E. F. ,

NUREG/CR-3154

ORNL/TM-8647

The In-Plant Reliability Data

Base for Nuclear Plant Components:

Interim Report-The Valve -Component

Raymond J. Borkowski

W. Keith Kahl

Thomas L. Hebble

Joseph R. Fragola

James W. Johnson

I

.,pOR SAFEl A"D

JAN2 0 1984

APPLIED PHYSICS (208) LUt{RY

Prepared for the U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Division of Risk Analysis

Under Interagency Agreement DOE 40-550-75
- -

Printed in the United States of America. Available from

National Technical Information Service

U.S. Department of Commerce

5285 Port Royal Road, Springfield, Virginia 22161

Available from

GPO Sales Program

Division of Technical Information and Document Control

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555

This report was prepared as an account of work sponsored by an agency of the

United States Government. Neither the United StatesGovernment nor any agency

thereof, nor any of their employees, makes any warranty. express or implied. or

assumes any legal liability or responsibility for the accuracy, completeness, or

usefulness of any information, apparatus, product, or process disclosed, or

represents that i ts use would not infringe privately owned rights. Reference herein

to any specific commercial product, process, or service by trade name, trademark.

manufacturer, or otherwise, does not necessarily constitute or imply its

endorsement, recommendation, or favoring by the United States Government or

any agency thereof. The views and opinions of authors expressed herein do not

necessarily state or reflect those of the United States Government or any agency

thereof.
NUREG/CR-3 154

ORNL/TM-8647

Dist. Category HG

Engineering Technology Division

TEE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT

COMPONENTS: INTERIM REPORT - TWE VALVE -COMPONENT

Raymond S. Borkowski

W. Keith Kahl

Thomas L. Hebble

Oak Ridge National Laboratory

Joseph R. Fragola

Science Applications, Inc.

James W. Johnson

U.S. Nuclear Regulatory Commission

Manuscript Completed - October 21, 1983

Date Published - December 1983

NOTICE This document contains information of a preliminary nature.

It is subject to revision or correction and therefore does not represent a

final report.

Prepared for the

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Division of Risk Analysis

Under Interagency Agreement DOE 40-550-75

NRC FIN No. B0445

Prepared by the

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, Tennessee 37830

operated by

UNION CARBIDE CORPORATION

for the

U. S. DEPARTMENT OF ENERGY

under Contract No. W-7405-eng-26

J
iii

CONTENTS

Pase

LIST OF FIGURES . ...................................... * ........... v

LIST OF TABLES ******.***.. **........................ * ** ........... * * .......... . vii

FOREWORD . ............ *...*..*.**.**..*.*..**.**............ ................. ix

ACKNOWLEDGMENT'S .... * o * *..............*** * *........................ xi

EXECUTIVE SUMMARY ............ .... 0 .............. ............... xiii

ABSTRACT . ......... **.*... ................ * .................. *............ 1

1. INTRODUCTION ........................................... e ................ 1

1.1 Program Description and Objectives ... .................... 1

1.2 Program Scope ... . . 0.. .............. ............. . ..... 2

2. METHODOLOGY .................................. e.oe .... ................... 3

2.1 Valve Boundary ...... a........ ................................ 0......... 3

2.2 Failure Mode Code Development Approach ................... 6

2.3 Cause Code Development Approach ......... . ................. 10

2.4 Classification of Failure Severity ............ ............ 10

2.5 Application of Failure Modes and Cause Codes ............. 10

2.6 Output Format .......... ..... *........................................ 12

3. FAILURE RATE CALCULATIONS ......... . . . . . ...................................... . 15

3.1 Recommended Point Value Estimation ............. . .......... 15

3.2 Interval Estimation . ......... . . . .. ................................... 16

4. DISCUSSION OF RESULTS .......... . .. . ........................................ . 18

4.1 Comparison of IPRDS with WASH-1400 and LERs .............. 18

4.2 Safety vs Nonsafety-Related Systems ....................... 2i

4.3 BWR vs PWR Valve Maintenance and Reliability Statistics ............. ........................................ 21

4.4 Repair Times .................................................. 24

4.5 A Technique for Studying Maintenance Histories - Corrective Maintenance Signatures ........................... 28

4.6 Plant Specific Information ............................... 30

4.7 Safety Valves and Power-Operated Relief Valves ............ 30

5. DATA BASE LIMITATIONS AND RECOMMENDATIONS ..................... 31

5.1 Relatively Short Time Span and Limited Number of

Plants ....**....................*................................... 31

5.2 Differing Maintenance Policies Affect Component Failure Rates .............. ...... ........................... 31
i

iv

Pa a

5.3 Underestimation of the Number of Annual Demands ....... . .. 31

5.4 Plant 2 and 3 Records ............ s *............ *........... 31

5.5 Information Documented in the Maintenance Work Requests .. 32

REFERENCES ..................................................... ,.. 33

APPENDIX A: SUMMARY OF PLANT 1 DATA .............................. 35

APPENDIX B: SUMMARY OF PLANT 4 DATA .............................. 41

APPENDIX C: SAFETY AND POWER OPERATED RELIEF VALVES .............. 47
V

LIST OF FIGURES

FiRure PaRe

1 Preliminary Catastrophic Failure Statistics of Plant

1 for: Pneumatically Operated Valves by Valve Type .... 20

2 Cumulative Distribution Function of Observed Repair

Times for Valves in Plantl ............................ 24

3 Exponential Plot of Repair Times of Plant 1 ............ 25

4 Log-Normal Plot of Repair Times of Plant 1 ..... . 26

5 Weibull Probability Plot of Repair Times of Plant 1 .... 26

6 Corrective Maintenance Signature of a Steam Generator Dump Valve ... *......***.......................... 29

C-1 Pressurizer Safety and Relief Valves Arrangement ....... 50
vii

LIST OF TABLES

Table Page

1 Date base status (September 1983) ........................ 2

2 Generic systems list .-. *...................... .......... 4

3 Valve failure modes .................. *.-- 8

4 Valve cause codes 11......... 11

5 General format for reporting IPRDS valve population,

failure and repair statistics ... ......................... 13

6 Example of IPRDS valve population, failure and repair

statistics ............ 0..................... .0 .***... 14

7 Comparison of some preliminary PWR and BWR catastrophic

failure statistics with WASH-1400 and LERs ... o............ 19

8 Preliminary valve reliability and maintenance statistics for safety vs nonsafety-related systems .............. 22

9 A comparison of preliminary BWR and PWR selected valve

maintenance and reliability statistics for one major

catagory of systems - nuclear systems .................... 23

10 IPRDS and WASH-1400 parameters of the log-normal distribution of repair times ................... ... .............. 28

11 Maintenance frequency and median repair times by valve

type for Plant 1 ..... .................................... 29

A-1 Valve populations, demands, and service hours for

Plant 1 *..........********* .......-...-..-........... o... 38

A-2 Valve failures by mode and severity for each valve type

in Plant 1 o..oo..... o....*........o.....*...*.....*00.* 39

B-1 Valve populations, demands and service hours for Plant

4 ...... ........................ ............. o 44

B-2 Valve failures by mode and severity for each valve type

in Plant 4 *.... ...... *q*....**.*..**............. oe ... 4

C-1 Corrective maintenance actions of pressurizer valves in

Plant 1 ..... 00.........000..00... . ... 51

C-2 Corrective maintenance actions of pressurizer valves in

Plant 2 ofva...fai...... . h m ... ...................... 52

C-3 Summary of valve failure mechanisms .. ....... ........ ...-. ....... 53
ix

FOREWORD

During the preparation of WASH-1400 and subsequent to its publication, the nuclear community recognized the need for more comprehensive

sources of reliability data. In response to that need several efforts

were undertaken:

1. An American National Standards Institute (ANSI)/Edison Electric Institute (EEI)/Nuclear Regulatory Commission (NRC)-sponsored effort to

collect safety component failure information was organized under the

auspices of the then-designated N18-20 Committee of the American Nuclear Society (ANS). This effort was called the Nuclear Plant Reliability Data System (NPRDS).

2. An Institute of Electrical and Electronics Engineers Subcommittee 5

sponsored effort to collect electrical and electronic equipment failure rates led to the publication of TEEE Standard 500.

3. An NRC-sponsored program with BG&G Idaho was undertaken to supplement

the failure-frequency information contained in the LERs. Estimates

of population and exposure (time and demands) were made to permit

failure rate estimates on major plant components (pumps, valves,

diesels, etc.).

These efforts greatly expanded the base of available information

although none of the data were extracted directly from records existing in

the plants.

An effort was organized under the auspices of the ANSI/Failure and

Incidents Reports Review (FIRR) Data Subcommittee to contact individual

plant sites and arrange for visits by data collection teams to extract

data from in-plant maintenance records, and to attempt to construct a base

of reliability data from these collected records. Because of the magnitude of each plant effort, the scope was limited to a few sample plants.

The initial data extraction, data encoding, and data analysis effort was

directed at the components considered to be most significant (viz., pumps

and valves). This effort was named the In-Plant Reliability Data System

(IPRDS).
xi

ACKNOWLEDGMENTS

We appreciate the fine work contributed by the members of the Computer Sciences Division of ORNL; Elmon Leach and Janice Trent for their

patience in entering changes to the data base and the many requests for

computer searches and lists. We would like to thank Carol Mason, Erin

Collins and Mia Fienemann of Science Applications, Inc., for their valuable technical assistance in reviewing the maintenance records.

