1- 

NUREG /CR-4302

Volume 1

ORNL-6193/V1

OAK RIDGE

NATIONAL

LABORATORY

P.. !,--A F 

Aging and Service Wear of Check

Valves Used in Engineered

Safety-Feature Systems

of Nuclear Power Plants

Volume 1. Operating Experience and

Failure Identification

W. L. Greenstreet

G. A. Murphy

R. B. Gallaher

D. M. Eissenberg

Prepared for the U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Under Interagency Agreement DOE 40-551-75

OPERATED BY

MARTIN MARIETTA*ENERGY SYSTEMS, INC.

FOR THE UNITED STATES

DEPARTMENT OF ENERGY
NUREG/CR-4302

Volume 1

ORNL-6193/V1

Dist. Category RV

AGING AND SERVICE WEAR OF CHECK VALVES USED

IN ENGINEERED SAFETY-FEATURE SYSTEMS

OF NUCLEAR POWER PLANTS

Volume 1. Operating Experience and

Failure Identification

W. L. Greenstreet

G. A. Murphy

R. B. Gallaher

D. M. Eissenberg

Manuscript Completed - October 25, 1985

Date Published - December 1985

Prepared for the

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

under Interagency Agreement DOE-40-551-75

NRC FIN No. B0828

Prepared by the

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, Tennessee 37831

operated by

MARTIN MARIETTA ENERGY SYSTEMS, INC.

for the

U.S. DEPARTMENT OF ENERGY

under Contract No. DE-AC05-840R21400
- 1

*iii

CONTENTS

Page

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

LIST OF TABLES i............x.. .......... oo ...o*........ .. ii

SUMEMARY O'.00.00.00.00.00..00.00 00O....0.0..00000.000. xi

ABSTRACT Ojcie.. ....... .. ......... 2 I

'1. INTRODUCTION ope ............................... ....... 1

1.1 Background I

1.2 -Objective oooo.ooooooooo.o. o.oooo. 2

1.3 Project Scope oo.0.0000.000.. 0...0000 2

1.4 Definitions .* ......... ... ***...**...*.*....**.******* 3

2. BASIC INFORMATION ***..*...*..... .. .... 00.0...... 4

2.1 Principal Types and Uses of CVs in BWWRs

and' PWRs ............ oooo...oo..oooo.o.......... 4

2.2 CV Types .... 5...................................... 5

2.3 Equipmen: Boundaries 8o.o.o ... ....... ... .o. 8

2.4 Functional. Requirements * ** * ** * * oo ... **o.*****. 8

2.5 Materials of Construction ........................ 9

3. TECHNICAL SPECIFICATION REQUIREMENTS ............ ....... 11

4. SUMMARY OF OPERATIONAL STRESSORS ........ o..o.ooo..*o...o 12

4.1 Electrical Stressors 14

4.2 Mechanical Stressors ... ooo.. .o........ .. o.......o 14

4.3 Thermal Stressors .* ... . .o.00 15

4.4 Chemical Stressors o.*...o 999 oo9ooo..9o9e9osooo9o9.. 16

4.5 Radiation Stressors ... **.eooo.*o..o .oo......... 17

4.6 Environmental StressorE .. o.oo.o .. *.0........ 0 18

5. OPERATING EXPERIENCE .... ......... ** .... *o.** 19

5.1 Summary of Failure Modes and Causes .................. 19

5.2 'Frequency of Failures 21

5.3 Methods of Detection ........... o..o..... o.... o. .oo' 21

5.4 Maintenance Actions 22

5.5 Modifications Resulting from Failures ............... 22

6. MANUFACTURER INPUT . 9 9 .99 9 9... 23

6.1 Failure Modes and Causes .. e.oo.o.o.o.o.. o..o.o....... 23

6.2 Failure Cause Analysis 23
iv

7. MANUFACTURER-RECOMMENDED SURVEILLANCE AND

MAINTENANCE PRACTICES ................. .......................

8. AGING AND SERVICE WEAR MONITORING ............................

8.1 Failure ModeandCause.Determination ...................

8.2 Measurable Parameters for Establishing Degradation

Trends*....... ......... ee...... 00....... 00.... ...0 

9. SUMMARY AND RECOMMENDATIONS ...... - ......................

REFERENCES ....... .... .. .....ER .A....R ............

APPENDIX A SUMMARY OF ASME BOILER AND PRESSURE VESSEL CODE

27

29

29

34

39

41

43

49

SECT. XI REQUIREMENTS FOR CHECK VALVES ..............

APPENDIX B. OPERATING EXPERIENCE DATA BASES AND REPORTS .........
v

ACKNOWLEDGMENTS

The authors gratefully acknowledge the continuing support and counsel of the NRC Nuclear Plant Aging Research Program Manager, J. P. Vora,

in the planning and implementing of this study.

The preparation of this report required the help of a number of individuals, and the authors acknowledge with gratitude the support given.

J. L. Hawley, Walthworth Company; E. J. Majewski, Rockwell International,

Flow Control Division; and R. Brennan, Atwood and Morrill Company, Inc.,

were very helpful in providing needed information as well as answering

questions in the initial phases.

Figures 2.1 through 2.4 were published by permission from:

Atwood and Merrill Company, Inc. (Fig. 2.1)

Aloyco, Inc. (Fig. 2.2)

Rockwell International, Flow Control Division (Fig. 2.3)

Jenkins Brothers (Fig. 2.4)
vii

LIST OF FIGURES

Figure Page

2.1 Swing CV *..........................*..*****..*. .. *.... 6

2.2 Swing CV, exploded view *................................. 6

2.3 Horizontal-piston-lift CV ..... .-. . . ... . . ... .o..... 7

2.4 Vertical lift CV ... *..00***..0 ...... **e****...... O... 7

2.5 Ball CV, showing ball and seat arrangement *o ............ 8

4.1 Exposure profile under accident conditions (IEEE 382)..... 12

9.1 NPAR Program strategy ......... o... *...... **....... 40
ix

LIST OF TABLES

Table Page

2.1 Summary of CV applications in nuclear power plants ...... 4

4.1 Equipment qualification tests for valve actuators ....... 13

5.1 Summary of CV failure information available from

operating experience and plant documents ............... 20

6.1 Check valve failure modes ........................... 23

6.2 Valve failure causes related to aging and service wear 25

7.1 Surveillance and maintenance recommendations ............ 28

8.1 Methods currently used to detect CV failure modes ....... 30

8.2 Methods for differentiating between failure causes ...... 31

8.3 Measurable parameters 35

8.4 Summary of valve part failure assessments .... 38

B.1 Check valve failures reported in LERs for period

1991 9 ... 9 - o.o...ooooe- 1 9 83.. . 50

B.2 Failure mode distribution . o. -o- .............. 63

B.3 Method of detection .......... 63 .............. 63

B.4 Maintenance activity .. o.o.ooo.................... 64

B.5 Identified failure cause ................................ 64
xi

SUMMARY

Practical and cost-effective methods are to be evaluated and identified for detecting, monitoring, and assessing the severity of timedependent degradation (aging and service wear) of check valves (CVs) in

nuclear plants under the Nuclear Plant Aging Research Program of the

Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research. These methods are to provide capabilities for establishing degradation trends prior to failure and developing guidance for effective

inspection, surveillance, and maintenance.

This report is the first-of three on CVs and addresses failure

modes and failure causes resulting from aging and service wear, recommended surveillance and maintenance practices, and measurable parameters

for detecting degradation prior to failure. The results presented are

based primarily on information from plant operating experience records,

plant operators, and equipment manufacturers. The two reports that follow will address, respectively, (1) assessment of inspection, surveillance, and monitoring techniques through testing and (2) recommendation

of guidelines for monitoring methods and maintenance to ensure operability under normal and emergency conditions.

This report briefly reviews typical CVs in boiling-water reactor and

:Pressurized-water reactor nuclear power plants in terms of functional

requirements, materials of construction, and operational stressors that

contribute to aging and service wear under both normal and emergency operating conditions.

Operating experiences reported in data bases for nuclear power

plants and in nuclear industry reports were examined. These data bases

included the Licensee Event Report (LER) file, Nuclear Plant Reliability

Data System (NPRDS), and the In-Plant Reliability Data System IPRDS).

Information was obtained from component manufacturers by reviewing

their literature and in direct discussions with their representatives.

The subjects addressed were failure modes, failure causes, and manufacturer-recommended surveillance and maintenance practices.' "Five' failure

modes were identified: failure to open, failure to close, plugged'' reverse leakage, and external leakage. Failure causes for each failure

mode were then identified at the subcomponent or subassembly level.

Manufacturer-recommended surveillance and maintenance practices are

general in nature, although detailed instructions for repair of internals

are sometimes provided. These practices include obturator movement and

external leakage checks, exercising, bonnet (or cap) joint inspection,

repair of internal parts, and reverse leakage repair.

Failure modes are examined in this study by identifying methods for

detecting failure modes and differentiating between failure causes. The

report identifies measurable parameters (including functional indicators)

currently used for inspection, surveillance, and monitoring. They consist of force or torque for obturator movement, pressure, temperature,

flow rate, reverse leakage rate, fluid level, and noise. The report also

identifies parameters potentially useful for enhancing detection of degradation and incipient failure; these parameters include dimensions, bolt

torque, noise, appearance, roughness, and cracking. The appropriateness

and utility of these and other parameters will be evaluated in subsequent

phases of the CV project.
- AGING 'AND SERVICE WEAR OF'CHECK VALVES USED

-:IN ENGINEERED SAFETY-FEATURE SYSTEMS

r'OF NUCLEAR POWER PLANTS

'Volume 1. Operating Experience and

Failure Identification

W. L. Greenstreet ' R. B.'-Gallaher' 

G.':A. Murphy-- ' D. M."'Eissenberg

ABSTRACT

This is the first in a series of three reports on check

valves (CVs) to be published under the Nuclear Plant Aging Research Program, and it addresses the subject of Detection of

Defects and Degradation.:Monitoring of Nuclear Plant Safety

Equipment. The program is concerned with the evaluation and

identification of practical and, cost-effective methods for -detecting, monitoring, and ass essing the severityof time-depen-

-dent degradation, (aging and service wear) ofCVs in nuclear

plants. These methods will allow degradation trends .to be

established prior to failure and allow guidance for effective

maintenance to be developed.

The topics of interest for this first report are failure

modes and causes resulting from aging and service wear, manufacturer-recommended maintenance, and surveillance practices,

and measurable parameters (including functional indicators) for

use in assessing operational readiness, establishing degrada-

*tion trends,'and'detecting incipient failure. The results

presented are based on-information derived from operating experience records, nuclear industry reports, manufacturer-supplied information, and input from plant operators.

Failuremodes' are identified for CVs. For each failure

*mode, failure causes are listed by subcomponent or subassembly,

and parameters potentially useful for detecting degradation,

which'could lead to arre'atabuated.' eailtre,' - ion

1. INTRODUCTION

1.1 Background7-'' -:

The Office of Nuclear Regulatory Research of the Nuclear Regulatory

Commission (NRC) has instituted-a `study'aimed at' unde'rstanding the timerelated degradation (aging) of nuclear power plant systems"and equipment.

It includes assessing the effectiveness of methods. of inspection and surveillance that monitor such degradation and establishing guidelines for
'. 

2

maintenance. The study is intended to provide technical bases for examining the ongoing operational safety-of operating;plants. The strategy

that will be followed should be useful to others interested in analyses

of equipment in nuclear applications.

This report addresses-the time-related-degradation of check valves

(CVs) - the second of eight components to be studied in the Nuclear Plant

Aging Research (NPAR) Program list of components. The others on the list

are motor-operated valves.(MOVs), auxiliaryfeedwater pumps, diesel generators, snubbers, batteries, -chargers, and inverters.

CVs are one of the most common components in a nuclear power plant'-

they are located in almost all plant fluid systems. The failure of these'

valves causes a significant amount-of plant maintenance and, more importantly, degradation of safety-related systems. In the last few years

considerable attention has been given to CVs by the NRC and industry

groups.

1.2 Objective

- The objective of this NPAR Program element is to review operating

experience and manufacturers' information, to identify failure modes and

causes resulting from aging and service wear of CVs in nuclear plant service, and to identify'measurable'parameters. These parameters are to be

suitable for detecting'and establishing trends in the time-dependent degradation of CV components prior to loss of function.

1.3 Project Scope

This report is Volume I of a three-part report to be prepared on

CVs. The contents of each of the three parts are summarized below.

Volume 1 - Operating experience, failure modes, and failure causes

1. Background-information- on CVs - boundary of -CVs to be studied,

types, uses, requirements, and materials of construction

2. Reviews of regulatory requirements,'guides, and'standards

3. Summary of operational--and environmental stressors

4. Summary of operating experience

5. Manufacturers' input

6. State-of-the-art aging and service wear monitoring and assessment

Volume 2 - Tests and assessments

1. Complete comprehensive aging assessment

Postservice examination and tests

In-plant assessments -

2.' Assessment of advanced monitoring techniques '
3

3. Controlled laboratory'testing

Aging assessment

Monitoring technique evaluation

Volume 3 - Analysis and recommendations

-1. Impact analysis

2. Recommendations of guidelines for monitoring methods and maintenance

1.4 Definitions

For the purpose of this report, the following definitions apply:

Failure mode - the way in which a component does not perform a function for which it was designed; that is, fails to actuate or leaks to

outside.

Failure cause - degradation (the presence of a defect) in a component that is the proximate cause of its failure; for example, bent shaft,

loss of lubricant, and loosening of a bolt.

Failure mechanisms - the phenomena that are responsible for the degradation present in a given component at a given time. Frequently, several failure mechanisms are collectively responsible for degradation

(synergistic influences). Where one major failure mechanism is identified, it has been called-the "root 'cause." Generic examples of failure

mechanisms (and of root causes) include aging, human error, and seismic

events.

Aging - the combined cumulative effects over time of internal and

external stressors acting on a component, leading to degradation of the

component, which increases with time.-Aging degradation may involve

changes in chemical, physical, electrical, or metallurgical properties,

dimensions, and/or relative positions of individual parts.

Normal aging - aging of -a component that has been designed, fabricated, installed, operated, and maintained in accordance with specifications, instructions, and good practice, and that results from exposure to

normal stressors for the specific application.. Normal aging should be

taken into account in component design and specification.

Measurable parameters - physical or chemical -characteristics of a

component that can be described or measured directly or indirectly and

that can be correlated with aging. Useful measurable parameters are

those that can be used to establish trends of the"magnitude of aging'associated with each .failure cause, that have well-defined criteria for

quantifying the approach to failure, and that are able to discriminate

between the degradation that leads to failure and other degradation.

Inspection, surveillance, and condition monitoring (ISCM) - the

spectrum of methods and hardware for obtaining qualitative or quantitative values of a measurable parameter of a component. The methods may be

periodic or continuous, may be in plant or may require-removal and installation in a test stand or disassembly, and may involve dynamic or

static measurements.
I

1.

4

2. BASIC INFORMATION

2.1 Principal Types and Uses of CVs in BWRs and PWRs

CVs are used extensively within pressurized-water reactor (PWR) and

boiling-water reactor (BWR) nuclear power plants for service in safetyrelated and balance-of-plant (BOP) systems. Sizes vary depending on service requirements and range between 0.5 and 28 in. (nominal pipe diameter).* The most commonly used types are swing, horizontal lift, vertical

lift, and ball CVs.

