINAL

N A Lind PD PR/D RISK RANK

1.OE-6 5.OE-12 -- -- -- -- 1

N: Risk importance was considered low for the fuel rack, and at an equivalent

level to the turbine and snubber risk significance.

A: A passive component constantly immersed in water, the fuel racks were seen

to be affected by aging similarly to the RPV or the turbine.

Eliminated from further consideration by screening criteria (ARc 1E-7 CD/yr.)

D.47
29a: Medium Pressure Tank

FINAL

N A Lind PD PR/D RISK RANK

6.OE-3 1.OE-12 12 0.10 0.70 5.4E-9 1

N: These were seen as tanks in the 50 to 100 psi range and were felt to be

very important since they are generally surge tanks and boron injection

tanks. Losing one was considered equivalent to losing a train of an important

support system.

A: Leaks occurred but not failures; assumed they were designed to withstand

contents such as boron, so the aging rank is higher than for accumulators, but

not by much.

Lind: Based on Tech Spec visual inspection requirements.

PD: Low because of inspection methods.

Until you get a leak, nothing happens.

Lagging on some.

PR/D: Good based on post-repair testing.

D.48
29b: Atmospheric Pressure Tank

FINAL

N A Lind PD PR/D RISK RANK

2.5E-2 2.OE-10 12 0.10 0.70 2.2E-6 1

N: Considered important to risk since the best examples of these tanks is the

Refueling Water Storage Tank and the Condensate Storage Tank, which supplies

water to many safety-related systems.

A: Only pinhole leaks were anticipated over time, so aging failure is not a

predominant failure issue.

L nd: Reflects Tech Spec compliance.

PD: Only visual checks are performed on these tanks. Exterior lagging may

hinder identification of leakage.

PR6E: Once detected, repair of this and any additional leakage is highly

pD able

D.49
29.c: High Pressure Tank

FINAL

N A Lind PD PR/D RISK RANK

1.OE-4 1.OE-12 0 -- -- -- 1

N: The original category of accumulators/tanks was subdivided on the basis of

the pressurization levels. The accumulator/high pressure tank category was

seen to have a risk importance at a low 2 rank since a total functional failure

(depressurization or loss of valves between tank and vessel) of one tank would

still leave others available for backup.

A: Experience extends to nearly 200 reactor years with few problems; aging is

not believed to be a significant factor.

Eliminated from further consideration by the screening criterion

(ARc 1E-7 CD/yr).

D.50
30a: AC Bus

FINAL

N A Lind PD PR/D RISK RANK

4.3E-2 1.1E-9 18 0.50 0.90 2.3E-6 2

N: The AC bus was felt to have a high risk significance due to its support of

plant status instrumentation and to the fact that its failure could lead to

widespread Instrument and Control (I&C) disablement. Still, the risk

importance was considered half an order of magnitude lower than the DC bus.

A: The age factors were considered the same for AC and DC buses and are

discribed under the DC bus (30.6).

Lind: Critical buses can only be inspected in a de-energized condition, i.e.,

during shutdowns.

PD: Specific bus deterioration tests are not currently established.

PR/D: Repair is generally by replacement, and follow-up testing is assumed.

D.51
30b: DC Bus

FINAL

N A Lind PD PR/D RISK RANK

1.1E-1 1.1E-9 18 0.50 0.90 5.9E-6 2

N: The DC bus was defined as the wiring that goes into the distribution box

and breakers. Due to the effects of the bus on a multitude of control and

command units, the risk importance was seen to be high.

A: Equipment qualification testing showed bus failures due to high humidity,

steam, and chemical factors but these failures were the same whether the bus

was one or thirty years old, so these mechanisms aren't aging related. The

failure mechanisms which are age-dependent are dust buildup to connective paths

(causing shorts) and how well the terminal block screws retain the torque

level.

Lind: 18 months. This bus must be de-energized to perform surveillances,

hence they correspond to the refueling shutdown interval.

PD: 50%--Based on current detection methods.

PR/D: Post-maintenance surveillance.

D.52
31: Bolts

FINAL

N A Lind PD PR/D RISK RANK

1.OE-4 5.1E-7 18 0.50 0.90 2.5E-6 2

N: Bolts were considered important to risk due to their widespread existence

in bracing primary and secondary side components such as valve bonnets, pumps,

and steam generators. The potential for the "zipper effect" of bolt failure

was also cited as an issue. The risk significance was assumed equivalent to a

small break LOCA, since that seems to be the most likely risk-significant event

resulting from bolt failure.

A: The aging contribution to bolt degradation was considered significant due

to steam cutting and wastage (erosion from chemicals such as boric acid) and

rusting.

Lind: Assumes inspection on disassembled, safety-related, bolted components.

PD: Depends on perceptible crack detection. Detection of a problem is assumed

to result in extended surveillance of similar fasteners.

PR/D: Flaw detection would result in fastener replacement.

D.53
APPENDIX E

EVALUATION OF AGING RESEARCH BENEFITS TO GSI RESOLUTION
APPENDIX E

EVALUATION OF AGING RESEARCH BENEFITS TO GSI RESOLUTION

E.1
TABLE E.1. Example of Format for Presentation of Prescreened GSIs

GSI A-il: REACTOR VESSEL MATERIAL TOUGHNESS

DESCRIPTION:

Decrease in vessel fracture toughness with accumulated neutron irradiation reduces safety margins.

STATUS:

Resolved, but work is continuing.

SAFETY ISSUE SCHEDULED RESOLUTION DATE:

RESOLUTION OF ESSENTIAL COMPONENT AGING TECHNICAL SAFETY ISSUES:

-RPV

COMMENT:

Resolved in October 1982. More work Is needed in areas such as vessel annealing.