The Institute of Electrical and Electronics Engineers (IEEE) Office

of Standards provided assistance in data handling and storage, as secretariat to the American National Standards Institute/Failure and Incidents

Reports Review (ANSI-FIRR) Committee. The TIEE Subcommittee on Reliability (SC-5) provided technical assistance as well as coordination of the

data collection plant visits.
Xiii

EXECUTIVE SUMMARY

This document details the data collection and preliminary analyses

related to valves in the In-Plant Reliability Data System. The data base

is developed primarily from historical records of corrective maintenance

actions obtained directly from nuclear plant maintenance files. A comprehensive valve population is also included. The results in this report

represent the data from one PWR and one BWR power plant in the data base.

The purpose of this report is to demonstrate the degree of distinction and refinement in-the-reliability statistics that is possible with

tdata from the IPRD and to suggest a general format for disclosure of suitable reliability statistics to satisfy needs within the nuclear data community. The examples given in the various tables and figures are suggested methods of comparing valve data and are representative of the

degree to which reliability statistics for any particular valve can be

ascertained. The refinement of the summary data available from IPRD as to

the precise valve (i.e., valve type, valve size, and operating parameters)

is compared to the refinement found in WASH-1400 1 and from LER2.

One objective of this report is to examine the improvement possible

using IPRD in refining the statistics to ultimately focus on the reliability of specific valve types and operators in specific operating environments in the U.S. nuclear power plants. The second objective is to generate comments from members of the nuclear data community as to the efficacy

of the suggested formats for documenting valve information and the various

methods used for comparison in this report. These comments will be used

to improve the reporting in a valve data manual which will cover information from an expanded data base in the IPRDS. The results presented here

should be treated as preliminary, and therefore, only as examples of the

statistics that could be made available in a valve data manual from an

enlarged data base.

Failure rate calculations are shown graphically for selected valves

and results are compared to failure rate estimates in WASH-1400 and LERs.

Presented in this report are breakdowns of failure rates by failure modes

and by failure causes showing calculated maintenance frequencies and repair times. IPRDS Repair time distributions, unavailable from LERs, are

also presented and evaluated. A short study of safety relief valves is

presented in the appendix.

The major observation in this report is that the preliminary results

obtained from the pilot data base indicate WASH-1400 statistics may be

nonconservative for reliability estimates for some valve types in certain

failure modes. Conclusive results are not possible due to the size of

this pilot data base.
THE IN-PLANT RELIABILITY DATA BASE FOR NUCLEAR PLANT

COMPONENTS: INTERIM REPORT T TEE VALVE COMPONENT

ABSTRACT

This report on valves in the IPRDS documents the type of

reliability information that could be generated using the current

IPRDS methodology on an expanded data base. Preliminary results

and various methods for their documentation are presented as suggested methods for reporting results in a data manual. Comparison of preliminary results within a plant, between plants, and

among other data sources are made to exemplify some of the alternate uses of the IPRDS information that would be possible with an

expanded data base.

1. INTRODUCTION

1.1 ProRram Description-and Objectives

The objective of the In-Plant Reliability Data (IPRD) program is to

develop a comprehensive, component-specific data base for probabilistic

risk assessment and for other statistical analyses relevant to component

reliability evaluations. This objective is being attained through a

cooperative'effort with several utilities, wherein each utility provides

access to the maintenance files and pertinent population information, and

in return, receives computerized listings and tapes of their component

populations (equipment lists) and the component maintenance records. This

data base includes (1) a component population list for each'plant including electromechanical and mechanical equipment and (2) comprehensive component failure and repair histories including corrective maintenance actions on each component, i.e., pumps (including drivers), valves (including operators), diesel generators, inverters, battery chargers and batteries.

This pilot -study was undertakenu to'estimate' the reliability characteristics-of valves in-two nuclear6power generating stations,-_a PWR unit and

a BWR-unit. The data sources used to develop the data base and, therefore, the component failure rates and mean repair times are the plant

valve equipment lists, plant drawings, and the maintenance work requests

on these valves. The data were entered into a computer data management

system developed for this project. Background information on the development of this data system is reported in "The In-Plant Reliability Data

Base for Nuclear Power Plant Components: Data Collection and Methodology

Report," NURBG/CR-2641,' and "The In-Plant Reliability Data Base for

Nuclear Power Plant Components: Interim Data Report - The Pump Component," NURB3/CR-2886.4
2

1.2 ProRram Scope

Currently, the valve population, failure, and repair records from two

PWR units and four BWR units have been entered into the data base (24 reactor years of information). Table 1 gives a breakdown of the maintenance

records currently in the data base. Differences in plant-specific information are described in Appendices A and B.

This report examines the reliability characteristics. of valves in

both selected systems and entire plants.' A sample of statistics on valves

from one PWR (Plant 1) and one BWR are developed in this report to illustrate the degree of refinement possible when using the IPRD. Plant 2 data

was not included because of the short time span which the collected data

cover. Plant 3 data was not included because of significant incompatibilities between population and failure records.

Table 1. Data bass status (September 1983)

IPRDS Plant

PIR BYR Total VASH-1400

1a 2 3 4a

Number of maintenance 30,000 10,000 50,000 30,000 120,000 700

records collected

Number of corrective 8.000 3,000 6,000 7,000 24,000 303

maintenance records

Number of valve 3,067 980 992 773 5,812 102

maintenance records

Time span of valve main- 5.0 1.6 10.9 6.0 23.S 17

tenance records (years)

Number of valve popula- 3.138 3,310 16,799 1,578 24.825 NA

tion records

NA - Not available.

aIdentifies plant data used in this report.

*1
3

2. METHODOLOGY

The procedure used for establishing the data base and calculating the

component failure rate is as follows: From the plant equipment lists,

piping and instrument diagrams, and process flow diagrams, a population

card was formulated for each valve containing information such as the component identification number, system, valve type (gate, globe, check,

etc.), type of operator, process fluid, and'valve size in inches. System

codes were assigned from descriptive information derived from-'plant equipment lists and piping & instrument drawings (P&ID's). The system codes,

universal for all IPRD components, are designated in Table 2. In cases in

which not all of the above information was readily available from the

plant records, these data fields were left blank.

The failure rate estimate is calculated after determining the appropriate numerator (number of failures) and denominator (component hours or

demands) from the'data base. To determine the numerator of the failure

rate estimate, the analysts reviewed all the corrective maintenance records collected from the plant visit for valve related failures. These records were separated, reviewed again, and classified. Analyzing the failure and repair text, the analysts assigned the following codes: failure

cause(s), failure severity, and failure mode. The data reported on the

maintenance record such as component name, failure date, failure and repair text, as well as the-code assignments from the analyst were entered

into the computer. A' computer program then matched the individual failure

and repair'record with the population record on the basis of the component

identification number. A population record/failure and repair record set

was thus generated for each population record, containing the failure and

repair history of each component. The total number of failures for a particular valve of interest was used as the numerator for-the failure rate

estimate. Information was gathered to determine the denominator of the

failure rate estimate: the total number of service hours for a timedependent failure rate or the total number of demands for a demand failure

probability. For each valve IPRD analysts assigned the service hours

(calendar hours in the system) and an estimated number of demands (12

actuations per year) to each valve. No valve specific estimates of the

number of demands were attempted for this interim report.

2.1 Valve Boundarv

The approach used to define the boundary around the valve component

was to"consider the valve body -and all- of its-internal parts, the valve

operator-(motor,'-solenoid, pneumatic,'7etc.), and the-limit and torque

switches:mounted on-the valve or-needediby the operat6rjto make the valve

function.- Supply or auxiliary systems to the valve (e.g., electrical,

air, or hydraulic) are considered outside the bounds'of the components.

This approach is consistent with the method used by plant maintenance personnel to create a'valve-maintenance work request action; typically, by -

the failure of the valve to function as designed.

J
4

Table 2. IPRDS generic systems list

BWR PWR

Nuclear Systems--N

I

I

NO1

N02

N02.A

N03

N04

N05

N06

N07

Reactor

Control

Control

system

Reactor

Reactor

Standby

Reactor

Neutron

core

rod drive system

rod drive hydraulic

control system

recirculation system

liquid control system

protection system

monitoring/nuclear

NO1 Reactor core

N02 Control rod drive system

N03 Reactor control system

N04 Reactor coolant system

NOS Emergency boration system

N06 Reactor protection system

N07 Nuclear monitoring/nuclear

instrumentation system

N08 Residual heat removal/low

pressure safety injection

system

N09 Chemical and volume control

system (CVCS)

instrumentation system

N08 Residual heat removal/low

pressure safety injection

system

N09 Reactor water cleanup system

Engineered Safety Systems S

SOl Reactor core isolation cooling

system

S02 Engineered safety features actuation system

S03 Engineered safety features

S03.A High pressure coolant injection/core spray system

S03.C Low pressure coolant injection

S03.D Low pressure core spray system

S03.E Automatic depressurization

system

S04 Remote shutdown system

S03 Safety injection system

S03.A High pressure safety injection subsystem

S03.B Safety injection tank/core

flood subsystem

S03.C Low pressure safety injection

subsystem

S04 Remote shutdown system

SOS Auxiliary feedwater system

Containment Systems--C

COl Primary containment and penetrations

C02 Reactor building

C03 Containment heat removal

C04 Containment isolation system

COS Containment purge system

C06 Standby gas treatment system

C07 Combustible gas control system

C08 Containment ventilation system

C09 Reactor building ventilation

system

Clo Containment spray system

C02 Reactor building/containment

C03

C03 .A

C04

Cos

C07

C08

and penetrations

Containment cooling system

Ice condenser system

Containment isolation system

Containment purge system

Combustible gas control system

Containment ventilation system

Clo Containment spray system

Cll Penetration room ventilation

system
5

Table 2 (continued)