A summary of the usage of CVs in typical BWR and PWR nuclear power

plant systems is given in Table 2.1. The table indicates numbers of

valves, size ranges, and types used in the various systems. The functions of the listed safety-related system may differ from plant to plant.

*In conformance with current nuclear power industry practice,

English units will be used in this report.

Table 2.1. Summary of CV applications.

in nuclear'power'plants

Valve

System Number size

Syse of CVs (in.)

BWR (typical)

Low-pressure 10-18 2-28

core spray

High-pressure 10-14 4-24

coolant injection

(HPCI)

Low-pressure 10-21 4-24

coolant injection

(LPCI) includes

residual heat removal

(RHR) and containment

spray]

BOP systems 200-400 1/2-24

PWR (typical)

Auxiliary feedwater 4-23 4-8

Containment spray 4-14 6-14

HPCI 12-28 2-1/2-4

LPCI/RHR 5-14 8-10

BOP systems 200-400 2-60
2.2 CV Types

Swing* CVs are the most widely used. of all CVs because they offer

very little resistance to the flow when in wide-open position. Generally

used on all piping where the pressure differential is of prime importance, swing CVs are used for flowing liquids and can be installed in

vertical or horizontal position. However, these CVs are not recommended

for applications where the reversal of flow i frequent; this causes the

valve obturator to fluctuate rapidly and result in "valve chatter." Some

swing CVs have an external lever and counterweight balance arrangement to

permit adjustments that make the valve obturator more sensitive to flow

and allow it to open under aminimum of fluid. pressure.

Horizontal-piston-lift CVs (Fig. 2.3) are quite frequently assembled

on the same valve bodies as those ised for the regular globe valves.

They are generally used for such applications where the reversal of flow

and pressure fluctuations are very frequent, because they have less tendency to develop "obturator slam".and valve chatter. Horizontal-lift CVs

are used for flowing steam, air, and gases on horizontal piping lines,

but!'they are not recommended for installation on vertical piping systems.

Vertical-lift CVs (Fig.'2.4) are similar in constructi6n to horizontal-lift CVs and are especially designed for installationon vertical

piping systems. Another modification of the vertical-lift CV is the

angle vertical CV, which is used on right-angle turns in the piping systems.

Ball CVs (Fig. 2.5) are designed to handle viscous fluids and for

services where scale and sediment are present. These valves, usually

made in vertical, horizontal, and angle designs, are particularly recommended for rapidly fluctuating flow because of their quiet operation.

During-the ball CV operation,.the ball rotates constantly, equalizing the

wear on the ball and seat, thus prolonging the life of the valve.

Further design variations of the CV include stop-check valves and

nonreturn valves.

Stop-check valves, sometimes called "screw-down" CVs, are actually

modifications of the globe or angle valves. This modification consists

of making a slip stem connection-to the valve obturator instead of using

the obturator lock nut. In this-design, the obturator can be closed by

%hand, but can be opened only by the CV action; that is, by the fluid

pressure under the obturator. Probably the most common application of

the stop-check valve is as-a safety nonreturn valve. The ASME Boiler and

Pressure Vessel Code specifies ue of these valves for the boiler nozzle

of 'every boiler when two or more boilers are connected to the same

header. These valves are also called boiler stop-check valves or boiler

"screw-down" checks. _

In nuclear power plants, CVs are frequently used as containment

isolation valves-in lines where, in normal operation, fluid flows into

the containment. 4If-the pipe outside the containment should fail or the

.*See Figs. 2.1 and 2.2, which show valves from two manufacturers.

The-nomenclature Zised is shown in Fig. 2.2.
ORNL-OWG 85-4714 ETD

CAP STUD FASTENER NUT

CAP STUD FASTENER BOLT

SPIRAL WOUND GASKET -

OBTURATOR FASTENER NUT PIN

OBTURATOR FASTENER NUT

OBTURATOR FASTENER NUT WASHER

OBTURATOR

HANGER PIN

HANGER

CAP

CAP PIN

IDENTIFICATION PLATE

BODY

ORNL-DWG 85-4713 ETD

Fig. 2.1. Swing CV. Fig. 2.2. Swing CV, exploded view.
.7

ORNL DG8S 71 ED

Fig. 2.3. Horizontal-piston-lift CV.

ORNL-DW 5-4716 ETD

. . . .

_ ,

.. . . . ..

.. .

. .

r ' . , . , _

_

i , . * -s

He . . . : . | . )

PART

I BODY

2 OBTURATOR HOLDER

3 OBTURATOR

4 OBTURATOR GUIDE NUT

Fig. 2.4.

5 SCREW-IN HUB |

Vertical lift CV.
--l~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

8

ORNL-DW15 85-4717 ETD

Fig. 2.5. Ball CV, showing ball and seat arrangement.

pressure inside the containment should increase during emergency conditions, flow into the containment.-ceases and the CV closes, thus preventing flow from the containment to the atmosphere or external systems. Because CVs work automatically to prevent backflow, they are ideal for this

situation. However, containment isolation valves are required to have a

very low through-seat leak rate, which is sometimes difficult to achieve.

2.3 Equipment Boundaries

For purposes of this report, the CV is defined as follows (see Fig.

2.2).

1. Body assembly - valve body, cap (bonnet), fasteners, and plugs;

2. Internals seat, obturator, locking devices, hanger pin, hanger,

and any other internal parts;

3. Seals - seals and gaskets for external position indicators and/or

operators, plus those employed to seal the cap.

The nomenclature for CVs given in Ref. 2 is used.

Remote external position indicator sensors or devices are not considered in this discussion. Their failure would not affect the operation

of the CV, and they are therefore not included here.

2.4 Functional Requirements

In the functional-grouping of the entire valve family, the CV will

be found in the group defined as valves designed to control the direction
I

9

of flow. Check valves differ considerably in their construction and

operation from the other groups of valves, designed either-to stop the

flow entirely (gate, plug, and quick-opening valves),-or throttle the flow

to a desired degree (globe,.angle, needle,-Y,- diaphragm,.and butterfly

valves).

-Check valves;are entirely automatic in their.-operation and are activated inteinallyiby the flow of fluid that'they..regulate.. Check valves

permit the flow.of fluids in:only.one:direction;('if-the flow stops or

tries to-reverse its direction,.the CV.closes immediately and prevents

backflow..;.When the operating pressure "direction" in the line is-reestablished, the CV opens and-.the flow is resumed in the same direction

as before. . -

-Nuclear power plant CVs typically meet the following requirements.

1. Ambient service conditions:- temperatures 32 to 140'F, pressures up

to 40 psig, possible vibration'from upstream or downstream connected

components.

2. Capability: the CV must operate reliably with a minimum of maintenance. r v ' ' -

3.. Differential pressure (d/p) to close: this depends on the valve service, but in general the valve should close on zero flow, which is

zero d/p. .. . ' . - - -

4. Position-sensing device: -not often required; may befound on some

swing CVs. .* . - '

5.; Minimum pressure drop AP) across the valve when.open: AP across' the

valve at the expected maximum flow rate.

6. Process fluid. temperature:and pressure: operating pressure'.up to

2600 psig and temperatures,:up to 650'F. -

7. Opening pressure must be less than the-pressure drop in the line at

the minimum flow. .- -

2.5 Materials of Construction

2.5.1 Body assembly -

- Body, cap. Caststainless-steel CVs are manufactured in sizes from

0.25 to 8 in. Working-pressure for-these types may range from-150 to2500 psig,-with the temperature limit from 500 to 1100'F, by ASTM Standard,.-depending on the alloy.used. -' .1- -.. - .I :

In addition to stainless steel,-.CVs are also:made-of bronze, cast

iron,`Monel, .nickel, polyvinyl- chloride,'and other corrosion-resistant

materials to withstand-the corrosive action of the fluids.in contact.

Only stainless-steel valves are considered.in this-.report. .

- Fasteners.. The cap.stud bolts:used-in nuclear service'valves are

generally of type 304 or 316 stainless steel to be compatible with valve

materials in expansion and contraction due to temperature. These materials offer.higher.strength for-a given-diameter but, due to their hardness, must be.-properly lubricated to prevent'galling:in use.: -
10

2.5.2 Internals

Seat. Nuclear service'CV seats are generally machined into the

forging on smallervalves 'up to about 3-in. nominal pipe diameter. The

seats in larger valves are generally replaceable and are constructed of

specially hardened'alloys'such as Stellite or Hastelloy. Some valve

seats are resilient materials (to provide better'sealing).

Because the.: pressure producing the flow in-the pipeline must be

sufficient to-lift the'CV-obturator from its seat,:designers have used

various seat angles toaid this lifting action. The'most commonly used

seat and obturator angles are 00, 6, 12-1/2, and 450 to the vertical.

In the 125-psi class, the obturator of swing CVs usually will be

found at an angle of 6 to the vertical. In the 200-psi and up class,

the obturators are usually placed at 450 angles, because sufficient pressure is available to lift the obturator and open the flow in the line.

Horizontal obturators, placed at 900 angles to the vertical, are found in

the lift CVs.

Obturator. Valve'obturators are-normally of the same material as

the valve body to accommodate thermal expansion and contraction. They

may have seating surface materials of'Stellite or another hard alloy to

resist etching or. "wire drawing."* Ball CV obturators may be made of

Stellite or another hard alloy to resist wear and scratching.

Hanger pin, hanger, and fastener.' Swing CV obturators are connected

to a hanger that, in turn, hinges on hanger pins; this arrangement permits movement of the obturator out of the flow stream. The hanger and

hanger pin in stainless steel-CVs are generally'stainless steel alloy for

strength and corrosion resistance. The obturator fastener nut, washer,

and pin (and optional spring) are also stainless-steel. If a valve pin

is equipped with a special bearing on the hanger, the bearing is usually

made of a hardened alloy such as Stellite.

2.5.3 Seals

Gaskets. Nuclear service valves may have (1) welded caps obviating

the need for a gasket; (2) pressure-seal construction utilizing a steel

sealing ring and bolt configuration that seals the cap; or (3) ordinary

machined surfaces. for asbestos-type gaskets. Flexitallice-type gaskets

consist of a stainless'steel V-shaped strip axially wound with alternating layers of asbestos to form achevron-like seal cross'section. Such

gaskets should only be used once because the steel "'V" shape is crushed

upon, tightening - which provides the seal function.

-Seals. Only'a few CV designs have packing or seals-of graphiteasbestos for the hanger'pin. (The hanger usually attaches to the cap.)

Where there is packing on hanger pin cap seals, the stressors will be

identical to those of the cap gasket; therefore, they will not be discussed separately.

- *Wire drawing.refers to the case of a minute leak across the seat

that, under high differential pressure, causes a straight-line eroded

leak path resembling a mark that might occur if a small-diameter wire

were drawn across the surface.
11

3. TECHNICAL SPECIFICATION REQUIREMENTS

In nuclear power plants, periodic'surveillance'tests are used to

ensure operability of safety-related components'.', Test requirements ncluded in the Technical Specification's for each plant describe, either

directly or by reference, the various in-service inspections to be performed on' the major components *of safety-related systems.' In-service

inspections of all ASME cla'ss 1, 2, 'and 3 components are specified 'to be

in accordance with Sect. XI f the 'ASME Boiler and Pressure Vse oe

In addition, the Code o Fderal'Regulations (CFR) provides leakage

requirements for some components.

Article IWV-3600',in Sect. XI of the ASME Code'desc'ribes in-service

inspection requiremenits' for CVs *This equirement consists primarily of

exercising,'the valve to verify obturator travel to or 'from the full'open

'and closed'p'ositions'as required t'fulf ill its safety function.- Confirmation of seating or opening may be by visual observation, a position 'indicator, observation of relevant pressures in the sys tem, or other psitive means.". Surveillance intervals or, frequencies are given in'the ASME

Code.'. A summary 'of Aticle'IWV-3000 is provided in Appendix 'A._

C6s used for, co'ntainment isolation are also required to be'tested-in

accordance with' 10'C'FR 50' 'Appendix J Rf ) These'tests involve pressurizing th-C lclyin th'aedrcina hnthe valve is required to perform its safety-function and comparing leakage'rate with the

specified standrd Tss're e ed at refueling outages or at'least

every 2 years.

The purpose of Technical Specifications requirements for surveillance testing s to demonstrate operability of the component within

specified limits.' The purpose thus does not specifically include monitoring abnormalities in the-component that at a later time may lead to

loss of operability.
31

12

4. SUMMARY OF OPERATIONAL STRESSORS

In this section the CV is divided into subcomponents and parts. The

significant stressors acting upon these parts are identified qualitatively and (where possible) quantitatively, under normal and accident

(emergency) conditions.

Stressors' have been divided into six categories: electrical, mechanical, thermal, chemical, radiation, and environmental. The origins

and magnitudes of these tressors depend on the specific valve and include those generated externally to the valve boundary and those generated internally.

Check valves used in nuclear plant safety systems are located inside

and outside the containment structure. Under normal conditions the valyes

insi'de the containment'structure are exposed to the same or slightly more

severe external stressors than' the valves outside the containment. Under

accident conditions, the external environmental stressors inside containment'are (depending 'on location and type of accident)'more'severe. For

some CVs under accident conditions, the internal stressors are also more

severe than normal. Therefore, it is impossible to define a unique set

of stressors'for CVs in safety systems-, particularly under accident conditions. Guidance as to possible values of various external stressors

can be obtained from'valve 'actuator equipment qualification standards

issued by Institute of Eectrical and Electronic Engineers (IEEE 382)

(Ref. 4). Excerpts from that standard are given in Table 4.1 and Fig.

4.1.