AGING RESEARCH IMPACT ON GSI: YES
TABLE E.2. Generic Safety Issues Scheduled for Resolution(a)

Issue

Number Title

m

.3

A.3,4,5 Steam generator tube

Integrity

A.17 Systems Interaction

A.40 Seismic design criteria

A.44 Station blackout

A.45 Shutdown decay heat

removal requirements

A.47 Safety Implications of

control systems

A.48 Hydrogen control measures

and effects of hydrogen

burns on safety equipment

23 Reactor coolant pump seal

failures

29 Bolting degradation or failures In nuclear power

plants

77 Flooding of safety equipment compartments by

back-flow through floor

drains

87 FaIlure of HPCI steam

line without Isolation

93 Steam binding of auxiliary feedwater pumps

94 Additional low-tenperature overpressure protection for light water

reactors

99 RCS/RHR suction line

Interlocks on PWR

Off ce/

Division/

Branch

NRR/DEST/EMTB

RES/DE/EIB

RES/DES/EIB

RES/DRPS/RPSI

RES/DRPS/RPSI

PrIority

UsI

USI

USI

UsI

USI

Draft

Resolution

04/83C

03/88

04/88

04/87C

10/87

01/88

12/87

01/89

ACRS

Review

10/83C

08/88

09/88

05/87C

12/87

CRGR

RevIew

10/83C

08/88

09/88

05/87C

02/88

Commission

RevIew

09/84C

N/A

N/A

11/87

04/88

N/A

Current

Resolution

Date

12/87( 

(04/89)

04/89

(03/88)

12/89

RES/DE/EIB USI 04/89

RES/DRAA/SAIB USI

07/88 07/88

01/88 N/A

05/89 05/89

03/89 05/89

01/88 01/88

RES/DE/EIB

RES/DE/EIB

RES/DE/EIB

RES/DRPS/RPS I

RES/DRPS/RPSI

RES/DRPS/RPSI

H

H 11/88

Prioritization

Date

01/78

01/78

01/79

01/79

12/80

12/80

12/80

04/83

08/82

N/A

N/A

12/88

N/A

(04/90)

10/89

H (To be Integrated Into A.17) 09/83

H 04/89 07/89 07/89

H (Complete except for Generic letter)

(12/90)

12/87

12/88

09/85

10/84

H 04/88 05/88 05/88 N/A 07/85

RES/DRPS/RPSI H 01/88 02/88 02/88 N/A 04/88 08/85
TABLE E.2. (contd)

Issue

Number

Office/

Division/

Title Branch

Draft ACRS CRGR Commission

Priority Resolution Review Review Review -

101 BWR water level

redundancy

105 Interfacing systems LOCA

at BWRs

RES/DE/EIB H 02/89

01/88

115 Enhancement of the reliability of Westinghouse

solid state protection

system

RES/DRPS/RPSI

RES/DRPS/RPSI

RES/DRA/ROB

H

12/89 12/89

04/88 04/88

02/89 02/89

N/A

N/A

N/A

Current

Resolution

Date

03/90

12/88

05/89

02/88

Prioritization

Date

05/85

06/85

07/86

09/85

H 11/88

121 Hydrogen control for

large, dry PWR

containment

H Recommendations on need for rulemaking

:.

122.2 Initiating Feed-and-Bleed

125.11.7 Reevaluate provision to

automatically isolate

feedwater from steam

generator during a line

break

128 Electrical power

reliability

113 Dynamic qualification of

large bore hydraulic

snubbers

130 Essential service water

pump failures at multiplant sites

134 Rule on degree and

experience requirements

for senior operators

B-56 Diesel reliability

C-8 Main steam line Isolation valve leakage

control systems

NRR/DEST/SRXB

RES/DRPS/RPSI H 10/88 03/89 03/89 N/A

RES/DE/EIB H 09/89 09/89 N/A

RES/DE/EIB H

12/88

08/89

02/89

H

05/90 06/91 N/A

08/89

12/89

09/91

09/89

09/89

12/88

02/89

09/86

11/86

03/85

03/87

07/86

11/83

11/83

RES/DRPS/RPSI

RES/DRA/RDB

RES/DRPS/RPSI

RES/DRPS/RPS I

H

H 10/88

05/89 06/89 N/A

04/89 04/89 07/87

06/88 06/88 N/A

04/88 05/88 N/A

H 04/88

H 11/87
TABLE E.2. (contd)

Issue

Number

I.A.4.2

(4)

I.F.1

I I.B.5

I I.B.5

(2)

Title

Review simulators for

conformance

Expand QA list

Behavior of severely

damaged fuel

Behavior of core-melt

Off Ice/

Division/

Branch

NRR/DLPQ/LOLB

PrIority

H

Draft ACRS

Resolution Review

03/85C 12/85C

CRGR

RevIew

02/86C

N/A

M

.1

II.C.4 Reliability engineering

II.E.4.3 (Containment) Integrity

check

II.H.2 Obtain technical data on

the conditions Inside

the TMI-2 containment

structure

HF1.l Shift staffing

HF4.1 Inspection procedure for

upgraded emergency

operating procedures

HF4.4 Guidelines for upgrading

other procedures

HF5.1 Local control stations

HF5.2 Review criteria for

human factors aspects

of advanced controls and

Instrumentation

NRR/DLPQ/LQAB H

RES/DRAA/AEB H

RES/DRAA/AEB H

RES/DRPS/RHFB H

RES/DRPS/RPSI H

RES/DRAA/AEB H

RES/DRPS/RHFB H

RES/DRPS/RHFB H

RES/DRPS/RHFB H

RES/DRPS/RHFB H

RES/DRPS/RHFB H

12/86C

05/86C

12/87

09/90

11/84C

11/87

N/A

Issue on hold(a)