BWR and PIR

Electrical svstems-E

EO1 Main power system

EO1.A Protective relaying and controls

E02 Plant AC distribution system

E02.A Essential power system

E02.B Non-essential power system

E02.C HPCS power system -

E02.D Protective relaying and

controls -

E03 Instrumentation and control

power systems

E03.A DC power system

* vital DC power subsystem

* plant DC power subsystem

E03.B Instrument-AC power system

* vital instrument AC power

subsystem

* plant instrument AC power

subsystem

E04 Emergency power system

E04.A Diesel-generator fuel oil

subsystem

E04.B Diesel-generator cooling water

subsystem

E04.C Diesel-generator air subsystem

E04.D Diesel-generator lubrication

oil subsystem

E05 , Plant lighting system

EOS.A Essential lighting

£05.B Non-essential lighting

B06 Plant computer

--E07 Switchyard

E07.A DC control power system

E07.B Protective relaying

D____Prvo- 4, <:. + at ._D rote _- vr~o ;>1-m seer - K

POl Main. steam system -

P02 Turbine-generator system

P02.A Electio-hydraulic control'

subsystem

P02.B Turbine gland seal subsystem

P02.C Turbine lubrication subsystem

P02.D Stator (hydrogen) cooling -

subsystem

P02.E Hydrogen seal oil subsystem

P03 Turbine bypass system

P04 Condenser and condensate

system

P04.A Condenser evacuation system

P04.B Condensate cleanup/polishing

P04.C

.P05

P05.A

system

Condensate heater drain subsystem

Feedwater system

Feedwater heater drain sub6 ,system

.P06 Circulating water system

P07 I Steam generator blowdown

(PWR)

P08 Auxiliairy steam system - I

Process Auxiliary Systems-W

101 Radioactive waste system

WO1.A Gaseous radwaste system

* offgas subsystem (DIR)

WO1.8 Liquid radwaste system

WO1.C Solid radwaste system

W02 Radiation monitoring system

W02.A Plant area radiation monitors

W02.B Environmental radiation

monitors

W02.C Process radiation monitors.,

W03 Cooling water systemsV03.A Reactor building cooling

water system

103.B Turbine building cooling

water system

W04 Service water systems

W04.A Demineralized makeup water

system

W04.B Station service water system

* Essential service water

system

* Non-essential service

water system

W04.C Chilled water system

W05 Refueling system

W06. . Spent-fuel storage system

V06.A Fuel pool cooling and cleanW107-

F07.A'

107.B8

W08

W09

W09.A

W09.B

up system

Compressed air system

Service air system

Instrument air system

Process sampling system

Plant gas system

Nitrogen system

Hydrogen system
6

Table 2 (continued)

I

BWR and PWR

Plant Auxiliary Svstems X

XO1 Potable and sanitary water

system

X02 Fire protection system

X02.A Water system

X02.B Carbon dioxide system

X03 Communications system

X04 Security system

XOS Heating, ventilating, and

air conditioning systems

XOS.A Control room habitability

system

XOS.B Turbine building ventilation

system

X05.C Diesel building ventilation

system

X05.D Auxiliary building ventilation system

XOS.E Fuel building ventilation

system

X06 Non-radioactive waste system

X06.A Gaseous waste subsystem

X06.B Liquid waste subsystem

X06.C Solid waste subsystem

2.2 Failure-Mode Code Development Approach

V

The encoding efforts for the valve component have relied upon previous LER related work, 'specifically, coding schemes for cause codes and

failure modes. The systematic development of these codes for the IPRD

valve data base produces a more useful coded informational base. This is

especially true in regard to the performance of reliability and risk

analysis.

The selected failure modes encoded in the IPRD data represent-the

only intermediary link (i.e., the only link without resorting, to review of

the individual failure record text) between the fault tree analyst and the

data analyst.- For this reason it is imperative that the failure modes

selected are consistent with the needs of the most commonly utilized fault

tree basic events. Research and experience indicated that basic events

for components are usually categorized according to a component type designation combined with a failure mode which indicated:

1. Loss of function of the component, or

2. Change of state without command, or

3. Failure to change state when commanded.

The significant valve component types identified in the risk assessment

outputs and based upon experience were:

Valve type....

!2. *'1.

3.

4.

6.'

7.

Manual valves

Air operated valves

Motor operated valves

Solenoid valves

Check valves

Safety valves

Relief valves
7

When the generalized failure modes were applied to the specific case

of valves, the following valve specific modes were systematically produced

by generating exhaustive binary state transition failures as would be done

in fault tree construction and applying these to a generic valve.

Mode 11 Loss of function of the component:

a. Valve leaks through

b. Valve plugged

Mode 2. Change of state without command:

a. Valve closed - fails open

b. Valve open - fails closed

Mode 3. Failure to change state when commanded:

a. Valve open - fails open*

b. Valve closed - fails closed

After generating the modes, they were applied to specific valve types and

the developed modes were tailored to each specific type for the catastrophic failure category. This application caused the development of

valve specific mode terminology in many cases. For example, when 3a is

applied to safety valves it becomes: "Valve is open (due to a previous

legitimate command); it is commanded to close (i.e., to reclose due to reduced system pressure), but it fails open (i.e., does not close)." This

long description can be simplified and summarized by the statement:

"Fails to Reclose," and this statement is just 3a tailored to safety

valves.

When this tailoring was completed for all valve types, certain questionable specific modes were generated. For example, although modes 2 and

3 can be developed for check valves, they would only be useful if the correlation between a failure and an actual demand or the lack of a demand

can be made (e.g., if the failure records indicate, "inlet check valve on

pump A fails to open when pump A is activated"). This correlation is

highly unlikely, and since the important system failures are contained

within mode 1 (i.e., fails to check, and plugged), modes 2-and 3 were

judged to be unnecessary for simple check valves (swing check valves are

exceptions).'

After the elimination of questionable types for specific valves was

completed, the remaining types were expanded for the degraded and incipient. Again, conversion to type specific terminology was made where appropriate, and the elimination of questionable specific modes was carried

out. In-the next step, the surviving modes were divided into those primarily time related and those primarily demand related. Finally, the

valve types and the valve specific mode categorizations were reviewed to

determine if category similarities would allow grouping of types. This

was attempted in order to reduce the final number of categories without

sacrificing the required mode specialization. The results of this process

are given in Table 3. In Table 3, each unique mode was assigned a unique

alphabetic single digit identifier. These unique identifiers represent

the suggested failure modes and their suggested encoding scheme.

*i.e., Valve is open, it is commanded to close, but does not close.
8

Table 3. Valve failure modes

I. Manual. Operated Valves and Solenoid Valves

1 - Catastrophic

A. Fails to operate

a) normally open - fails open

b) normally closed - fails closed

B. Spurious operation

a) normally open - fails closed

b) normally closed - fails open

C. Plugged

D. Leaks through (disabling internal

leakage)

2 - Degraded

E. Improper operation (operates out of

specification)

F. Leaks through (debilitating internal

leakage)

I. Plugged (partial)

3 - Incipient

G. External leakage

H. Faulty indication

II. Check Valves

1 - Catastrophic

C. Plugged

D. Leaks through (disabling internal

leakage)

2 - Degraded

E. Improper operation (operates out

of specification)

F. Leaks through (debilitating

internal leakage)

I. Plugged (partial)

3 - Incipient

J. Chattering

G. External leakage

H. Faulty indication

Time/Demand Related

Demand

Time

Time

Time

Time

Time

Time

Time

Time

Time/Demand Related

Time

Time

Time

Time

Time

Time

Time

Time
9

Table 3 (continued)

0

III. Safety and Relief Valves

1 - Catastrophic

A. Fails to operate (significant

delayed operation)

B. Spurious operation

D. Leaks through (significant

internal leakage),

K. Fails to reclose

2 - Degraded

E. Improper operation

1. Premature operation

2. Delayed operation (operates

out of spec)

F. Leakage

3 - Incipient -

L. Small external leakage

H. Faulty indication

Time/Demand Related

Demand

Time

Time

Demand -

Time

Time

Time

Time

Failure Mode Su Larv

Mode

A. Fails to operate

B. Spurious operation

C. Plugged

D. Leaks through (significant internal

leakage)

E. Improper operation (operates out of

spec)

F. Leaks through (internal leakage)

G. External leakage

H. Faulty indication

I. Plugged (partial)

J. Chattering

K. Fails to reclose

L. Weepage

Time/Demand Related

Demand

Time

Time

Time

Time

Time

Time

Time

Time

Time

Demand

Time
10

2.3 Cause Code Development Approach

A systematic attempt was made to develop the cause encoding scheme

for valves. The thrust of the approach was to allow the maintenance record descriptions to specify the scheme. A sample (several hundred) of

representative failure and repair records were reviewed by the data analysts. The analysts were instructed in each case to extract the essential

cause description contained in each record. The analysts were trained to

key on certain cause descriptors such as piece part failures, control

failures, environmental failures, and installation failures. They were

instructed to construct new cause descriptions from the data only when the

essential cause of the description was not listed and was significantly

different from those listed.