ORNL-DWG 85-4513 ETD

~~~3 min

- CHEMICAL SOLUTION SPRAY 41 (SI5)

_4.0 385F (196°C) WITHIN 45 s DBE STEAM/TEMPERATURE

_ ~75 psig MINIMUM EXPOSURE STIMULATION

350 34f 7071 M7l~WTHIN 45 INDICATES 350 ~~~70psig MINIMUM OE-LS

320 F 160 C)/60 psig MINIMUM OPERATION

UL 300 300 F 160 0C)/60 psig MINIMUM 00

265 F (129 C)/40 psig MINIMUM

20 F-121 C)/30 psig MINIMUM

250 2300F (1 100C)/20 p~~~~~sig MINIMUM

HANGE CONDITIONS |

AT 1l/min TYPICAL 2 F s 00

CASE I, 11 AND III TERMINATES EACH TRANSIENT TO I

150 */ BEGIN FROM 120°F

10

0 1 2 3 4 .5 6 7 8 9 101112 18 24 5 10 15 20 25 30

| * HOURS * | - DAYS

TIME

Fig. 4.1. Exposure profile under accident condtions (IEEE 382).
13

Table 4.1. Equipment qualification tests for valve actuators

a. Environmental aging (parameters derived from Arrhenius formula) -

1380 C for 73 days, 400 cycles (all' cycles-defined as one stroke open

and one stroke close against one-third rated load with torque switch

operation at rated torque in the close direction),._,

b. Mechanical wear aging - 2,000 cycles (400 included in environmental

aging).

c. Pressurization aging - 15'cycles of 3 min duration at 65 psig.

d.- Radiation aging adesign basis event (DEE) radiation - 2.04 x 108

rad.

e. -Plant-induced vibration aging,- Biaxial sinusoidal motion of 0.75,g

with a frequency of 10 to 100 to'10 Hz at a rate of two-octaves-per

minute. 'Ninety minutes of vibration.in each orthogonal axis.

f. Resonant search - A-low-level (0.2-g) resonant search from 1 to 35

Hz and at-one octave per minute.'-

g. Seismic - A random multifrequency test witfia 30-s duration simultaneous-horizontal and vertical phase - incoherent inputs of random

motion consisting of fr'equency band widths spaced one-third octave

apart over the frequency range off to 40 Hz as necessary to envelope theirequired response-spectra. Five operating basis--earth-

,quake (OBE)'level tests [three-fourths of safe shutdown earthquake

(SSE) level] and one SSE level test in each orientation.

h. DBE environmental test - A steam exposure profile (see Fig. 4.1) for

an LOCA simulation representing PWR and BWR in-containment service.

i.' Steam impingement test -A steam exposure profile to 4920 F (2550 C)

to simulate'a steam-line break 'in a'PWR 4inlet steam temperature of

4920 F (255%C) obtained in 4 s].''

J. Seismic - required input motion test - Two OBE tests with a

sinusoidal sweep from 2 to 35 to 2 Hz in each axis-at a rate of one

octave per minute and a levelof,two-thirds of. therequiredinput

motion." One' SSE in each axis consisting of a 'continuous seriesof

single frequency since'beat tests a the one-third octave interval

test frequencies and'test levels indicated in IEEE 382-1980 (see

Fig. 4.1).

Source: Ref. 4.

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

14

4.1 Electrical Stressors

Check valves do not have electrical operators or position switches,

nor are they subject to electrical current. Therefore, there are no

electrical stressors associated with them.

4.2 Mechanical Stressors

4.2.1 Body assembly

Body, cap. Mechanical stressors on the valve body consist of

(1) pressure of the internal fluid; (2) vibration, including seismic

forces; (3) flow-induced forces; and (4) forces'resulting from the connections to the piping system. Operating experience has indicated few

valve body failures resulting from mechanical stressors during normal

conditions except where structural flaws have existed. Water hammer

-stress is a rapid'pressure pulse'that momentarily increases the tensile

stress in the wall. It tends to cause crack propagation from areas where

stress risers exist. In te event of an earthquake, the valve bodies may

be subjected to seismic low-frequency, low-magnitude vibrations of a

short duration. Emergency conditions, such as an LOCA, may result in

flow-induced vibration in the CV and its associated piping, but a vibration of no greater magnitude than that experienced under periodic test

conditions. Valves' are'also subjected to stresses from downstream

equipment-induced vibration.

Fasteners.' Cap stud olts are normally under tensile stresses from

tightening with mode'rate-to'-high shear forces exerted under potential

seismic loads. For stop-check valves there may be moderate'shear forces

exerted when the valve is losed with the manual'operator. Vibration,

.either flow-induced or other, may cause the cap stud bolt nuts to relax,

thus reducing the tensile stress in the bolt and nuts. If the bolts become sufficiently loose, the shear forces may increase considerably because of loss of' restraint of the friction forces of the bolted mating

surfaces, and fastener failure may result.

4.2.2 Internals

Seat. Valve seats are subject to compressive forces when closed

tightly in normal or accident-conditions.' Such forces are transmitted to

the valve body because' the seat is often a machined portion of the body

< or, in the case of replaceable seats, fits into a machined recess. Foreign objects lodged between the seat and obturator can prevent the valve

from sealing and can lead to scratches and subsequent erosion of the

seat.

Obturator. The portion of the valve that moves to control fluid

passage is subjected to hydraulic forces when in the closed position.

Additional mechanical stresses arise in stop-check valves if the obturator is forced onto the seat to prevent it from opening. When the valve

is open, low-to-moderate stresses occur from operating fluid passing
15

around or over the obturator. The stresses are essentially the same

under normal or emergency conditions. -

Obturator stress also results from vibration, rapid movement because

of flow transients (including water hammer), and pressure differential

during potential backflow conditions.--

Vibration, both flow-induced and transmitted from nearby equipment,

r stresses the hanger pin plugs and-their bearings and the areas where the

obturator attaches to the hanger (see Fig. 2.2). n these two Aoint locat'ions vibration can\caus excessiye wear/and high stress \i cenqationshat\ay/ resulW in_crackingof the part sj

Under conditions of pulsati flow, the CV may cycle with each

pulse. Ball checks are not appreciably stressed under these conditions

except for'the spring, which eventually may'fail from fatigue. 'The swing

CV obturator may be heavily stressed;' in'fact, rapid flow increases may

cause the obturator assembly'to open'suddenly, impacting the stop located

on the valve body. This impact can induce stresses throughout the assembly. -

When there is a potential for backflow (the closing of the CV prevents actual backflow), a pressure-differential-across the seating sur-

.face occurs. Low stress then develops in the ob'turator. The seating.

surface stress may result in (1) distortion of the surfaces, (2) damage

to soft-seated seals, or (3) jamming of the obturator in the seat. The

forces involved depend on the mating surface angle, as well as the pressure differential.

4.2.3 Seals

Gaskets. Valve gaskets are generally flat or 0-ring type.' Such

devices are .compressed to form a pressure-tight seal-to prevent fluid

leakage. The mechanical (compressive) stresses placed on these-,parts

normally do not degrade the part.- Thermal cycling could cause loosening

of the bolts, thus decreasing the compressive-forces on the gasket.

Loosening of the adjacent parts may permit the process fluid to-leak

through,-quickly erode the material, and destroy the sealing ability of

the gasket. A gasket properly located between-secured-mating components

can withstand the normal and potential emergency loads., -

4.3 Thermal Stressors

4.3.1 Body assembly

--- Body,' cap>. Thermal stressors applied to a valve during'normal

operation originate primarily-from the heat of* the-process fluid. Some

stainless steel alloys used in valve forgings or- castings-may-be-susceptible to corrosion under certain chemical and temperature conditions [intergranular stress corrosion cracking-(IGSCC)]., In some instances,-the

valve-temperature -may. change very rapidly when the system changes from a

-no-flow condition to a flow condition. Stress results from the temperature gradient across the valve wall.
I

16

During emergency conditions valve parts may be subject to slightly

higher process fluid temperatures but still considerably lower than the

degradation 'level.

Fasteners. Fasteners are subject to heat conducted from the valve

body but are not adversely affected under normal: or emergency conditions,

except for possible loosening because of thermal cycling. Under temperature cycling, fasteners may loosen because of differential thermal expansion.

4.3.2 Internals

Seat, obturator. These valve internal parts are generally stainless

steel alloys and are subject to the same thermal stresses as the valve

body. Such stresses normally do not degrade these components.

4.3.3 Seals

- Gasket. Thermally induced degradation of valve gaskets is a significant aging-related effect, even during normal operation. Heat acts

on valve gasket materials'to cause degradation of sealing capability

because of embrittlement.

4.4 Chemical Stressors

4.4.1 Body assembly

Body, cap. Other than chloride stress-corrosion cracking similar to

that experienced in piping, the only appreciable chemical stressors on

valve bodies result from contacts with borated water.

Under normal conditions most valves are not subject to borated waterinduced chemical stressors,' the exceptions being those valves that are a

part of the boron injection system or'are otherwise in'contact with primary-system, brated water in PWRs. Under accident or surveillance test

conditions, many other systems may be subject to-boric acid corrosion

either internally or (from sprays or leaks) externally.

Fasteners. Bolts may experience some external chemical stress from

borated-water spray or leakage under normal and accident emergency conditions.

4.4.2 Internals

Seat, obturator. Valve'seats may be subject to erosion or corrosion

' stressors from the working fluid and flow velocity. The chemical composition or the presence of-particulates (including impurities) in the

'fluid can affect the'corrosion rate.

Most valve obturators arethick enough to resist chemical attack so

that seat leakage will occur before obturator failure.' Erosion or corrosion of the internal components may, however, cause obturator binding
17

because of the roughness of sliding surfaces, particularly when the valve

is not operated frequently.

There is no appreciable difference in chemical stressor levels for

normal vs emergency operation.

4.4.3 Seals

Gasket. Most valve gaskets are fabricated from materiaisthat-are

relatively impervious to the water encountered in nuclear plant systems.

Borated water may impose additional chemical stress on gaskets with metallic components (Flexitallic8),but'failure would most likely occur

when degradation allows the borated water to breech the gasket.

4.5 Radiation Stressors

4.5.1 Body assembly

''Ionizing radiation has little effect-on the metallic parts-of valves

used in nuclear power plants. Valves-in safety-related service are ''

qualified to about 2 x108 rads foir-a 40-year'integrated-dose'(inside

~containment) and'2 x 107 'rads for-'outside containment.-"Few valves are

subject to'these levelsin normal operation so that-'for aging purposes

radiation-stress'can be'considered-negligible.' Under accident conditions

high radiation levels may be present for a short time but would ot'significantly affect the aging of the body assembly materials.

4.5.2 Internals

Same as body assembly above.

4.5.3 Seals

Valve seals and gasket materials can degrade because of ionizing radiation. The effects of radiation, combined with elevated temperature

and humidity, can shorten the life of such nonmetallic materials by a

combination of oxidation and free radical reactions that decrease

strength and elasticity. The damage increases with increased radiation

dose. The effects of radiation, temperature, and humidity appear to be

synergistic, and the order of exposure may affect the amount of damage.

The typical integrated radiation dose qualification limits are 2 x

108 rads for valves. In normal service inside (or outside) containment,

valve gaskets are exposed to levels considerably below this - typically

on the order of hundreds of millions. Thus, it is not expected that they

will suffer significant degradation because of radiation alone. However,

since the combined effects of radiation, temperature, and humidity are

not well known, it is possible that damage may occur because of a combination of stressors that includes radiation. Radiation stressors during

transient or emergency conditions are not expected to be different from
.

18

,normal except-for'LOCA 'conditions. During such'conditions, if fuel damage occurs such that there are high-radiation dose rates near valves inside containment, damage due'to radiation may occur. Such damage may be

exacerbated if elevated temperatures or humidity is also present.

In other transient and emergency conditions not involving the release of radioactive material, any existing elevated temperatures may increase the combined damage, including the radiation effect, but such damage should not be significant.

4.6 Environmental Stressors

The overall atmospheric environment that a valve may be subjected to

affects mainly the outside surfaces of the valve. Effects on individual

parts are negligible unless'the integrity of the valve is degraded; then

other stressors (discussed earlier) become dominant.

High humidity may cause unprotected external surfaces to rust or

corrode. Outdoor CVs at coastal plants may be subject to chemical stressors from salt spray or mist. In general, however, the effects should

not impair the operation of-the CV.

In a postulated emergency environment, such as produced by an LOCA,

a combination of high temperature, steam (humidity), pressure, and radiation can act synergistically on the valve. This action is in addition to

the operational stressors imposed by the altered hydraulic conditions

caused by an LOCA.
'19

5. OPERATING EXPERIENCE

The purpose of this section is to identify CV aging information obtained from various sources of nuclear power plant operating experience.

Several LER-based valve failure studies were examined for relevant CV

operating and-failure'information.' In addition, a number of special reports and studies that addressed valve problems in the nuclear industry

were examined. While these'documents-d6not'always contain specific CV

age-related failure data, the operating experience summaries and failure

cause data, included with the overall analysis results,,are helpful in

understanding the aging degradation-of CVs.

There are a number of operating experience data bases for nuclear

power plants. The data bases examined for'this report include

1. LER file

2. NPRDS

3. IPRDS

Specific information needed for CV failure characterization includes

(1) failure modes, causes, and mechanisms; (2) frequencies of failure;

(3) methods of failure detection; (4) maintenance actions;'and (5)'modifications resulting from failures. Each of the above items serves to

build a failure "signature" that,' i.hen taken totally, can provide a comprehensive assessment of the component failure.

Unfortunately, no single data base provides all of the information

desired fr each failure. But each data base does possess some useful

data elements that can be extracted for CV failure study. 'Additionally,

several studies on valve failures (including CVs) that provide-backup

information have been conducted by the NRC and industry organizations.

Table 5.1 lists he information available from various sources of operating experience and plant-specific documents. A summary of'CV failure

information available from' several da'ta'sources and a special study is

contained in Appendix B.

5.1 Summary of FailureModes and Causes

1. Valve seat leakage is a widespread problem in power plant applications. Causes of valve seat leakage'include the accumulation of dirt

and scale on 'the surfaces,' foreign objects lodging between the surfaces,

wear and/or wire cutting, deterioration ofielastomers, and insufficient

pressure differential for seating.

2. Wear or damage to valve internals is the next most frequent

problem. Vibration loosens the fastener holding the obturator to the

hanger pin, allowing the obturator to move out of position or fall free

into the valve body. A rapid start of flow through the valve may cause

the obturator to open suddenly, impacting the stop located on the valve

body. This impact may cause the obturator to become dislodged so that it

does not seal properly. In some cases it has broken free of the hanger,

and in other cases the hanger pin or its bushing has broken on one or
I

20

Table 5.1. Summary of CV failure information available from

operating experience and plant documents

Operating experience Plant-specific

data basesa documents Data source -