12/86C A

Commission

RevIew

04/86C

SA

SA

N/A

N/A

N/A

N/A

N/A

A N/A

Current

Resolution

Date

10/87

(TED)

06/94

06/94

06/88

02/88

03/91

06/88

07/88 (b)

06/89

12/90

12/90

N/A N/A

Prioritizatlon

Date

11/83

11/83

11/83

11/83

11/83

11/83

11/83

10/84

10/84

10/84

10/84

10/84

08/86C 02/85C N/A

Ti 2515/79 Issued on 06/15/86

Transfer to RES not completed

07/89

07/89

10/89

10/89

02/90

02/90

08/90

08/90

HF8 Maintenance and surveillance program

51 Proposed requirements for

improving reliability of

open cycle service water

systems

NRR/DLPQ/LPEB H (Policy statement to be Issued) 01/88

03/90

10/84

RES/DE/EIB M 01/89 05/89 06/89 N/A 06/83
TABLE E.2. (contd)

Issue

Number

Of f ce/

Division/

Title Branch

Draft ACRS CRGR Commission

Priority Resolution Review Review Review

70 PORV and block valve

reliability

79 Unanalyzed reactor vessel thermal stress during

natural convection

cooldown

82 Beyond design bases

accidents In spent fuel

pools

RES/DE/EIB M

RES/DE/EIB M

02/88

02/88

RES/DRPS/RPSI 04/88

RES/DRPS/RPSI

M

A-29

B-5

m

rn

B-17

B-55

B-61

Nuclear power plant

design for the reduction

of vulnerability to

Industrial sabotage

Ductility of two way

slabs and shells and

buckling behavior of

steel containments

Criteria for safetyrelated operator

actions

Improve reliability of

target rock safety

relief vaIves

Allowable ECCS equipment

outage periods

RES/DE/EIB M

RES/DRPS/RHFB M

RES/DE/EIB M

RES/DRAA/PRAB M

RES/DE/MEB M

RES/DRPS/RHFB M

RES/DRPS/RHFB M

RES/DRAA/AEB M

03/88

12/88

09/88

06/88

02/88

06/90

M 06/89

05/88 06/88

N/A N/A

05/88 05/88

N/A N/A

08/88 08/88

03/89 03/89

01/89 N/A

09/88 09/88

11/88 10/88

12/90 09/90

07/88

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

N/A

Current

Resolution

Date

10/88

07/88

12/88

08/89

03/89

06/90

03/89

10/88

05/89

09/91

12/87

09/88

Prior it izat ion

Date

05/84

07/83

12/83

11/83

11/83

11/83

11/83

11/83

11/83

11/83

11/83

11/83

I.D.3 Safety system status

monitoring

l.D.4 Control room design

standard

l.D.5(5) disturbance analysis

systems

II.B.5 Effect of hydrogen burn-

(3) Ing and explosions on

containment structure

Compiete closecut memo to EDO

12/86C A N/A SA
TABLE E.2. (contd)

Issue

Number

Off ce/

Division/

Title Branch

II.E.6.1 Test adequacy study

II.F.5 Classification of

Instrumentation, control,

and electrical equipment

66 Steam generator

requirements

RES/DE/EIB

RES/

NRR/DEST/EMTB

Draft ACRS CRGR Commission

Priority Resolution Review Review Review

M 04/89 02/90 02/90 N/A

M Complete closeout memo to EDO

Current

Resolution

Date

05/90

01/88

Prioritization

Date

11/83

11/83

NR 04/83C 10/83C 10/83C 09/84C TBD(a) 11/83

mn

75 Generic Implications of

ATWS events at the Salem

nuclear plant

83 Control room habitability

84 CE PORVs

86 Long range plan for

dealing with stress

corrosion cracking In

BWR piping

91 Main crankshaft failures

In TDI diesel

102 Human error In events

Involving wrong unit or wrong train

103 Design for probable

maximum precipitation

124 Auxiliary feedwater

system reliability

B-64 Decommissioning of

nuclear reactors

l.D.5(3) On-line reactor

surveillance systems

RES/DRA/ARGIB NR 09/88

08/896

Deferred7

02/89 04/89 N/A

RES/DRAA/SAIB

NRR/DEST/SRXB

NRR/DEST/EMTB

NR

NR

NR

10/89 12/89 02/90 04/90

TBD

-- 08/87 09/87 10/87

01/90

NRR/DEST/EMTB

NRR/DLPQ/LPEB

RES/DE/EIB

Completed 09/87

10/83

06/84

02/85

10/84

09/85

02/85

09/85

02/86

11/83

11/83

Integrated Into policy statement3

NR 01/88 09/88 09/88

NRR/DEST/SRXB NR Schedule has plant specific

milestone dates

RES/DE/MEB NR 11/86 06/87 06/87

RES/DE/EMEB NR Complete closeout memo to EDO

N/A

12/87

TBD

10/88

02/88

01/89

06/89
TABLE E.2. (contd)

Office/ Current

Issue Division/ Draft ACRS CRGR Commission Resolution Prioritization

Number Title Branch Priority Resolution Review Review Review Date Date

II.J.4.1 Revise deficiency report AEOD/ NR __ 10/87 10/87 11/87 12/87 11/83

requirements

(a) Received 11/9/87 from R. C. Emrit NRC/RES/DEA.

(b) Resolution date may be reevaluated depending on outcome of North Anna steam generator tube rupture evaluation that Is

currently in progress.