The resulting cause categories were reviewed in an attempt to restructure them so as to reduce their number without significantly affecting their cause content. Cause codes which were clearly outliers, (i.e.,

appeared only once) were eliminated and the remaining codes were grouped

according to logical sets. Each of the codes within the sets were assigned unique, two digit, numerical identifiers. Blank entries were introduced between groups and also given identifiers. These blanks were

reserved for cause codes which might be uncovered by further analysis of

the data during the data encoding process. The suggested cause codes for

valves which resulted from this analysis are given in Table 4.

2.4 Classification of Failure Severity

The failure severity of the component was classified in one of the

following categories.

Catastrophic: The component is completely unable to perform its function.

Degraded: The component operates at less than its specified performance level.

Incipient;: The component performs within its'design envelope but exhibits characteristics that, if left unattended, will probably develop into a degraded or catastrophic failure.

2.5 Application of Failure Modes and Cause Codes

The use of the valve cause codes in Table 4 in many instances is

through a combination of two or more codes to specify both the part or

subassembly of the valve and the cause of the failure. Therefore,'code's

-14,through-41--in Table 4 identify valve parts-whose failure can bendescribed by. codes 53 through 60. For example, a binding or sticking valve

stem th'at>causes a sluggish valve stem movement would be assigned a "Degraded" failure severity, an E (improper operation) failure mode, with

cause codes 33 and 55 to specify the cause and type of failure. In other

cases, a failure may be described by assigning a single cause code with

the failure severity/mode. For example, a common external leak through

the valve packing can be encoded with an "Incipient" failure severity, a G
I.

Table 4. IPRD valve cause codes

C(F

V

w~

-",P

00 jAy, ';01

02

03

a' 04

A0S

06

/t ~~07

08

t4z 09

10

\,11

12

4y -_N \ ;, 13

'.14

15

\ 16

7

/18

19

20

21

2

13

*5

: 7

28

29

30

31

Unknown -

Design error

Personnel error

Fabrication error/construction

Procedural discrepancy

Blank

Blank

Blank

Leakage/general, unspecified

Leakage/air. gas. steam

Leakage/liquid, hydraulic fluid

Leakage/lubricant, oil. grease

Seals/gaskets, O-rings, lantern ring

Damaged seal surface

Coupling/shaft, reach rod, rocker arm, arm, universal

joints

Unions/connections, connecting pipe, elbows

Welds

Fasteners, bolts, nuts, set screws, bonnet bolts, lugs,

studs

Packing

Diaphragm

Cam

Solenoid

Motor

Actuator

Valve operator ';-

Gear/pinion, bevel gear, gear box

Gate

Flange

Bushing/bearing

Handwheel/handle

Disc/bellows rupture

Linksge

32

L- _ 33

; 34

35

36

,37

38

39

k* 40

41

42

43

44

45

46

47

48

49

50

51

) 52

54

155

~j56

57

.58

59

60

'61

,_62

Seat

Stem

Spring

Bonnet

Collar

Orifice

Nipple

Damper

Pins/shear pin, Cotter pin, retainer pins

Hoses/sample lines, sensing line. EH lines. air lines,

flush lines, copper tubing

Blank

Blank

Control circuit failure (electrical)/position Indicator.

relay. positioner, lights, contracts. accumulator, dead

band controls, alarm, loop controller, pilot valve

Fuse failure

Switch failure/microswitch

Limit switch failure

Tire/leads

Transducer/transformer

Faulty mechanical controls/regulator

Blank

Blank

Corrosion/erosion

Foreign material contamination/plugged

Binding/bound/selied/sticking

Cracked/pierced

Out of adjustment

Misaligned

Improper clearance

Trips on overload

Blank

Blank

I.-.

'-A

-1 if, 5 ..e- P I-e a

jl_> 
12

(external leakage) failure mode, and a single cause code at 18 to identify

the location of the leakage.

2.6 Output Format

The data format presented in Table 5 is intended to be the means for

documenting the valve reliability statistics of the IPRDS in a valve data

manual. The rationale behind the format development is to allow hierarchical aggregation of the basic, valve-specific statistics to yield the

more general valve statistics, e.g., the aggregation of basic statistics

from all the tables on globe valves would yield general reliability statistics for globe valves with all operator types, valves sizes and in all

process fluids. An example of the use of this format is given in Table

6. Terms shown in this format and other tables are defined as:

Annual demands:

Component class:

Failure cause:

Failure demand probability:

Failure mode:

Failure rate:

Failure severity:

Maintenance frequency:

Failure population

(Pop.)

Operating period:

Plant:

Plant type:

Population:

Population demands:

Population service

hours:

Primary class:

Service hours:

Subclass:

Average number of annual demands per valve

(estimated at 12 per valve/year for this

report).

Valve (includes operator)

The principal failure causes as found in

Table 4.

The probability determined according to

equations in Sect. 3.

The IPRD mode classification found in Table

3.

The rates calculated according to equations

in Sect. 3.

One of the three IPRD classes: catastrophic (D), degraded (D), incipient

(I).

The total number of failures divided by the

valve population divided by the population

service hours.

Total number of failures assigned to the

valves.

Years between commercialization and date of

last record collected from plant.

IPRD identification number.

BWR or PWR.

Number of valves.

Average annual demands per valve times

operating period (in years) times population.

The period of observation (in hours) times

population.

Valve operator.

Length of time covered by data multiplied

by the number of valves.

Hierarchical information including operator

type, system, type and size of valve, and

process fluid.

a1
-I I

Table S. General format for reporting IPRDS valve

population, failure and repair statistics

Plant type

Plant no.

Operating period yrs.

Population Information

Primary Class: Operator type

1st subclass: Valve type

2nd subclass: System

3rd subclass: Size

4th subclass: Process fluid

Component population:

Annual demands/valve:

Population demands:

Population service hours:

- - Maintenance frequency: -A .

wJ

Time-Related Failure and Repair Stat-stics

Failure Failure

mode severity

Failure

JDDUlti On Failures/l0 h Failure cause

Low recommended high Code (Failure Pop.)

Repair time (h)

low median high

Demand-Related FailurO and Repair Statistics

Failure

mode

Failure

severity

Failure

population Fallures/103 cycles

Low recommended high

Failure cause

Code (Failure Pop.)

Repair time (h)

low median high
Table 6. Example of IPRDS valve population, failure and repair statistics

Plant no. 1

Plant type PWR

Operating period S yrs.

Primary Class:

1st subclass:

2nd subclass:

3rd subclass:

Pneumatic

Globe

Condensate - P04

Water

Component population: 6

Annual demands/valve: 12

Population demands: 360

Population service hours: 2.62*10' h

Maintenance frequency: 1.26 E-4/h

4th subclass: 6 in. (total failures all severities - 33)

Time-Related Failure and Repair Statistics (modes B. C. D. E. F. G. H I)

Pailures/10' h

Failure Failure Failure

mode severity population Low Recommended High

Failure cause

Codes (Failure Pop.)

Repair time (h)

Lol Median EiJh

B (spurious

operation)

E (improper

operation)

F (internal

leakage)

G (external

leakage)

H (faulty

indication)

C 1

C 9

0.20

18

3.1

1.4

12

3.8

34

11

7.6

27

D 3

18 48 (1)

60 3. 57, 47(1); 57,

47(1); 57, 29(1);

14, 48(1); 44(1);

43(1); 24(1);

21(1); 17(1)

30 33, 32(1); 12(1);

10(1)

24 17, 41(1); 41(1)

50 3, 44, 48(1); 44,

58(1); 2, 48(1);

28, 29(1); 44(1);

59(1); 0(1)

3.0

0 2.5 25

I-'

0 40 52

I 2 0 25

I 7 0 2 6

Demand-Related Failure Statistics (mode A)

FailuresllO' cycles Failure cause

Failure Failure Failure

mode severity population Low

Repair time (h)

Recommended High Codes (Failure Pop.) Low Median HiRh

A (fails to

operate)

C 11 51 92 151 45(3); 47, 57(2);

45, 48(1); 3,

44(1); 17, 21(1);

48(1); 21(1);

0(1)

2 6 23

., J 4 . o
15

3. FAILURE RATE CALCULATIONS

3.1 Recommended Point Value Estimation

The equation used to estimate the probability of failure on demand

(Qd )is

n

Qd D

where

n = the number of failures observed and

D = the total number of demands experienced.

The equation used to estimate the failure rate (At, per hour) is

n

t T

where

n = the number of failures observed and

T = the total operating time of the components.

In the data tables these values of-Qd and At are listed under the

column labeled "mean." When no failures were observed (n = 0), the point

estimates Q and A in this column were determined using the median of a

chi-square variable with one degree of freedom

At x2o.s(1)/ 2 T

= 0.227/T

Qd X20.s0(l)/2D

= 0.227/D

For (D - n) < 40, the F-variate at the 50% point with one degree of freedom was used to calculate

F

n

d 2(D - n) + F + 1

n D

where Fn = Fn (1, 2D + 1).
I

16

3.2 Interval Estimation

The confidence limits for the hourly failure rates were calculated on

the assumption that the component times to failure are exponentially

distributed. Although for Q the number of failures n is binomially distributed, the Poisson distribution may be used to approximate the distribution of this variable when the number of failures is small compared

to the number of demands. The equations for estimating the 90% confidence

bounds on the failure rates when n > 0 and D - n > 40 are:

t ~2T

* X2,s (2n + 2)

5% =0

t 2T

X^.os (Zii)

Qd ' 2D , and

95% o.9s (On + 2)

d 2D

where

x~oss(2n) = the chi-square variate at the 0.05 level with 2n

degrees of freedom and

Xl~(2n+ 2) = the chi-square variate at the 0.95 level with

(2n + 2) degrees of freedom.