LER NPRDS IPRDS SAR SD TS ISI/IST

Valve type and description 0 X X X

Manufacturer and model No. 0 X X

Operating environment X X X

Failure cause 0 X

Failure mechanism 0 0

Discrete failed part 0 X

Maintenance action 0 0 X

Modification to prevent 0 0 X

recurrence.

Failure trend data X

Incipient failure X X

detection

Specific application X X

aAcronyms

IPRDS = In-Plant Reliability Data System

ISI/IST = In-Service Inspection/In-Service Testing Program

LER - Licensee Event Report

NPRDS = Nuclear Plant Reliability Data System

SAR = Safety Analysis Report

SD (Plant) System Description.

TS - Technical Specification/Surveillance Test Program

bo = Occasionallytincluded in failure report

X = Generally available

X

X

X
21

both sides. A loose obturator can result in restricted flow through the

valve.

3. Hanger pins have corroded, causing binding and resulting in the

failure of the valve'to open. Pins also have failed because of both excessive valve'movement during off-design flow rates and fatigue from the

impact of the obturator against 'the body during the off-design flow conditions.

4. Failure of the seal between the valve' seat and the valve body

occurred in a few cases.

5. -Accumulations 'of dirt and scale in the valve body have caused

binding of the valve internals so that the valves do not open.

6. In some stop-check valves, the valve seating angles are such

that if excessive pressure'is'used in seating the valve in the stop mode,

the valve will bind and fail to open with the pressure available from

normal operation.,

7. After installation-and initial testing of the CVs, very few

problems have occurred with the valve bodies. A few. through-wall leaks

have occurred. Small structural flaws in the valve body can act as

stress risers, resulting in crack propagation. through the wall. No

record of catastrophic failure of -the body was found; only small leaks

were found., Most flaws have-been found during installation, testing, or

routine surveillance. -- -

8. Small leaks through gaskets are also-a minor problem. In boric

acid systems, such leakage can-cause corrosion of the bolts that, if not

found, could result in larger leaks..

5.2 Frequency of Failures

Two data bases contain failure frequencies: the NPRDS and the

IPRDS. The NPRDS data are contained in the quarterly and annual reports

prepared by the Institute of Nuclear Power Operations (INPO). In the

1981 Annual Report, for CVs up to 4 in., 3.04 failures/106 calendar hours

were reported during the time period of 7/1/74 to 12/31/81 for leak and

failure to stop. For 4- to 12-in. valves, 3.96 failures/106 calendar

hours were reported for the same period. Data for CVs of 12 in. and

larger were insufficient to calculate failure rates.

The IPRDS data base has insufficient data on CVs to arrive at a

failure frequency.

5.3 Methods of Detection

The principal method of detecting CV failure is testing. In an LER

survey, surveillance testing found 32%, while 10 CR 50 Appendix J (Ref.

3) leakage testing found 27%. (These'tests are described in Appendix A.)

Another 28% of the failures were detected during normal operation. Only

1% of the failures occurred during an operational demand.
I

22

5.4 Maintenance Actions

The IPRDS, which extracts repair information from plant maintenance

records, is the only data base that contains detailed information on

maintenance actions performed on failed CVs. However, because of insufficient entries for CVs, this data base could not be included.

Maintenance activity is sometimes stated briefly in the LERs. Based

on these reports, the valves were repaired 54% of the time and replaced

11%. About 25% of the LERs did not indicate any maintenance activity.

5.5 Modifications Resulting from Failures

The operating experience data bases do not contain detailed descriptions of postfailure modifications. Some IE publications have outlined a

few CV modifications, which are summarized below:

1. Improved soft-seated valve seals - Hard seat valves were modified to

a combination soft'and hard seat configuration. Several types of

soft rings were tried before a molded (one-piece) seal provided a

satisfactory leaktightness.

2. Obturator attachment -The locking device that secures the obturator

to its hanger wore sufficiently to allow the obturator to fall free

of the hanger. Modifications to the design reduced this wear to an

acceptable level.
23

6. MANUFACTURER INPUT

This section summarizes CV failure modes and failure causes information, which was determined primarily on the basis of information provided

by valve manufacturers. Swing and lift CVs were examined; those in the

last category include piston lift and ball. Each valve is assumed to be

made up.of a bodyassembly, internals, and seals.

6.1 Failure Modes and Causes

The failure modes associated with the three CV designs are listed

in Table 6.1, which gives the modes and clarifying, or defining, remarks. A number of failure causes are associated with each failure

mode. The causes of interest in this report are those due to aging and

service wear.

Table 6.1. Check valve failure modesa

Failure mode Remarks

Failure to Open Valve failed to open fully

Failure to Close Valve failed to close fully

Plugged This failure mode refers to any event that would

stop or limit flow through a normally open valve;

valves that fail to open are not considered

plugged valves

Reverse Leakage Valve leaks through (measurable leakage past seat),

even though the valve indicates closed

External Leakage A leak or rupture of the valve that would allow the

contained medium to escape from the component

boundary

aAdapted from Ref. 5.

The following paragraphs describe the procedure used to identify

failure modes and causes with the assistance of manufacturers. The resulting modes and related causes are then given in tabular form.

6.2 Failure Cause Analysis

To obtain manufacturer input on failure modes and causes, studies

were done on manufacturer-supplied information, and telephone discussions were held with company representatives. Lists of failure causes

for each failure mode then were compiled for each valve type.
24

Visits were made to Rockwell International, Flow Control Division,

and to the Walworth Company to discuss failure modes and causes and recommended surveillance and maintenance practices. The compiled lists

were used as bases for discussion.

Failure causes are correlated with failure modes for each valve design in Table 6.2. The failure causes listed are self-explanatory.

.Operating experience indicates that Foreign nmterial is an important cause of Failure to Open and Failure to Close failure modes. Obturator and seat wear and erosion are important causes of Reverse Leakage. Also prominent are obturator fastener loosening and hanger pin

corrosion and fracture.
I

Table 6.2. Valve failure causes related to aging and service wear

Failure modes

Subcomponent Failure cause Failure Failure Reverse External

to Open to Close Leakage Leakage

CV type: Ball lift

Body assembl-y

Internals

Seals

Body guide rib corrosion

Body guide rib wear, erosion, corrosion

Body wear, erosion, corrosion

Body rupture.

Fastener..loosening,, breakage

' Obturator corrosion

Obturator. wear,erosion,,corrosion

Seat, corrosion

Seat.wear,,erosion, corrosion

Foreign aterial .

,Cap or~bonnet seal deterioration.

X X

X

x X

x

x

X X

X

X On X X

X X

X

CV type: Piston lift

,; I I I I' . . ! ,8

Body assembly Obturator'gulde'wear, erosion, corrosion

Body'wear, erosion,''corrosion

Body rupture

Equalizer plugged

Fastener loosening, breakage

Internals Obturator wear, erosion, corrosion

Seat corrosion

Seat wear, erosion, corrosion

- Foreign-material

X X X

x X

X

X

X

x

'C

K

X

X

X

'C

x

X

Seals Cap or bonnet seal deterioration -' X
Table 6.2 (continued)

Failure modes

Subcomponent Failure cause Failure Failure Reverse External

to Open to Close Plugged Leakage Leakage

CV type: Swing

Body assembly Body wear, erosion, corrosion X

Body erosion, corrosion X

Body rupture X

Fastener loosening, breakage X

Internals Hanger pin.wear,. erosion, corrosion, fracture X X X X

Hanger pin fracture X

Hanger pin bearing wear, fracture, corrosion X X X

Obturator hanger wear, fracture X X

Obturator hanger wear X

Obturator fastener loosening, tightening, X X X

breakage

Obturator wear, erosion, corrosion X

Seat wear, erosion, corrosion X

Foreign material X X X

Seals Cap or bonnet seal deterioration X

Hanger pin seal wear, deterioration X

on

V
27

7. MANUFACTURER-RECOMMENDED SURVEILLANCE AND

MAINTENANCE PRACTICES

Recommended surveillance and maintenance practices are contained in

manufacturer-supplied manuals. Much of the coverage is given only in

general terms because the products may be used in a variety of applications and be subjected to a broad spectrum of service conditions. Recommendations given by three manufacturers are outlined in Table 7.1.
Table 7.1. Surveillance and maintenance recommendations

Manufacturer Valve Reference Maintenance and troubleshooting

Atwood and Morrill Co., Inc. Bleeder check with side- 6 Preventive maintenance

closing cylinder Shaft binding check

Disk movement check

External leakage check

General maintenance

Inspection, repair, and. replacement

Exercising of valve

Rockwell International, Piston-lift check 7 Troubleshooting

Flow Control Division Bonnet (or cap) seal leakage

Seat leakage

Body rupture

Body guide rib wear, corrosion

Foreign material

Maintenance

Valve body repair

Seat leakage repair

Walworth Company Swing check 8 Inspection, repair, and replacement

Hanger pin

Hanger

Cap or bonnet

Obturator

Body

Troubleshooting

Cap or bonnet seal leakage

Reverse leakage

..
8. AGINGAND SERVICE WEAR MONITORING

Failure modes and causes along with associated inspection,,surveillance, and monitoring aspects are addressed in this section. The discussion is based on information derived from ASME in-service inspection

codes and standards, manufacturers, .and this study. The areas covered

are failure mode detection, cause determination and identification of

parameters for degradation trending, and incipient failure-detection.

8.1 Failure Mode and Cause Determination

Failure mode detection is'described in'terms of currently used parameters and methods. Candidate methods are'-also-identified. Failure

cause determination embraces both methods for cause differentiation and

*use'of measurable parameters for detailed evaluation. Methods for'differentiation' are discussed in-this6-subsection, while measurable parameters are discussed in Subsect. 8.2.

Technical Specification' reqiirements invoke use of the ASME Boiler

and Pressure Vessel Code, Sect. XI rules for in-service inspection of

CVs. These rules employ valve exercising for'assessing operational

readiness. Exercising 'is defined as'the demonstration, based on direct

or indirect visual or other positive 'indication,' that the moving parts

of the valve function satisfactorily. For CVs, the exercising tests are

to verify that the obturator travels t 'the full-open and/or full-closed

position.

Internal leakage rate testing is required for CVs used for containment isolation.-'This testing is conducted'in'-accordance'with Ref.-3.

The ASME Committee on Operation and:Maintenance is preparing a

standard, ANSI/ASME OM-10,'Inservice Testing of Valves, which is expected to supercede Subsect. IWV-of Sect. XI.- It is expected that exercising will again be used to measure operational-readiness.

In conducting the exercising tests in-accordance with either ASME

Code'Sect.- XI'-or'ANSI/ASMEOM-1O-a mechanical exerciser can be used to

move the obturator. When such an exerciser is used, the applied force

or torque is measured. For normally closed valves whose function is to

open on reversal of pressure differential, flow rate and pressure differential are parameters that, can be used for confirmation of obturator

movement under Sect. XI.,

The requirements of the ASME Code are related-to failu're'modes described in Table 6.1. Measurable parameters, in addition to force or

torque, as well as methods for monitoring .aging and service wear,,can .be

identified by considering methods for detecting .CV failures and ascertaining failure causes. Methods now used for detecting failure modes,

are listed in Table 8.1; methods .for differentiation between failure

causes are listed, in Table 8.2. ' - ' '

Table 8.1, shows ,that surveillance testing; process instrumentation

measurements of fluid .levels, pressure, temperature,,and flow-rate

changes; and disassembly' to verify operability are prominent means for
1

30

Table 8.1. Methods currently used to detect CV failure modes

Failure mode Means of identification

Failure to Open Surveillance testing in accordance with ASME Code

Sect.' XI (Technical Specification requirement)

Failure to Close Process instrumentation measurements of fluid level,

pressure, temperature, and flow-rate changes or

lack thereof

Operational'abnormality as 'shown by position

indicator (if equipped)

Disassembly to verify operability

X-ray examination

Plugged Process instrumentation measurements of fluid level,

pressure, temperature, and flow-rate changes or

lack thereof

Disassembly to verify operability

Reverse Leakage Surveillance testing in accordance with 10 CFR 50,

Appendix J (Ref. 3) (Technical Specification

requirement for containment isolation valves)

Leakage rate testing

Process instrumentation measurements of changes in

system pressure, level, or temperature

External Leakage Environmental changes in vicinity of valve; that is,

flooding-and high humidity

- routine surveillance

- incidental observation

Area sump-monitoring

Hydrostatic testing

failure mode identification. Other important means are position indicator signals, leakage rate testing, and X-ray examination. Nondestructive examination (NDE) methods other than X-ray examination that merit

consideration9 are ones based on eddy-current and ultrasonic techniques.

Candidate methods for Reverse Leakage identification include acoustic

monitoring,1 .

0

'll infrared remote detection, and dedicated downstream

temperature measurement.

Only piston lift and swing CVs'are addressed in Table 8.2. Because

compilations for ball valves do not add to the methods for differentiation, they' are omitted. The table shows that 'cause differentiation is

heavily dependent on'valve'disassembly and inspection. Visual examination and inspection during maintenance are applicable to External Leakage.
Table 8.2. Methods for differentiating between failure causes

Failure mode Subcrmponent 'Failure causes Methods for differentiation

Failure to Open Body assembly

Internals

CV type: Piston lift

Obturator guide wear, erosion, corrosion

Equalizer plugged

Obturator wear, erosion, corrosion

Seat corrosion

Foreign material

,Obturator. guide wear, erosion,.corrosion

Obturator wear, ,erosion,,corrosion

tSeat wear, erosion, corrosion

Foreign material

Foreign material

Disassembly and

Disassembly and

Disassembly and

*Disassembly and

Disassembly and

Disassembly and

*Disassembly'and

* Disassembly and

. Disassembly, and

inspection

inspection

inspection

inspection

inspection

Failure to Close Body assembly

Internals

inspection

inspection

inspection

inspection IPlugged Internals

Reverse Leakage Body assembly

Internals

Obturator guide wear, erosion, corrosion

Body weajr, erosion, corrosion

Obttirator wear, erosion, corrosion

Seat wear, erosion, corrosion

Body wear, erosion, corrosion

Body rupture

Disassembly and inspection

Disassembly and inspection

Disassembly and inspection

Disassembly and inspection

Disassembly and inspection

Disassembly and inspection

Visual examination, inspection

during maintenance

Visual examination, inspection

during maintenance

Visual examination, inspection

during maintenance

External Leakage

I ! ..

Body assembly

Fastener loosening, breakage

Seals Cap or bonnet seal deterioration
Table 8.2 (continued)

Failure mode Subcomponent Failure causes Methods for differentiation

Failure to Open Internals

CV type: Swing

Hanger pin wear, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Foreign material

Disassembly and inspection

Disassembly and inspection

Disassembly and inspection

Failure to Close Internals Hanger pin wear, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Obturator hanger-wear, fracture

Obturator fastener loosening, breakage

Foreign material

Disassembly

Disassembly

Disassembly

Disassembly

Disassembly

and inspection

and inspection

and inspection

and inspection

and inspection

Plugged Internals Hanger pin fracture

Obturator hanger wear, fracture

Obturator fastener loosening, breakage

Foreign material

Disassembly

Disassembly

Disassembly

Disassembly

and

and

and

and

inspection

inspection

inspection

inspection

Reverse Leakage Body assembly

Internals

Body erosion, corrosion