A = Annual review; SA = Semi Annual review.

(c) On hold pending resolution of Important to Safety and Safety-Related Issues.

(d) Inspection procedures will be developed based on experience gained during performance of Ts.

(e) The staff Is considering dropping two unresolved Items and closing this Issue with no further action. Based on anticipated

transfer of lead responsibility from NRR to RES before 10/01/87. Pending Resolution of Generic Issue A-45.

(f) Issue date for final rule, final resolution schedule will be provided by NRR after transfer.

M

.o
APPENDIX F

NRC/NRR USER NEED LETTER
APPENDIX F

NRC/NRR USER NEED LETTER

F.1
0s UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON. D. C. 20555

APR9 19P7 ** 

MEMORANDUM FOR: Eric S. Beckjord, Director

Office of Nuclear Regulatory Research

FROM: Harold R. Denton, Director

Office of Nuclear Reactor Regulation

SUBJECT: USER NEED LETTER - NUCLEAR PLANT AGING RESEARCH PROGRAM

RES is currently formulating a major research effort to study the effects of

aging on nuclear poster plant structures, systems, and components. We are in

full agreement that aging effects are a major concern of this agency and we

erdorse the general objectives of your program.

My staff has been working closely with yours over the past year and we are

currently reviewing the details of your aging program plan. The purpose of

this memo is to provide you with some additional guidance regarding the scope

of your aging research programs to ensure that their usefulness to the

regulatory staff will be maximized.

The usefulness of any research result is in a large part measured by its

quantitative safety significance. For aging-related research to be of maximum

benefit to NRR, we must know not only the effects of aging on structures,

systems, and components, but also the risk significance to public health and

safety of the aging process in structures, systems, and components if aging is

allowed to proceed uncorrected. Knowledge of potential risk reductions due

to various corrective actions, such as maintenance and replacement, would be

extremely useful in assisting us to determine proposed regulatory actions.

The aging data should also include information which permits extrapolation of

the aging process and associated risk significance into a time frame

appropriate to the periods expected to be requested for license renewal. We

would like your aging program to include these additional elements in it so

that research results could be more easily and justifiably implemented into

the regulatory process. Our staffs should work closely together regarding the

detailed implementation of these suggestions.

I also ask you to consider implementing a program of establishing the risk

significance of results in other appropriate research areas.

rold R. Denton, Dir tor

0 fice of Nuclear Reactor Regulation

CONTACT: D. Cleary, EIB, DSRO

49-28547

F.2
EEICB, DET, RES PROGRAM FOR

NUCLEAR PLANT AGING RESEARCH/LIFE EXTENSION

Policy and Planning Guidance, Technical Integration, User Needs

J. Vora

EEICB, ET, RES

F.3
EEICB, DET, RES PROGRAM FOR

NUCLEAR PLANT AGING RESEARCh/LIFE EXTENSION

Policy and Planning Guidance, Technical Integration, User Needs

A. Excerpts from the Proposed 1986 Commission's PPG 1986

A.1. Aging

A.2. Plant Life Extension

A.3 Mothballing/Reactivation of Nuclear Power Plants

B. Technical Integration - "Aging"

Memorandum from the Office Directors to EDO 11/13//85

C. Closeout of Generic Issues B-58 and C-11

Memorandum from H. R. Denton to T. P. Speis 07/09/85

D. RES Assistance on Generic Issue No. 70 - "PORV and

Block Valve Reliability"

Memorandum from J. P. Knight to G. A. Arlotto 06/27/85

E. NRR Review of RES FY 1987 Internal Budget

Presentation to ACRS and BRG

Memorandum from H. R. Denton to R. B. Minogue 05/29/85

F. Relicensing of "Mothballed" Plants

F.l. Memorandum from J. R. Carter to B. M. Morris 04/02/85

F.2. Memorandum from D. Crutchfield to D. Eisenhut 01/24/85

G. NRR Input for the Long Range Research Plan (LRRP) FY 1986-1990

Ilemorandum from H. R. Denton to R. B. Minogue 12/06/84

F.4
Page 2 of

H. DHFS, NRR Maintenance and Surveillance Plan 10/10/84

1. Use of Signature Tracing Techniques to Detect Degradation or

Incorrect Adjustment of Safety Related Motor Operated Valves

Memorandum from H. R. Denton to R. B. Minogue 06/21/84

J. Draft Maintenance Program Plan: RES Resources

Memorandum from H. L. Thompson to G. A. Arlotto 05/24/84

K. Review Comments on RES' Program Plan on Nuclear Plant

Aging Research

K.1. Memorandum from J. N. Grace, E. L. Jordan to B. M. Morris 06/15/84

K.2. Memorandum from W. T. Russell to L. S. Rubenstein 04/11/84

K.3. Memorandum from K. V. Seyfrit to B. M. Morris 03/01/84

L. Staff Involvement in Shippingport Decommissioning

L.1 Memorandum from Chairman Palladino to EDO 11/29/83

L.2 Memorandum from EDO to Chairman Palladino 12/16/83

L.3 Memorandum from E. L. Jordan to G. A. Arlotto 04/17/84

L.4 Memorandum from D. Eisenhut to R. Mattson, T. Speis, 04/19/84

H. Thompson, R. Vollmer

M. Potential Areas of Research in Equipment Qualification

Memorandum from V. S. Noonan to W. F. Anderson 07/15/83

N. NRR's Comments on the Draft RES LRRP for FY 84-88

Memorandum from H. R. Denton to R. B. Minogue 03/25/82

0. ACRS Report to Congress (NUREG-0864) 02/:982

F.5
RES Contractors Currently Involved in the NPAR Program Address:

ORNL - B0828

PNL - B2865

INEL - A6389

BNL - A3270

A, B, C, D, F H I, J K, M, N, 0

A, B, F H, J K, L, N, 0

A, B, E, F, G H, J N, 0

Ar B F, H J K, N, 0

F.6
Plant Aging Research/Life Extension

Policy and Planning Guidance, Technical Integration, and Users Needs

A. Excerpts from the Proposed 1986 Comfnission's PPG

A.1 "The NRC will continue to seek to understand the effects of aging and

irradiation of materials and components in reactor containments.