For the cases where D - n < 40, the Poisson approximation to the binomial distribution is not adequate, and the following equations are used

when n > 0:

nF

5* ~~~i

Ad D -n + 1d+gnei and

(n + 1) F

5% n=

d D-n + (n + 1) Fu
17

where

Fi = F 0. 0 5 (2n, 2D - 2n + 2)

which is the F variate at the 0.05 level with 2n and 2D - 2n + 2 degrees

of freedom, and

Fu = F0O (2n + 2. 2D- 2n) , ,

which is the F variate at the 0.95 level with 2n'+ 2'and 2D - 2n degrees

of freedom.

When n = 0, no estimates were made for the 5% 'values of A or- Q

The upper confidence level when n 0 was calculated using t d

At X0ss(2)/2T and

Qs= x20*,5 (2)12D
I

18

4. DISCUSSION OF RESULTS

INTRODUCTION

This is a limited presentation of selected preliminary results derived from the data of two nuclear power plants. This only touches upon

the numerous uses for the statistics and the various comparisons possible

with IPRDS results. The tables and figures should be viewed as proposed

methods or means for documenting future IPRDS information and are purposely noncomprehensive to inhibit direct use of the preliminary statistics.

IPRDS results are compared with the overall catastrophic failure statistics of WASH-1400 and LERs. In addition there is a comparison of valve

reliability and maintenance data for safety and nonsafety-related systems

for various specific valve operator types. Failure statistics from one

BWR plant (Plant 4) are contrasted with those of one PWR (Plant 1) for one

specific system category. Repair times from Plant 1 are presented on the

basis of three distributions. A cumulative distribution of repair times

is shown and compared to the WASH-1400 results. The parameters of the

lognormal repair time distribution-are given as well as the maintenance

frequencies and median repair times for valve types and valve operators.

4.1 Comparison of IPRDS with WASH-1400 and LERs

Upon initial review, the preliminary sampling of IPRDS results found

in Table 7 tends to indicate that differences exist with WASH-1400 and LER

values for the overall demand failure probabilities of valves. Although

this may be implied by the results, certain caveats should be considered

when evaluating this table, as well as other figures. First, the estimate

of individual valve demands is one demand per month, or twelve per year

for this report. This first order estimation is applied to all valves in

all systems of the plant, and may be significantly different than the

actual number of demands incurred by any particular valve. Also, the

tables shown are meant to represent the results that are possible from

analyses of data in the IPRDS and to present suggested formats for a computer-generated data manual. Finally, the general overall reliability

statistics on valves as documented in Table 7 may not be considered as

reasonable from an engineering standpoint. The reliability of valves can

be affected by- their operating and environmental conditions, and it likely

varies for different valve types. Thus, combining data from different

valve types (check, relief, gate, etc.) from all systems within the plant,

gives results as in Table 7 that are comparable to WASH-1400 values; but

based on engineering judgement, a more reasonable approach to valve failure data reporting would be Fig. 1.

Figure 1 graphically depicts a sample of preliminary reliability results for pneumatically operated valves by valve type from one PWR. Failure rates are depicted along with the population of valves and the number

of failures that were used to calculate the failure statistics. The

bounds of WASH-1400 estimates are shown by the dotted lines. The IPRDS
. V

Table 7. Comparison of some preliminary PWR and BWR catastrophic

failure statistics with WASH-1400 and LERs

for one mode of failure

IPEDS

Valve - WASH-1400 LERs Vpealve Failure mode pwOa BWRb

Operator Failure mode DIR Low Mean High Low Mean High

Low Mean High Low Mean High

Demand-related failure probabilities Qd (1/108 d)

Pneumatic Failed to operate 2.5 4.8 8.4 1.8 2.8 4.3 0.1 0.3 0.9 0.19 0.7 1.8

Solenoid Failed to operate 0.13 8.4 40 0.12 2.3 11 0.33 1.0 3

Motor-driven Failed to operate 3.5 6.4 11 1.9 3.7 6.4 0.33 1.0 3 3.6 4 4.4

Manual Failed to operate 0.15 0.42 0.88 0.39 0.61 0.90 0.02 0.08 0.21

'0

aIncludes

bIncludes

only valves in nuclear systems.

only valves in process auxiliary systems.
ORNL-DWG 83-5347 ETD

FAILURE MODE

10-1

A. FAILS TO

OPERATE

5

B. SPURIOUS C. PLUGGED

OPERATION

_---

10-2 1

10-3

MIDPOINT1 -4

1. BALL

2. BUTTERFLY

3. CHECK

4. DIAPHRAGM

5 7

G1l

5. GATE

6. GLOBE

7. DIRECTIONAL

D. INTERNAL

LEAKAGE

3

7 i ~~7-

24 I-1*10-5

2

4-

5

6

10-6

0

I -MIDPOINT

10-7

10-8

--- WASH-1400 BOUNDS

1 T 5 AND 95% CONFIDENCE BOUNDS

* MEAN

I I I I I --- I I I 1 I I 1 L-5 _I I I I I

POPULATION 42 26 5 364211811

FAILURES 4 0 0 0 25 28 0

I I I I I I _ . . . . . . . . . . . _

42 26 36 42 118 11

3 0 1 5 6 0

42265 364211811

0 00 0 0 0 0

4226 5 3642 118 11

3 1 5 0 1 2 0

Fig. 1. Preliminary Catastrophic Failure Statistics of Plant 1 for:

Pneumatically Operated Valves by Valve T~rpe.

.,I 1,




25

distributions so that the plotted points can be compared to a straight

line. These plots are given in Figs. 3-5. Of the three, the log normal

most closely resembles a straight line. Note that not all of the points

are plotted; only the 99 different repair times. The first 39 repair

times that are equal toO0.5 h are represented by the middle rank of 20 (=

39/2 + 1/2). The mode of 2 h (most frequently occurring repair time) is

represented by the middle rank of 598.5. At least one-half of the observed repair times are less than or equal to 4 h (median value).

Exponential Distribution (Fig, 3)

The density is given by

f(t) = Xe t for tŽ0

where parameter A is the failure rate. An

obtained from the mean time to failure by

A1 N

I = (mean time to failure) F

i=l

estimate of I, denoted i, is

t,: = 0.0767

i/

where

t. = ith repair time, and

N = total number of repair times = 2809.

ORNL-DWG 83-5349 ETD

9

8

7

6

5

4

3

2

60 120 180 240 300 360 420 480 540 600 660 720 780 840 91

LN REPAIR TIME (h)

Fig. 3. Exponential Plot of Repair Times of Plant 1.
26

ORNL-DWG 83-5350 ETD

4-U . . .._, I I I I I I I I l

A

A - 3.0 1-

2.0 I1.0 1-

'< 0

-1.0 Aa_

A -2.0__

XA

-3 .

0 I I I I I I I 1 1111 I

-0.5 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0 6.5 7.0

LN REPAIR TIME (hi

Fig. 4. Log-Normal Plot of Repair Times of Plant 1.

ORNL-DWG 83-5351 ETD

3-

2-

0-

-1

-2

-3

-= -4 -

- 5-

-6 -

-7 -

-8

-9

-10

I1.