Hanger pin wear, erosion, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Obturator hanger wear

Obturator fastener loosening, tightening,

breakage

Obturator wear, erosion, corrosion

Seat wear, erosion, corrosion

Disassembly

Disassembly

Disassembly

Disassembly

Disassembly

Disassembly

Disassembly

and:

and

and

and

and

inspection

inspection

inspection

inspection

inspection

and inspection

and inspection

.4
Table 8.2 (continued)

Failure mode Subcomponent Failure causes, Methods for differentiation

CV type: Swing (continue

External Leakage Body assembly

Internals -

Seals I .

Body wear, erosion, corrosion

Body rupture

Fastener loosening, breakage

Hanger pin wear,.corrosion, frac

Cap or bonnet seal deterioration

Hanger pin seal wear, deterioratJ

. . Disassembly and inspection I

ture Digassembi and'inspection

'Visual 

Auring 

Lo'n, Visuai'examina'tion; inspection

'during 

monitoring 

led) .- 

Disassembly and inspection

Visual examination, inspection

during maintenance

Visual examination, inspection

during maintenance

ture Disassembly and inspection

¢ - Visual examination, inspection

during maintenance.

Lon * Visual examination; inspection

'during maintenance, acoustic .'. monitoring for packing tightness, measurementof applied

force or torque forobturator - - - movementw

I:

.. ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 I 

. . . . I .

- I 1

.. . I .. 

I . .
34

8.2 Measurable Parameters for Establishing

Degradation Trends

In the preceding subsection, failure mode determination and failure

cause differentiation were considered. Measurable parameter use was

also discussed. As stated in the NPAR strategy, the objective of this

subsection is to enlarge on that use by introducing measurable parameters that have the potential for being combined with those already identified to enhance capabilities for examining degradation trends and

detecting incipient failure.'

Measurable parameters identifiable for evaluating operational readiness include force or torque applied to move the obturator; fluid level,

temperature, pressure, pressure differential, and flow rate; reverse

leakage rate; humidity; and noise. Additional parameters are necessary

both for positive failure cause identification and enhancement of capabilities for degradation tracking and incipient failure detection. Suggested parameters for fulfilling these needs, dimensions, appearance,

roughness, cracking, packing gland position, and bolt torque, are given

in Table 8.3; these parameters require further investigation. Leakage

rate, noise, and applied force or torque are included in the table as

well as in the list given previously. Although appearance is not

clearly a measurable parameter and is a term whose meaning depends on

the application, it is included because it can be used to fulfill a

major requirement of monitoring, that is, imparting useful information

for establishing trends and assessing aging and service wear.

A summary of valve part failure assessments as addressed in this

report is given in Table 8.4., which illustrates relationships between

materials, stressors, failure causes, and measurable parameters.

The utility of the parameters identified in this report will be

evaluated, and other parameters may be introduced in subsequent phases

of the CV investigation. A companion need to. that of measurable parameter identification and evaluation for inspection, maintenance, and

monitoring use is the development of criteria for accepting or rejecting

components or assemblies for further service. The decision criteria

will ensure that the component performs its function during system normal operating transients and emergency conditions. Development of such

criteria will be an evolutionary process requiring cooperative efforts

with users and, thus, is beyond the scope of the NPAR Program.
Table 8.3. Measurable parameters

Failure mode Subcomponent . Failure causes Measurable parametersa

Failure to Open Body assent

Internals

CV type: Piston lift

bly Obturator guide wear, erosion, corrosion,

Equalizer plugged

Obturator wear, erosion, corrosion

Seat corrosion

Foreign material

ly Obturator guide wear, erosion, corrosion

Obturator wear, erosion, corrosion

Seat wear; erosion, corrosion

Foreign material

Dimensions, appearance, roughness

Pressure differential, flow rate

Dimensions, appearance, roughness

Dimensions, appearance

Appearance

Failure to Close Body assent

Internals

Dimensions, appearance,

Dimensions, appearance,

Dimensions, appearance,

Appearance

roughness

roughness

roughness

W

inPlugged

Reverse Leakage

External Leakage

Internals

Body assembly

Internals

Body assembly

Seals

Foreign material

Obturator guide wear, erosion, corrosion

Body wear, erosion, corrosion

Obturator wear, erosion, corrosion

Seat wear, erosion, corrosion

Body wear, erosion, corrosion

Body rupture

Fastener loosening, breakage

Cap or bonnet seal deterioration

Appearance

Dimensions, appearance, cracking

Dimensions, appearance, cracking

Leakage rate, dimensions, appearance, cracking

Leakage rate, dimensions, appearance, cracking

Dimensions, appearance, cracking

Dimensions, appearance

Torque, appearance

Appearance
Table 8.3 (continued)

Failure mode Subcomponent Failure causes Measurable parametersa

Failure to Open Internals

Failure to Close Internals

CV type: Swing

Hanger pin wear, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Foreign material

Hanger pin wear, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Obturator hanger wear, fracture

Obturator fastener loosening, breakage

Foreign material

Dimensions, appearance, roughness

Dimensions, appearance

Appearance

Dimensions, appearance, roughness

Dimensions, appearance

Dimensions, appearance, roughness

Torque, appearance

Appearance

Plugged Internals Hanger pin fracture

Obturator hanger wear, fracture

Obturator fastener loosening, breakage

Foreign material

Appearance

Appearance

Torque, appearance

Appearance

Reverse Leakage Body assembly

Internals

Body erosion, corrosion

Hanger-pin wear, erosion, corrosion, fracture

Hanger pin bearing wear, fracture, corrosion

Obturator hanger wear

Obturator fastener loosening, tightening,

breakage

Obturator wear, erosion, corrosion

Dimensions, appearance, cracking

Dimensions, appearance

Dimensions, appearance

Dimensions, appearance

Torque, appearance

Leakage- rate, dimensions,

appearance, cracking

Leakage rate, dimensions,

appearance, cracking

Seat wear, erosion, corrosion
Table 8.3 (continued)

Failure mode Subcomponent Failure causes Measurable parametersa

CV type: Swing (continue )

External Leakage Body assembly Body wear, erosion, corrosion' Dimensions, appearance, cracking

Body rupture Dimensions, appearance

Fastener loosening, breakage Torque, appearance

Internals Hanger pin wear, corrosion, fracture Dimensions, appearance

Seals : -Cap or bonnet eal-deterioration Appearance'

Hanger pin seal wear, deterioration Appearance,' noise. force or torque

-- ', . . . ;. - .; . . . , .applied foe obturator-movement,

packing gland position

" aThe measureable parameterslisted in this table reflect,primarily the methods for differentiation given in

Table 8.2. . * *, 

I

4.
Part

Body, cap

(bonnet)

Fasteners

Seat

Obturator

Obturator

hanger

Hanger pin

Hanger pin

bearing

Seals, gaskets

Materia.

Stainless stei

Stainless stei

Stainless stei

hardened alla

Resilient mats

Stainless stew

hardened all!

ing surface

Stainless ste4

Stainless stei

Hardened allo

Asbestos type

Stainless stew

Resilient mati

Table 8.4. Summary of valve part failure assessments

ls Significant stressors and

failure causes

el Mechanical: obturator guide Dimensi

wear, galling, body wear, cracki

rupture

Chemical: corrosion, erosion

el Mechanical: loosening, breakage Torque,

Chemical: corrosion

el or Mechanical: wear Leakage

3y Chemical: erosion, corrosion cracki

erial

el with Mechanical: wear Leakage

3y seat- Chemical: erosion,, corrosion cracki

el Mechanical: wear, fracture Dimensi

Chemical: erosion, corrosion

el Mechanical: wear, fracture Dimensi

Chemical: erosion, corrosion

y Mechanical: wear, fracture Dimensi

Chemical: erosion, corrosion

Mechanical: distortion, compres- Externa

el pression torque

erial Thermal: hardening, embrittle- tor mo

ment (nonmetals) tion

Chemical: corrosion

Measurable parameters

ons, appearance, roughness,

ing

appearance

rate, dimensions, appearance,

ing

rate, dimensions, appearance,

ing

ons, appearance, roughness

ons, appearance, roughness

ons, appearance

.l leakage, appearance, noise,

or force applied for obtura-

,vement, packing gland posico

.4
39

'9. SUMMARY AND RECOMMENDATIONS

The objective of this study was to identify failure modes and

causes resulting from aging and service wear of CVs in nuclear plant

service and to identify measurable parameters that are suitable for detecting and establishing time-dependent degradation trends prior to

failure, as well .as giving input for effective maintenance. To this

end, operating experience information, nuclear industry reports, manufacturer-supplied information, and results from discussions with manufacturers and-plant operators have been used. -

The dominant'failure mode shown -by-operatingzexperience records is

Reverse Leakage past the'seating surfaces., These.records also show that

Failure-to Close and PZugged are'frequent failure modes. These results

were not unexpected, and many'possible failure causes can be identified

with the three modes. It is;thesejcauses and-those associated with other

failure modes that were the focus of this study. Having identified failure causes, potentially useful-parameters for degradation tracking and

incipient failure detection were listed. The effectiveness and acceptability of these parameters will be'evaluated in subsequent phases of the

CV project.

The major methods used for failure-cause identification are valve

disassembly-and inspection, visual examination, and 'inspection during

maintenance. -Thus, periodic inspection and surveillance are expected to

continue for CVs.

Beyond cause determination and degradation monitoring are assessments of the extent of aging and service wear. These assessments will

be made- in terms of acceptance or rejection-criteria for further service, with synergistic effect influences factored in., Decisions will be

based on criteria that will ensure that the component performs its function during normal'system operating transients and emergency conditions.

This broader-perspective will be addressed in subsequent phases of the

CV study, and it is recommended that review-and development of acceptance criteria (in cooperation' with'users) be given -attention in keeping

with the prominent roles these play.

The relationship.of:the first-phase study reported here to the NPAR

Program'strategy is illustrated by the cross-hatched portions of the

diagram in Fig. 9.1. - . ; ' -
40

ORNL-OWG SS-4512 TO

C ,1\\dIat Ion "OS:NPROS. L~~~~~~~~~~~~~~l~~~s Intract~~~~~~~~~~~ns ~~for Lit.

Injer#Cjjonj Exsensuons

Esperts ~ R/ NPE* atc.5~/,/ Cadet. Standards. Practical Cost with Code

Eoplcte KnowledLcgeig nd nards fti sPot Effactive ormanice Valueimts Study SsndfoAtp Ad

nspectiInusty ratiesIndicators Committees Gieie

Prioritiration ,-Es and Technology for'~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~fo

Study v l Knyritde - 0 esigns / Postservice . emw and neretions Interactions Inoatv

L S W Eanr Speiiain Esaiain lnpcH - Veiicto of wm t C h Material

Fyig. Advanc.d Indtr NrogrCm Staftegn

ms ,Ifldi..stry Operalionat/ Surveillance Methods~~~~~~~~~~~~~~~~~~~~~~Preicton

i ~ Patc~ aantr oratoryspeests MGintenance

Aging and 'Ongo~~~~~~~~nij Aimatytes Artificlafl~~~~~~~~~~~~~~~~cceteratedrdsan

Fig. 9.1. strategy.~ NPAR Program ~ ~ ~ ~ ~ ~~~~~Gide
41

REFERENCES

1. Nuclear Plant Aging Research for Operating Reactors - Inspection

Surveillance and Maintenance Program Plan, U.S. Nuclear Regulatory

Commission Draft Report, Washington, D.C., July 1984.

2. P. A. Schwertzer, Handbook of Valves, Industrial Press Inc., New

York, 1972.

3. Code of Federal Regulations, Title 10 - Energy, Part 50, App. J,

"Primary Reactor Containment Leakage Testing for Water-Cooled Power

Reactors," Jan. 1, 1984.

4. "IEEE Standard for Qualification of Safety-Related Valve Actuators," .IEEE 382-1980, The Institute of Electrical and Electronics

Engineers, Inc., New York, Oct. 31, 1980.

5. C. F. Miller et al., Data Sumnmaries of Licensee Event Reports of

Valves at U.S. Commercial Nuclear Pants, Jan. 1, 1976-Dec. 31,

1980, NUREG/CR-1363 (EGG-EA-5816) Rev. 1, October 1982.

6. Reference Manual for Bleeder Check Valves, Atwood and Morrill Co.,

Inc., prepared for the Kansas Power and Light Company, Jeffrey

Energy Center Unit 3, undated.

7. Maintenance Manual for Rockwell-Edward Pressure-Seal Valves, V-377

R1, Rockwell International Flow Control Division, Raleigh, N.C.,

February 1983.

8. Manually Operated Gate and Globe Valves and Self-Actuating Swing

Check Valves, Walworth Company Aloyco Plant, Linden, N.J., May

1977.

9. W. L. Greenstreet, Oak Ridge National Laboratory, Oak Ridge, Tenn.,

personal communication with R. W. McClung, Oak Ridge National Laboratory, July 30, 1984.

10. "A Sound Method of Testing," Compressed Air Nag. 88(9) (September

1983).

11. J. W. Allen, W. F. Hartman, and J. C. Robinson, Acoustic Monitoring

of Power Plant Valves, EPRI NP-2444-SY, June 1982.
43

Appendix A

SUMMARY OF 'ASME BOILER AND PRESSURE VESSEL CODE

SECT. 'XI:REQUIREMENTS FOR CHECK'VALVES

The ASME Code Sect. XI requirements'for valves are given in Subsect.

IWV, Inscrvice Testing of Valves in Nuclear Power Plants. This subsection discusses in-service testing--of Nuclear Class 1, 2, and 3 valves,

including their activating and position-indicating systems,.which are required to perform a'specific-function in bringing a reactor to the cold

shutdown condition or in mitigating the consequences of an accident. 'Although some valves,' such as manual vent, drain, instrument, maintenance,

and test valves, are excluded from test requirements, they do not comprise a significant percentage of'those valves that fall within the categories requiring testing.

A.1 Valve Categories

Tht valves selected for in-service testing are placed in one or more

of the following categories. When one or more distinguishing category

characteristic is applicable, all requirements of each of the individual

categories are applicable,-although duplication or repetition of common

requirements is not done.

1. Category A - valves for which seat leakage is limited to a specific

maximum amount in the closedposition for fulfillment of their functions.* ' ' p t f of thi fuc'

Category B - valves for which'seat leakage inthe closed psition'is

inconsequential for' fulfillment of their function.

'3. Category C - valves that are self-actuating in response to some system-characteristics, such'as pressure (relief valves) or flow'direction (check valves). -

4. Category D - valves that are actuated by an energy source capable 'of

only one operation,'such as- r-ture'disks or explosive-actuated

valveso'

In additiontd' these categories,ithe valves-are'further classified

as being active or-passive."Active-valves are-those requiring a change

in position to accomplish their functions, while passive valves do not

requirearchange'in position to accomplish their functions.'-

The requirements applied to CVs (Category C)-are` discussed below.

''2 . T a i , ' IvesP * _, _ 5 

'-' A.2 Testing of Category C Valve's

t CVs, safety valves, and relief valves which, by design, must change

position fr fulfillment of their function, are Category.C valves. Distinction is important betweenvalves that serve a-safety-related function
I

I

44

and valves that are employed only for overpressure protection under system functional test conditions or for control of flow distribution within

interconnected systems undergoing tests. The latter valves are not subject to tests. The intent of Category C tests is to confirm the freedom

of obturators in CVs to prevent reverse flow where such reverse flow

would impair the fulfillment of a safety function. (Safety and relief

valves are tested in accordance with ASME PTC-25.3.) In either case,

these valves must be tested on a regular schedule as directed in Subsect.

IWV with each valve tested at least once in 5 years.

When a valve fails to function properly during a regular test, all

valves in the system in.that particular category must be tested. Additional valves are determined by an arbitrary assumption that a 12-month

operating period has passed to another refueling period and the additional valves to be tested will make the cumulative total at least N/60

times the total valves in this category.

For example, if there are 10 valves in this category and it is the

twelfth month after startup:

N/60 x number of valves to be tested = (12/60) x 10 =

(1/5) x 10 2 valves