- The staff will conduct research to identify measures which can be

taken to correct deficiences attributable to aging and irradiation

and to reduce risks inherent to degraded equipment."

A.2 "The Commission intends to begin development of the policies and

criteria to define requirements for operating license extensions to

help assure that industry's efforts in this area are focused on the

primary regulatory concerns.

- In view of industry initiatives to address plant life extensions,

the staff should propose policy guidance and develop licensing

criteria to define requirements for operating license extensions.

The staff should work with industry to ensure that key regulatory

issues are identified."

A.3 "The staff should propose policy guidance and develop procedures and

requirements for mothballing and for proceeding with such projects."

B. Technical Integration "Aging" (Ref. Memorandum from the Office Directors

to EDO dated November 13, 1985)

'(1) Aging - Potential involvement by all major offices. This effort

will study such time-related issues as the mechanisms of aging and degradation, methods of examination and testing to determine the condition of

components, and interpretation of results of these tests for appropriate

action. Work will provide the bases by which the staff can initiate requirements for test and examination of components. 'These assessments will

also provide bases for licensing decisions on whether operating plants

continue to meet health and safety requirements and, if needed, will facilitate relicensing to extend plant life beyond that originally anticipated."

F.7
C. Closeout of Generic Issues B-58 and C-1l (Ref. Memorandum from

H. R. Denton to T. P. Speis, dated 07/09/85)

"The scope of each issue proposes comprehensive, systematic studies with

the purpose of identifying significant sources of pump and valve

unreliability, and implementing maintenance, redesign or replacement

programs. A comprehensive systematic study of this type has not been

undertaken thus far in NRR. This type of program is more suitable for

Office of Nuclear Regulatory Research (RES) work. Nevertheless, there has

been a considerable amount of related pump and valve reliability work

performed to date.

Under the Nuclear Plant Aging Research (NPAR) program systematic studies

will be performed to (a) identify aging and service wear effects

associated with mechanical components that could impair plant safety, and

(2) identify techniques that will be effective in determining aging and

service wear effects prior to loss of safety function so that proper

maintenance and timely repair or replacement can be implemented. Although

this is a program intended to encompass many component types it is

envisioned to include various safety-related valve types and pump

components. RES intends to evaluate LWR operating experience and identify

aging trends. One of the specific benefits cited is improved reliability

and availability.

From our evaluation of the NPAR program we have concluded that it

generally encompasses the scope of the programs inferred by generic issues

B-58 and C-li. When this program is completed it is expected that

recommendations will be made by RES for maintenance, repair, or

replacement, according to component type. At that time these

recommendations will be grouped into manageable tasks and considered by

NRR staff for making changes to regulatory requirements.

CONCLUSIONS

A substantial amount of work related to pump and valve reliability has

already been done to reduce risk by identifying and resolving specific

F .8
issues. Concerns identified by AEOD related to this subject will continue

to be prioritized and worked as appropriate. The aging research program

includes systematic studies as called for in B-58 and C-1l and the results

will be used by NRR to make appropriate changes in regulatory

requirements. On this basis B-58 and C-1l will be closed out and further

resources related to these generic issues will only be used to monitor the

aging research program."

D. RES Assistance on Generic Issue No. 70- "PORV and Block Valve Reliability"

(Res: Memorandum from J. P. Knight to G. A. Arlotto, Dated 06/27/85).

"In recent discussions between DE and RES staff regarding the RES effort

to study component aging and service wear, we became aware of a current

effort at Oak Ridge National Lab (ORNL) which involves various types of

acquisition of operational data which might be of use to NRR in several

areas of study. One of the possible uses which we would have for such

data is in the resolution of Generic Issue No. 70 (GI-70). As shown in

the attached task action plan (TAP), the Mechanical Engineering Branch

(MEB) has the lead for completing subtasks 1.1 and 1.2 of GI-70 which

involves in part the acquisition of PWR pressurizer PORV and block valve

data. Among the items which we believe we could use RES/ORNL assistance

in completing these subtasks are in providing the following information:

(1) a complete list of the failures and failure rates of PORVs and block

valves, (2) component and associated control system failure causes, (3)

any other PORV or block valve failure event contributors, and (4)

subsequent corrective actions.

We have already discussed this with Mr. J. Vora in RES, NRC Project

Manager for the ORNL aging program. Based on Mr. Vora's discussion with

ORNL, we understand this work can be started June 1985 under the scope of

the current RES contract FIN B-0828. In order for us to be most able to

evaluate the information provided, we request that the data be categorized

in at least the following three ways: (1) by plant name, (2) by valve

vendor, and (3) by failure mechanism. In addition to LER information,

Nuclear Plant Reliability Data System (NPRDS) and Foreign Event File (FEF)

information should also be used. Because of recent specific comments

F.9
made by the ACRS Sub-Committee on Valve Reliability, ORNL should pay

particular attention to any reported PORV failures resulting from

incompatibility with boron in the primary coolant fluid.

Because of the short time period allowed for subtasks 1.1 and 1.2 in the

TAP, we request that this work be initiated as soon as possible."