I I I I I I I I I I I I I I I

- ~~~A&A&

-A&&

A

I I I I I I I I I I I I I I I

0 0 1 2 3 4

LN REPAIR TIME (hi

5 6 7

Fig. 5. Weibull Probability Plot of Repair Times of Plant 1.
27

Figure 3 plots the pairs

[in (N + 0.625) *

where

ri = rank of ith ordered repair time.

Log Normal Distribution (Figt. 4)

When the logarithm of a random variable has a normal distribution,

the random variable is distributed log normal. The density is

1 r (n t/A)3l

f(t) = a olq exp - 2a1 I ' ° 1

where the parameters P and a are measures

spectively. Estimates are given by

t < -l, i < o < -, as s > d0,

of the location and spread, reP = 3.91 h

A

a = 1.55.

Figure 4 plots the pairs

-1 __i __0.37S

G N + 0.25 / ln(ti)

where

(r 0..375 -1 i

G N + 0.25

(r - 0.37Sth

Vsth N + 0.25 percentile value from the

normal distribution

Weibull Distribution (Fig. 5)

The Weibull density for two parameters is given by:

f (t) = P(a) [eXip (a)] v a, p > 0, t I 0

with parameters a and P. No estimates are given.
28

Figure 5 plots the pairs

En 100] 1n' t

i i 100 - (r /N) 100 in (t)

The parameter estimates for the log normal distribution of repair

times and other statistics are given in Table 10. Of particular interest

is the range of valves (0.5 to 880 h) and the median for all valves (4 h).

Table 11 further breaks down the repair times and maintenance freqencies

by valve type for each operator. Again of interest is the range of median

repair times (2 to 10 h) with the majority of valve types requiring

between 4 and 6 h for repair.

Table 10. IPRDS and WASH-1400 parameters

of the log-normal distribution of

repair times

IPRDS WASH-1400

Number of observations 2809 28

Mean, h 5.2 24

Median, h 4.0 NA

Mode, h 2.0 NA

Standard deviation, h 3.2 NA

Maximum, h 880 350

Minimum, h 0.5 1

NA - not available.

NOTE: These preliminary and most general

IPRDS parameters have been determined

using all valve types, all failure severities and modes, and all valve sizes.

4.5 A Technique for Studving Maintenance Histories -

Corrective Maintenance Sianatures

The technique of corrective maintenance (C. M.) signatures is to

portray the entire corrective maintenance history of a particular component on a time line and graphically represent the failure and its corresponding severity as shown in Fig. 6. To complete the failure history,

the causes of each failure can be associated with the corresponding line.
29

Table 11. Maintenance frequency and median

repair times by valve type

for Plant 1

Valve type Operator Maintenance frequency Median type (No. of failures/10 h) repair time (h)

Ball All 7.36 8

- pneumatic 7.61 8

- others - 6.76 3

Butterfly All 9.64 4

- pneumatic 35.1 4

- motor-driven 28.2 4

- others 3.40 3

Check All 9.65 6

Diaphragm All 4.52 5

Gate All 17.6 6

- pneumatic 97.8 8

- motor-driven 62.0 4

- hand 28.5 4

- others 4.42 3

Globe All 17.2 4

- pneumatic - 43.1 4

- solenoid 182.0 4

- motor-driven 48.0 2

- hand 25.1 10

- others 4.19 2

Relief/Safety All 14.5 6

Directional Control All 14.6 3

- pneumatic 18.7 3

- solenoid 4.15 5

- motor-driven 68.2 2

- others 5.71 3

ORNL-DWG 83-5352 ETD

FAILURE SEVERITY:

C CATASTROPHIC

D DEGRADED

I INCIPIENT

C

D

1975 1976 1977 1978

YEAR

1979 1980 1981

Fig. 6. Corrective Maintenance Signature of a Steam Generator Dump

Valve.
30

This technique enables the relationships between time-variant factors

affecting component reliability (i.e., plant status, component environment, preventative maintenance) and component failures to be evaluated by

overlaying the C. M. signature with the history of the appropriate factor.

Changes in preventative maintenance policies and their effects on component reliability are particularly evident using this technique.

4.6 Plant Specific Information

The majority of plant specific information can be found in the Appendices A and B, representing Plants 1 and 4 respectively. The first table

(i.e., Table A-1 and B-1) in both appendices gives the valve population,

the estimated demands, and service hours for each type of valve and specific operator in the plant. Tables A-2 and B-2 provide background information that was used to develop the failure statistics for Plants 1 and 4,

respectively. They provide the number of failures for each mode by valve

type and operator type.

4.7 Safety Valve and Power-Operated Relief Valves

A ministudy on safety valves and power operated relief valves is included in Appendix C.
31

5. DATA BASE LIMITATION AND RECOMMENDATIONS

The data and calculated values in this pilot study must be considered

preliminary in nature and should be used only as screening values. The

calculated values may be subject to substantive changes as the data base

expands.

5.1 Relatively Short-Time -Span -and-Limited

- Number of -Plants

The IPRD system currently has valve population data from four nuclear

power plant stations (six units). Although the maintenance records from

these six units (four BWR and two PWRs) span almost 24 reactor-years of

commercial operation, the number of reactor-years of data from each unit

is relatively small (1.6 to 6 years). The failure rates and mean repair

times calculated in this report are from two of the six units and should

therefore be considered preliminary values. In many cases, the time span

of the data collected and the number of failures, most importantly catastrophic failures, were small.

It is recommended that (1) data from additional plants be collected

and (2) updating of the four plants currently in the data base continue.

5.2 Differina Maintenance -Policies-Affect

Component -Failure Rates

The differing maintenance policies of these two particular plants may

not reflect the overall population of nuclear power plants in the United

States. This could lead to plant-specific component failure rates and

maintenance frequencies which are not representative of the nuclear industry. Until data from additional plants are available, it should not

be assumed that these preliminary results are applicable to the general

population of nuclear valve components.

5.3 Underestimation of the -Number-of-Annual-Demands

It is recommended that for selected valves, the operator logs should

be reviewed to ascertain the actual number of demands.

5.4 Plant -2 and 3 Records

Plant 2 equipment lists were insufficient in documenting the valve

type and size. Additional information from the plant P&ID's is necessary

for developing the necessary hierarchical structure for each valve to

enable proper statistical analysis.
32

Plant 3 failure and repair records were extracted from the monthly

maintenance summary reports. As such, the component identification numbers were frequently omitted or recorded erroneously. This made matching

with Plant 3 population records a difficult task. Ultimately it became

evident that less than one third of the failure and repair records could

be matched. This did not yield a suitable sample to perform statistical

analysis upon. The original failure and repair records are necessary for

proper data base development.

5.5 Information Documented in-the Maintenance Work Requests

In reviewing and classifying the maintenance work request (MWR) records of the four nuclear stations, additional information on the MWR about

the components' failure mode, failure severity, and repair or unavailable

time would be helpful in using the failure and repair document for data

base development.
33

REFERENCES

1. U.S. Nuclear Regulatory Commission, Reactor Safety Study: An Assessment of Accident Risks in UJ.S. Commercial Nuclear Power Plants, WASH1400 (NURFG-75/014) (October 1975).

2. H. Hubble and C. F. Miller Data Summaries of Licensee Event Reports of

Valves at U.S. Commercial Nuclear Power Plants, NURBJ/4CR-1363, Vol. 1

(June 1980).

3. T. P. Drago et al., The In-Plant Reliability Data Base for Nuclear

Power Plant Components: Data Collection and Methodology Report,

NUREG/CR-2641.

4. J. P. Drago at al., The In-Plant Reliability Data Base for Nuclear

Plant Components: Interim Data Report - the Pump Component NUREG/CR2886.
35

APPENDIX A

SUMMARY OF PLANT 1 DATA
37

APPENDIX A (PLANT 1)

Plant 1

Population data. Equipment lists containing: component identification number, valve location, type and size of valve, operator type, the

number (population) of such valves, and the operating mode (normal valve

position). This information was available on 1051 of 3138 (33%) population records.

Maintenance work request data. The plant component and system summary cards of the individual work requests were the input to IPRD. Each

summary card contains the component identification number, an abbreviated

description of the failure, repair actions, repair time derived from the

original maintenance work request, dates of the failure and repair action,

and report number. Of 3078 total failure and repair records, 2942 matched

with 1347 population records.
38

Table A-1. Valve populations, demands, and service hours

for Plant 1

Valve type Operator Total Total service Maintenance frequency type Population demands hours (104 h) (No. of failures/106h)

Ball All 59 3,540 2.58 7.36

- pneumatic 42 2,520 1.84 7.61 - others* 17 1,020 0.740 6.76

Butterfly All 251 15,100 11.0 9.64 - pneumatic 26 1,560 1.14 35.1 - motor-driven 30 1.800 1.31 28.2 - others* 195 11,700 8.54 3.40