where N = 12 months. After one valve has been tested, the second valve

fails the test. N now becomes 24, therefore, (24/60) x 10 = (2/5) x 10

4 valves total or 2 additional valves. If either of these two additional

valves fail the test, then all ten valves in this category have to be

tested. The'exercising tests for CVs are identical to those stated for

Category A and B valves; that is, they are simply to show that the valve

can obtain the position required to fulfill its function. If the fullstroke position is not practical duringtplant operation, then the valve

will be part-stroke exercised, followed.up by a full-stroke'exercise during cold shutdown. If a valve cannot be exercised at all during plant

operation, then it will also receive a full-stroke exercise during cold

shutdown.

Check valves that are normally closed during, plant operation and

whose function is to open on reversal of pressure differential will be

tested by proving that the obturator moves promptly away from the seat

when the closing pressure differential is removed and flow through the

valve is initiated or when a mechanical opening force is applied to the

obturator. This test can be made with or without flow through the valve;

however, a mechanical exerciser shall be used to move the obturator if a

no-flow test is conducted.'

Confirmation that the CV obturator is either on its seat or has

moved away from its seat will be by (1) visual observation, (2) an electrical signal indicated by a position-indicating device, (3) observation

of appropriate pressure indications in the system, or (4) other positive

means.

- The corrective action for CVs is also identical to Category A and B

valves in that, if the CV fails to exhibit the required change of obtu-

*rator position, corrective action is to be taken immediately. If the
45

condition is not corrected within- 24 h, the valve is declared inoperative. If this occurs during a cold shutdown, the valve's condition shall

be corrected prior to startup.

A typical in-service testing program outline for selected valves is

shown in Table A.l. The following information is given:

1. Valve Number lists the valve identification number as shown on the

piping and instrument drawing (P&ID). The first digit of the valve

number usually indicates the appropriate power plant unit.

2. Coordinates references the P&ID on which the valve appears and its

coordinates.

3. Class is the In-service Inspection (ISI) classification of the

valve. All primary containment valves'are included in the program,

even though some do not'have an ISI classification. These valves

are designated as Class NC (not classified).

4. Valve Category-indicates the category assigned to the valve based on

the defin-tions'given previously.

5. Valve Size lists the nominal pipe size of the valve in inches.

6. Valve Type lists the valve design as indicated by the following abbreviations: Gate - GA,' Globe- GL, and Check - CK.

7. Actuator Type lists the type of valve actuator as indicated by the

following abbreviations: Motor Operator - MO and Self-Actuated - SA.

8. Normal Position indicates the normal position of the valve during

plant operation; either normally open (0) or normally closed (C).

9. Stroke Direction indicates-the direction that an active valve must

stroke to perform its safety function.' Also, the direction in which

the valve will be stroked to satisfy the ISI exercising requirements. This may be specified as .open (0), closed (C),' or both

(O&C).

10. Test lists the test or tests that will be performed for each valve

to fulfill the requirements of Subsect. IWV. The following tests

and abbreviations are used:

Seat Leak Test (AT)

Valve will be seat leak tested at the appropriate functional differential pressure.

Full-Stroke Exercise Test (BT)

Valve will be full-stroke exercised for operability in the direction

necessary to fulfill its safety function.,'

Check Valve Exercise'Test (CT-i)-

Check valve will be exercised fully open, closed, or both, depending

on the safety function of the valve.

Position Indication Check (PIT)

All valves with remote position indicators-that are inaccessible for

direct observation during normal plant operation must be checked to

verify that remote valve indications accurately reflect valve operation.

11. Test Mode indicates the frequency at which the above-mentioned tests

will be performed. The following abbreviations are used:

Normal Operation (OP) -

Valve tests with this designationf will be 'performed once every 3

months.
Table A.I. In-service testing program for Class 1, 2, and 3 valves

(Nuclear Power Station Unit-1, System: Residual

Heat Removal, P1D: M-29, Sht. I)

Maximum

Valve Coordi- Class Valve Valve Valve Actuator Normal Stroke Tet Test stroke Relief

No. nates category size type type position direction es mode time request Remarks

(s)

-- AT RR

2-1001-29A A-5 I A 16 GA M0 C 0 BT OP 25

AT RR 2-1001-29B A-7 I A 16 GA W) C 0 aT OP 25

AT- RR 2-1001-47 C-5 1 A 20 GA MO C O&C BT CS 40 VR-9 Group 2

AT RR isolation

2-1001-50 B-5 1 A 20 GA MO C O&C BT CS 40 VR-9 Group 2

AT RR isolation

2-1001-60 A-7 I A 4 GA MO C O&C BT CS 25 VR-9 Group 2

AT RR isolation

2-1001-63 A-6 I A 4 GA MO C O&C BT CS 25 VR-9 Group 2

PIT RR isolation

2-1001-68A A-5 I C 16 CK SA C 0 CT-I CS NA VR-7

PIT R

2-1001-68B A-6 I C 16 CK SA C 0 CT-I CS NA VR-7

2-1001-16A D-2 2 B 18 CL MO O&C 0 BT OP 125

2-1001-16B D-10 2 B 18 GL MO O&C 0 BT OP 125

2-1001-18A B-4 2 B 3 GA MO 0 C BT OP 30 VR-8

4-
47

Cold Shutdown* (CS)

Valve testing at cold shutdown is testing that commences not later

than 72 h after cold shutdown and continues until required testing

is completed or plant startup, whichever occurs first. Completion

of all required valve testing is not a requisite to plant startup.

Valve testing that is not completed during a cold shutdown will be

performed during subsequent cold shutdowns to meet the ASME Codespecified testing requirements. No valve needs to be tested more

often than once every 90 d.

Reactor Refueling (RR)

Valve tests with this designation will be conducted at reactor refueling outages only.

12. Maximum Stroke Time lists the maximum allowed full-stroke time in

seconds for valves requiring a BT test.

13. Relief Request references the relief request that applies to the particular valve.

14. Remarks lists clarification remarks or indicates that a valve receives an automatic isolation signal.

*NOTE: Most required valve testing is normally completed in 96 h

following cold shutdown. However, completion of all valve testing during

cold shutdown is not required if plant operating conditions will not permit the testing of specific valves.
49

Appendix B '

OPERATING EXPERIENCE DATA BASES AND REPORTS

B.1 ORNL-Nuclear Operations Analysis

Center LER Survey

B.1.1 Introduction

Abstracts of all Licensee Event'Reports :(LERs) (and event reports

issued prior to the LERs) that were issued by U.S. utilities are stored

on the Department of Energy (DOE) RECON data base or the NRC's Sequence

Coding and Search Sstem (SCSS). Both data bases can be accessed at the

Oak Ridge National Laboratory (ORNL) 'Nuclear Operations Analysis Center.

A search was made of these data bases'for all events indexed as check

valves (CVs), excluding main steam isolation valves (MSIV). Each event

abstract was reviewed to determine (1) mode of failure, (2) mode of detection, (3) maintenance activity', and'(4) ause of failure.

This review found 472 events;that span the time frame from 1969

through 1983 (1981-1983 events were' obtained via the SCSS). Results are

summarized in Table:B.1. Of the events, 51% occurred at pressurizedwater reactors (PWRs), 47% at boiling water reactors (BWRs), and 2 at

advanced or research reactors. The reactortype 'should. have little or no

effect on CV operation. During the review, certain types of events were

not included in this study of the 472 events. The types not included are:

1. failure to test the CV,

2. foreign reactor'events,

3. design errors of including or omitting CVs,-

4. incorrect seismic analysis, C

5. ventilation system check dampers,

6. BWR torus vacuum relief valves, and

7. valve body defects found during construction or initial testing.

Each report might include failure of more than one CV or multiple

failures of the same valve. Thus, the count does'not reflect the total

number of CV failures but is the iumber of reports sent to the Nuclear

Regulatory Commission (NRC). Also, the-utilities are not required to report all failures but only those that meet conditions as specified in

10 CFR 50. Some failures are not reported asLERs and,.therefore, are

not included in this review. 

B.1.2 Discussion of Results

B.1.2.1 Mode of failure. As would beexpected, leakage past the

seating surfaces of the CVs was the dominant failure mode (52%). The description of the event would state that leakage through the CV caused

certain conditions to occur. The most reported event was leakage into

accumulators resulting in high fluid level and/or low boron concentration. In most of these reports the failure cause and description of the
50

Table B.1 Check valve failures reported

in LERs for period 1969-1983

Percent

I. Mode of Failure

Leakage

Seating surfaces 52

Gasket 4

Seat-to-body 3

Internals 32

Body 2

Slow response time 1

Operational error 2

Other/unknown 4

II. Mode of Detection

Surveillance testing 32

Leak.rate testing 27

Normal operation 28

Maintenance 9

Demanda 1

Other/unknown 3

III. Maintenance Activity

Repair 54

Replace

In-kind 8

Different 3

Modification 9

Other 1

Unknown 25

IV. Cause

Wear 8

*Crud 15

Corrosion/erosion 5

Failure to seat 4

Design error 6

Crack/fatigue - 2

Installation/fabrication 9

Binding 3

Other/unknown 48

aResulting from emergency or accident condition.
51

maintenance activity to' repair were not included - most CV failure's were

listed as being detected during'normal operation. Level alarms notified

the operators of high level s,^ while routine sampling found low concentrations. Gasket leaks (pressure boundary leaks) occurred in 4% of the report, with seat-to-body leaks-(internal leaks)_acounting for only 3%.

Problems with the'valve' internals were rep6ted in$32% ofthe

events. Failure of parts often allows the valve obturator of,-the swing

CVs (see Fig. 2.1) to move out of position, thus preventing seating or

allowing it to fall free. The free obturator could move so as to throttle or block flow through-the line.

Only a few (2%) valve body'problems were reported and only a part of

these resulted in pressure boundary leakage. No catastrophic failure was

reported. - - " 

Slow valve response time.(1%) -and-errors in operating the valves

(2%) account for-the other identified modes ,of failure. A few reports,

-(4%) did;not describe-the-failure'mode-or give a mode not fitting the

-above-discussed categories.

B.1.2.2. Mode:of detection. In 59% of the events, failures were

-found during testing. Leak rate testing (10 CFR 50, App. J).was-involved

*in 27% of the'events, while regular surveillance testing of-various systems found 32% of -the events. Surveillance testing of diesel generators

-found CVs-that leaked through the seat, allowing the fuel to drain out of

-the fuel line. This resulted in-excessive starting times for the diesel

generators.' - -

The next most frequent mode of detection was normal operation (28%).

Pumps failed to start because of CVs sticking closed. Accumulator problems were found during'routine sampling or reading of instruments.

Only 9 of the events were discovered during maintenance activities

where, while repairing equipment, a CV failure was discovered. Some

loose obturators were found-this -way.- 

In only a very few cases' ()-was a system called on to'function in

an emergency condition, and a CV failed to operate properly. Thus, the

present test programs and maintenance'activities have found most of the

problems before the emergency demand occurred.

B.1.2.3 Maintenance activity. The major stated-maintenance activity was repair of the valve (54%), including cleaning,'lubricating,'

and replacing defective parts. 'In 11% of the cases the CV was replaced,

with 8%of the replacements being the same model, while 3% were replaced

with a different typ'e'of'CV. To solve the failure problem, modifications

were made in 9 of the events.' -This includes change of materials','change

of sizes, and a different way of securing the part. Repair procedure was

not stated 25% of the time.

B.1'. 3' Cause - :

Foreign material-(rust scale, sand, weld slug, etc.) accumulating in

'-the valve body caused 15% of-the events. This crud prevented seating

surfaces from sealing or'caused binding of the valve-internals.- Installation-and fabrication errors caused 9 of the failures., Valves were installed backwards or in vertical piping instead of horizontal runs.,,

Proper dimensional tolerances were not met.
52

Wear of internal components resulted in 8% of the failures.. These

were mainly the obturator stud and the hanger pin failures that resulted

in loose or mispositioned obturators. De'sign errors, were the cause of

failure 6% of the time, while corrosion/erosion caused 5%. Some reports

(4%) only recorded that the valve failed to seat. Binding of internals

caused another 3%, while 2% were caused by cracks or fatigue. However,

48% of the time no cause was given.

B.2 IE Bulletins, IE Circulars, and IE

Information Notices

When an incident occurs at a nuclear facility or several similar incidences occur at one or more than one nuclear facility that might have a

similar effect on other facilities, the NRC Office of Inspection and Enforcement may issue an IE Bulletin, IE Circular, or IE Information Notice

to those facilities that might be similarly affected. (IE Circulars have

not been issued since 1981.) These notifications briefly describe the -

incident(s) with emphasis on the cause of failure. A solution to correct

the failure(s) and/or prevent recurrence may be given. IE Bulletins require the licensees to take certain specified action and to provide

written response'to IE concerning the results of such actions. IE Information Notices may suggest action(s) to be taken, but no written response

is required. A review of all of these IE publications found 3 IE Bulletins (IEBs), 2 IE Circulars (IECs), and 12 IE Information Notices (INs)

that concern CV failures. Each of these publications is summarized below.

B.2.1 IEB 83-03, Check Valve Failures in Raw Water Cooling

Systems.of Diesel Generators, issued March 10, 1983

.A review of available operating experience data and LERs shows that

numerous CV failures have occurred in systems important to safety in nuclear power plants. A series of IE generic communications has been issued that describes a broad range of CV failures involving various designs, causes, and applications. The NRC has evaluated CV failures in

consideration of the need to request generic action by licensees. The

focus of this bulletin isdirected primarily at the failure mode of disassembly or partial disassembly of CV internals; for example, the CV disk

becomes separated from. the hinge.

Although most CVs in systems important to.safety are included in

current in-service testing (IST) program reviews, most are not required

to be reverse-flow tested or disassembled to detect gross internal failure because licensees have identified each of these valves as having a

single safety function: the open position. However, forward-flow tests

to verify the open position are inadequate for detecting internal disasse'mbly. Effective CV testing techniques are necessary to the development

of-a more meaningful and productive IST program. Operating experience

provides a basis -for determination of what areas of IST CV surveillance

need to be improved.
i 53

The specific requirements of this bulletin stem from analysis of CV

failures in the raw cooling water supply -to the diesel generators at the

Dresden and Quad'Cities 'nuclear power stations and other events.- At

Dresden and'Quad Cities', it was found that six'of six CVs in the raw

cooling water systems for the 'diesel generators had failed, with the disk

becoming detached from the pivot arm.-

For all valves,' the most dominant failure mode 'was caused by-a combination of abrasive and corrosive wear of valve internals. In particular, the valve disk was held to the pivot arm by a stud with washer and

nut. Apparently, flowconditions at the valves were such that the disks

vibrated (fluttered),' causing local abrasive wear attheari'bore of the

hinge where it joins the disk. This same action'also-resulted in severe

degradation of the washer used to retain the disk on the hinge, and, once

the degree of 'degradation at:the' hinge bore and washer was sufficient,

the two' components separated. The stud and nut wore such that the stud

and nut assembly pulled through the'enlarged hole. in the pivot arm and

became' detached.