E. NRR Review of RES FY 1987 Internal Budget Presentation to ACRS ar'. BRG

(Ref: Memorandum from H. R. Denton to R. B. Minogue, Dated 05/29/85)

"Systems Aging and Service Wear Effects," (A-6389). Furthermore, we would

like to see by the end of FY 1986 a summary report which presents the

results of this study of whether and where equipment aging is a safety

problem."

F. Relicensing of "Mothballed" Plants (Ref: Memorandum from J. R. Conti to

B. Morris, Dated 04/02/85; Memorandum from D. Crutchfield to D. Eisenhut,

Dated 01/24/85.)

"RES currently has a Nuclear Plant Aging Research Program in progress.

RES will be asked to identify any type of components or technical concerns

with respect to preservation/ maintenance. This will be used as one

aspect of whether a facility or stored equipment in a facility may be

reactivated and if so, what testing may be necessary."

"...identify aging concerns that would exist if a reactor facility had the

construction delayed or prolonged. Are there any precautions that must be

taken if NRC were to ultimately license the facility for operation?"

G. NRR Input for the Long Range Research Plan (LRRP) FY 1986-1990

(Ref: Memorandum from H. R. Denton to R. B. Minogue, Dated 12/06/84)

"We also request that RES initiate a program to develop guidance for

license renewal applications. As plants near the end of their operating

licenses, we expect to receive applications for operating license renewals

or extensions. At present, there is no guidance for making or reviewing

such applications. Our preliminary priority ranking is 'medium'."

F. 10
H. DHFS. NRR Maintenance and Surveillance Program Plan (Ref: Memorandum from

H. L. Thompson to R. B. Minogue, Dated 10/10/84)

Problem and Objectives include:

-- To detect the causes and effects of equipment degradation, and to

identify corrective action to minimize equipment failures and

unavailability.---"

Technical Issues include:

o Indicators of maintenance effectiveness

-- Measures of maintenance effectiveness may include indices of the

effects of aging.

o The role of preventive maintenance in counteracting aging and service

wear effects."

Related Projects include:

---"The nuclear plant aging research of RES/DET will provide information

related to the role of maintenance in counteracting the effects of plant

aging and make recommendations as appropriate for criteria and standards

development."

-3.3.6 Equipment Qualification

---"The equipment qualifications program requires information regarding

age-related performance degradation of equipment and components and

maintenance and surveillance activities which affect equipment integrity.

Results of the nuclear plant aging research will provide implications for

the environmental integrity of electrical and mechanical components and

equipment qualifications."

I. Use of Signature Tracing Techniques to Detect Degradation or Incorrect

Adjustment of Safety Related Motor Operated Valves (Ref: Memorandum from

H. R. Denton to R. B. Minogue, Dated 06/21/84)

F .11
"Bac kg round

Past NRC studies have identified numerous events in which motor operated

valves (MOVs) have failed to operated during required plant inservice

tests. A significant portion of the events are symptomatic of aging

mechanisms and associated with making incorrect valve adjustments. This

trend is continuing as is evidenced in current license event reports.

These types of valve malfunctions were highlighted in a report recently

published by Oak Ridge National Laboratory under an RES funded contract,

NUREG/CR-3543 "Survey of Operating Experience from LERs to identify Aging

Trends" dated January 1984.

Information Needs

We request that RES provide funding for the expeditious development and

implementation of a limited test program utilizing this new equipment.

The overall objective of the program should be to learn exactly what the

equipment can provide about safety related MOV operational readiness above

and beyond the currently used ASME Section XI methods.

We believe that this requested program would be a natural fit with the RES

current program "Operating Reactors Inspection, Maintenance and Repair,

Nuclear Plant Aging Research" (FIN B-0828). In particular the subtask at

ORNL covering detection of defects and degradation monitoring seems

appropriate. This request has been discussed with W. Morris and J. Vora

of the RES staff who are responsible for the referenced program.

We believe that, in addition to NRR, the results of this program will be

of interest to the Office of Inspection and Enforcement and Analysis and

Evaluation of Operational Data, and that they should also be kept informed

of the progress of the program.

Schedule

Because of the potential for near term improvement Iin safety-related valve

reliability through use of this type of test equipment, we request,

F.12
funding permitting, that a limited program be undertaken in FY 84

consistent with the medium priority categorization of generic issue

II.E.6.1 and scheduled resolution of the issue no later than the end

of FY 1986.

Other NRC Office Concurrence

This proposed program has been discussed with the Office of Inspection and

Enforcement, Division of Emergency Preparedness and Engineering Response,

and they have concurred in this memorandum."

J. Draft Maintenance Program Plan: RES Resources (Ref: Memorandum from

H. L. Thompson to G. A. Arlotto, Dated 05/24/84)

"Two RES projects have been identified in the Maintenance Program Plan

which appear to be most appropriately conducted as part of your N~uclear

Plant Aging Program (NPAP). These are:

3.4 Investigate Relationship between Plant Cycling and Maintenance

3.6 Establish Preventive Maintenance Requirements and Techniques

My staff and yours (J. Vora) have been coordinating these efforts. Please

advise as soon as possible whether it is your intent to cover the above

Maintenance Program Plan requirements within your existing Plant Aging

Porgram or whether additional resources are required."

K. Comments on Nuclear Plant Aging Research Program Plan

K.1 Memorandum from J. N. Grace, E. L. Jordan to B. M. Morris, Dated

06/15/84

K.2 Memorandum from W. T. Russell to L. S. Rubenstein, Dated 04/11/84

K.3 Memorandum from K. V. Seyfeit to B. M. Morris, Dated 03/01/84

Comments in the aforementioned memoranda were resolved - to the

satisfaction of the staff from the respective offices (NRR, IE, AEOD).