Check All 116 6,960 5.08 9.65

Diaphragm All 353 21,200 15.5 4.52

Gate All 752 45,100 32.9 17.6 - pneumatic 42 2,520 1.840 97.8

- solenoid 1 60 0.044 NC - motor-driven 95 5,700 4.16 62.0 - hand 20 1,200 0.876 28.5 - others* 594 35,600 26.0 4.42

Globe All 496 29,800 21.70 17.2 - pneumatic 118 7,080 5.17 43.1

- solenoid 2 120 0.088 182.0 - motor-driven 29 1,740 1.27 48.0 - hand 10 600 0.438 25.1 - others* 337 20,200 14.8 4.19

Needle All 1 60 0.044 NC

Plug All 53 3,180 2.326 NC

- pneumatic 2 120 0.088 NC - others* 51 3,060 2.23 NC

Safety/Relief All 131 7,860 5.74 14.5

Directional All 28 1,680 1.23 14.6

control - pneumatic 11 660 0.482 18.7 - solenoid 11 660 0.482 4.1S - motor-driven 2 120 0.088 68.2

- others' 4 240 0.17S S.71

*Indicates that valve equipment

jority of manually operated valves.

NC - not calculated.

list did not specify operator type. This includes the ma-
39

Table A-2. Valve failures by node and severity

for each valve type in Plant 1

Number of failures

Severity

Valve typOperator Valve type typC Catastrophic Degraded Incipient

Mode

A B C D K E F I G B 3 L

Ball

Butterfly

Chece

Diaphragm

Gate

Globe

All

- pneumatic

- othersa,

All

- pneumatic

- motor-driven

- othersa

All

wA1l

All

ci - pneumatic V[0'/ | motor-driven

- hand

- others*

All

- pneumatic

- solenoid

- motor-driven - hand

- others*

ety All

22

4

13

5

N/A

7

106

25

61

3

17

55

2g

2

16

2

7

6

5

3

2

N/A

2

16

5

10

1

9

6

1

2

.3 N/A *4

3 3

1

1 N/A *38

1 9

16

13

'5 N/A '4

1 .3 N/A 17

'2 N/A *155

1 62

58

8

1 27

-5 N/A 71

2 44

3

9

1

3 14

S/A N/A 4 '11

N/A 3

2

)I 13 )313

2

2

129

2

2

.7

20

.55

19

12

3

21

, 49

32

2

5

10

* 30

2

1

1

/-77

N/A 6

3

3

N/A *23

11

4

8

27

N/A r16

N/A '215

54

106

10

45

N/A t141

79

8

26

3

25

N/A N/A

N/A 11

4

. 4

3

1

.5

3

1

1

N/A

.4

-28

14

10

4

'43

32

2

6

3

2

2

2

N/A N/A - I D

6 /AA / 9

N/A N/A -7 1

N/A N/A If77 15t b

N/A N/A 1 c

W/A N/A 3 23

N/A 30 p3

N/A N/A g

Relief/Safi N/A N

Directional Control All

- pneumatic

- solenoid

- motor-driven 6

t1p1q STA 6'

Mode Codes: (A) - Fails to Operate

(B) - Spurious Operation

(C) - Plugged

(D) - Significant Internal Leakage

(K) - Fails to reclose/reseat

(E) - Improper Operation

(F) - Internal Leakage

(I) - Partial Plugging

(G) - External Leakage

(B) - Faulty Indication

(3) - Chattering

(L) - Weepage

*Indicates valve equipment list did not specify operator type. Includes majority of manually operated

valves.

N/A - Not applicable.
41

APPENDIX B

SUMMAY OF PLANT 4 DATA
43

APPENDIX B (PLANT 4)

Plant 4

Population data. An equipment list and plant manual containing:

component name and identification number, valve type and size, operator

type, and system. These data were available for 523 out of 1578 (33%)

population records.

Maintenance work request data. Copies of the original maintenance

records were obtained. Each record contains the component identification

number, failure description, repair action, dates of failure report and

repair, and report number. All 547 total failure and repair records were

matched with 263 population records.
44

Table B-1. Valve population. demands. and service hours

for Plant 4

Maintenance

Operator Total Total service frequency

Valve typo type Population demands hours (10' h) no. of failure\

10' h

Ball All 3 216 0.158 12.7

- pneumatic 3 216 0.158 12.7

Butterfly All 230 16,600 12.1 14.8

- pneumatic 98 7,060 5.15 29.1

- motor-driven 30 2,160 1.58 12.0

- chain 18 1,300 0.946 0

- others 84 6,050 4.42 2.26

Check All 194 14,000 10.2 3.43

- pneumatic 170 12,200 8.94 2.58

- others 24 1,730 1.26 2.38

Diaphragm All 79 5,690 4.15 0.962

- pneumatic 1 72 0.053 0

- solenoid 1 72 0.053 0

- others 77 5,S40 4.05 0.988

Gate All 547 39,400 28.8 5.82

- pneumatic 8 576 0.420 0

- solenoid 3 216 0.158 0

- motor-driven 93 6,700 4.89 26.2

- chain 1 72 0.053 0

- hand 8 576 0.420 9.52

- others 434 31,200 22.8 1.54

Globe All 118 8,500 6.20 8.23

- pneumatic 20 1,440 1.05 1.90

- motor-driven 17 1,220 0.894 39.1 - others 81 5,830 4.26 3.29

Plug All 18 1,300 0.946 8.47

- pneumatic 8 576 0.420 14.3

- hand 2 144 0.105 9.52

- others 8 576 0.420 2.38

Safety/Relief All 49 3,530 2.58 9.69

- pneumatic 41 2.950 2.15 6.94

- others* 8 576 0.420 23.8

Angle All 10 720 0.526 7.60

- Motor-driven 7 504 0.368 10.9

- others 3 216 0.158 0

*Indicates that valve equipment list did not specify operator type. This includes the

majority of manually operated valves.
45

Table B-2. Valve failures by mode and severity

for each valve type In Plant 4

Number of failures

Severity

Volve type Operator type Catastrophic Degraded Incipient

Mode

A B C D K E F I G B J L

Ball All

- pneumatic N/A 2 N/A N/A

Butterfly All 32 2 2 N/A 38 22 2 14 73 N/A N/A

- pneumatic 19 1 2 28 17 2 11 70

- motor-driven 9 1 4 2 1 2

- chain

- others* 4 2 1 2 1

Check All N/A NIA N/A 11 3 18 12 NIA

Diaphragm All NIA 3 N/A NIA

Gate All 12 7 N/A 26 10 95 17 N/A NIA

- pneumatic

- motor-driven 7 7 21 10 74 9

- hand

- others* 5 5 18 7

Globe All 2 NIA 2 4 40 3 NIA NIA

- pneumatic 1 1

- motor-driven 2 3 28 2 - otherse 1 1 11 1

Plug All N/A 1 7 N/A N/A

- pneumatic 6

- hand 1

- other 1

Relief/Safety All 1 N/A 1 7 5 1 NIA 1 N/A

- pneumatic 1 1 6 4 1 1

- other 1 1

Angle All 1 N/A 3 NIA NIA

-motor-driven 1 3 - others

Unknown All 50 15 57 10 6 54 73 3 2

- penumatic 3 8 2 3 15

- solenoid 1 1 1

- hand

- others 46 15 49 10 4 50 57 3 2

Node Codes: (A) - Fails to Operate

(B) - Spurious Operation

(C) - Plugged

(D) - Significant Internal Leakage

(K) - Fails to reclose/reseat

(E) - Improper Operation (G) - External Leakage

(F) - Internal Leakage (B) - Faulty Indication

(I) - Partial Plugging (1) - Chattering

CL) - Weepage

*Indicates valve equipment list did not specify operator type. Includes majority of manually operated

valves.

N/A - Not applicable.
47

APPENDIX C

SAFETY AND POWER OPERATED RELIEF VALVES
49

APPENDIX C

SAFETY VALVES AND POWER OPERATED RELIEF VALVES

1. Purpose and Scope

The purpose of this task was to review and categorize maintenance

records of the two PWR plants participating in the In-Plant Reliability

Data program for failures of the ASME code safety valves and the poweioperated relief valves (PORV) located at the pressurizer. Emphasis was on

the failure mode, "Failure to close given the valve is open." Summaries

of the failure and repair actions from the maintenance records are provided.

2. Observation-and Conclusions

* Because of the short time span for which failure data are available

(5 years of commercial operations for Plant 1 and 1.6 years for Plant

2) and the small population size, the conclusions drawn from reviewing the maintenance records should be considered preliminary.

* No failures of the ASME code pressurizer safety valves (PSV) either

to open on demand or to reclose were found.

* Most of the failures of'the PSVs, the power operated relief valves

(PORV) and the motor operated isolation valves (MOV) were external

leakage.

* No information was available from the maintenance records on the

total number of actual demands on any of the three types of valves

(PSV, PORV, and MOV) and therefore a failure rate for the failure

mode "Fails to reclose" for the PSVs was not calculated.

* The PORYs are operated to relieve-reactor coolant system pressure and

limit the undesirable opening of the spring-loaded safety valves.

Because of this design feature it is likely that the code'safety

valves have not been demanded to open during plant operation. Any

demands on the PSVs were most probably due to'the testing requirements of the ASME code. Because test interval is 5 years only one or

two demands on the PSVs in 6.6 years of commercial operation are

likely. This number of demands is insufficient to justify calculating a failure rate.

3. System Description

A simplified schematic of the pressurizer and the safety and relief

valves for the two PWR plants are shown in Fig. C-l.-

There are three pressurizer safety valves (PSV) at each plant. The

PSVs are totally enclosed pop-type valves. The valves are spring-loaded,

self-activated and with back-pressure compensation designed to prevent

system pressure from exceeding the design pressure by more than 110*, in
so

ORNL-DWG 83-5353 ETD

PLANT 1 PLANT 2

i99rP/-SV X 99rPV

P ORV MOV C PRESSURIZER PORV \ MOV PRESSURIZER

PRESSURIZER PRESSURIZER

RELIEF TANK RELIEF TANK

REACTOR STEAM REACTOR STEAM

GENERATOR GENERATOR

PORV = POWER OPERATED RELIEF VALVE

PSV = PRESSURE SAFETY VALVE

MOV = MOTOR OPERATED VALVE

Fig. C-1. Pressurizer Safety and Relief Valves Arrangement.

accordance with the ASME Boiler and Pressure Code, Sect. III. The set

pressure of the valves is 2485 psig.

The pressurizer is equipped with power-operated relief valves (PORV)

which limit system pressure for a large power mismatch and thus prevent

actuation of the fixed high-pressure reactor trip. The relief valves are

operated automatically or by remote manual control. The operation of

these valves also limits the undesirable opening of the spring-loaded

safety valves. Remotely motor operated stop valves (MOVs) are provided to

isolate the power-operated relief valves if excessive leakage occurs. The