B.2.2 IEB 80-01, Operability of ADS Valve Pneunatic Supply,'

issued January 14, 1980

Engineering evaluation for Peach Bottom 2 and'.3-has disclosed.that

the Automatic-Depressurization System (ADS) pneumatic'supply (either nitrogen or air) may not be operable for all possible'events'because of a

combination of misapplication ofCV, a lack of leak testing of the accumulator'systemibacking up eachADS valve operator,.and questions about

the continued operability of the pneumatic supply in a seismic-'event.

The CV'nearest the accumulator is a PAL, 3/4-in., stainless steel,

socket-welded CV with a'hard seat. -

-B.2.3. IEB 79-04, Incorrect Weights'for 'Swing"Check VaZves

'anufactured by-Vean Engineering Corporation,

issued March 30, 1979

North Anna No. 1, Beaver Valley No. 1land Salem No. l have reported

to -the NRC 'that theyihad been'provided incorrect'weights'for the 6-in.

swing CVs-provided-by ~Vlan Engineering Corporation. The 6-in. valve'

weight provided on`'the'drawing was-225 b,'whereas the'actual weight has

been determined to be 450 lb. In additlon~to'the 6-in' valves, drawings

for'3-in. valves have specified-60-lb weight,' while the'measured weight

by the manufacturer was 85 lb, Jand drawings for 4-in. valves have'specified 100-lb weight, while the measured weight was135 lb. 'In some'cases,

incorrect valve weights derived from engineering drawings were used in

piping stress analyses. ' ;: ' ' - ' '' 

B.2.4 -I C 78-15, Tilting Disc Check Valzves Fail to Close With

Gravity in Vertical Position, issued July 20, 1978

'At the San Onofre Nuclear Plant, an 8-in., 1500-lb tilting disk CV

failed to close with gravity because it was' installed in a vertical

rather than a horizontal pipeline. The valve disk was counter-weighted
i.

54

to close with the force of gravity when installed in a horizontal pipe.

The manufacturer did not determine the reverse flow necessary to close

the improperly installed valve. The CV is located in the Low Pressure

Safety Injection System as the first valve inside the containment and

may not have closed as required to maintain the containment integrity.

Tilting disk CVs can be designed for either horizontal or vertical

piping but not for both. Improperly installed tilting disk CVs will not

function properly..

B.2.5 IEC 77-08, Failure of Feedwater Sample Probe, issued

April- 15, 1977

During surveillance testing at the Cooper 'station on January 21,

1977, a high-pressure coolant injection (HPCI)'system'CV was found to be

nonfunctional. Inspection of the valve revealed a length of feedwater

sample probe lodged in the valve preventing the CV from fully closing,

which allowed feedwater to flow backward into the HPCI' system injection

line. However, the blocked CV would not have prevented the HPCI system

from supplying coolant to the feedwater system in the event it was required at the time.

B.2.6 IN 84-12, Failure of Soft Seat Valve Seals, issued

February 27,1984

This information notice is provided as a notification of the failure

of soft seat valve seals to meet the leakage limits of Appendix J of

10 CFR 50.

On September 29, 1983; the Commonwealth Edison Company reported (LER

83-107) that the inboard feedwater CVs at LaSalle Unit had 'failed to

meet the leakage limits of Appendix J of 10 CFR 50. When the CVs were

opened for inspection, the soft seat showed damage around the pressurerelieving vent grooves, some wear on the soft seat face, and slight wear

on the body seat.

These CVs had been modified before initial plant operation from a

hard seat valve to a combination soft and hard -seat configuration. This

was accomplished :by modifying the valve disks to allow the installation

of the soft seat seals. The seals were of molded ethylene-propylene rubber obtained through the valve manufacturer, Anchor/Darling Valve Company, from the Stillman- Rubber Company.

The reason these soft seat valve seals failed has not been definitely determined at this-time, but failure is believed to be due to one

or more of the following.

1. Sharp edges around the pressure-equalizing ports located in the disks

had cut the soft seal material in many locations. The sharp edges

apparently had not been properly removed when the valve disks were,

modified. It' is possible that air bypassed the seal through these

cuts. _'-

2. The machining of the soft seals for proper fit may have affected

their sealing capability.

3. The service conditions- encountered by the valves during lant startup

'and shutdown may have damaged the soft seals;
55

The damaged olded seals were replaced in September with new soft

seals of an extruded-vulcanized design obtained 'through the valve manufacturer from Stevens Associates. The licensee reported (LER 83-146)

that, following approximately one month of operation, the-inboard feedwater CVs again failed to pass the local' leak rate tests.' It was -determined that the excessive leakage was a result of gaps on the perimeter of

the disk seal material, one about' 1/2.in.' long and the other about 1-1/2

in. in length. These gaps appeared at the seam, or vulcanized," points

of the seal. The utility has 'replaced the vulcanized'seals with molded

(one piece) seals similar to those in the original design.

B.2.7 IN 84-06, Steam Binding of AuxiZiary Feedwater Punps,

issued January 25, 1984

This information notice provides notification of a problem pertaining to steam binding in the auxiliary feedwater (AFW) pumps duetoleakage from the main feedwater. system.

'The discharge piping' from the motor-driven AFW train is connected to

the main feedwater piping near the steam generator. 'Hot water, about

4250 F, from the main feedwater system' leaked back through the first CV,;

the motor-operatedvalve, and' the second CV to the, pump and flashed to

steam because of the lower pressure in the AFW, system. "(A significant

amount of steam was vented from the pump casing during the testing to determine 'the' cause of the trip'.)'' When the motor-driven pumps started, the

instrumentation sensed a low discharge pressure. The steam binding re-

-duced flow and prevent'ed discharge pressure from increasing above the low

pressure set point in the' 30's before the 'instrumentation tripped the

pump. Condensation could have further lowered the pressure to the sensors.

Leakage into the AF 'from, the feedwater system 'constitutes a common

mode failure that can lead to the loss of all AFW capability." Further,

there is the potential for water hammer damage if an AFW pump discharges

relatively cold water'into a region' of'the piping system that contains

steam.

B.2.8 IN 83-06, Nonidentical Repacement Parts, issued, , February 24, 1983 ,,

In October 1980,'Beaver Valley 1 filed an .LER reporting' the failure

of a pump discharge CV'to seat, properly when the' pump was' shutdown. 'The

licensee attributed..the problem to biiding between an antirotation device

on the'valve'disk and the'disk' swing'arm. Because thiswas' the'third

time' the,licenseehad experi'en'cedsimilar problems with this styleof

Velan valves, Additional efforts were;,directedtoward longer term resolution'after, correcting the'imediate problem. The licensee found that replacement disks, installed s partof leakage correction maintenace,

differed enough frm ,the original' disks to cause the problem. ' A total of

24 v alves 'of this'make'and type are installed at the facility.'

,., , , ,,. . - ,'. - . : .
56

B.2.9 IN 82-35, Failure of Three Check Valves on High Pressure

Injection Lines-to Pass.Flow, issued August 25, 1982

At Davis-Besse Unit 1 on June 4, 1982, a stop-check valve (HP-57)

in the normal makeup system failed to pass flow although 120 psid was

applied across the valve. Normal opening pressure is about 5 psid. The

problem was discovered while filling the reactor coolant system (RCS)

using a small low-head pump following a refueling and maintenance outage.

Normal makeup at Davis-Besse is via one of the four 2-1/2-in. highpressure injection (HPI) lines. Upon further investigation, HPI valves

HP-48 and HP-56 also failed to pass flow at 120 psid. Each HPI line has

a stop CV and a swing CV in series.

According to the manufacturer, all Velan 2-1/2-in. stop-check valves

are of the same basic design. The internals'consist' of- a disk that is

lightly spring-loaded against the valve seat. The disk opens to allow

'flow at pressures sufficient to overcome'spring tension. A valve stem,

which'is not connected to the disk, can be turned down on the disk via a

handwheel to block it against the seat. In this mode, the valve provides

an'isolating'function. '

The causes for valve failure are thought to be a combination of

overtorquing by operators and7a steep valve seat angle. Wear may have

also been a contributing factor; however, no obvious signs of wear have

been detected by visual'inspection.

Because the stem packing of the valves was so tight, the operators

used a 1-1/2-ft valve wrench rather than the handwheel to close the

valves. The valve manufacturer recommends that no more than 150 ft-lb of

torque be used to close the valve. With the valve wrench, the operator

could have easily overtorqued the valve.

B.2.10 IN 82-26, RCIC and HPCI Turbine Exhaust Check Valve

Failures, issued July 22, 1982

A number of reactor core isolation cooling (RCIC) turbine exhaust CV

failures have occurred during the past 20 months.

On December 10, 1980, Carolina Power and Light Company reported (LER

80-101/03L) an RCIC system turbine trip at Brunswick Steam Electric Plant

Unit 2 while conducting an-RCIC'system test. The turbine tripped on high

turbine exhaust pressure due to the turbine exhaust swing CV failing in

the closed position. Inspection revealed the CV disk stem had broken off

where it connects to the'valve hinge assembly, allowing the' disk to fall

into the discharge part of the valve and isolate flow. '

"On May.29, 1981, Pennsylvania Power and Light;Company reported

(LERs 100450/100508)'the failure of the RCIC turbine exhaust'swing CV at

Susq'uehanna Steam'Electric Station Unit l while conducting'an RCIC system

test. The stud'(integrally'cast with the disk) that attached the disk to

the valve hinge broke-off. In a subsequent report on February 5, 1982,

.they indicated that turbine exhaust-steam flow conditions experienced

during testing caused the'valve disk to cycle violently open and close.

On December10, 1981, Georgia Power Company reported (LER 81-

112/03L) an RCIC isolation at Edwin I. Hatch Nuclear Plant Unit 2 while

conducting an RCIC rated flow test. An investigation revealed that the
57

turbine exhaust CV 'had internal damage, creating a block 'in the line

causing the rupture diaphram to fail.

General' Electric identified the 'possible causes of failure as improper system operation,'improper CV'sizing, inadequate CV design, or inadequate exhaust line'design.

B.2.11 IN 82-20, Check VaZve Probl ems, issued-June 28, 1982

During required modifications of the low-pressure coolant injection

system-at the Palisades NuclearPlant, Consumers Power-Co. of Michigan

reported that two of the four LPSI 'swing CVs were found to have internal

damage. In both valves the disk nut washer and -the disk nut pin were

missing, and the valve-body,- clapper arm, disk clapper arm shaft, and

clapper arm support were severely worn. The disks were still attached to

their clapper arms; however, valve seat and disk sealing surfaces were

damaged, and leaks from the valves could have been excessive.

During start-up testing at the Susquehanna Steam Electric Station'

Unit 1, Pennsylvania Power and Light reported three problems with Pacific

check valves:-- (1) disk assembly-to-body interference and excessive packing friction, (2) excessive wear'at hinge arm/disk stud interface,' and-

(3) disk stud breakage. The Pacific 'check valves 'are used in many nonsafety systems as well as the residual heat removal,'reactor core isolation cooling, and core spray systems.

B.2.12 IN 82-09, Cracking in Piping of Makeup CooZant Lines

at BW Plants, issued March 31, 1982-- -

-A visual inspection inside the reactor building revealed a leak associated with a 2-1/2-in. CV (MOV-43) in the makeup line to the 26-in.-

reactor coolant loop A inlet line. This line is used for normal makeup

of reactor coolant but is also part of the redundant HPCI system. After

the insulation was removed from- the' affected valve, a' 1400-'circumferintial crack in the CV body near the valve-to-safe-end weld (i.e., valve

end toward RC inlet nozzle) was found. The leak was nonisolatable.

'A metallurgical investigation of-the affe'cted'valve body indicated

two 'crack initiation' sites. One was inside on'the valve body at a ma- ' chine mark (i.e., wld counterbore area) 'and one was on'the'outside'diameter (OD)'at the valve-to-weld transition (geometrical discontinuity).

The cracks progressed through the wall on a'slightly different piane and

merged'about midwall' of the valve body.' Scanning electron microscope examination-of the fracture features digclosed'the cracks propagated transgranularly and exhibited clearly defined grain structure striations'characteristic of cyclic'fatigue'failure.'' '

B.2.13 IN'82-08,-Check VaZve aitures on DieseZ GeneratorX.

.Engine CooZinq Systems,' issued March 26, 1982 ; --

,_S-;,'i ' v ' .-''., -, -a'r

,During-a monthly diesel generator surveillance test, the diesel generator was started normally from the control-room but soon tripped on

high engine temperature. Cooling water, flow to the--diesel generator heat
58

exchanger was, found to be inadequate. A surveillance test was then commenced on a second diesel generator where indications of insufficient

cooling water flow were also observed. A broken CV on the discharge of

the second diesel generator was found and replaced. The valve disk had

broken free of the pivot arm and was lodged in the discharge side of the

valve, restricting nearly all flow. The licensee inspected the discharge

CV on the first diesel generator pump and found it was broken also. As

was the case with the second pump CV, the disk' had broken free of the

pivot arm.

These failures were not adequately characterized by operator observations and instrument-readings during diesel generator surveillance

tests but were discovered by-direct-inspection of the internals of the

CV. It is not known how long these CVs were broken before their condition was detected because the broken valve disks were free to move within

the valve bodies.

8.2.14 IN 81-35, Check Valve Failures, issued December 2, 1981

Metropolitan Edison Company reported loose valve internals in the

high-pressure injection pump discharge CVs. The valves are Crane 3-in.

1500-lb tilting obturator CVs. The initial cause of the loose valve internals was traced to the corrosion of the seat holddown devices of the

valves.

Metropolitan Edison also found many fabrication inconsistencies

that may have initiated and/or contributed to these failures. These inconsistencies ranged from the use of materials, other than those specified in procurement documents,' to poor workmanship, particularly in the

case of welds. Thus, the CV failures can be attributed to two main

causes: (1) poor retaining device design and (2) poor quality control on

the assembly of the valve internals.

B.2.15 IN 81-30, VeZan Swing Check VaZves, issued

September 28, 1981

While a CV'leakage test at the Point Beach Nuclear Plant Unit 1 was

being performed, the CVs closest to the rea'ctor coolant system in the

low-head safety injection lines were found to be leaking more than allowed'by the leakage acceptance criteria. The valves are Velan 6-in.

1500-psig ASA swing CVs (Velan Drawing No. 78704).

" 'The valves'were.disassembled and the disks were found to be stuck in

the full-open position due to interference between the disk nut lockwire

(disk wire) and-the valve body. The disk nut and its shaft can rotate

freely, and, in certain random rotational positions, this interferenceis

likely to occur.

While a leak in the bonnet of a swing CV in the steam supply to the

turbine-driven auxiliary feedwater pump at Salem Generating Station Unit 2

was being repaired, the valve'was found' to be internally damaged. The

valve is a Velan 6-in. swing CV (Type B14-2114 B-2TS).

The valve disk stud had broken and the valve' disk was in the bottom

of the'valve body. The-valve also had cracks in the disk, cracked bushings, and a warped hinge pin, and all hinge pin holes were elongated.
; 59

The licensee inspected the corresponding swing CV in the other, steam supply line and discovered similar damage.

B.2.16 IN 80-41, Failure of Sing Check Valve in the Decay

Heat Remova Stem at Davis-Besse Unit No. 1, -

issued November 10, 1981 ..

The.licensee performed leak rate tests.and identified excessive

leakage through decay heat:removal system CV CF-30. Valve CF-30 is the

inboard-one of two in-series CVs that is-used to isolate the RCS from-'the

7low-pressure decay heat removal-system. On further investigation, the