F.13
L. Staff Involvement in Shippingport Decommissioning (Ref: Memorandum from

Chairman Palladino to EDO, Dated 11/29/83)

L.1 "I am interested in knowing the extent to which the staff is or is

not participating in deciding what data we should be seeking from

Shippingport as part of its decommissioning. In particular, I would

like to know if each staff office (RES, NRR, I&E, etc.) is giving

overt attention to this data gathering opportunity."

L.2 (Ref: Memorandum from EDO to Chairman Palladino, Dated 12/16/83)

"Nuclear Plant Aging Research

Enclosures 1 and 2 represent an exchange of correspondence between

Mr. Minogue and Admiral McKee regarding the use of Shippingport for

conducting research. Mr. Minogue initiated this correspondence

because we have begun a research program to evalute the potential

impact on safety associated with aging, including wear, of commercial

nuclear plant equipment and structures and believed that Shippingport

could be fruitful in supplying important information. A meeting was

held on November 30, 1983 at Shippingport to begin the process of

selecting equipment which is sufficiently similar to LWR equipment

that examination or tests would be beneficial in meeting the

objectives of the aging research program. Our staff is staying in

close contact with Admiral McKee's staff as we move forward in

getting the needed technical information from Shippingport.

Decommissioning Research

RES has also initiated a research program to provide data to allow an

assessment of radiation exposure during decommissioning and of

implementation of ALARA techniques. Radioactive inventories,

exposures, dose rates, contamination levels, ALARA techniques and

results, waste shipment and disposal costs and public dose reports

will be obtained by observing decommissioning activities at various

facilities, including Shippingport. Current NRC activities related

F.14
to Shippingport include collection of engineering data and preliminary exposure estimated. These will be used for comparison to

data obtained during the actual decommissioning which is not expected

to begin until late 1984 or early 1985.

Although the other NRC offices are not currently conducting specific.

data gathering activities related to Shippingport, appropriate

technical staff from NRR, IE, and AEOD have participated in the

planning of the research activities discussed above and will be

participating in the review of the planning and implementation as the

research goes forward. This process should provide an adequate

mechanism to assure that the staff takes advantage of potential

opportunities, in addition to aging and decommissioning studies, for

obtaining data from Shippingport."

L.3 (Ref: Memorandum from E. L. Jordan to G. A. Arlotto, Dated 04/17/84)

"We urge that particular attention be directed toward procuring

metallurgical samples from the reactor coolant system. These samples

would have high value in bench marking metallurgical research now in

progress under the auspices of your materials engineering branch. We

have discussed this with Mr. Serpan, and he is enthusiastic.

We particularly urge that you obtain sizable samples from the thermal

shield and from the reactor vessel in the core belt line region. We

recommend that a specimen at least 12-inches in diameter be removed

by core drilling through both the thermal shield and the vessel wall.

A location should be chosen such that the intersection between a

girth and longitudinal weld is included. Sectioning of the specimen

would provide an unequalled opportunity to evaluate radiation damage

to the vessel all the way through the wall. We are advised that the

vessel material is the same as that used in licensed reactor vessels

now in service. We recognize that the reactor coolant system

operating temperature at Shippingport is slightly lower than the

typical commercial plant temperature, but we d not consider that to

be a significant impediment. The neutron damage rate in the

Shippingport vessel may be somewhat higher but that would only make

the results slightly conservative.

F.15
We also urge that samples be obtained from at least one loop of the

reactor coolant system. These should include the wrought stainless

steel pipe and cast stainless steel valve or pump bodies. We understand that some of the pumps may have been replaced since initial

construction. The chosen samples should be from original equipment,

since our area of interest relates to long-term effects.

Our primary concern at this time is assurance that the necessary

samples will be obtained while they are still available. We will be

glad to work with your staff in development of an appropriate program

and schedule for detailed analytical work."

L.4 (Ref: Memoeandumn from D. Eisenhut to R. Mattson, T. Speis,

H. Thompson, R. Vollmer, Dated 04/19/84)

"As you may know, the decommissioning of the Shippingport reactor

will commence in late 1984 or early 1985. This activity represents

an unusual opportunity for gathering useful data regarding light

water reactors. The Office of Research (RES) has initiated

discussions with the Navy regarding the use of Shippingport for

conducting research. RES expects to obtain data useful in both their

Nuclear Plant Aging Research Program and Decommissioning Research.

In order for NRR to participate and benefit from this data gathering

opportunity, we must communicate our needs and desires to RES.

Therefore, I request that each division review its technical program

(e.g., USI's, generic issues) and identify areas where information

and data from plant decommissioning activities may be useful."

M. Potential Areas of Research in Equipment Qualification (Ref: Memorandum

from V. S. Noonan to W. F. Anderson, Dated 07/15/83)

"In the Equipment Qualification Branch's continuing review of electrical

and mechanical equipment qualification used in operating plants and plants

receiving new operating licenses, several technical issues arise. The

attachment provides an outline of the areas of concern in connection with

seismic and environmental qualification, including pump and valve

operability.

F.16
B. Surveillance of Age-Related Degradation of Electrical Equipment

BACKGROUND INFORMATION

Equipment installed in operating plants has, in many cases, been qualified

to older standards and requirements not requiring the use of accelerated

aging prior to exposure to accident conditions. In addition, the

artificial aging utilized on newer equipment may not simulate the type and

extent of aging degradation experienced under actual plant conditions.

As a result, it is necessary to establish methods for monitoring the

condition of equipment located in a harsh environment to determine if

significant age-related degradation is occurring.