11DVs are normally in the open position and the PORVs are normally in the

closed position.

The relief valves are designed to limit the pressurizer pressure to a

value below the high-pressure trip set-point for all design transients up

to and including the design percent step load decrease with steam dump but

without reactor trip. The set pressure of the PORVs is 2335 psig. Plant

1 has two parallel lines of PORVs; Plant 2 has three parallel lines. The

discharge ports of the PORVs and PSVs are routed to the pressurizer relief

tank.

4. Plant Data

The time frame of the data from Plant 1 is 5 years of commercial

operation; for Plant 2 is 1.6 years of commercial operation. Due to the

relatively short time span of the collected data and the small population

(2 plants), all conclusions drawn from reviewing these maintenance records

should be considered preliminary.

The corrective maintenance actions for the PSVs, PORVs, and MOVs for

Plants 1 and 2 are presented in Tables C-1 and C-2. A summary of the
I<

Table C-1. Corrective maintenance actions of pressurizer valves in Plant 1

(time frame of data:i 5 years of commercial operation)

Valve Failure description Repair description epi (h)

PSV-1 Leaks. (Failure occurred prior to commercialization date.) (No documentation.)

PSV-1 Leaks past seat. (Failure occurred prior to commercializa- Replaced gasket and lapped seat. 48

tion date.)

PSV-1 Safety valve appears to leak thrn seat. (Something under Lifted seat and reset. OK now. 45

seat.)

PSV-2 Possible leak past seat. Removed plug. 3

PSV-3 Remove rust. (No documentation.) 8

PORV-1 Valve leaks by (failure occurred prior to commercialization Replaced gasket and lapped seat. 30

date.)

PORV-1 Excessive leakage. Beveled and lapped seat - replaced 38

gasket.

PORV-1 Leaking., Polished both seats and replaced 40

gasket.

PORV-1 During test, cycled once but not twice. Installed gaskets and one screen 4

in regulator.

PORV-1 Regulators leak. Renew gaskets and gages. 4

PORV-1 Limit switches need adjustment. Adjusted limit switches. 4

PORV-1 Valve leaks through. Adjusted spring tension-cycled. 80

PORV-1 Leaks through. Loosened lock and adjusted valve. 8

PORV-1 Air leak in inlet to PORV nipple. Installed solenoid, tested. 8

PORV-1 (No documentation.) Changed diaphragm. 4

PORV-2 Leaks slightly. No leaks at normal pressure.

PORV-2 Leaks by. - Machined seat, straightened. 40

PORV-2 High temperature alarm indicating seat leakage. Replaced stem and flex gasket. 34

PORV-2 Limit switches requires setting. Adjusted limit switches. 4

PORV-2 Regulator leaks. Renewed gaskets and gages. 4

PORV-2 Stem plug and cage assembly removed during shut down. Machined stem plug face, and cage 12

seat. Lapped plug and seat.

PORV-2 Limit switches out of adjustment. Adjusted upper limit switch. 2.5

PORV-2 Valve leaks through. Inspected and repaired valve. 80

PORV-2. Diaphragm on operators. Leaking. Repair as instructed. 16

PORV-2 Air regulator for PORV. Replaced regulator. 4

MOV-1 Small body to bonnet leak. Retorqued and welded seal. 51

MOV-2 Small body to bonnet leak. Retorqued and seal welded leak. 42
Table C-2. Corrective maintenance actions of pressurizer valves in Plant 2

(time frame of data: 1.6 years of commerical operation)

Valve Failure description Repair description Rpi

PSY-1 (No failures reported.)

PSV-2 (No failures reported.)

PSV-3 (No failures reported.)

PORV-1 Valve opened for preoperations test crew, it did not Adjusted pre-load tension on 4

reset. Incorrect preload tension on valve spring. valve spring and functionally

(Failure occurred prior to commercialization date.) checked.

PORV-1 PORV-1, -2, -3 lift prematurely. (Failure occurred Found bad solenoid valve on PORV-3. 3

prior to commercialization date.) Replaced solenoid and calibrated.

PORV-1 Valve leaks thru. Seat and plug wire drawn. Installed new seat and lapped plug 20

to it. Now gaskets, repacked,

functionally checked.

PORV-2 Valve is leaking by. (Failure occurred prior to com- Valve not seated. Seat valve and 2

mercialization date.) stroked to insure properly seated.

PORV-2 Valve leaking by at normal pressure because disc is Deterioration from service. In- 12

ruined. stalled new stem and disc. Replaced seat ring gasket and bonnet gasket. Replaced packing.

PORV-3 Valve failed to open. Solenoid valve no good. Replaced 28

solenoid valve.

NOV-1 (Not documented.) Retorqued packing gland per pro- 12

cedure spec.

NOV-1 Packing leak. Natural end of packing life. Re- 16

packed valve.

NOV-1 (Not documented.) Valve was jammed shut as clearance 4

point.

NOV-1 Packing leak. Natural end of packing life. Re- 6

packed valve.

MOV-2 (Not documented.) Valve was jammed shut as clearance 4

point.

NOV-3 Valve wedge jammed in seat. Over torqued by motor operator Pulled bonnet and freed wedge. 52

and by hand to effect isolation for another job. Stem reassembled and repacked.

NOV-3 Won't open electrically. Broken terminal on switch. Broken terminal on benchboard 8

switch repaired.

tA

4 .. 4 ...
53

valve failure mechanisms is presented in Table C-3. Most of the valve

failures are seat leakage. No maintenance records for the PSVs "Failing

to close, given the valve is open" were found. A failure of a PORV to

reset was observed in Plant 2 (PORV-1). However, this failure should not

be considered as a random failure since it occurred in preoperation testing, that is prior to commercialization of the plant.

Table C-3. Summary of valve failure mechanisms

Plant 1 Plant 2

Valve type

PSV PORV 11O PSV PORV MO

Failure mechanism

Valve seat leakage 4 10 2 0 3 3

Limit switch 0 3 0 0 0 0

Air/regulator leak 0 4 0 0 0 0

Operator failure 0 2 0 0 0 0

Failed to reset 0 0 0 0 1 0

Lifted prematurely 0 0 0 0 1 0

Solenoid failure 0 0 0 0 1 0

Other 1 1 0 0 0 4

Total 5 20 2 0 6 7
55

NURBG/cR-3154

ORNL/TM-8647

Dist. Category RG

Internal Distribution

1-5. R. J. Borkowski 17. H. E. Knee

6. J. R. Buchanan 18. A. P. Malinauskas

7. T. J. Burns 19. G. A. Murphy

8. T. E. Cole 20. F. R. Mynatt

9. W. B. Cottrell 21. H. E. Trammell

10. D. J. Downing 22. D. B. Trauger

11. G. F. Flanagan 23. ORNL Patent Office

12. D. S. Griffith 24. Central Research Library

13. P. M. Haas 25. Document Reference Section

14. M. J. Haire 26-27. Laboratory Records Department

15. W. 0. Harms 28. Laboratory Records, RC

16. J. Jones, Jr.

External Distribution

29. Shahid Ahmed, Babcock & Wilcox, P.O. Box 1260, Old Forest Road,

Lynchburg, VA

30. F. Balkovitz, BG&G Idaho, Inc., P.O. Box 1625, Idaho Falls, ID

83415

31. R. Bari, Brookhaven National Laboratory, Dept. of Nuclear

Energy, Upton, NY 11973

32. J. P. Drago, Connecticut Yankee Atomic Power, Co., RR #1, Box

127E, E. Hampton, cr 06424

33. Ronald Feit, NRC, MS 5650, Washington, DC 20555

34. J. R. Gracia, CRBR Project Office, Oak Ridge, 7N 37830

35. J. A. Hartung, Rockwell International, 8900 DeSota Ave., Canoga

Park, CA 91304

36-101. J. W. Johnson, Risk Methodology and Data Branch, Division of

Risk Analysis, Office of Nuclear Regulatory Research, U.S.

Nuclear Regulatory Commission, NL-5650, Washington, DC'20555

102. J. Minarick, SAI, 800 Oak Ridge Turnpike, Oak Ridge, TN 37830

103. J. Penland, SAI, 800 Oak Ridge Turnpike, Oak Ridge, TN 37830

104. Samuel D. Reynolds, Jr., NRC, 631 Park Ave., King of Prussia, PA

19406

105. W. Sullivan, EG&G Idaho, Inc., P.O. Box 1625, Idaho Falls, ID

83415

106. Office of Assistant Manager for Energy Research and Development,

Department of Energy, ORO, Oak Ridge, TN 37830

107-108. Technical Information Center, DOE, Oak Ridge, TN 37830

109-357. Given distribution as shown in Category 1G (10-NTIS)
NRC FoRM 335 1. REPORT NUMeER (Assignedby DDC) NRC Fll U.S. NUCLEAR REGULATORY COMMISSION NUREG/CR-3154

BIBLIOGRAPHIC DATA SHEET ORNL/TM-8647

4. TITLE AND SUBTITLE (Add Volume No.. ifappropriate) 2. (Leave blank)

The In-Plant Reliability Data Base for Nuclear Plant

Components: Interim Data Report - The Valve Component 3. RECIPIENT'S ACCESSION NO.

7. AUTHOR(S) Raymond J. Borkowski, W. Keith Kahl, 5. DATE REPORT COMPLETED

Thomas L. Hebble, Joseph R. Fragola, James W. Johnson MONTH September YEAR 1983

9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) DATE REPORT ISSUED

Oak Ridge National Laboratory MONTH |YEAR

P.O. Box X

Oak Ridge, TN 37831 6. [Leave blankl

8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code) 10. PROJECT/TASK/WORK UNIT NO.

Division of Risk Analysis

Office of Research 11. FIN NO.

U.S. Nuclear Regulatory Commission B0445

Washington, DC 20555

13. TYPE OF REPORT PERIOD COVERED (Inclusive dares)

Topical

15. SUPPLEMENTARY NOTES 14. (Leave olank)

16. ABSTRACT (200 words or less)

This report documents the methodology and some of the preliminary results attained

from the valve data in the In-Plant Reliability Data System. The methodology used

for data base development and failure rate calculation is described along with the

present limitation of and recommendation for data system development.

A sampling of preliminary results is presented to exemplify the potential usefulness of an expanded data system.

17. KEY WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS

17h. IDENTIFIERS'OPEN-ENDED TERMS

18 AVAILABILITY STATEMENT 19. SECURITY CLASS (rh's report 2. NO. OF PAGES

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