-licensee found that the valve.disk and arm had separated from the valve

-body and was lodged ust.under the valve.cover plate. The:two-2-5/8-by

5/8-in.' bolts and'locking mechanism for-the bolts that holds the'arm to

-the.valve body were-missing and have.not been located. The CF-30 valve

is a 14-in.'swing CV manufactured by Velan.Valve'Corporation. The cause

-.of the failure has not been identified. :

B.2.17 IN 79-08, Interconneition of Contaminated Systems

with Service Systems Vsed as the Source of .,

Breathing Air, issued March 29, 1979 '

One'of'the functions of the, service air system at Peach Bottom is to

provide a source of breathing air'for personnel using supplied air respiratory protective equipment. By means of an interconnection-to the-radwaste system,-the'facility also-uses the service-air syste m to provide.a

source of compressed air during the backwash cycle.of the demineralizer

''filter-element. The comp'ressed air provides the-motive force for reverse-water flow'through thefilter element and was.beingused to perform

this function when two'incidents occurred wherein'liquid from the radwaste system leaked past a CV and a process valve.

The examinations 'revealed'the presence of dirt deposits intheCV

'and air-operated ball valve. The specific ause'of the leakage ;was attributed to these dirt deposits, which prevented the proper seating of

the valves.,

* B.3 ALO-75;'Pilot Proqramito Identify Valve Failures"

Which Impact the Safety and Operation of 'LWR '

* ! NuclearPower-Pants, published April.1980 . .:.

' , This 'paper presents the results 6f a'pilot program initiated by

Sandia Laboratories under'the'Department of Energy, Light Wite'r Reactor

Safety Research and Development Program. The program was conceived as a

result of earlier LWR safety and reliabiifty'studies thatindicated that

a substantial number-of;'plant trip incidents were caused'byuftilure of

system components such as valves. The specific'objectives of 'this program were to (1) identify the principal types and causes of failures in

valves, valve operators, and their controls and associated hardware that

lead to or could lead to 'plant trip and (2) suggest possible remedies for
- - ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~i

;4

60

the prevention of these failures and recommend future research and development programs that could lead to reducing these valve failures or to

mitigating their effect on plant operation.

The data surveyed cover incidents reported over the 6-year period,

beginning 1973 through the end of 1978. Three sources of information 'on

valve failures were consulted: (1) failure data centers, (2) participating organizations in the nuclear industry, and (3) technical documents.

The results of this study indicate that frequent failure modes in

valves include lack of leaktightness in'both stem packing seals and valve

seats and operational malfunction resulting from problems with actuators,

their power controls, and instrumentation. Specifically for CVs, the

study concluded that main seat leaktightness in main steam isolation and

feedwater CVs was reported-as a major source of maintenance work. These

hard seat valves require long periods of work onsite and may involve removal of bonnets and welding; grinding, and lapping of seat surfaces.

Valve seat leakage is a universal problem in electric power generation. In nuclear applications, this problem becomes more acute because

of the severe restrictions imposed on permissible leakage rates. Utility

personnel report that this problem occurs most during off-power testing,

when pressure differentials across the disk are low. Therefore, it is

not a major disruptive factor during plant power operation, but it is a

major source of maintenance activity during outages. Opinions of utility

personnel attribute this problem primarily to the severity of leakage

limitations and changing leak test requirements, coupled with the difficulty in obtaining leaktight repeatability in valves. In addition, leakage indicated'in a gaseous test medium, such as nitrogen, is not considered to necessarily indicate excessive leakage under the LWR'operating

medium.

This study found 5 CV failures out of a totak of 138 valve failures.

Two of these were in PWRs and three in BWRs. Failure modes were one seat

leakage, one-packing leakage, one stuck valve, and two procedures. A

program to address'seat leakage'has been recommended in a previous study

of this problem. MPR Associates in a report entitled Assessment of Industry Valve Problem, EPRI NP-241, November 1976, recommends, in summary,

the following.

1. Find improved methods of achieving seat tightness for MSIV and feedwater CVs in BWRs and containment isolation valves in PWRs and BWRs.

2. Develop leak testing methods and techniques that are directly applicable'to nuclear stations. -

3. Sponsor a long-range program to develop technology-for achieving

leaktight seating-designs in steam, gas, and- high-pressure, hightemperature water applications. This program would address material

combinations,' seat geometry,, surface wear, corrosion, radiation damage, and alignment of moving parts-.

4. Develop maintenance procedures, tooling, and techniques for restoring

seat tightness while keeping radiation exposure to maintenance personnel at'a minimum.
2 1

61

'-B.4 'SAND 80-1887, Proceedings EPRI/DOE'Workshop, 'Nuclear

Industry Valve Problel, Washington, D.C., c'-May 20-21, 1980 -

A workshop on nuclear industry valve problems was held at the Electric Power Research Institute (EPRI) offices in Washington, D.C. The

following recommendations were developed in working sessions on key

valves and on valve stem and seat leakage: ,(1) establish~a small permanent expert staff to collect ,analyze, 'and'disseminate information about

nuclear valve problems; -(2) perform generic."key"''valve programs for PWRs

and BWRs and several-plant-specific."key" valve' programs, the latter to

demonstrate the cost effectiveness-of such studies; (3) confirm the identity of, define, and initiate needed longer-term research and development

programs dealing with seat'and stem leakage; and (4) establish anindustry working group"to review and advis6 on these efforts.

Valve problems-are discussed'in general'terms with no data'-given.'

Concern is focused'on valve problems that resulted in'reactor trips orshutdowns. Four other reportsrare included-as appendices. Parts of'one

of these that concern'-CVs is'as follows.' '"

EPRI Report NP-241, Assessment of Industry Vatve Problems,

November '1976 (prepared by MPR Associates,' Inc.), '

Maintenance burdensassociated with)CVs include renewal of-pivot pin

seals and relapping or.replacing disk-to-bodyseating surfaces., Onewof

the problems is misapplication of specific valve type in using, CVs where

leaktightness of the seat is demanded. Leaktightness of valve seat to

flapper is a generic technical problem. Another problem is awarding of

purchase tothe lowest bidder.:'"Themisapplication'can-result in excessive maintenance requirements and/or high radiation exposures to qualified maintenance personnel.,- ,

B.5 R. L. Scott and R. B. Gallaher, Swmry and Bibliography

of Operating Experience With-Valves in-Light-Water-Reactor

NucZear Power Pants-for the-Period 1965-1978,

.NUREG/CR-0848 (July 1979) - ,

4 ' - ;. . ' ~ .'.;': . ' ' 

Operating experience with all types of valves in LWRs is summarized

for the period 1965-1978. 'Tables'are presented :givi the' - causes of

valve failures, time of occurrence, systems involved, and the equipment

in which the valve failures occurred. Check valves are included as part

of the whole but are not tabulated separately.

* * - _'-n,*.,.,,t-....

B.6 W. H. Hubble and C F. Miller, Licensee Event Report

Analysis for Selected Safety System Valves,

-';- - ; IDO-1570-Ts (1979) -

This analysis utilized the NRC LER-file to estimate LER-based failure rates for 'selected safety-system valves in operating nuclear power
62

plants. In general, the selected safety systems. included PWR and BWR

emergency-core cooling system (ECCS)-valves, AFW valves, and primary

safety/relief valves. LER'rates were calculated for reverse leakage of

check valves in both the ECCS'and AFW systems as well as other types of

valve rates. The time frame used for this analysis was January 1976-

October 1978.

B.7 W H. Schmidt,.An Analysis of Nuclear' Power

Plant Valve FaiZure rom Licensee Event

Reports 1975-1978, SAND80-0743,

(April 1980)

-A computer analysis of the NRC data file, compiled from LER data

sheets, has been performed to characterize and highlight valve failures

in LWR nuclear power plants and provide guidance for valve improvement

programs. The analysis is based on data from 1975 through 1978. For

PWRs, the second most important identified component failure category is

one-way flow; for BWRs the third category is one-way flow.

B.8 In-Plant Reliability Data System (IPRDS)

* A search was'made of the IPRDS for CVs. Two plants were included in

the search, one PWR and one BWR. The data available'were insufficient to

include in this review.

B.9 Nuclear Plant Reliability Data System (NPRDS)

* The NPRDS, operated by the Institute for Nuclear Power Operations

'(INPO), contains component engineering and failure data that can be obtained upon special request. Such data do have some limitations - no

plant identification'or'failure-event reference is permitted, preventing

correlation with-other data bases such as LERs.' 'But generic..population

failure data can'be'obtained-byutilizing a specified sort strategy. For

a failure event, certain information can be obtained from the data base

if computer searching techniques are applied, such as

1. severity - incipient, degraded, immediate;

.2. failure symptom;

3. failure detection;

4. cause description;

5. environment - internal and external;

6. manufacturer and model number of failed component;

7. material; -

8. size; and

9. narrative of failure cause, descrfption, and corrective action.

Because of a lack of computer searching capability in response to a

special request, INPO'provided hard copy of a data search, which yielded
63

585 CV failure events. The event data had to be manually reviewed and

sorted to extract actual aging-related check valve failures. The following types of failure reports were eliminated:--

1. main steam check valves,

2. vacuum relief valves,

3. design errors,

4. maintenance errors that were immediately identified,

5. nonaging events,

6. operational errors,

7. instruments attached to the check valve,

8. installation errors, and

9. Fort St. Vrain reports.

After elimination-of the above-event types, 382 check valve failures

remainded. Each event was reviewed and data were collected as to failure

mode, method of detection, maintenance activity, and identified failure

cause. Tables B.2.-B.5 summarize the results of this effort. The NPRDS

annual reports contain data on cumulative component reliability. Copies

of NPRDS annual reports are available from INPO to NPRDS participating

members only.

Summary tables follow of 382 events involving NPRDS component VALVE,

component engineering code C (check valves).

Table B.2.--Failure.mode

distribution

Failure mode Percent

Seat leakage 70

External leakage 16

Failed to close 8

Failed to open 2

Damaged internals 4

Table B.3. Method of detection

Detection Percent

In-service and surveillance 67

test

-Incidental observation 4

Routine observation 14

Operational abnormality 11

Maintenance 2

Special inspection 2
64

Table B.4. Maintenance activity

Activity Percent

Repair/replace 93

Modify/substitute 4

Temporary measure 3

Table B.5. Identified failure cause

Failure cause Percent

Aging/cyclic fatigue 7

Normal/abnormal wear 50

Binding/mechanical damage 6

Lubrication problem 2

Previous repair/installation 2

Corrosion 4

Weld related 2

Dirty 14

Particulate contamination 1

Out of adjustment 3

Foreign/incorrect material 3

Unknown 1

Connection defect/loose part 3

Material defect 2
0

65

- NUREG/CR-4302

-Volume 1

ORNL-6193/Vl

- *'. *' .Dist. Category RV

Internal Distribution

1-4. D. M. Eissenberg 19-22. G. A. Murphy

5-6. R. B. Gallaher -23. H. E.'Trammell

7. J. A. Getsi 24. ORNL Patent 'Office

8-14. W. L. Greenstreet ' 25.' Central Research Library

15. H. D. Haynes - '26. Document Reference Section

16. J. E. Jones Jr. 27-28.. Laboratory.Records Department

17. R. C. Kryter' i' '29; Laboratory Records (RC)

18. A. P. Malinauskas

External Distribution

30. J. L. Hawley, The Walworth Company, 1400 W. Elizabeth Avenue,

Linden, NJ 07036

31. E. J. Majewski, Jr., Flow Control Division, Rockwell

International Corporation, P.O. Box 1961, Raleigh, NC 27602

32. R. Brennan, Atwood and Morrill Co., Inc., 285 Canal Street,

Salem, MA 01970

33. B. P. Brooks, Electric Power Research Institute, P.O. Box 10412,

Palo Alto, CA 94303

34. R. L. Simard, Institute for Nuclear Power Operations, 1100

Circle 75 Parkway, Atlanta, GA 30339

35. J. A. Hunter, EG&G Idaho, Inc., P.O. Box 1625, Idaho Falls, ID

83401

36. J. H. Taylor, Engineering and Risk Assessment Division,

Department of Nuclear Energy, Brookhaven National Laboratory,

Upton, NY 11973

37. A. B. Johnson, Pacific Northwest Laboratory, P.O. Box 999,

Richland, WA 99352

38. S. P. Carfagno, Franklin Research Center, 20th & Race Streets,

Philadelphia, PA 19103

39. J. Bothwell, Florida Power and Light Co., P.O. Box 029100,

Miami, FL 33102

40. J. W. McElroy, Philadelphia Electric Co., P.O. Box 8699,

Philadelphia, PA 19101

41. B. M. Morris, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

42. J. P. Vora, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

43. G. C. Millman, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852
66

44. G. H. Weidenhamer, Division of Engineering Technology, Office of

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission,

5650 Nicholson Lane, Rockville, MD 20852

45. C. Michelson, ACRS, 20 Argonne Plaza, Suite 365, Oak Ridge, TN

37830

46. R. E. Schnurstein, Energy Technology Engineering Center,

Rockwell International Corporation, P.O. Box 1449, Canoga Park,

CA 91304

47. J. Dickey, David Taylor Naval Ship R&D Center, Code 274R,

Annapolis, MD 21402

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

Department of Energy, Oak Ridge Operations Office, Oak Ridge,

TN 37831

49-50. Technical Information Center, Department of Energy, Oak Ridge,

TN 37831

51-425. Given distribution as shown in NRC category RV (10 - NTIS)
'RCONu35 U.S. NUCLEAR REGULATORY COMISSIO I REPORT NUMBERIAssg'e Dv TC lOdd V do.vl

NACM 1102 NUREG/CR-4302, Vol. 1 3201.3202 BIBLIOGRAPHIC DATA SHEET ORIL-6193/V1

SEE INSTRUCTIONS ON THE REVERSE Dist. Category RV

2 TITLE AND SUBTITLE 3 LEAVE BLANK

Aging and Service Wear of Check Valves Used in Engineered

Safety-Feature Systems of Nuclear Power Plants * DATE REPORT COMPLETED

MONTH YEAR

5 AUTHOR SI October 1985

W. L. Greenstreet R. B. Gallaher b DATE REPORT ISSUED

G. A. Murphy D. M. Eissenberg MONTH YEAR

December 1985 I PERFORMINGORaANIZATION NAMEAND MAILINGADDRESSvI/su. ZwCI B. PROJECTITASKANORK UNIT NUMBER

Oak Ridge National Laboratory

P.O. Box Y 9 IN OR GRANT NUMBER

Oak Ridge, TN 37831 B0828

10 SPONSORING ORGANIZATION NAME AND0 MAILING ADDRESS Mck ZE Codol laTYPE OF REPORT

Division of Engineering Technology

Office of uclear Regulatory Research Topical

U.S. Nuclear Regulatory Commission D PERIOD COVERED II0cI-ea

Washington, DC 20555

12 SUPPLEMENTAR NOTES

5 ABSTRACT ax .ero. o' .eu

This is the first in a series of three reports on check valves (CVs) to be

produced under the U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Research

program. This program addresses the evaluation and identification of practical and

cost-effective methods for detecting, monitoring, and assessing the severity of timedependent degradation (aging and service wear) of CVs in nuclear plants. These

methods are to provide capabilities for establishing degradation trends prior to failure

and developing guidance for effective maintenance.

This report examines failure modes and causes resulting from aging and service

wear, manufacturer-recommended maintenance and surveillance practices, and measurable

parameters (including functional indicators) for use in assessing operational readiness,

establishing degradation trends, and detecting incipient failure. The results

presented are based on information derived from operating experience records, nuclear

industry reports, manufacturer-supplied information, and input from plant operators.

I DOCUM4ENT A%ALVS4S - KE'WoRDS DESCRIPTORS

Valves, check valves, aging, service wear, maintenance, degradation,

failure mode, failure cause, measurable parameter, incipient failure,

inspection, surveillance, -monitoring, functional indicators, operating

experience.,

D IDENTIFIERS CPE EEO TERMS

I' AVAILABILITY

STATEMENT

Unlimited

16 SECURITYCLASSIPiCATION

o,. s AI

Unclassified

eTVj rtrj

Unclassified

I NUMBER Of PAGES

Id PRICE