OBJECTIVE:

Identify practical methods for monitoring and measuring the age

degradation of electrical equipment in a harsh environment.

SCOPE

1. Identify equipment expected to experience a significant age-related

degradation and which is located in a harsh environment.

2. By use of experience data, previous testing and analysis, identify

the most promising methods for measuring degradation, including an

indication of precursors to failure modes.

3. Conduct testing to improving the understanding of the

phenomenological process to provide visibility to degradation.

4. Recommend guidelines and criteria for use by the licensing staff to

assess if the applicant's program is adequate to assure function is

not compromised during and after an accident."

F.17
N. NRR Comments on the Draft Long Range Research Plan for FY 84-88

(Ref: Memorandum from H. R. Denton to R. B. Minogue, Dated 03/25/82)

"We have reviewed the problem areas of concern to NRR that could benefit

from research early in the FY 84-88 time frame, and have identified the

following high priority areas:

Research into aging of plant structures, systems, and components,

including material degradation, valve behavior, flaw detection,

maintenance, and inservice inspection."

0. Report to Congress (NUREG-0864 Safety Research Program for FY 1983

3.4 Aging

As nuclear plant operating life advances, some degradation in its

equipment and systems must be expected. Pressure vessel embrittlement has

been recognized for many years to increase with accumulated fast neutron

fluence requiring adjustment of the operating temperature and pressure

limits to assure rupture resistance. The frequency of steam generator

tube leakage also increases with operating life. Organic coverings of

electrical and control cabling, elastomer seals on equipment closures, and

potting compounds for electrical connections are all known to deteriorate

under extended exposure to humidity and temperature conditions and low

level gamma and beta irradiation. Cyclic operating conditions affect the

performance reliability of valves, motor operators, ventilation machinery,

and comparable equipment. Emergency diesel power equipment and DC power

sources also are subject to degradation with operating use. The safety of

nuclear power plants depends on making certain that such degradation with

age is recognized and accommodated before it can cause significant

reduction in safety.

There is inadequate knowledge regarding the effectiveness of the surveillance programs related to aging and the effects of operating

transients on equipment that has deteriorated with age. As highlighted

by the recent concern for thermal shock on some older pressure vessels,

it appears important to apply some research effort to improve

F.18
understanding of the safety significance of loss of capability through such

aging. The RC should initiate a comprehensive, systematic investigation of

safety-related effects of aging for LWRs."

F.19
NUREG/CR-5248

PNL-6701

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NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I REPORT NUMBER Ass'gedbv 00C. odd Val No. da@,yj

42 841

NRCM 102 BIBLIOGRAPHIC DATA SHEET NUREG/CR-5248

SEE INSTRUCTIONS ON THE REVERSE PNL-6701 2 TITLE AND SUBTITLE 3 LEAVE BLANK

Prioritization of TIRGALEX-Recommended Components for

Further Aging Research 4 DATE REPORT COMPLETED

MONTH YEAR

5 AUTHORIS) June 1988 I. S Ly J. Wreathal , SAIC 6. DATE REPORT ISSUED D. S. Levy11 G. DeMoss, SAIC MONTH YEAR

D. Jarrell B. ~~A.WofrSII

E. P. Collins, SAIC Wolford, SAIC November 1988 7 PERFORMING ORGANIZATION NAME AND MAILING ADDRESS IIocNud.Zpo Cod} S. PROJECTITASKIWORK UNIT NUMBER

Pacific Northwest Laboratory

P .. 8oX 999 . IN OR GRANT NUMBER

Richand, WA 99 352 Science Applications International Corporation B2865 Richlad, WA99352 2941 Kenny Road, Suite 210B26

Columbus, OH 43221

10 SPONSORING ORGANIZATION NAME AND MAILING ADDRESS f(I,.d*Zio Code) Ila TYPE OF REPORT Division of Engineering

Office of Nuclear Regulatory Research Technical

U.S. Nuclear Regulatory Commission oPERIOD COVERED O/W/ dam)

Washington, DC 20555

12 SUPPLEMENTARY NOTES

13 ABSTRACT 200 words or !.#.i

The "Plan for Integration of Aging and Life Extension," developed by Technical Integration Review Group for Aging and Life Extension (TIRGALEX) in

May 1987, identified the safety-related nuclear power plant structures and

components (S/C) that should be prioritized for further evaluation by the

NRC's Nuclear Plant Aging Research Program (NPAR).

This report documents the results of an expert panel workshop established to perform the S/C prioritization activity. Prioritization was

primarily based upon criteria derived from a specially-developed risk-based

methodology. This methodology incorporates the effect upon plant risk of

both component aging and the effectiveness of current industry aging management practices in mitigating that aging.

An additional set of criteria used to categorize the S/C is the

importance of aging research on S/Cs to the resolution of generic safety issues (GSI) and/or to identify NRC/NRR user needs. The resultant S/C

categorization was to provide additional information to decision makers, but

was not used to calculate final S/C ranks.

14 DOCUMENT ANALYSIS - a KEYWORDSIOESCRIPTORS 15 AVAILABILITY

STATEMENT

nuclear plant aging, components, prioritization, risk methodology,

component aging, aging management practices, overhaul interval, Unlimited

detection, mitigation, inspection frequency 16 SECURITYCLASSIFICATION

b IDENTIFIERSIOPENENDED TERMS Unclassi fied

(I'TSo, r]Ot

Unclassified

17. NUMBER O PAGES

18 PRICE
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

SPECIAL FOURTH-CLASS RATE POSTAGE & FEES PAID

USNRC

PERMIT No. G-67